7;' 2 e ORNL-4451 UC-70 — Waste Disposal and Processing JUL - 81970 QMTE [SSUED: s iSO SITING OF FUEL REPROCESSING PLANTS AND WASTE MANAGEMENT FACILITIES OAK RIDGE NATIONAL LABORATORY operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMISSION Printed in the United States of America, Available from Clearinghouse for Federal Scientific and Technical Information, Nationo! Bureau of $tandards, U.S. Department of Commerce, Springfield, Virginia 22151 Price: Printed Copy $3.00; Microfiche $0.65 LEGAL NOTICE This report was prepared os an account of Government sponsored work. Neither the United States, nor the Commission, nor any person acting on behalf of the Commission: A, B. Makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the informotion contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or Assumes any lichilities with respect to the use of, or for damages resulting from the use of any information, apparatus, method, or process disclosed in this report. As used in the above, '‘person acting on behalf of the Commission'’ includes any empleyee or contractor of the Commission, or employee of such contractor, te the extent that such employee or contractor of the Commission, or employee of such contractor prepares, disseminates, or provides access to, any information pursuant to his employment or contract with the Commission, or his employment with such contractor, ORNL-L451 Contract No., W-7405-eng-26 CHEMICAL TECHNOLOGY DIVISION SITING OF FUEL REPROCESSING PLANTS AND WASTE MANAGEMENT FACILITIES Compiled and Edited by the Staff of the Oak Ridge National Laboratory JULY 1970 e OAK RTIDGE NATTIONAL LABORATCRY Oak Ridge, Tennessee operated by UNLON CARBIDE CORPORATION for the - U, 5. ATOMIC ENERGY COMMISSION iii CONTENTS IntrOduCtion . . . . . . * . . * » . . . . . . » » . . . . . Summary and Conclusions . . ¢ v v v v v v e e b e e . e 2.1 2,2 2.3 2.4 2.5 2.6 2.7 2.8 Key ISSUBS & & 4o ¢ o 4 o o o o o o o o o o o o o o o Conclusions . . ¢ v v v 4 v o v 6 e e e 4 e s e e e e Magnitude of the Problem . . . . . . . + ¢« ¢ v « « « . Technical Conslderations . . « « &+ 4o ¢ o v o o o o« Transportation Considerations . . .« v v ¢« ¢ v ¢ &« « & & Economic Considerations . . . « « & ¢« o ¢ v o o o o o & Siting Considerations . . . . . ¢« « v v v 4 v o 4 4 . 2.7.1 Environmmental Considerations . . . « « « o & o . 2.7.2 Geographic Considerations . . . . . . . . « . . Health and Safety Aspects of Plant Siting . . . . . 2.8.1 Routine Release of Radioactive Materials » . . . 2.8.2 Releases from Upper Limit Accidents . . . . . Magnitude of the Problem , . . . . . . . ¢« « ¢ ¢ v ¢« ¢« « o . 3.1 3.2 3.3 3. 3.5 3.6 Projected Nuclear Power Buildup and Reprocessing Loads Reactor Design and Performance Characteristics Radiation Characteristics of Irradiated Fuels and of Wastes Generated During Spent-Fuel Processing . . . . . 3.3.1 Diablo Canyon Reference LWR . . . . . . . . . . 3.3.2 Atomics International Reference Oxide IMFBR ., . Shipments of Spent Fuel « . . . v ¢ v v ¢ ¢ v ¢« o o & & Waste Management Projections . . . « ¢« « ¢« ¢« « o« 3.5.1 High-Level Wastes . . .« v v ¢ ¢ o « o o o o o & 3.5.2 Management of Solidified High-Level Wastes . . . 3.5.3 Intermediate- and Low-Level Liquid Wastes . . 3.5.4 Solid Wastes . v v v v v v v v b e e e e e e e Ref erenc eS . . . . . . . o . . * . * . . . . . . - o . 2-1 2-2 2-3 . 2-8 2-10 . 2-1 2-16 . 2-18 . 2-18 2-20 2-22 2=23 . 2-25 . 3-1 . 3-1 . 3-11 3-11 3-15 . 3-23 3-43 . 3-55 3-55 3-62 3-62 . 3-62 . 3-68 L. iv Technical Considerations . « .+ « o ¢ o ¢ o ¢ o o o o o o o .1 Design of Fuel Reprocessing Plants . . . . . « + « « ,.1.1 Preventive Measures and Containment Criteria . ., . l.1.2 Probable Trend of Plant Design . . . . . . . . . . ;. 1.3 Plant Decommissioning Considerations . . . . 4.1.4 Design Criteria for Resistance to Earthquakes and Tornadoes + 4 « o o ¢ + o o 5 ¢ s s e & 4 s e 4.2 Waste Management Technology: General . . .« « « « & o + o 1.2.1 Applications of the Dilution-Dispersion Principle ,.2.2 Applications of the Delay-Decay Principle ., . . . 4.2.3 Applications of the Concentration-Containment Principle . . v ¢ ¢ v ¢+ s o o s s s e 0 e s L.2.4 Applications of ICRP Recommendations to Waste Releases & . ¢ ¢ v ¢ o ¢ o o o o o o o o o o o ,.2.5 Assessment of Current Waste Disposal Practices . . 4.,2.6 Definitions of TermsS . . «v « o o o « o « o o o i.3 Waste Management Technology: High-Level Wastes . . . . . h.3.1 Liquid Wast@s v v« v v v o o o o o o o o o o o o » 1.3.2 Solidification . . . . . v ¢ v v 4 v 4 4 e s 0 4 s ,.3.3 Interim Storage of Solidified Wastes . . . . . . . L.3.4 Disposal of Solidified Wastes in Bedded Salt Formations . « ¢« &« ¢ v ¢ ¢ o o o o « o o o o o ;.3.5 Disposal of Solidified Wastes in Rock Types Other Than Bedded Salt ., . ¢ + « v v ¢ o ¢ o « o o s & . Waste Management Technology: Intermediate- and Low-Level Wastes . ¢« v 4 ¢ ¢ ¢ o v o ¢ ¢ ¢ o o o o o o o L.h.1 Treatment of Liquid Wastes . « v v v v v « + « . L.1.2 Problems with Tritium . v ¢ o o « o o o o o o o o L4.h.3 Disposal by Hydraulic Fracturing . . . . . . . . i.5 Waste Management Technology: Solid Wastes . . . . . . . L]-.S.l Land B.L]'rial ® » * * . * . . . L J . » . . * . - . . 1.5.2 Disposal in Salt Mines . . « v v v v o o o & o Page . ,-I»-l . L-17 . L-17 . 4-18 . 4-18 . L-19 . 1-20 . L-25 . -52 . L-55 . L-67 . L-71 . L-75 . L-81 . 4-86 . L-86 . L-92 L.6 References « o« o« ¢ ¢ o o o o o o o o s o o o o o o s o Transportation Considerations . . ¢« ¢ & ¢ ¢« o ¢ ¢ o ¢ o o ¢ 5.1 5.3 5.l 5.5 Cask Requirements and Design Considerations , . . . . . 5.1.1 Release of Radionuclides o+ + o v o o o o o o o o 5.1.2 Increased Dose Rate . . + ¢« v v v v v ¢ v v o o & 5.1.3 Temperature Limits . . v v & o v v 4 o ¢ ¢ o « o & 5.1.4 Contamination of the Primary Coolant ., . . . . . . 5.1.5 FEmission of Neutrons from Spent Fuels and Wastes . Shipments of Spent Fuel and Waste « « . ¢ ¢« ¢ ¢« ¢« ¢« v ¢ o . 5.2,1 Effects of Source Design on the Design of Spent Fuel Casks « v v v ¢ ¢ o v o v 4 4 o o o o o o 4 . 5.2.2 Shipment of Wastes . « v v v v v & o o « o o Shipment of Product Material . . . . . . . . . . . .+ .+ .. Conclus iOnS . . . * . * . * . . . . . . . . . . Ref er enc es » L ] . . . . . . . . . . * . ® . . . . * . . . Fconomic Considerations . . . . . 6.1 Reprocessing Costs . . v v ¢ ¢ ¢ & ¢« o + o o o o o o o 6.3 6.1, 1 Economies of Scale . .« + v « ¢ ¢ ¢ o o s o « o o 6.1.2 Unit Reprocessing Costs . . . . 6.1.3 Optimization Studies . . « « v ¢« « « o & & o & Fuel Shipping Costs « . + & ¢ v ¢ ¢ ¢ o o o o o ¢ o o &« 6.2.1 Costs of Shipping Spent Fuel . . . . . . . + . . . 6.2.2 Recovered Fissile and Fertile Materials . . . . . Waste Management Costs . . . . ¢« ¢« ¢« ¢ & ¢ o o ¢« ¢« o & 6.3.1 Basis and Fconomic Model . v v v v v + o o o & o & 6.3.2 Perpetual Tank Storage . 6.3.3 Solidification and Disposal in Salt . . . . . . . 6.3, Comparison of Salt with Concrete Vaults and Granite Page -9k 5-1 . 5-3 5 5-5 . 5-5 . 5-6 5-17 5-8 5-9 5-10 5-11 6-1 6-1 . 6-6 . 6-8 6-1l . 6-15 . 6-21 6-23 . 6-23 . 6-26 . 6-31 6-42 6.3.5 Salt Mine Repository Charges . . . . . . « . . . . . . 6-U} vi 6. Site Costs v v v v v 4 v e e e e s e e e e e 6.5 Costs of Engineered Safeguards . . . . . . 6.6 References . . . « v o o o o = + o o o o Siting Considerations . . . .« « o ¢ ¢« o « o o« ¢ o 7.1 Envirommental Factors . . . ¢« ¢« ¢« ¢« & o & « & T.1.1 Meteorology .+ « « o v o o o« v o o o o 7.1.2 Geology and Hydrology . . . . . . T.L.3 Geoselsmology « ¢ o v v o o o o o o o o 7.2 Geographic Factors . . . .. . . « . . T7.2.1 Site Size . . ¢ & & o ¢ ¢ ¢ o ¢ o o o 7.2.2 Population Density of the Surrounding Area . 7.2.3 Land Usage . . +« ¢ ¢« ¢« o « o « & 7.2.L Relation of the Plant to Other NMuclear Facilities. 7.2.5 Regional Distribution of Potential Sites in the United States . . . « . &« +« « ¢ o « o & 7.3 References . . . . « ¢ ¢ v ¢ ¢« ¢ o o o o o o o Health and Safety Aspects of Plant Siting . 8.1 Buildup of 85Kr and 3H in an Expanding Nuclear Power Industry . + ¢ ¢ o o o o « o ¢ o o o o o 8.1.1 85Kr Distribution and Dose Equivalent 8.1.2 Tritium Distribution and Dose Equivalents 8.2 Routine Release of Radionuclides to the Atmosphere 8.3 8.2.1 Sources of Routine Releases . v ¢ o o « o o o 8.2.2 1%ocal Environmental Consequences of Releasing SKI' and 3H . * . - . - . . . o . » * » . . . 8.2.3 Local Environmental Consequences from All Routine Releases . . . . « 4 ¢ ¢« ¢ ¢ ¢ ¢ o & Accidental Releases of Radioactive Materials . . . 8.3.1 Assumed Properties of Fuel Reprocessing Plants . 8.3.2 Analytical Models and Mechanisms of Accidental Release . & & v ¢ 4 ¢ ¢ o o o o o o o o o o & Page 6-46 6-48 6-50 7-1 7-1 7-1 7-5 7-8 7-13 7-1h 7-15 7-15 7-16 7-16 7-26 8-1 8-4 8-l 8-7 8-11 8-11 8-18 8-35 8-39 8-L1 8-L5 vii Page 8.3.3 Method of Analysis of the Downwind Consequences of a Unit Release of Radioactive Material. . . . . 8-57 8.3.4 Downwind Consequences of Upper Limit ACCiden'tS . . . - . . . . . . . - . . . . . - . . 8"'82 8.3.5 Maximum Theoretical Accident . . . + « + + « « . . B8-92 8.3.6 Consequences of the Leakage of High-Level Wastes tO the G‘I'Ound . . o e . . . e o @ . . . o o . . 8 -97 8.l Requirements for Treatment of Routine Effluents . . . . . 8-108 8.5 RefEreNCeS « « o o o o o o o o s o o o o o o o « « « « o« 8-110 > 1-1 1. INTRODUGCTION This report contains background information which is believed to be pertinent in establishing a policy with respect to the siting of spent-fuel processing plants and their radioactive waste management facilities. It contains much basic information that has been published previously; how- ever, this material has been reviewed and revised, where required, to serve present needs. 1In addition, much new information has been included, particularly on the health and safety aspects of the problem, The information is'organized to conform generally with an outline transmitted to ORNL by the USAEC (letter from Milton Shaw, USAEC, to F. L. Culler, Jr., ORNL, dated February 16, 1968). It was developed in cooperation with Battelle-Northwest, the Idaho Nuclear Corporation, the Savannah River Plant and the Savannah River Laboratory, the Atlantic Richfield Hanford Company, and the Atmospheric Turbulence and Diffusion Laboratory of the Enviromnmental Science Services Administration. The information is analyzed and discussed principally within the context of the subject matter contained in individuwal sections; however, an attempt has been made to interpret a number of key issues more compre- hensively in the Summary and Conclusions, Sect. 2. Section 3 contains the basic data on reactor and fuel characteristics, projections of spent-fuel processing loads, fuel shipping requirements, and waste characteristics and production. In Sect. li, fuel reprocessing is discussed very briefly and waste management technology is considered in significant detail; considerations of cask design as related to safety in transporting spent fuel and solidified waste are discussed in Sect. 5; fuel reprocessing and waste management costs are considered in Sect. 63 envirommental and geo- graphical considerations of siting are reviewed in Sect. 7; and health and safety aspects are presented in Sect. 8. e 2-1 2. SUMMARY AND CONCLUSIONS The principal objective of this study is to identify and characterize the factors that may influence the growth patterns of the fuel reprocessing industry. Emphasis is placed on the siting of reprocessing plants and waste storage and disposal facilities, particularly those for high-level waste. Another purpose is to explore the need for an AEC policy on siting, which, while fully meeting the requirements imposed by considerations of public health and safety, would not present an impediment to the growth of economic nuclear power. In this section, a compilation of the key issues under consideration and the principal conclusions of the study are presented. Then, the technical information found in the body of the report relating to these issues and conclusions is summarized. In this study, it has been assumed that future fuel reprocessing plants and their associated waste management facilities will be located, built, and operated subject to the following bases, which are believed to be practical and reasonable: (1) The secondary confinement barriers (the cell, vault, water in the storage pool, and ventilation-filter system) and the tertiary barrier (the building) will be designed, tested, and routinely inspected to ensure that their confinement potential is maintained following exposure to any credible internal forces. (2) Process and confinement systems will be designed, tested, routinely inspected, and maintained so that exposure to credible external events or forces (loss of power, earthquakes, tornados, floods, hurricanes, impaction by moving vehicles, etc., but not including acts of war) will not impair the ability to shut down the plant safely and maintain safe shutdown conditions. (3) While the circumvention of administrative measures (as well, in general, as those involving instrument systems) for prevention of accidents is considered credible, it is consldered incredible that the obvious remedial measures for mitigation of the consequences 2-2 of accidents would not be instituted within hours following a clear notification of the occurrence of an accident, 2.1 Key Issues key issues of this study were considered to be the following: Are new federal regulations needed to govern the siting of fuel reprocessing plants and waste management facilities, or should licensing procedures continue to be performed using existing federal regulations for protection of the public against radia- tion (LOCFR20), siting of nuclear power reactors (10CFR100), and licensing of production and utilization facilities (1OCFR50)? Do routine releases or potential releases from accidents control the siting of fuel reprocessing plants and waste storage facilities with respect to site boundaries and population centers? After what period of time will it be necessary to limit the release of noble-gas fission products and tritium to the atmosphere to pre- vent worldwide pollution of the troposphere? What local restric- tions are imposed by the routine release of radiocactive materials to the enviromment? Current fuel-cycle economics favor the use of large-capaclity fuel reprocessing plants, Are there technical and safety factors which indicate preference for either a few large-capacity, or more numerous small-capacity, fuel reprocessing plants (sites)? Are there limitations, either inherent or as a matter of prudence, which should be imposed on the capacity of fuel reprocessing plants (independent of site size and geography) from a public safety standpoint? Is the risk to the public increased by higher inventories of hazardous materials? Does the storage of high-level liquid waste in subsurface tanks represent an acceptable waste management approach? (In this report, "storage! connotes intended retrievability and a high degree of surveillance, whereas '"disposal! connotes the reverse.) 2-3 What are the technically acceptable alternatives to tank storage of high-level waste? Is a significant economic penalty involved in providing greater assurance of containment than has been demon- strated by tank storage of waste? Does immediate solidification of ligquid waste result in an apprecilable decrease in risk to the public? What are the considerations that affect the decisions to dispose of radioactive waste on other than government-owned land? Can the reprocessing plant ever be decontaminated to the degree necessary to permit subsequent abandomment? If not, is govermment ownership of the property required? Are the hazards or the economics of shipping spent fuels, solid wastes, and fissile materials of such magnitude that these ship- ments should be limited to specified routes within regional boundaries or that shipping off-site should be precluded? 2.2 Conclusions Minimal impediments to the growth of economic nuclear power, while meeting the requirements imposed by considerations of public safety, may result from the promulgation of standards or regulations that establish (1) the acceptable chronic and acute radiation exposure of each of the critical organs of men, women, and children, both in individuals and in critical popula- tion groups, and (2) performance criteria for engineered safety features. Information is presently available to allow substantial progress toward these goals through revision of existing AEC regu- laticns., Any revisions should attempt to provide an appropriate balance of risk vs benefit on the basis of current technological alternatives, should be subject to periodic upgrading, and, preferably, should be sufficiently inclusive to apply to all nuclear fuel-cycle installations including their waste storage and disposal facilities., The criteria for chronic exposure of members of the public should be related to maximum acceptable 2-0 doses and to body organs rather than to permissible concentra- tions of radicactive effluents in air and water, The latter do not explicitly consider perhaps more limiting pathways of radia- tion exposure than those caused by submergence in (or inhalation of) air and ingestion of water. Given acceptable doses and dose rates, the designer (with the assistance of experts in the field of radiation protection) can evaluate all important pathways of radiation exposure, However, it may be desirable to retain the "maximum allowable” concentrations in air and water as point-of- departure reference values to facilitate monitoring and inspection. The criteria for acute or emergency exposure of members of the public surrounding a nuclear facility should provide guidelines for acceptable doses and dose commitments to all organs and be developed in conformance with the recommendations of authorita- tive agencies such as the Federal Radiation Council (FRC) and the National Council on Radiation Protection and Measurement. The acceptable acute doses and dose commitments for members of the public would presumably be applicable to the gquantitative determination of a site boundary and the required distance from a large population center, The performance criteria for engineered safety features in fuel reprocessing plants and waste management facilities would pre- sumably be similar to those proposed for nuclear power reactors in the proposed Appendix A of 10CFR50 entitled, "General Design Criteria for Nuclear Power Plant Construction Permits." These studies indicate that, based on the current technology of systems for cleaning off-gas streams from fuel reprocessing plants, routine releases tend to control the site boundaries. It is estimated that on-site waste storage facilities do not materially increase either the rate of routine release of radio- active material or the potential release of such material as a result of accidents, provided these facilities are designed to ensure containment following exposure to internal and external forces. For large plants, the estimated site boundaries are of 2-5 such a size that economics will probably favor the installation of noble-gas removal equipment in plants handling more than a few tons of fuel per day. The development of off-gas systems having a capability of routinely removing iodine by a factor of about 108 is necessary if FBR fuels are to be processed after decay periods approximating only 30 days. Study indicates that the worldwide distribution of 85Kr and 3H in the year 2000, assuming the complete release of these nuclides during fuel reprocessing, results in dose equivalents to man that are small (<1%) compared with current guidelines for population exposure. In other words, these nuclides will constitute radiation problems to the local enviromment long before they cause worldwide pollution hazards, These studies indicate that the confinement barriers of fuel reprocessing plants in the size range of interest, including their waste storage facilities, can be designed to maintain their confinement potential following exposure to credible internal or external forces (excluding acts of war or sabotage). Regardless of size, plants that are sited and constructed within a given set of acceptable criteria for chronic and probably acute exposure of members of the public at the site boundary are con- sidered to be equivalently safe. The costs of preventive measures and the relatively expensive confinement systems are estimated to scale in such a way that larger plants are favored, while the costs of off-gas treatment facilities required to achieve practi- cal site sizes in large plants are estimated to be modest. Conse- quently, the conclusion that economics favors fewer larger plants is valid. High-level liquid wastes can be stored safely in tanks that have been provided with adequate engineered safety features. These features include, as a minimum, two independent cooling systems (e.g., submerged coils and a reflux condenser); reinforced concrete vaults, lined with steel, which are designed either to withstand credible internal pressures without rupture or to 2-6 relieve these pressures safely by ventilation to a contaimment system with large capacitance or to a pool of water for steam suppression; installation of a contaimment structure,located above the waste vaults, that is ventilated through a condenser and filter; provision of spare tankage; and the capability for prompt, efficient transfer of the waste from any tank to a spare. Because of the regquirement for the continuous removal of heat, the effectiveness of the contaimment system will require a very high degree of surveillance. Liquid waste storage can be con- doned only as long as the reprocessing plant remains fully active. In this context, '"storage! does not constitute disposal, and 'perpetual tank storage,! even under govermment auspices, is not an acceptable substitute for disposal. The only current, technically acceptable alternative to tank storage of high-level liquid wastes is immediate solidification of the wastes, Currently, the disposal of solidified wastes by emplacement in bedded salt deposits is believed to be the safest method and has been shown to be technologically feasible. Eco- nomic studies indicate that the series of operations consisting of immediate solidification, storage of the solid wastes on-site for 3 to l years, and shipment and disposal in salt mines, could be carried out for about 0,038 mill/kwhr (electrical). This is about 20% more than is estimated for perpetual storage of liquid wastes in tanks. If the solidified wastes are shipped to salt mines after storage on-site for only one year (the earliest time believed to be feasible), the total cost would be about 0.04L mill/kwhr. Disposal of wastes of low specific heat generation rates by hydrofracturing or by emplacement in bedrock caverns may be acceptable at sites with suitable geology. The applica- tion of properly engineered safety features, together with a high degree of surveillance, can result in low risk to the public, regardless of whether the waste is stored as a solid or liquid. 2-7 Considerations of the long-term hazard of the wastes and the nearly prohibitive costs for reclaiming large areas of contami- nated land militate against any disposal (or burial) of wastes on privately owned land, All radioactive wastes must be main- tained in a retrievable condition as long as they are retained on-site. In-tank solidification of wastes, as practiced at Hanford and SRP, is not an acceptable form of storage on privately owned land because of the difficulties that would be encountered at the time of removal, Government ownership must extend to any subsurface geological formation used for disposal, as well as to the land areas above, Control of the land surface must be malntained to prevent unauthorized explorations of the formations utilized for disposal, although the surface per se can be put to agricultural or recrea- tional use. Plants and storage facilities built with proper foresight can be decontaminated and/or made sufficiently inaccessible (e.g., by grouting) so that they do not represent hazards to public safety. If it can be stipulated that all contaminated equipment and mater- ials outside the massively shielded concrete canyons and vaults be removed from the premises before abandomment of the site, then government ownership is not required. Private ownership of the site should be permitted, however, only if the site, with all its facilities, appurtenances, buildings, tarnks, cribs, and lands, can be returned to unrestricted use within some finite time (perhaps 10 to 50 years) after plant retirement. These studies indicate that shipping of all nuclear materials, except high-level liquid wastes, can be conducted safely and economically. The costs of shipping will tend to favor location of the various fuel cycle and waste disposal facilities in close proximity. The shipment of liquid wastes is considered to be unwise because of considerations of steam-pressure buildup within casks following a loss-of-cooling incident, 2-8 2.3 Magnitude of the Problem Projections of the Civilian Nuclear Power Program (Table 2.1) indi- cate that the nuclear economy will expand from about 1l gigawattis (elec- trical) in 1970 to about 153 gigawatts by 1980, and to about 735 gigawatts by the year 2000, It is expected that most of the nuclear power stations will be located in FPC Power Supply Regions III (southeastern states) and I (northeastern states) by the year 2000, and that the fewest will be found in Regions VI and VII (the western plains and mountain states). The fuel shipping industry will also expand at a very rapid rate. The number of casks to be shipped annually will increase from 30 in 1970 (an average of one in transit on any given day) to 1200 in 1980 (1l in transit on any day) and to 9500 in 2000 (85 in transit). Approximate total fuel reproc- essing rates (in metric tons/year) will increase from 100 in 1970 to 3500 in 1980, and to 15,000 in the year 2000. The heat-generation rate of FBR core fuels at the time of processing, i.e., after 30 to 75 days of cooling for FBR fuel and after 150 days for LWR fuel, will be 10 to 6 times as high as that for LWR fuel. The gross beta activity of FBR core fuels will be 8 to 5 times that of LWR fuels. The total radioactivity due to beta emitters in the accumulated wastes will increase from 210 megacuries in 1970 to 18,800 megacuries in 1980 and to 209,000 megacuries in 2000. The annual generation of high- level wastes will increase from 17,000 gal in 1970 to 1,000,000 gal in 1980 and to 4,600,000 gal in 2000, If these wastes are stored as liquids, 60,000,000 gal is expected to accumulate by the year 2000, On the other hand, if they are converted to solid forms, volumes may be reduced by a factor of about 13. Another significant type of solid waste will be spent-fuel hulls. Induced activity will be produced in either stainless steel or Zircaloy by (n,y) or (n,p) reactions; in each case, shielding will be required to handle or to ship these hulls., In addition to the induced activities, up to 0.1% of the plutonium in the fuel can be associated with the cladding. 2-9 Table 2,1, Projected Fuel Processing Requirements and High-Level Waste Conditions for the Civilian Nuclear Power Program Calendar Year 1970 1980 1990 2000 Installed capacity, Ma(e)® 10,000 153,000 368,000 735,000 Flectricity generated, 107 kwhr/year® 7L 1000 2410 4420 Spent fuel shipping Number of casks shipped annually 30 1200 6800 $500 Number of loaded casks in transit 1 1l 60 8s Spent-fuel processed, metric tons/yeara 9l 3500 13,500 15,000 Volume of high-level liguid waste generatedb’c Annually, 106 gal/year 0,017 0.97 2.69 L.60 Accumulated, lO6 gal 0.017 L.4o 23.8 60.1 Volume of high-level waste, if solidified”’@ Annually, 10° £t5/year 0.17 9.73 26.9 146.0 Acéumilated, 10° £t3 0.17 .0 238 601 Solidified Waste Shipping® Number of casks shipped annually o 3 172 L77 Number of loaded casks in transit 0 1 L 10 Significant radioisotopes in wasteg’h Total accumulated weight, metric tons 1.8 Lgo 21,00 6200 Total accumvlated beta activity, megacuries 210 18,900 85,000 209,000 Total heat-generation rate, megawatts 0.9 80 340 810 P generated annually, megacuries 4.0 230 560 770 P05¢ accumulated, megacuries b0 960 1600 10,000 13705 generated annually, megacuries 5.6 320 880 1500 13703 accumulated, megacuries 5.6 1300 6500 15,600 1291 generated annually, curies 2.0 110 Lho 670 1291 accumulated, curies 2.0 L80 2700 7600 85Kr generated annually, megacuries 0.6 33 20 150 85Kr accumulated, megacuries 0.6 2oL 570 1200 3y generated annually, megacuries 0.0k 2.1 6,2 12 3H accumulated, megacuries 0,0 7.3 36 90 238p, generated anmually, megacuries 0.0007 0.0l 0.2 0.6 238p,, accumulated, megacuries G.0007 1.20 8.3 3l 239Pu generated anmally, megacuries 0. 00009 0.005 0.05 0.2 3%,y accumulated, megacuries 0.00009 0.2 0.2 1.3 2hOPu generated annually, megacuries 0.,00012 0,007 0.06 0.21 205, acoumuilated, megacuries 0.00012 0.l 0.l 1.9 2hlAm generated annually, megacuries 0,009 0,5 L.4 15 2m‘.‘im accumulated, megacuries 0. 009 2.3 23 120 2430 generated annually, megacuries 0.00021 G.01 0.1 0.5 2)'LBAm accumulated, megacuries 0.00021 0.23 1.5 5.2 Zthm generated annually, megacuries 0.13 7.4 18 23 Zhhcm accumulated, megacuries 0.13 30 Lo 260 Volume of cladding hulls geenex‘atedfL Annually, 10° 1t 0.3 8 L0 90 Accumilated, 10° £t° 0.3 Lo 320 1030 ®Data from Phase 3, Case L2, Systems Analysis Task Force (Apr. 11, 1958). Based on an average fuel exposure of 33,000 Mwd/ton, and a delay of 2 years between power generation and fuel processing. ®Assumes wastes concentrated to 100 gal per 10,000 Mwd (thermal). Q) ssumes 1 ft5 of solidified waste per 10,000 Mvwd {thermal). ®Assumes 10-year-old wastes, shipped in thirty-six 6-in,-diam cylinders per shipment cask. fOne—way transit time is 7 days. Eprssumes TWR fuel continuously irradiated at 30 Mw/ton to 33,000 Mwd/ton, and fuel processing 90 days after discharge from reactor; LMFBR core continuously irradiated to 83,000 Mwd/ton at 148 Mw/ton, axial blanket to 2500 Mwd/ton at l.6 Mw/ton, radial blanket to 8100 Mwd/ton at 8.4 Mw/ton, and fuel processing 30 days after discharge. By ssumes 0.5% of Pu in spent fuel is lost to waste. ‘Based on 2.1 ft3 of cladding hulls per ton of LWR fuel processed, and 8.7 ft3 of cladding hardware per ton of LMFBR mixed core and blankets processed, 2-10 2.1, Technical Considerations Present-day fuel reprocessing plants make use of organic-aqueous solvent-extraction processes to separate U, Pu, and Th from mixtures of fission products and inert materials., Volatile fission products are separated during dissolution of the fuel. These fission products, and radioactive particulates from the process are removed from the plant off-gas, as required before discharge, by sorption, chemical interactions, - and filtration., In addition to the treatment of normal radiocactive effluent streams, special consideration must be given, during the design and operation of these plants, to the contaimment of radiocactivity in the event of accidents or natural phenomena such as earthquakes and tornados. The future trends in plant design for the nuclear power industry must take both safety and economy into account while reprocessing fuels containing higher quantities of fissionable materials and fission products at shorter cooling times. This implies more severe problems at almost all stages of reprocessing, including shipment and management of the waste effluents. "~ Finally, in designing these plants, consideration must be given to the problem of eventual decommissioning of the plants and the return of " the site to other uses. Much of the technology for resolving these prob- * lems either exists or is belng developed. This includes the design of - carriers for safe transport of fuels, efficient mechanical head-end equipment, continuous dissolution equipment, high-speed solvent-extrac- tion contactors, methods for improved separation and containment of fission-product gases and particulates, and improved methods of waste management, High-level wastes originate mainly from the first cycle of solvent extraction and contain greater than 99.9% of the nonvolatile fission products. The present practice is to concentrate and store these wastes on an interim basis in underground carbon and stainless steel tanks, which are equipped with devices for removing decay heat if necessary., More than 80,000,000 gal of waste are now in storage at AEC production sites. . 2-11 Although corrosion data indicate tank lifetimes in excess of 100 years might be expected, there have been 15 known instances of tank failure, all in carbon steel systems at Hanford and Savannah River. Eleven of the failures have occurred at Hanford, where it is estimated that liquid waste containing 140,000 curies of 137Cs has leaked to the ground and been retained in the soil about 10 ft below the tank bottoms, In one of the four tank failures at the Savannah River Plant (SRP), about 700 gal of waste may have escaped the liner, although ground water has shown contamination levels equivalent to only a few gallons of waste. The causes of these failures are established as stress-corrosion cracking and/or thermal stress of the reinforced concrete structures, and these factors are being taken into account in new tankage under construction; however, it is clear that many of the liquid waste storage facilities now in existence do not merit confidence in their long-term integrity. Waste management plans at Hanford call for separating about 95% of the 9OSr and 13708 from the waste and concentrating the residue, after a suitable decay period, by in-tank evaporation until the residual salts solidify into a massive cake. The strontium and cesium fractions are to be solidified and packaged for interim storage in on-site storage basins pending decisions on their long-term disposition., At SRP, the most prac- tical, safe, and economical long-term alternative to present tank storage practices is believed to be storage of these wastes in vaults excavated in erystalline bedrock about 1500 ft beneath the plant site, Toward this end, exploratory drilling has been done, hydrologic data have been collec- ted, and safety analyses have been made. As presently conceived, the storage facility would consist of tunnels, about 30 ft wide, 15 ft high, and 1000 to 2000 ft long, radiating from a central access shaft that extends vertically from the surface, At ICPP, all stored waste solutions are converted to granular solids in the Waste Calcining Facility (WCF). These solids are stored in underground stainless steel bins. The storage of liquid wastes from power-reactor fuel reprocessing will be even more difficult than the storage of current production wastes because of their higher heat-generation rates, significant rates of radio- lytic hydrogen production, and corrosive nature. Nevertheless, it should 2-12 be possible to store them safely for a limited period of time and at an acceptable cost, provided adequate engineered safeguards are built into the storage systems. The alternative to long-term or perpetual storage of wastes in tanks is conversion of the wastes to thermally and radiolytically stable solids of low solubility for burial in selected geologic formations or storage in man-made vaults., Processes for conversion of these wastes to solids are being developed both in the United States and overseas. The four U.S. solidification methods currently emphasized are the pot, spray, phosphate- glass, and fluidized-bed processes. The pot, spray, and phosphate-glass processes have been demonstrated for the AEC on a full-radioactivity-level, engineering scale in the WSEP at Hanford., The fluidized-bed process has been demonstrated at the ICPP in a large-capacity plant operating on inter- mediate-level feeds since 1963, Within the next few years, the AEC's waste solidification development program of currently known concepts will be completed. The processes will have been demonstrated using wastes from advanced fuels, and effects of severe temperature and radiation on the properties of the solidified waste products will have been measured and evaluated. This technology will provide a reliable basis for the design and safe operation of waste solidification plants. Once solidified, the wastes may be stored safely on-site (prior to disposal) and at less expense than can the corresponding liquid wastes. Conceptual designs have been published for the storage of encapsulated, solidified wastes in water-filled canals and air-cooled vaults, and for the storage of granular solids from fluidized-bed processing in air-cooled bins.,. The most promising method for disposal of the solidified high-level wastes imvolves their placement in natural salt formations. In this regard, a 19-month demonstration disposal of high-level radiocactive waste solids was carried out in a salt mine at Lyons, Kansas, using spent reactor fuel in lieu of actual solidified wastes, In the course of this program, most of the technical problems related to disposal in salt were resolved. The feasibility and safety of handling highly radiocactive materials in an underground enviromment were demonstrated; salt was shown to be stable el 2=-13 under the effects of heat and radiation; and data on the creep and plastic flow characteristics of salt were obtained, thereby making possible the design of a safe disposal facility. Cost studies indicate that this method is economically acceptable, The 2000 acres of salt that may be committed to disposal purposes by the year 2000 is only a small fraction of the 500,000 square miles that are underlain by salt in the United States. Dry openings that could be utilized for the storage of radiocactive solid wastes can be excavated in rocks other than salt; however, investi- gations are needed to delineate the effects of heat and radiation on the rock media, as well as to define more precisely the geological conditions that determine the usefulness of local sites within the most desirable geographic regions. Intermediate- and low-level wastes are usually large in volume and are handled by storage in tanks, by disposal to the ground, or by partial decontamination and release to surface waters. The release of large quantities of these wastes to the enviromment has been controlled so that the exposure of members of the public from this source has been consider- ably less than the limits recommended by the ICRP and other authoritative bodies. However, the trend is toward less dependence on envirommental disposal and greater emphasis on methods for concentration and containment of the radiocactive material. Evaporation, ion exchange, and coprecipita- tion and coagulation processes are frequently used for concentrating the radionuclides, and waste-water recycle schemes have been studied. The radioactive concentrates from treatment may be insolubilized by incorpora- tion in asphalts or certain plastic materials for long-term storage, land burial, or disposal in salt mines. Disposal of intermediate- and low-level wastes by a method based on the technique of hydraulic fracturing has been demonstrated to be both safe and economical. This method prevents radionuclides from being released, via any credible accident, into the biological enviromment by depositing them deep underground in a solid matrix. The technique is limited, however, to use at sites that are underlain by suitable geological formations of low permeability. 2-1L Tritium causes difficulty in waste management because it is unre- sponsive to separation and concentration by conventional procedures, For example, the 75 to greater than 99% of the tritium in spent fuel that appears in the low-level liquid wastes cannot be sufficiently diluted with process water in the plant to obtain the concentration specified in 10CFR20 (i.e., 3 x 107> c/cc) before discharge to surface waters. Tritium can be released more effectively as a gas to the atmosphere by vaporizing the tritiated water up the stack; under this condition, the tritium would be dispersed widely and diluted well below acceptable concentrations. Currently, from 2,000,000 to 3,000,000 ft3 of low- and intermediate- level solid wastes are buried annually above the water table on state or federal land; about one-fourth of this volume is from commercial sources. Projections of future land requirements for burial of the solid wastes that will accrue from power-reactor fuel reprocessing indicate that land consumption will increase from 1 acre/year in 1970 to 80 acres/year in 2000, and that the accumulated area of land devoted to this purpose should increase from L acres in 1970 %o 940 acres in 2000. In the interests of land conservation, it may be desirable to store part or all of this material in salt mines, Sufficient space already has been mined in bedded salt to contain all solid wastes that are expected to be generated through the year 2020. It should also be possible to utilize part of the space that may be mined for disposal of high-level solidified wastes. 2.5 Transportation Considerations The transportation of radiocactive materials to and from the reprocess- ing plant is an important consideration in plant siting, Fuel reprocessing plants receive fuel elements from the reactor, export purified fissile and fertile materials to fuel fabrication plants, and transport wastes to desig- nated disposal sites. Heavily shielded containers are used for shipping both spent fuel and solidified waste, The main difference is in the integrity of the material that is being shipped. Available evidence, based on experience, is that all types of spent-fuel shipping casks can be designed to meet present 215 contamination requirements. Ruptured spent fuel must be encapsulated prior to shipment, while fast reactor fuel will probably have to be encapsulated with sodium for heat dissipation purposes. A canister and closure can be designed such that containment of the contents is main- tained even under accident conditions., Containment may be lost due to relative deflections of the 1lid and cask body in a 30-ft impact; however, tests have shown that feasible shock-absorbing members can sufficiently dissipate energy and distribute the impact load in such a manner that seals are maintailned, A reprocessor has more control over the solid wastes leaving his plant than he has over the spent fuel entering it. Decay times of the wastes are easily varied without incurring the economic penalties that exist for spent fuel. In addition, waste containers can be designed for shipment via either truck or rail, whereas there may be little choice available for transporting spent fuel. The waste product will be doubly contained, first by a welded steel container and then by the shipping cask itself. The calcined or glass waste product is relatively immobile; although the 30-ft impact accident condition could create some fracturing of the product, this amount would be of little consequence. The 1L75°F fire accident condition could increase the center-line temperature of calcined wastes above 1650°F, but the consequences of this thermal tran- sient do not appear to be severe, Pressure increases would be small, certainly within the resistance capabilities of the steel pot whose maximum temperature will not rise more than 300°F above normal. In short, the degree of control over solid waste shipments, coupled with the fact that the fission products are in a relatively nondispersible form, indi- cate that such waste shipments should be safe, The shipment of high-level liquid wastes is not considered safe because of the possibility of radlolytic gas explosions or steam-pressure buildup within casks following loss-of-cooling incidents. Considerable experience has been accumulated in the shipment of fissile material in both liquid and solid forms. Shipments are made in a birdcage-type package, often a 55-gal drum in which a central cavity is formed by metal, wood, or other support. Since the material is free 2-16 from most fission products, little or no shielding is required; and, since there is negligible heat evolved from the material, substantial insulation may be installed to protect the material from external fires. For these reasons, the shipment of fissile and fertile products in either liquid or solid forms is feasible, Designs of product containers that will meet (and exceed) the requirements of the shipping regulations are available. Potential damage resulting from severe accidents may be expected to be minimal and thus should not affect siting of the plant. 2.6 Economic Congiderations Present-day spent-fuel processing costs, including waste disposal, are approximately 0.2 mill/kwhr (electrical) for standard light-water reactors (IWR's). Unit reprocessing costs are expected to decrease significantly as plant size increases; unit waste disposal costs will also decrease, but not as rapidly as reprocessing costs. The combined total reprocessing cost for LWR fuel is projected to decrease to 0.1 mill/kwhr (electrical) by 1985-1990 and to 0,05 mill/kwhr (electrical) by 2010, assuming that our cost estimates are valid up to about a LO- metric ton/day capacity for LWR fuel or a 20-metric ton/day capacity for FBR fuel and that plant size i1s permitted to increase to these levels by about the year 2010, (By 2020, there should be about ten reprocessing plants in operation in the U,S., with capacities ranging from 20 to 40 metric tons/day for LWR fuel or 10 to 20 metric tons/day for FBR fuel.) In making these estimates, we have used 1970 dollars and made no allow- ance for escalation. Reprocessing costs for FBR fuels are projected to be about twice those of LWR fuels on a weight basis, but can be about the same on a mills/kwhr (electrical) basis if the (core-plus-blanket) FBR burnup averages about 60% higher (and the thermal efficiency averages 25% higher) than for LWR's. If individual reprocessing plant sizes are limited to 10 metric tons/day for IWR fuel or to 5 metric tons/day for FBR fuel, the cost will stop decreasing by about 1990. In this case, about 30 reproc- essing plants would be needed in the United States by the year 2010, at 2-17 a cost penalty of 75% as compared with ten larger plants ($1.3 billion vs $0.8 billion per year in 2010). Present-day spent-~fuel shipping costs for IWR fuels are about 0,020 to 0.025 mill/kwhr (electrical) for 700-mile shipments (estimated average distance in 1970). Our estimates for 1000-mile shipments of spent FBR fuel vary from 0.0 to 0.11 mill/kwhr (electrical), for a variety of pro- posed designs. The costs for 700-mile shipments would be about 15% less. Assuming that reprocessing plants can be built in all geographical regions of the United States (as required by economic optimization of shipping and reprocessing cost totals), shipping costs should decrease about 20% by the year 2000 as the average shipping distance decreases from 700 miles (in 1970) to 350 miles; they should decrease an additional 10% as a result of technological improvements. Shipping costs in the year 2000 are projected to be $120 million for spent fuel, plus $15 million for recovered uranium and plutonium., If siting policies are sufficiently restrictive to increase the average shipping distance to 1000 miles, the total costs for the year 2000 would increase from $13% million to $200 million (not including an estimated $6 million increase in inventory charges associated with increased shipping time). The current cost for perpetual tank storage of neutralized wastes at Nuclear Fuel Services, Inc., (NFS) has been reported to be about 0,012 mill/kwhr (electrical); however, this does not include operating costs or any interest or return on investment during the first 15 years. On a somewhat more conservative basis, we estimate a total of 0.031 to 0,032 mill/kwhr (electrical) for perpetual tank storage of acid wastes in a plant reprocessing 688 metric tons/year of spent fuel irradiated to 33,000 Mwd/ton, 0.03L to 0.039 mill/kwhr (electrical) for waste manage- ment by a series of operations consisting of interim liquid storage, bot calcination, interim storage of solids, shipment, and disposal in a salt mine, Waste management unit costs decrease only slowly as the plant size increases, perhaps 35% as the size increases by a factor of 10. Thus, in 1970, waste management may contribute 15% of a total reprocessing cost of 0.20 mill/kwhr (electrical), but may contribute 25% of a total of 0.07 mill/kwhr (electrical) in the year 2010, 2-18 These reprocessing and waste management cost estimates probably should be revised upward about 10% to allow for improved containment systems costs to cover enhanced removal of rare gases and iodine, improved containment of internal explosions, and earthquake-resistant design and construction, This alternative appears to be more economical than accept- ing the extremely large and remote sites that would otherwise be required for large reprocessing plants, especlally for those handling short-cooled FBR fuel. We have not estimated the cost of inspection to safeguard against the diversion of fissile material to unauthorized use; instead, we have assumed this to be a national or international policing cost that would not be charged directly to the electric power industry. This cost should, however, scale in such a manner that fewer larger reprocessing plants, rather than many small ones, would be favored. 2.7 Siting Considerations In general, except possibly for dispersive events caused by acts of sabotage or war, engineered safety features can be devised that will miti- gate practically all of the envirommental or geographical deficiencies of a site. However, in some cases (e.g., those involving the location of a plant on a known active fault or in the center of a metropolitan city), an economic analysis of the costs of development, design, construction, and testing of special, engineered safety features will dictate against a radical departure from the conservative norm., The following sections will discuss envirommental and geographical factors in site selection. 2.7.1 Environmental Considerations The envirommental factors of principal concern in site selection are meteorology, geology, hydrology, and geoseismology. Meteorclogy. - An understanding of the meteorology of a site is important because the atmosphere provides a potential means of conveying an active, and practically unavoidable, threat to the safety of persons 2-19 downwind, Conversely, it can serve as a very large sink for the safe dispersal of radioactive materials if local problems can be avoided. Fortunately, meteorology is perhaps the best understood and most easily quantified of the envirommental factors that influence siting. The methodology for estimating concentrations and deposition of materials is relatively well established, and appropriate data for a given site may usually be obtained by relatively simple measurements, complemented with data from local or regional weather stations. Geology and Hydrology. - The geology and hydrology of the site of a nuclear fuel reprocessing plant can influence: (1) the foundations of the plant, (2) the emplacement of underground waste-storage tanks, (3) the water supply, (L) the routine disposal of liquid and solid radioactive wastes, (5) the danger from earthquakes, and (6) the consequences of an accidental release of significant quantities of radioactive materials. Geologic conditions that would be favorable for one of these consideratlons might be unfavorable for another; therefore, an ideal enviromment does not exist, and the selection of any actual site will require compromise. Per- haps the only valid generalization is that all of these considerations will be easier to evaluate if the geology and hydrology of the site are simple and predictable, In comnection with the consequences of accidental release, simplicity in the hydrologic and climatologic enviromment is particularly desirable. Only in cases where the conditions can be analyzed in detail and with con- siderable confidence can predictions of the possible results of an accident be made. These predictions will allow proper precautions to be taken against such an eventuality, as well as suggest effective remedial measures in the event of an accident. A simple geologic and hydrologic envirorment also makes it possible to determine, with confidence, the most effective local methods for ultimate disposal, the maximum quantities of radioactive material that may be released to the enviromment, and the best methods for monitoring the enviromment to make certain that safe levels of discharge are not being exceeded. 2-20 Geoseismology. - Faults, vibrations, and tsunamis are the major earth- quake-induced phenomena to be considered in the siting and the design of nuclear facilities (including fuel reprocessing plants). All of these are important for some sites along the West Coast of the United States; on the other hand, vibratory effects are generally the sole concern in the eastern part of the country. In many regions of the United States, it appears that earthquake-induced phenomena can be adegquately considered through currently acceptable engineering practices; however, in some highly seismically active regions, the high degree of geoselsmological conservatism requires that unique and presently improved designs be considered, 2.7.2 Geographic Considerations The primary congsideration in acquiring a site for a fuel reprocessing plant is to provide sufficient distance between the plant and private lands to ensure that the general public will not be harmed by either normal oper- ations or by credible accidents. Second, the site should be located at a place where the aggregate cost of raw materials, transportation of materials to the plant, manufacturing, and transportation of finished products to the market will be at a minimum. In present plants, the basic raw materials are water, nitric acid, solvent, and aggregate for concrete. Either a railroad spur or a waterway with barging facilities 1s a practical necessity since some spent-fuel shipping casks weigh 50 to 100 tons. Paved highways are necessary for trucking smaller casks, raw materials, finished products, and waste. Manufacturing costs are dependent on an adequate supply of skilled labor and on the prevailing wage scales in the vicinity. Conven- iently located housing and community facilities are desirable. Long commuting distances, poor social facilities, and undesirable climates all tend to result in a large labor turnover, The plant must have adequate acreage for possible future expansion, suitable soil or rock foundations for heavy concrete structures, and reliable electric power, preferably from two independent sources. Ideally, the plant should be located relatively near power reactors and sites designated for the disposal of high- and low-level wastes,. 2-21 Site Size. - The site boundary is determined most accurately and restrictively by the requirement that the direct exposure of the surround- ing public to radiocactive gaseous or liquid effluents must be maintained at allowable levels. Penetrating radiation that escapes through the shielding used in the plant is not normally a consideration. Studies at Hanford indicate that controlled areas extending 0.5 to 1 mile from the plant are desirable for the control of "nuisance contami- nation" resulting from a temporary loss of control of relatively small quantities of radioactive materialg. Such minor releases might result from outside decontamination operations on large pieces of process equip- ment or shipping casks. This is not an absolute limitation; it is possible (i.e., at increased cost) to house those facilities that would potentially disperse low-level contaminants. It was found that the routine release of noxious nonradioactive chemicals to the atmosphere (most significantly NOZ) would dictate a site boundary about 1 mile from the stack. This is also not an absolute limitation, since such gases may be removed from stack effluents to practically any extent required using present technology. The discharge of low-level liquid radioactive effluents is determined primarily by the relative flow rate of groundwater and surface water as a function of distance from the plant and the subsequent use of the water. Surrounding Population Density. - Federal regulations (1OCFR100) specify that there shall be a zone of low population (presently not quan- titatively defined) surrounding a reactor plant. The primary concern is to prevent the general public from receiving somatically or genetically significant doses of radiation. The cost of indemnification is also of concern; claims resulting from overexposure to radiation during an accident would probably be directly proportional to the number of persons involved. Land and Water Usage. - Special considerations are required when fuel reprocessing plants are located in areas where there are mechanisms for reconcentration of the radiocactive effluents and pathways for ingestion 70 137Cs) are known by the public. Since certain radionuclides (e.g., 7 Sr, to concentrate in crops and fish, the restrictions on the discharge of low-level liquid waste effluents containing these nuclides to surface waters subsequently used for irrigation or fishing may be more severe 2-22 than if the water were used only for drinking., Deposition of radioiodine from gaseous wastes on grass, followed by the cow-milk pathway to the thy- roids of small children, may result in maximum permissible air concentrations which are lower by a factor of 500 to 1000 than those for inhalation. Relation of the Plant to Other Nuclear Facilities. - The fuel reproc- essing plant should be designed and located to take into account adjacent nuclear facilities, including reactor plants, other reprocessing plants, and waste disposal sites, Effluents from the plant must not mask nuclear instrumentation at adjacent sites. Accldents in the plant should not cause undue haste and unsafe evacuations of adjacent sites. In addition, the effluents from each plant must be restricted in such a way that their combined effect will not endanger the safety of the public. In practice, the effect of these restrictions has been minimal at the production plants and national laboratories; the incremental costs of additional engineered safety features are generally offset by the decreased costs resulting from shared personnel, services, and facilities. Regional Distribution of Potential Sites. - Results of a rather general study (see Sect. 7.2.5), which takes into account the results presented elsewhere in this report, indicate that there are many potential sites for fuel reprocessing plants in each of the electric utility districts in the United States. Of the districts that are predicted to have a large concentration of power reactors, it appears that the least difficulty would be encountered by siting in the Southeast because of the low popu- lation density, adequate access to railroads, and low selsmic probability; the most difficulty should be encountered in siting near the West Coast, primarily because of the high seismic probability. 2.8 Health and Safety Aspects of Plant Siting The principal criterion for judging the adequacy of a site for a fuel reprocessing plant is the provision that no undue risk exists with regard to public health and safety in the surrounding areas. Presgsent and foresee- able technology requires that such plants routinely discharge small quan- tities of radiocactive materials to the atmosphere; for this reason, and o i 2-23 also because of the large inventory of hazardous materials, there is always a small, but finite, probability of a major discharge. 2.8.1 Routine Release of Radioactive Materials The consequences of, and the site boundary distances dictated by, routine releases from fuel reprocessing plants were estimated by assuming the following: (1) ORNL meteorclogical conditions, (2) the complete release of noble gases and tritium, (3) iodine decontamination factors (DF's) of 1000 (present technology) and 107 in plants for processing highly irradiated fuels after cooling periods of 150 and 30 days, respec- tively, and (i) a particulate-release-rate model that agrees satisfactorily with existing data. For reference purposes, the acceptable concentrations at the site boundary were selected as one-third of the air concentrations listed in 10CFR20, Appendix B, Table II, Column 1, with the exception that the 1311 concentrations were reduced by a factor of 700 to account for the grass-cow-milk pathway to the thyroids of small children, Table 2.2 compares the average annual air concentrations of radio- nuclides at the (dictated) site boundaries of conceptual plants with those estimated for the NFS, MFRP, and BNFP plants. The downwind doses resulting from the normal release of radionuclides from plants are estimated to be controlled by the noble gases and iodine, The magnitude of the distances to the site boundary estimated for plants of large capacity indicates the need for at least partial removal of the noble gases and removal of a larger fraction of the iodine than was assumed for the analysis. On the basis that the site boundaries dictated by routine releases should be no greater than those dictated by "upper 1limit accident," equipment for removing 50 to 99% of the noble gases appears necessary for plants having capacities of more than a few tons per day; an iodine removal capability greater than that demonstrated in present technology will be required for LWR plants having capacities greater than about 6 to 10 tons/day, while DF's as high as 108 will be required for FBR plants if the FBR fuel is to be processed after decay times of only 30 days. Table 2.2. Fraction of Maximum Permissible Average Annual Air Concentrations Resulting from the Routine Release of Radionuclides at the Site Boundaries of Existing, Proposed, and Conceptual Private Industrial Fuel Processing Plants (7?60 days of operation per year) R Average Fuel Characteristics Distance Annual Fraction of 1/3 x(LOCFR20) Concentrations at Site Boundary®’® Plant Specific Decay to Site Aeolian Capacity Burnup Power Period Boundary Dilutign 8 159. 141 Fission Product Actinide Plant (metric tons/day) {Mwd/ton) (Mw/ton) {(days) (1cm) (sec/m”) 5Kr~133Xe 34 911311 Solids Solids NFS 1 20,000 3 150 1.5 2,2 x 1077 0.23 0.002 0.47 0.0007° - (3,300, 000) (18,000) (3.1) (~1) - MFRP 1 h3,800 30 160 0,6-3 1.1 x 107 0,12 0.005 0.23 <0,0005 <0.11 (3,300,000) {100,000} (3.1) (<2.2) (<0.63) BNFP 5.8 35,000 4o 160 2 5.7 x 1078 0.24 - 0,02 0.27 0.003 0,017 (L4 x 10"} (600, 000) (21) (60) (3.5) LR 1 33,000 30 150 <0.6 6.3 x 1077 0.58 0.05h 0.15 0.003 0.021 (2.9 x 10°) (180,000) (0.56) (13) (0.43) LWR 6 33,000 30 150 0.5-6 1.8 x 1077 Lo 0.093 0.25 Q,002 0.018 (1.7 x 10) (1,100,000) (3.1) (L1) (1.3} LWR 36 33,000 10 150 5-29 3.0 x 1078 1.0 4 0,093 0.25 0,001 0.009 (1.0 x 107) (6,500,000) (20) (120) (3.8) FBR 1 33,000 =8 30 <0.6 6.3 x 1077 0.92 0.073 0.52 0.0003 0.008 (4.6 x 10%) (2140,000) (3.6) (4.8 (0.16) FBR 6 33,000 58 30 1.5-10 1.1 x 1077 1.0 . 0.079 0.56 0. 0001 0.003 (2.8 x 10") (1,450,000) (22) (5.0) (0.31) FBR 36 33,000 58 30 7-L2 1.9 x 1070 Lo ¢ 0,079 0.56 0.0001 0.003 (1.7 x 107} (8, 700,000) (130) (54) (1.9) “The reference values selected are one-third of the concentrations found in 10CFR20, Appendix B, Table II, Column 1, They are 1 x 10_7, 7T x 10-8, 1x 10'10, 3x 10'10, and I} x 10713 for 85Kr -133Xe, 3H, mixed LWR fission products, mixed FBR fission products, and mixed actinides respectively. The LOCFR20 value for 311 gas reduced by a factor of 700, resulting in a reference concentration of 1.l x 10'13. The 10CFR20C value for 1311 was reduced by a factor of 700, resulting in a reference concentration of 1.l x 10743, The 10CFR20 value for 1°°T was reduced by a factor of 7000, resulting in a reference concentration of 3 x 10_15. b . . : . Release rates, in curies/year, are given in parentheses, T2-2 2-25 2.8.2 Releases from Upper Limit Accidents The consequences of upper limit accidents were estimated assuming that the acceptable annual dose commitments resulting from exposure to the radioactive cloud or inhalation at the site boundary are values recommended by the National Committee on Radiation Protection for annual occupational exposure. The dose commitment analysis was based on the agssumptions of flat downwind terrain and exposure to the radiocactive cloud. The consequences of downwind ground contamination and additional exposures by such phenomens as reentrainment were not considered as mechanisms that would limit plant siting. Excessive levels of ground contamination would cause inconveniences, require expensive decontamina- tion procedures, and result in property loss; however, they probably would not present an unavoidable threat to the health and safety of the public. In Table 2.3, the total dose commitments resulting from various upper limit accidents at the accident-dictated site boundaries of the conceptual plants are compared with estimated dose commitments at the site boundaries of the NFS, MFRP, and BNFP plants. Confinement and ventilation systems in fuel reprocessing plants remove particulates of nonvolatiles dispersed under accidental conditions to such an extent that the upper limit acci- dents are controlled by the release of such volatile and semivolatile materials as the noble gases, lodine, ruthenium, cesium, and tellurium, The maximum site boundaries for all plants are estimated to be determined by the whole-body dose resulting from the release of volatile "fresh" fission products from a nuclear excursion (30% and 1% release of iodines from IWR and FBR plants, respectively, plus 100% release of the noble gases). Credible upper limit accidents in well-designed facilities for the interim storage of either ligquid or solid wastes are estimated to be inconsequential with respect to those from processing operations in the plant, It is assumed that future liquid waste storage facilities will be designed to maintain their containment potential when exposed to credible internal (e.g., a hydrogen-air explosion) or external (e.g., loss of power, earthquake, etc.) forces. The consequences of a liquid Table 2.3. Estimated Lifetime Dose Commitments to Critical Organs Resulting from Upper %imit Accidents at NFS, MFRP, BNFP, and Conceptual Plants for Processing LWR and FBR Fuels®’ Conceptual LWR FPlants of Capacity: Conceptual FBR Plants of Capacity: 1 Metric 6 Metric 36 Metric 1 Metric 6 Metric 36 Metric Type of Release NFS MFRP BNFP Ton/Day Tons/Day Tons/Day Ton/Day Tons,/Day Tons/day "Fresh" fisgsion products Total number of fissions 10°0 1020 1018 2.7 x 10°0 1.6 x 10°T 1.6 x 10°% 8.0 x 10°° 1.6 x 107 2.k x 10°% Thyroid dose commitment, rems ~2 26 - 9.4 30 30 0.65 1.0 1.3 Whole-body dose commitment, rems 0.09 0.002° 5.0 5.0 5.0 5.0 5.0 5.0 Noble gases (85Kr and 133Xe) Release, curies - - - 70,000 420,000 2,500, 000 350,000 2,100,000 13,000, 000 Whole-body dose commitment, rems - - - 0,054 0.18 1.0 0.18 88 L.k Halogens (1311 and 1291) Release, curies 1.7 1.7 1.1 3.1 18 55 1100 6500 3700 Thyroid dose commitment, rems - 0.017 - 0.05 0.2 0.5 Lh.6 22 ?7 Semivolatile fission products A Releagse, curies - - 15900 760 L500 L5000 3600 7300 11,000 ' 0%, curies - - 1500 110 2500 2500 1300 2600 3900 pY Lung dose commitment, rems - - ~0.0007¢ 2.7 8.9 8.9 5.0 7.9 13 Nonvolatile fission products and transplutonics Release, curies 1.1 5 120 3.3 20 20 37 h 111 Ce, curies - - 23 0.58 3.5 3.5 2.3 h.7 7.1 2LL2Cm, curies 1.7 0.011 0.068 0.068 Lung dose commitment, rems - - <0.0007° 0.008 0.03 0.03 0.0k 0.06 0.07 Bone dose commitment, rems (~0,02) 0.075 - 0.02L (0.005) 0.077 (0.0L7) 0.077 (0.017) 0.060 (0.024) 0.10 (0.0L) 0.12 (0.05) Plutoniuvm Release, alpha curies 0.65 <3 0,11 0.16 0.98 0.98 0,30 0.61 0.91 Bone dose commitment, rems 13 <0,0007% 6.7 (0.26) 22 (0.8) 22 (0.8) 8.6 (0.3) 1L (0.5) 18 (0.7) Distance to site boundary, km 1.5 0.6 2 o.LL 2.0 2.0 1.2 2.0 2.8 %The underlined numbers are those that fix the radial distance to the site boundary. bThe numbers in parentheses are the first-year dose commitment for those cases in which the first-year dose commitment is not equal to the lifetime dose commitment. ®The Allied Chemical Corporation reports the external exposure dose from beta and garmma radiation. 2=27 waste tank boildown that occurs over several days (assuming that no remedial action is taken) with the accompanying release of radioactive material directly to the atmosphere by entrainment in the steam, or a loss of canal water with resultant meltdown and entrainment of calcined waste, are sufficiently serious that they must be rendered incredible by the provision of adequately engineered safety features, 3-1 3. MAGNITUDE OF THE PROBLEM This section contains the data characterizing the fuel reprocessing and waste management operations associated with the civilian nuclear power economy that is projected for the United States over the next three to four decades. Much of the material serves as the basis for further calculations and considerations in subsequent sections of the report. A recent projection of nuclear power growth and of fuel reproc- essing requirements for the entire nation is broken into components corresponding to the geographical regions of the Federal Power Commission; design and performance characteristics are summarized for a typical light- water reactor (LWR) and a liquid-metal-cooled fast breeder reactor (LMFBR); isotopic compositions and radiation characteristics of the irradiated fuels from these reactors, and of the wastes generated by the reprocessing of these fuels, are tabulated; and projections of spent-fuel shipping requirements and waste management operations are made. For the primary purposes of this report, only projections through the end of this century are emphasized; however, in many of the following tables and figures, the forecasts have been extended an additional 20 years as a matter of general interest. 3.1 Projected Nuclear Power Buildup and Reprocessing Loads The projection of nuclear power growth and fuel reprocessing require- ments that served as a bagis for this study was taken from Phase 3, Case j2, a study made by the AEC Systems Analysis Task Force (SATF) in April 1968.% This particular case considers power generation by only two reactor types. Light-water reactors predominate until the early 1990's, but fast breeder reactors go on-stream during the 1980-1981 period and assume an increasingly significant role thereafter (Table 3.1 and Fig. 3.1). The %More recent projections have been made by the AEC (see USAEC Report WASH-11)49, in press), but these forecasts were not available at the inception of this study. The differences between them and Phase 3, Case ;2 are not of sufficient magnitude to affect the fundamental thesis and conclusions of this report. 3-2 Table 3.1. Projected Installed Nuclear Capacity in the United Statesa Installed Capacity [gigawatts (electrical)]b Period LWR IMFBR Total 1970-1971 1L 0 1 1972-1973 32 0 32 1974-1975 52 0 52 1976-1977 17 0 77 1578-1979 112 0 112 1980-1981 149 b 153 1982-1983 181 12 193 1981;-1985 203 28 231 1986-1987 211 60 271 1988-1989 223 95 318 1990-1991 223 145 368 1992 -1993 223 201 Lol 1994-1995 223 265 1,88 1996-1997 223 337 560 1998-1999 223 120 643 2000-2001 209 526 735 2002-2003 192 655 8l,7 200l-2005 201 768 969 2006-2007 238 861 1099 2008-2009 27 990 1237 2010-2011 360 1023 1383 2012-2013 387 1150 1537 201 ~2015 368 1329 1697 2016-2017 506 1357 1863 2018-2019 541 1493 2034 *Taken from Phase 3, Case L2, Systems Analysis Task Force (April 11, 1968). bThe installed capacities given here correspond to those in existence at the midpoint of the respective two-year periods. 3-3 ORNL-DWG 69 - 67B6R 108 INSTALLED NUCLEAR CAPACITY (megawatts) (¢) TOTAL NUCLEAR 5 (6) FAST BREEDER REACTORS (¢} LIGHT-WATER REACTORS 103 1970 1980 1990 2000 2010 2020 CALENDAR YEAR ENDING Fig. 3.1. Installed Nuclear Electric Generating Capacity in the United States (SATF Phase 3, Case L2). 3-b total installed nuclear generating capacity increases from 1l,000 Mw (electrical) in 1970 to 153,000 Mw in 1980, and reaches 735,000 Mw in the year 2000, The quantities of spent fuel discharged by reactor and by fuel type are presented in Table 3.2, In the case of LWR's, enriched uranium and plutonium recycle fuels are listed separately; the IMFBR | estimates include both core and blankets. The Phase 3, Case L2 projections for the entire United States were apportioned into the elght geographical power supply regions of the Federal Power Commission (FPC), as shown in Fig. 3.2.1 This was done by using previous AEC estimates of nuclear power growth through 19802 and a distribution proposed by Searl3 for the year 2000, For the pres- ent study, the AEC data were regrouped according to FPC region to serve for the 1970-1980 period. For the period between 1980 and 2000, the data were smoothed and normalized in order to yield the same distribution in the year 2000 as was forecast by Searl. Finally, for the years following 2000, the assumption was made that the nuclear power disitribution remained unchanged., Table 3.3 presents the resulting projections of installed nuclear power capacity for the FPC regions, and these data are presented graphically in Fig. 3.3. The projected regional distribution of spent fuel is given in Table 3.4 and Fig. 3.4. These data were generated by assuming a time lag be- tween power generation and spent-fuel discharge computed on the basis that the distribution in any year is proportional to two-thirds of the power distribution one year earlier, and to one-third of the power distri- bution two years earlier. Mathematically, 1 L = T [=f + =f t,r B (t-1),r 3 (t—2),r] , " where Lt,r = load generated in region, r, at time t, - T = total load generated at time t, f = fraction of power generated in region r. 3-5 Table 3.2. Projected Spent Fuel Discharge Schedule by Reactor and Fuel Type? (Metric tons discharged during the two-year period indicated) Period ILWR-U LWR-Pu Recycle ILMFBR Total 1970-1971 L5 0 0 15 1972-1973 1,291 3L 0 1,325 1974-1975 2,238 16l 0 2,1,02 1976-1977 3,307 386 0 3,693 1978-1979 5,276 509 0 5,785 1980-1981 6,308 1,445 91 7,840 1982-1983 6,483 L,,104 359 10,946 198L-1985 7,028 7,211 L9 14,988 1986-1987 7,621 9,118 2,475 19,21l 1988-1989 7,284 9,57h 5,439 22,297 1990-1991 7,981 9,943 9,221 27,145 1992-1993 7,965 8,911 10,612 27,488 1994-1995 7,553 7,100 11,994 26,647 1996-1997 6,863 6,822 1,477 28,162 1998-1999 6,76k 5,897 16,1.35 28,796 2000-2001 6,610 5,640 17,872 30,122 2002 -2003 L, 98l 4,803 21,232 31,019 200-2005 L,h3lk 5,299 25,0LL 3k, 777 2006-2007 1,168 6,467 26,118 36,753 2008-2009 3,037 10,018 27,082 10,137 2010-2011 0 15,299 32,693 17,992 2012-2013 0 18,107 30,973 119,080 201 -2015 0 20,727 33,708 5L, L3k 2016-2017 0 20, 785 36,26l 57,0u9 2018-2019 0 23,813 10,221 6l1,03L #Taken from Phase 3, Case 42, Systems Analysis Task Force (April 11, 1968). 3-6 ORNL—DWG 66 —4670 "—fl"l]b ---rl- \ e N LEGEND X [ 77 Federal Power Commission Power Supply Area - m Regional Grouping of Power Supply Areas wpgpy ST m Fig. 3.2. Federal Power Commission Electric Power Supply Areas. 3-7 Table 3.3. Projected Geographical Distribution of Nuclear Power Capacity (Gigawatts installed as of beginning of year) FPC Region Designation Total in Year I II ITI v v VI VIT VIIT U.S.A.2 1970 2.8 0.3 0.5 2.3 0.0 0.0 0.6 0.4 7 1971 5.9 0.5 2.8 k.0 0.0 0.0 0.7 0.4 1 1972 8.8 0.8 6.6 .8 0,0 0.5 0.8 0.5 23 1973 10.3 2.9 8.0 6.6 0.0 1.3 0.8 2.0 32 1974 12.9 2.9 10.8 7.7 0.7 1,2 1.6 3,6 h1 1975 16.1 3,9 14.0 9.0 0.7 1.6 2.1 h.6 52 1976 19.7 L.2 16.9 10.7 0.7 1.6 3.5 6.5 61 1977 23.14 5.4 20.7 12.3 1.4 1.5 L.l 8.5 77 1978 28.3 5.9 25.7 1h.L 1.4 2.3 5.1 10.9 oL 1979 33.8 7.1 30,5 16.9 1.5 2.0 6.2 13,7 112 1980 39,1 8.9 6.4 19.1 2.3 3,2 7.2 16.6 133 1982 Lol 12,7 7.3 23,7 .1 1.2 9.4 22.7 173 198 58.3 16.8 57.4 28,1 6.3 5.1 11.5 28.5 2172 1986 66.1 21.6 66.9 32.1 9.2 6,0 13.7 34.3 250 1988 73.8 27,2 77.5 37.6 13.8 7.1 16.3 1.0 29 1990 81.2 33,9 88.54 43.2 20,2 8.3 19.2 L8.1 342 1992 88,0 L1.6 98.6 49,3 29.6 9.6 22,7 55.8 395 199, 95.8 50,9 110, 56.6 39,7 11.1 26.7 6.1 455 1996 104.7 61.5 122.5 6L L 51.3 12.8 31,7 73.7 523 1998 115.3 73.9 136.3 73.h 6L.6 14.8 37.8 8.2 600 2000 127.5 87.4, 151.8 83.3 78.8 16.9 L5.6 95.2 686 2002 143.6 103.2 170.7 oh.6 95.7 19.L cl,.2 108, 790 2004 161.9 121.,L 192,99 108.0 113,8 22,3 6L.3 122.3 907 2006 181.2 141.3 216.6 122.8 133.6 25.) 75.3 136.7 1033 2008 202.L 162.5 2l1.7 138.3 1s5h.L 28.8 87.3 151.5 1167 2010 225,0 185.,3 269.2 15L4.7 176.9 32.5 99.5 165.8 1309 2012 2h9.7 208.3 297.6 172.2 200.3 36.3 113.2 181.5 1459 201 275.2 233.3 325.,5 190.2 225.,3 LO.L 127.1 197.1 1616 2016 301.7 259,0 358.L 209, 250.9 LL4.7 1L42.0 213.8 1779 2018 328.6 285.8 390.8 229.4 277.1 L9.0 156.5 230.L4 1948 2020 357.2 313.3 250.1 304.2 53,6 172.1 2L47.0 2122 L2L.0 “Phase 3, Case L2, Systems Analysis Task Force (April 11, 1968). ORNL—DWG 68— 4165RAR 1000 oo INSTALLED NUCLEAR CAPACITY (gigowatts) o o4 1970 1980 1990 2000 2010 2020 CALENDAR YEAR BEGINNING Fig. 3.3. Projected Geographical Distribution of Installed Nuclear Electric Generating Capacity in Eight FPC Power Supply Regions. i 3-9 Table 3.li. Projected Geographical Distribution of Spent Fuel Discharges (Metric tons discharged during year) FPC Region Designation Total in Year T IT III v v vi VII VIII U.S.A2 1970 56 2 0 12 0 0 15 9 9l 1971 156 11 16 88 0 0 3L 20 321 1972 218 20 82 156 0 0 32 20 528 1973 31l 28 206 185 0 12 32 20 797 1974 357 75 275 216 0 35 29 52 1040 1975 1130 105 351 263 16 L3 L6 109 1362 1976 1,90 116 21 281 23 L6 69 139 1585 1977 653 1h6 562 358 25 55 105 20L 2108 1978 805 181 705 L2 L1 56 138 282 2635 1979 950 205 857 189 50 72 169 358 3150 1980 1064 228 963 535 18 78 193 L2l 353L 1982 149 3,0 1359 701 98 120 270 631 1,968 1984 1966 52, 1893 92 177 167 378 918 6966 1986 2506 752 2487 1210 20l 221 501 1250 9222 1988 27hl 930 2803 1360 k21 25k 578 1h53 10,542 1990 3332 1278 3542 1722 685 328 750 1892 13,530 1992 3211 1400 3543 17h2 887 337 778 1952 13,849 1994 2866 1411 3247 1636 1037 320 759 1857 13,134 1996 2897 1596 3353 173L 1271 343 830 1978 1k,002 1998 2803 1703 3302 17h9 1hhh 349 877 2004 14,232 2000 2839 1863 3362 182 16L6 368 956 2093 1,949 2002 2767 1931 3299 1816 1763 370 1016 2078 15,040 2004 3092 2261 3685 20L8 2105 L20o 1192 2338 17,110 2006 3182 2422 3803 2138 2277 Lh2 1285 2LoL 17,953 2008 3335 2629 3982 2267 2487 L70 1L03 2513 19,087 2010 12l 3340 L9°B8 2828 3182 589 1794 3075 23,860 2012 Looh 3387 LB93 2819 3249 593 1830 3007 23,871 2014, L593 3857 sht6 3171 3710 671 2096 3318 26,891 2016 L70, Loo6 5591 3256 3868 692 2187 33L9 27,651 2018 5378 L4639 6393 3741 LL4BO 799 25LO0 3788 31,757 2020 5556 L8L9 6605 3884 L701 832 2661 3868 32,956 ®Phase 3, Case lj2, Systems Analysis Task Force (April 11, 1968). SPENT FUEL DISCHARGED {(metric tons} Fig. 3.h4. 3-10 ORNL-DWG 69-6788R 104 N o ol (o2} N 1970 1980 1990 2000 2010 2020 CALENDAR YEAR ENDING Projected Discharges of Spent Fuel in the Eight FPC Regions. 3-11 The projected annual discharge of fissile plutonium isotopes in the eight FPC power supply regions during the period 1970-2020 is presented in Table 3.5 and Fig. 3.5. Again, a time lag, which was computed in the same mammer as that used to estimate the distribution of spent-fuel dis- charges, was applied. 3.2 Reactor Design and Performance Characteristics Two 1000-Mw (electrical) reactors whose design and performance charac- teristics have been previously defined were chosen as representative types for this study (Table 3.6). The LWR is the reference pressurized-water type described in a recent AEC-sponsored task force stu.d;)r.)'L Fueled with Zircaloy-clad UO, (3.3% 235 34.8 Mi/metric ton and achieves a fuel exposure of 33,000 Mwd/metric ton, U), it operates at an average power level of The IMFBR is the reference oxide design that was developed by Atomics International (AI) for the Systems Analysis Task Force Study.5 It is fueled with stainless-steel-clad UOE--ls.é% PuO2 less-steel-clad, slightly enriched UO2 in the axial and radial blankets. Fuel exposures of 80,000 Mwd/metric ton at a specific power of 175 Mw/metric ton, 2500 Mwd/metric ton at 5,5 Mw/metric ton, and 8100 Mwd/metric ton at 10 Mw/metric ton are achieved in the core, the axial blanket, and the in the core, and stain- radial blanket respectively. The projected refueling cycle is once every 153 days, when one-third of the core and the axial blanket and about three- sixteenths of the radial blanket are discharged. 3.3 Radiation Characteristics of Irradiated Fuels and of Wastes Generated During Spent-Fuel Processing The masses, radioactivity, and thermal power of fission products, actinide isotopes, and activation products present in the irradiated fuels from the LWR and the LMFBR described above, and in the wastes generated during spent-fuel processing, were calculated as a function of decay time using the computer program ORIGEN. The nuclear characteristics of the Diablo Canyon Nuclear Power Plant reactor were used in the calculations for the reference LWR since some of the required data were not given for Table 3 .5. (Metric tons discharged during year) 3-12 Projected Geographical Distribution of Fissile Plutonium Discharged by Reactors FPC Region Designation Total in Year I IT ITT IV v VI VII VIII U.S.A.2 1970 0.5 0 0 0.2 0 0 0.1 0.1 0.9 1971 0.5 0 0 0.3 0 0 0.1 0.1 1.0 1972 1.4 0.1 0.5 1.0 0 0 0.2 0.1 3.3 1973 1.5 0.1 1.0 1.1 0 0.1 0.2 0.1 L.1 1974 2.1 0.4 1. 1.2 0 0.2 0.2 0.3 6.0 1975 2.9 0.8 2.2 1.8 0.1 0.2 0.3 0.7 9.0 1976 3L 1.0 2.8 2.0 0.1 0.3 0.4 1.2 11 1977 4.0 1.1 3.5 2,2 0.2 0.4 0.7 1.8 1 1978 5.2 1.2 L.6 2.6 0.2 0.4 0.9 2.0 17 1979 6.8 1.4 5.8 3.1 0.3 0.8 1.3 2.8 23 1980 8 2 7 L 0.4 1 1 3 26 1982 13 3 12 6 1 1 2 6 L) 198l 21 6 20 10 2 2 L 10 7h 1986 26 8 26 13 3 2 g 13 97 1988 37 13 38 18 6 3 8 20 140 1990 55 21 59 29 11 5 12 31 220 1992 66 29 72 36 18 7 16 L0 280 199 73 36 82 I 26 8 19 L7 330 1996 8L U6 97 50 37 10 2l 57 1100 1998 91 55 107 57 17 11 29 65 - L60 2000 97 N 115 62 56 13 33 71 510 2002 111 77 132 73 70 15 L1 83 600 200 13l 98 160 89 91 18 52 101 740 2006 140 106 167 oL 100 19 56 106 790 2008 1h7 116 175 100 109 21 62 110 840 2010 180 146 215 123 139 26 78 13L 1040 2012 182 151 218 125 145 26 81 13l 1060 201L 198 166 236 137 160 29 90 143 1160 2016 202 172 21,0 140 166 30 9l 1l 1190 2018 2138 205 282 165 198 35 112 167 11,00 2020 o0 213 290 171 207 37 117 170 1500 #Phase 3, Case L2, Systems Analysis Task Force (April 11, 1968). 3-13 ORNL-DWG 69-6789R PLUTONIUM DISCHARGED (metric tons) 109 4970 t980 1990 2000 20410 2020 CALENDAR YEAR ENDING Fig. 3.5. Projected Discharge of Fissile Plutonium in the Eight FPC Regions. 3-1k Table 3.6. Summary of Reactor Design and Performance Characteristics LWR IMFBR Fuel form Power, Mw (thermal) Thermal efficiency, % Core Avg. sp. power, Mw/metric ton Burnup, Mwd/metric ton Charge, metric tons Enrichment, % Refueling interval, full-power days Refueling fraction Fuel element Rods/element Elements/reactor Rod length, with plenum, in, Cladding Outside diameter, in. Wall thickness, in. Axial blanket Avg, sp. power, Mw/metric ton Burnup, Mwd/metric ton Charge, metric tons Enrichment, % Radial blanket Avg, sp. power, Mw/metric ton Burnup, Mwd/metric ton Charge, metric tons Enrichment, % Refueling interval, full-power days Refueling fraction Fuel element Rods/el.ement Elements/reactor Rod length, with plenum, in. Cladding Outside diameter, in. Wall thickness, in. Oxide pellets 3083 35.4 34.8 33,000 88.6 (U) 3.3 (2350) ~365 1/3 square 20l 193 148 Zircaloy-l (Inconel spacers) 0,427 0.0243 Oxide pellets 2500 L0 175 80,000 12.6 (U + Pu) 15.6 (237py) 1532 1/3% Hex2 2178 2504 142 304 Ss@ 0.252 0.015% 5.5 2500 g:g2(§%%U) 10 8100 : 26.7 (U 1.96 (Q%SU) 153 ~3/16 Hex 169, 91 39, 87 8L, 72 304 SS 0.35, 0.51 0.015 *Also applicable to the axial blanket which is an integral unit with the core assembly. W 3-15 the reference reactor that was described in the task force report."L The LWR was assumed to operate at a constant average specific power of 30 Muw/ metric ton (equivalent to a load factor of 0,85), In the case of the AL Reference Oxide LMFBR, the core was assumed to operate at a constant average specific power of 148,15 Mw/metric ton for ShO days (equivalent to a load factor of 0.85). The specific power of the axial blanket was input as a step function, varying from 2,27 Mw/metric ton (at startup) to 6.99 Mw/metric ton (at a discharge time of 540 days) and averaging L.63 Mw/metric ton, The specific power of the radial blanket varied from 2.32 Mw/metric ton (at startup) to 1l.38 Mw/metric ton (at discharge) and averaged 8.4 Mw/metric ton. In this study, it is assumed that the core and blankets are mixed proportionately ("homogenized") prior to processing, yielding a fuel mixture having an average burnup of 33,000 Mwd/metric ton, Transient conditions of about 700 nuclides in the current data libraries of ORIGEN were calculated for each reactor, and the results are presented in the form of summary tables of the most significant isotopes present in spent fuels and wastes in terms of mass, activity, and thermal power. These properties are tabulated for each isotope and for each element. All resulis are based on one metric ton of uranium charged to the LWR, and on one metric ton of uranium-plus-plutonium originally charged in the "homogenized" IMFBR core and blankets. 3.3.1 Diablo Canyon Reference LWR Fission Products, - Tables 3.7 through 3.12 present the calculated masses, radioactivity, and thermal power of significant fission products present in the wastes generated by the processing of spent Diablo Canyon reference fuel (or in the spent fuel before reprocessing) as a function of postirradiation decay times of 90 to 365,250 days. Tables 3.7, 3.9, and 3.11 give the welght, activity, and thermal power, respectively, for individual isotopes; these same data, summed for each fission-product element, are given in Tables 3.8, 3.10, and 3,12 respectively. 3-16 Table 3.7. Masses of Fission-Product Nuclides Calculated to Be Present in Spent Diablo Canyon Reference LWR Fuel and in the Wastes Generated by the Reprocessing of This Fuel DIABLO CANYON REFERENCE LWR - WASTE DECAY TIMES (PROCESSED 90 DAYS) PNWER= 20.00 MW/MT, BURNUP= 132000. MWD/MT, FLUX= 2.91E 13 N/CM*%2-SEC NUCLIDE CONCENTRATIONS, GRAMS / METRIC TON FUEL CHARGED TO REACYOR CHARGE 9C.D 150.D 26%.D 3652.D 36525.D 365250.D SE 78 0.0 2.55F 00 2.55E 0C 2.55E 00 2.55F 00 2,55E 00 2.55F OO SFE 790 0.0 S.65E 00 S,65E 00 5,65E 00 5,65E 20 S.64F 00 5.59F 05 SE 8G L0 1.03E €1 1,.63€ €1 1.02F 01 1,.03F 01 1.03F 0! 1.03E 01 BR 81 (.0 1.51F 01 1,51 €1 1.51f €1 1.51F Q1 1,51 01 1.51E Ol SE 82 0.0 2.25E 01 3,258 01 3,25F 01 2,25€ (1 2.25F 01 2,25 () KR 82 .6 4,M8F N1 4,088 31 4,08FE £1 4,08F N1 4,08% 0) 4.08F 01 KR 24 0.0 1.11€ 02 1,11F 02 1,11F 02 1,11F 02 1.11F 02 1.11F (2 KR 85 (0.0 2,906 01 2.87E Q) 2.,76E 1 1.55E 91 4,69E-02 [.D RB 85 (.0 Q.41 D1 9,456 N1 9,55F 01 1,88E 22 1.23F (02 1,23€ 02 KR R& CL.0 1,02 02 1.92F 02 1.92F 02 1.92F 02 1,92F 02 1.92F (2 RB 87 (.0 2395 02 2,398 02 2.3%9E 02 2,39F 02 2,395 02 2.39F (02 SR B8R N, 2,51€E £2 3,51F 02 3,51F N2 2,51F 2 3,51F 002 2,.51F 02 SR 89 0,0 7.57E CO 2.40F 00 1.94E-01 1.81FE=20 0.0 .0 Y 89 0.0 4.5TE 02 4.61F 02 4.,65F 02 4.65F 02 4.65F 02 4.65F 02 SR 9 Do 5.44F 2 5.41FE 02 54.24E 02 4,278 €2 4,64F D) 1,06E-C8 IR 90 0.0 2.64% 01 2.86E 01 3.64F 01 1.43F 02 5.24F 02 5.70F C2 Y 91 0.0 1.32€ 01 £.50€ 00 S.,16£-01 7,70€E-18 0.0 0.0 IR 91 r,C 5.96E N2 6.03F 02 6.09E €2 6.,N9E 12 6,09 N2 6,N9E D2 IR 92 0.0 6.,63E 02 6.63F 02 €.63F 02 6.63F 02 H.63F 02 6.63F 02 ZR 92 0.0 7.35F 02 7.35F 02 7.35E 02 7,35FE 02 7.35F 02 7.34F 02 IR Q4 € .0 7.89% (2 T.BIE 02 T.89F N2 T,89F 02 T.,89E N2 T.R9E (2 IR 95 0.0 2.475 01 1.31€ 01 1.32F OC 7.94F=-16 0.0 c.C NR 95 (0.0 2.21E C1 1,326 01 1.51F OC 4.81€E-10 0.9 0.0 MO @8 N0 Te23E (2 TL44F 002 T.6TE 02 7.7T0E 02 T.70E 02 7.70F 02 IR 96 .0 2.30F 02 8,30F 02 B,3CE 02 8,30F 02 8.30F 02 8.30F 02 MO 96 0.0 3.95¢ 01 3,05E 01 2,86Ff Q1 3,9%FE 01 3.95F 01 3,95F Q1 MO Q7 0,0 8.38E 02 8.38F 02 8.38F 02 8,38F 2 8,38E 02 8,38F (2 MO 98 .G Be49E 02 B.A9E 02 B.49E 02 B.49FE 02 B,49F 02 R,49E 02 TC 99 0.0 8.25E 02 8.35F 02 8.35E 02 8.35F 02 B8.35F 02 8.22F 02 MoIon 0.0 9.71% ©2 9.,71F 02 9, 7IE 72 9.71E 52 9,.71F 02 ©.71F 02 RUICY 1,0 5.56F C1 5,56F (1 5.56FE 01 5.56E 01 5.56F 01 5.56F 01 RUICl 0.0 T.76E 02 T.76F 02 T.76E 02 T.76E 02 T.76E 02 7.76F 02 RU102 0.0 T.68E 02 7+68E C2 T.68F 02 T.68E 92 T7.68F 02 T7.68E 02 rRUIC3 .0 T.95E 00 2.78E OO0 6,45E-02 0.0 0.0 0.0 RH103 (.0 2.84F 02 3.89F 02 3,92€f 02 3.,92€ 02 3,92€ 02 2,92F 02 RUIC4 0.0 5.38E 02 5.38E 02 S.38FE 02 5.38E N2 5.38F 02 5.38F D2 PDICH D.0 2.46F £2 2,46E 02 2.46E 02 2.46F 02 2,458F 02 2.46F )2 PDICS 0.0 2e94E 02 2.94F 02 2.94F (02 2.94% 02 2.94F 02 2.94F 02 RUI0S 0.0 1.37€ 02 1.22E 02 8.13F 01 1.64E-01 0.0 0.0 PDICEs 0.0 2.12E 02 3,26F 02 3.67E 02 4.48F N2 4.48F 02 4,48F (2 PNIOT 0.0 2.36E 02 2.36F 02 2.36F 02 2.36F 02 2.36FE 02 2.36F 02 PDIC8 0.0 1.56E 02 1.56FE 02 1.56E 02 1.56€ 02 1.56F 02 1.56F Q2 AGINO 0.0 6.,00F D) 6.00FE G 6.00F 01 6,00F N1 6.N03E N1 6.00F §1 PD110 0.0 3.26F 01 3,.36E 01 3,26E 01 3.36FE 01 3.36F C!1 3.36F 0Ol cD110 (0.0 4.10F Ol 4.10FE 0 4.11E 0 4,11E 01 4,11F 01 4.11F 01 €o111 re.o0 1.71E €1 1.71E 01 1.71E 01 1.71€ 01 1.71F 01 1,71€ O} €H112 0.0 9.17E €O 9.17E 00 9.17€ 00 9.17F CO 9.17E 0C 9.17F 00 cD114 0.0 1.22F C1 1.22E 0Y 1,22E 0) 1422E 0! 1.22F D1 1.22F 01 IN11S 0.0 1,27 €O 1,20F €0 1.20FE €0 1.20F 00 1.20F 00 1.20F 00 cdlle 0.0 3.78FE 00 3.78E 00 3.78FE 00 3.78E 00 3,.78F 00 2,78F 00 SN116 0.0 2.6TE 00 2.6TE DQ 2.67F 00 2.67E ON 2.67FE 30 2.67% 00 SN117 a.n 3.94F 00 3,94E OC 3,94F 0C 3.94F 20 3,94FE 00 3,94E 00 SN118 0.0 4,02€ 00 4.02F 00 4, C2E Q0 4.02E 00 4.02F 00 4.02€E 0O SN119 0.0 4,16E 00 4.,16E 00 4.17€E 00 4.17F OO0 4.17F OO 4.17E QO SN120 9.0 4,35 DO 4.35E 00 4.35E CC 4.35E 00 4.35F 30 4.35% OO DTABLO CANYON POWER= 3-17 Table 3.7 (Continued) 30.C0 MW/MT, BURNUP= NUCLIDE CONCENTRAYIONS, $SBI21 SN122 SB123 SN124 $812% TE125 SNY26 1127 TF128 XF128 1129 TE13N XF13n XF131 XF1132 €S1212 X €134 CS12a BAl124 €sS135 X F134 BAl36 €S127 BAY3T RA128 LA 20 CEl140 Crl41 PR141Y CF142 ND14?2 ND143 CFl144 ND144 ND145 ND1 44 PM147 SM147 NDY 48 SM148 SM149 ND1SO SMY &N $M1 51 SM15? FY1 &3 SM154 FU) sS4 GD154& EU1S&S GDYSh GN15A8 TR1%9 SUBTOT TOTALS DTOODIOOIVOINDNOITIVAODNIINOOIDO IODOICD 3OO IIQOONDNOIADODID O CHARGF .0 ‘e @& & 9 ¢ & & 8 & & & w CODOOCIDIOGODVDOOIVDITNDODNODDODIOOTTOODOOIICIVIIVIOODIOIVDCICOLO a » &« ® 2 5 @ a & & 0 & 5 ® * e * & & s & @ 0 * 9 & & & ¢ @ 0.0 90.D 4&.54F 00 5.,10F 00 5,.45F T.62E 8.NQE 1,25F 1.94F 3,90F 1.34E 2.98F 2.32E 4,26E 1.07€ 4,00F 1.15F 1.01E 1.53F 1.73F B,4NF 2,26 2.32F 2.57E 1.23F 5.01F 1.21F 1.27E 1.46F 7,15F 1.'9F 1.18F ? JRF 8.08F ?2.79F 1.06EF T.0AE 7.11€ 1.12E 5,99F 2,78F 2.51¢ £.19F 1,83 2,15 4.23F 9,20F 1.20F 3,70€ 4,73€ 2.22F S.22F R,45F 1.33F 1.77E 1,51F 3.51E REFERENCE LWR - WASTE DECAY TIMES (PROCESSED 90 33C00. MWD/MT, FLUX= 2,91F 13 N/CM*%2-SEC DAYS) GRAMS / METRIC TON FUEL CHARGED TO REACTOR 150.0 4 ,54F 5.10F 5.47E T.62F T.6TE 3,58F 1.94F 2,93F 1.34F 2.98¢F 2.33F L,26F 1.07€ 4, 09€ 1.15F 1.01F 1.53F 1.64F 9.33F 3,26F 2.32F 2.57€F "1.22€ S .48F 1.21F 1.27F 1.46F 1.98F 1.20F 1.18F 2.08F 8,08EF 2e41F 1.10E T.06% T.11E 1.07¢ 6.4hE 3.78F 2.51F 6.19E 1.83E 3.15E 4.,23E 9,20F 1.30E 3.70E 4,70% 2.55E 5.00F R.45E 1.33E 1.77€ 3.51F 2.51F 00 00 eo 365,0 4,54E 00 5.10E 00 5.50F OC T7.62F 00 &, 59F 00 4, 68F 0O 1.S4F 01 3,98F 0} 1.24F 02 2.98F 0C 2 .33F 02 4o 26 Q2 1.07e 01 4,09E Q2 1.15E 73 1.01E 03 1.52F 03 1.34EFE (2 1.238 02 2. 265 02 2,32 02 2.57E 21 1.21t 03 1.21F 123 1.27¢ 03 1.46E 03 1.20F 03 1.19F 032 2.08F 1 8,08 C2 » 1. 43F 02 1.20F 03 T. 06E 02 7.11F 02 Q.18 N1 g, 018 O 3. 78E 02 2.51E Q2 &€.19E CC 1.82E 02 3,155 €2 4.21F 1 Q.20 1 1.3CE 2 2, 0E 01 4,50 Q) 3. T4E €D 3.99E 00 8.45E 01 1.23E 1 1.77E €O 3,81F 04 3.51F 04 3652.0 4.54E 00 5.10€ 00 5.51E 0O 7.62F 00 6.54E-01 1.08€ 01 1.94€ 01 3,99€ 01 1.34€ 02 2.98E 00 2.33¢ 02 4.26E 02 1.07€ 01 4,09F 02 1.15€ 03 1.01F 03 1.53€ 03 6.40E 0N 2.51F 3.26E 2.32F 2.87F 9.81E 2.98E 1.21F 1.27¢ 1.46€ 2.0 1.20¢ 03 1.18€ 03 2.088 N R.08E 02 4.68E-02 1.34E N3 7.06E 02 7.11€ 02 R,4TE NO 1.62F 02 2.78E 02 2.51E 02 6.19F 00 1.81€ N2 3.15F 72 3.92€ 01 9.,20€ 01 1.30E 02 3.70€ 01 3,10E 01 1.85€ 01 1.27E-01 2.,45€ 01 1.338 01 1.77€ 00 1,51F 04 2.51F 74 36525.D 4,54 DO 5.10E O 5.51E 22 T.62E 00 6.03E-11 1.14E 01 1.94F 01} 3,99F 01} 1.34€ 02 2.98E OC 2.33F 02 6,26F 02 1.07€ 01 4,09F 02 1.15€ 02 1.01F 03 1.53FE 03 3oq8E-l3 2.57E 02 2,26E Q2 2.32F 02 2.57¢ 01 1.23F C2 1.16F 02 1.,21F 03 1.27F 03 1.46F 03 D.0 1.20F 1.18E 2.08E R,08E 0.0 1.34F T.06E 02 T.11F 02 1.72F Q2 3,785 02 2.51F 12 6.,19F QC 1.83E 02 3.15€E €2 1.91F 21 9.21F 01} 1.30F 72 3.7 01 6.29FE-01 4 ,89F 0?2 1.37E=16 B8.45F D1} 1.32F 1 1.778 CC 3.50F N4 02 c3 a1 02 02 3.51F 04 365250 .D 4.54E 00 5.10F 00 5.51E 00 7.62E 00 0.0 1.14E 01 1.93F 01 3.99€ 01 1.34E 02 2.98E 00 2.32€ 02 4h.26F 02 1.07€ 01 4.,09€ 02 1.15€ 03 1.01E 03 1.53F 03 0.0 2.57€ 02 3.26F 02 2.32E 03 2.57€ 01 1.14F-07 1.28€ 03 1.21F 03 1.27€ 03 1.46F 03 3.0 1.20€ 03 1.18E 03 2.08E 01 8.08E 02 0.0 1.34E 33 7.06F 02 7.11€ 02 n.e 1.72F G2 3.78F 02 2.51F 02 6.19€ 00 1.83E 02 3.15F €2 1.47E-02 9.21F C1 1.233E G2 3,70E 01 7.28E-18 4.96E 01 0.0 2.45€ 01 1.33€ 01 1.77F 00 3.50E 04 3.51F Q4 Table 3.8. 3-18 Total Masses of Significant Fission Product Elements Calcu- lated to Be Present in Spent Diablo Canyon Reference LWR Fuel and in the Wastes Generated by the Reprocessing of This Fuel NITABLO CANYON REFERENCE (WR - WASTE DECAY TIMES (PRDCESSED 9C DAYS) POWNFR= 20,rN MW/MT, BURNUP= 32000, MWD/MT, ELEMFNT CONCENTRATIONS, GRAMS / METRIC TON FUEL CHARGED CHARGE 90.D 150.0 265.D0 3652.D H ol To20E=02 T 14E=[2 €,90E=C2 4,16E-02 GA 0.0 1.04E=19 0,0 0.0 0.0 GF c.0 3,79F=01 2,79E-01 2,79E-01 3.79F-01 AS ~a0 A,78E-02 8, 7RE-02 R, TRE=-02 R,TAF-N2 SF 2.0 5.20F Q1 5.20F 01 5.20E 01 5,20F 01 BR 0.0 1.51F 01 1,51F 01 1.51F 01 1.51F 01 KR 0.0 2,74E 02 2,7 N2 2, T2E 02 3.60F 02 RB ¢.r 3,276 02 3,33F €2 2,34F 0?2 3,46E 02 SR 0.0 ©,02E 02 8,965 02 R.85€ 02 7.78F 02 Y roC G,71F 02 L,6RF 02 4,65F 02 4.65F 02 1% £on 2A,66F 03 3,66F 02 3,66F 03 3,77F 03 NB 0.0 2.22E 01 1.32F €1 1,51F 00 3.93E-03 MO 0.0 2,42F 03 3,44F D2 3,46F 02 3,47F 03 TC ©on R.25F 02 R, 2SE 12 8,35F N2 R,35F 02 RU 0.0 2.28F 03 2.26F 03 2.22F 03 2,14€ 02 RH 0.0 2,R4F 02 2,R9F N2 2,92F N2 3,92F 02 PD L0 1.285 £3 1,29F 22 1,32E 02 1,.41F 03 AG 0un 6.01F 01 6,01F 01 6,01F 01 6,00F 01 cn 0.0 8.256 01 8,35F 01 8,26F 01 8,36F N IN a0 1,206 70 1,20F A€ 1,20F A0 1,29E £0 SN 0.0 5,155 01 5,15F 01 S5.15¢ C1 5,15€ 01 SR 0.n 1.80E 01 1.776 01 1,66E C1 1.07€ O1 TF A 5.66F 02 5,65F N2 5,665 02 5,72F 02 I 0.0 2.7T1F 02 2,72F 02 2.73F 02 2.73EF 02 X F 0.0 S,42F 03 §,42F 02 5,425 02 5,42F 02 cs N 2.74F 03 2,72E 03 2,6RE 02 2,32C 13 BA C.0 1.378 03 1.30F 03 1,42F €2 1,.79E 03 LA 0.0 1.27F 03 1,27F 02 1,27F 02 1,27E 02 re Fal 2.92F £3 2,88F €2 2,78F 02 2,64F N3 PR 0.0 1.195 02 1,208 02 1,20F €2 1,20F 02 ND 0.0 2,87E 03 3,91F 02 4,00F 02 4,15F 03 PM "t 1,128 €2 1,07E 02 9,16F 01 8,47E N0 SM ~ .0 R,03F 02 B,08F 02 8,23F 02 9.04F 02 FU 0.0 1,826 02 1,82F €2 1,80F 02 1.64F 02 6D e 1.028 02 1,03E G2 1,058 C2 1,22F 02 T8 ron 1.R2E N0 1,80F 00 1,7RE OC 1.77F 00 DY 0.0 9.93£-01 1.02E OC 1.C7F 00 1,12F 00 HO 0.0 RL4OF=(12 R,40F-02 £,49E-02 2,40F-02 ER N 2.79F-02 2,.79F-72 2,79F-02 2,79E-02 TOTALS 0.0 3.51F 04 2,51F 04 3,51F 04 3,51F 04 36525,.D 2.61E-04 D.O 3,79F=01 B.78BE-0i? 5,20F 01 1,51 01 3,45F 02 3,625 02 3.97F 02 4 ,65F N2 4,15 072 3 45FE-02 3.,47F N2 B.35F (2 2.14F D7 3,92€ 02 1.41F 03 6.00F 01 A.36F 31 1.,20F nr £.,15e M 1.00E M 5.,72F n? 2.72F (02 S.42F 02 1.454F 22 2.65F 02 1.27F 032 2.hGFE 272 1.20F ¢ 4,15 03 2.86F-11 3,928 02 1.54F (2 1.54F 27 1.77% CC 1,12F 0OC 2 ., 79E=02 3.51F 04 FLUX= 2.91F 12 N/CM%%2-SEC TO REACTOR 3565250.0 0.0 0.0 3,79E=-01 8,7RE=D?2 5.19F 01 1.51E 01 3,45F 02 3,62F 02 3,51F 02 4,65F 02 4,20F 03 3,40E=-01 2.,4TE (03 R,22F 02 2.14F 03 3.82F (2 1.41F 03 £.00F 31 8.36% )1 1,2°F Or 5.14F 01 1.00F 31 5.T2F €2 2.73F 02 5.42F 03 1.24F 03 2.77¢ 03 1.27F Q2 2.64F 03 1.20F 23 4,15F 03 Cor f8,73FE 02 1.72E 02 1.54F 32 1.77F O 1.12F 00 8 ,40F=02 ?.80E=02 3.51F Q4 3-19 Table 3.9. Calculated Radioactivity Levels of Significant Fission Product Nuclides Present in Spent Diablo Canyon Reference LWR Fuel and in the Wastes Generated by the Reprocessing of This Fuel DIABLD CANYON REFERENCE LWR -~ WASTE DECAY TIMES (PROCESSED 9T DAYS) POWER= 20,00 MW/MT, BURNUP= 32000, MWD/MT, FLUX= 2.,91F 13 N/CM*%2-SEC NUCLIDE RADIOACTIVITY, CURIES / METRIC TON FUEL CHARGED TO REACTOR CHARGE 90.0 159.D 265 .0 3652.0 136525.0 265253.0 H 2 (.0 6.98F 02 6,92E 2 6.69E 02 4.03F 02 2.53E 0C 0.0 KR 8= 0,0 1.13€ 04 1.12F 04 1.CRE 04 6,05F 02 1,.83F 0! C.0C RR /% (0,0 1.72€ 01 1.85E N0 A.49E-04 0.0 .0 C.0 SR 89 (.0 2.14F 05 9,60F 24 5,47 03 5,12E-16 2,7 o.C SR 9C 0.0 T.69E 04 T.66F 04 T,55F 04 5,04E 04 6,.56F 03 1.50E-Q6 Y Q0 ¢.0 ThRAE 04 T.66E 04 T.S55E 04 H,DFF D4 6,57F 02 1,80FE-D6 Y @1 0.0 3,22 £5 1,59E 0% 1.26FE 4 1.88E-13 2.2 0L0 IR 92 0,0 1.88E 00 1.8R8F 00 1,RRFE 00 1,.,88% 00 1,.88F 0C 1.R8F 00 IR 65 0.0 5.24F 05 2,76E 0% 2,.79E (& 1,.68F~-11 0.0 0.0 N8 Q5M (.10 1.11F N4 5,86F D3 5,92 02 3,56E-13 0,0 Lol TC 9¢ 0.0 1.42E DY 1.,42F 01 1,42F 01 1,42E 01 1.42F Q1 1.42F 01 RUIC3 2.0 2.55E 05 8,91E €4 2,07E 03 NN D.0 c.r0 RHIO2M 0.0 2.55F 0S5 R,91F 04 2,C7F 02 0.0 0.0 0.0 RUICE C.O 4,59 05 4,10 05 2.72F 08 5,5QF 02 0.0 c.0 RHYDA 0,0 L,59F £S5 4,1NFE 085 2,73 0F §,80F N2 0,0 0.0 AG1'0M C.0O 3,08F 02 2.61F 02 1.4FFE €2 1.78E~02 0.0 C.C AG11C D.C 4,01FE ¢1 2,40F 01 1,89 01 2.31E-02 Q.0 C.0 CR11ISM €0 1,178 N2 4,472F N1 1,39 €C 0,0 2,7 g.r SN11gM 7.0 1,29 01 1.,09F C1 6,02F CC H.62E~04 0.0 C.0 SNI23IM 0,0 S.11F 02 2,66FE 02 1.11F 02 1,25F=-06 0.0 0.0 SB124 0.0 1.,73E €2 B,63F 1 7.,2CE 00 2.33F-16 02,7 D.0 SN12% r,0 1.67¢ 01 2.00E-01 2. ¢€1E-0f 0.0 0.0 0.0 SR125 N,0 R,4RF 03 BJ13FE 03 £,99FE 03 4,93F 02 6.29F-08 0.0 TEI2SM 7,0 2,228 N3 3,28 03 2,89F 03 2,87F 02 2,65E-DB 2.C TE127M 0,0 Q,N4E €3 6,18 €3 1,87F 02 1,22E~-06 0.0 C.C TF127 0.0 R,94E 03 6,11FE 03 1,56F €3 1,30F=06 0.0 0.0 TE1?29M 0.0 2.278 €4 6,69F N3 B,26F N1 N0 N.0 e TE1?20 .0 T46F N4 4,29F 73 S,36F N1 2,0 0.0 C.C 1121 C.O 3.81F 2 2.17¢ 0C 1,Q8F-08 0.0 0.0 C.0 YEIZIM 0,0 1.068 02 2.,27E CC 1.08BE-25 1,0 D3 Dl €sS134 1,0 2.28F 08 2,13 N8 1_,75F 0F 8,23E 03 5,18F-10 0.0 CS!'3s 0.0 BE.I0FE 02 2.08FE 01 2,18E-04 0.0 0.0 0.0 cs137 0.0 1.07¢ 05 1.06F 05 1,05F (% 8.53F D4 1,27F 04 9,93E-036 BATI2TM (0 2,29F N4 9,946F N4 9,82F 4 T7.98F 04 §,97E N3 0,29F=06 BAI40 QL0 1.11F 04 4,208 02 3,7RE-Q2 3.0 0.0 0.0 LAY&C C.0 1,28F 04 4,G5FE 02 Q,75E 01 0.0 0.0 J.0 Cr141 70 ?.N6E 05 S,67E N4 S, TCE 02 0,0 N0 CaC PR142 n.0 1.44F 04 6,94F 02 1,.21E-02 0.0 0.0 0.0 CrEl44 (L0 8.92% C5 7.70F 05 4,5€F 05 1,50F 02 0.0 C.C PRILL4 7,0 B,92F N8 T,TAE 08 &,86F 05 1,.50F D2 0,0 CL.0 ND14T 0.0 2.16F 03 5,10F 01 7,54E-0% 0.0 0.0 0.0 PMI&T (.0 1.04F 05 Q9.94F 04 R,51F 04 T7.,87F 03 3,59E~07 (C.0 PM14RM (0 1.MAF 03 3,92F 02 1,.13E M1 N0 D0 W0 pMisg 0,0 2,828 01 3,15F 01 9,(7E-01 Q.C 0,0 0.0 SM151 0.0 1,18F 02 1,15F 02 1,15F 02 1,07F 03 5,21 0? 4.C0F-01 EYLs? r.n 1.16F 01 14158 71 1.11F ) 6.59E 0N 3 ,64F-072 N,0 GD1%2 0,0 2.66F 01 2.,24F 01 1,21F 01 9.85F-04 0.0 G.C cfyYl1s4e 0,0 £.8TE 02 6£.,82F 03 6,65F 02 4,50F N2 9,12F J1 1.MBE~-15 EUIsSE N0 £.TOF N3 £,27F 02 5,N9E O 1,62F 72 1,75F-13 0.0 FUYE6 0.0 2,518 03 2.19F 02 1.,06F-0C2 0.0 0.0 c.0 TBYAD 0.0 B.34F 02 3,00F 02 2,80F Q1 T7.59F=13 0.2 D0 GME2 1.0 1.BAF N2 1,66FE 02 1,115 N2 2,16%-N1 D ,.N Tl TRIA2M C,0 1.856F 02 1.66F 02 1.11F 02 2.16F~01 0.0 0.0 SURTOT 0.0 5.,19F 06 &.,39F Q6 2.22F 06 2,17E 05 3.44F D& 1,.65F 01 TOTALS ©,0 6.19F N6 4,39F 06 2,.22F 06 3,178 N5 3,44F 04 2,07F 01 DIABLO CANYON REFERENCE LWR - WASTE DECAY TIMES POWER= 24,00 MW/MT, ELEMENT RADIDACTIVITY, H GA SF KR 2R SR Y IR NB MO TC RU RH AG ce IN SN SR TE I Ye €S RA LA CF PR ND PM SM EU Gn T8 DY HO Table 3.10, CHARGF r.n 0.0 CeD x ” 0 » . » . » . . - 3D QDTS DO DIDDTIOD 3 5 o . e & & o ® & & & e . ¢« & o+ 9 . QOO TN D THO DD IDDIAOOIDDTITIDDTNOD DO TOTALS .0 3-20 Calculated Radiocactivity (Total) of Fission Product Elements Present in Spent Diableo Canyon Reference IWR Fuel and in the Wastes Generated by the Reprocessing of This Fuel an . n 60q8F ?2 3.20E-113 2,092E-01 1,13F 54 1.728 01 2.90F 05 2,99 N5 B.24F 05 9,80F S 2,M2E-04 1.478 01 T.14F 05 7,145 0§ 2,58 (2 .17 €2 T.10E-005 S.41F C2 f.A5F 03 DeBAFE N4 2.21F Q2 1.17F 02 1,238 05§ 1.11F 05 1.285 04 1.198 6 QQfiéE "R 2.15F O3 1.06F 08 1.15E 03 1.72€ C4 2.13F 02 T.21F N2 1.11E-10 7.71E~C5 6.19F D6 BURNUP = 180,00 £.Q2E 02 O.C 1,93F=-01 1,128 T4 1.85F CO 1.73¢ 8 2.258 5 ?2.T65F OF S.24F 0OF IOCZF-IF 1.,42F 0 4.99F 05 4 ,99F MK 2.95F (2 L.44F (01 3.13E—fi5 2,788 (02 B.22E G2 2.65F 4 2+.21E 00 2.28F OC 2.20F N5 1.00F C5 4,28F (2 B.27E % T«7T1E 08 5.10F 01 9.,9R8F a4 1.15F 12 1.348 04 1,83 02 4L,66F 2 S.EQF-lfi T« T1E-CS 4,29F 06 32050, CURTES /7 MFTRIC TON 245,.0N &, 69F D2 C.D ?,92E -1 YL.CRE 04 €. AOFE =04 g,.09F N4 R,RBIE 24 2. 70F D4 5 +Q0FE Q& C.0 1.42F 01 2. 15E 0% 2. 15 5 1.64E Q2 1.42F QC 1.8596=06 1.18F (02 7.00F 02 6, 16F £32 2. 80F-02 ] . DRE =08 2. 8CE CF 2, A2F Q4 Q, 75E 01 L, ,56F N§ 4 .56F Q% TeS4E~(QF R.B1F D& 1.15F 03 1.17F 04 1.22¢€ 2 1.49F 12 0.0 7. TIE=-DE 2.22F 06 MWD/MT, FLUX= {PROCESSED 9f DAYS) 2.91F 12 N/CM*%2-SEC FUEL CHARGED TO REACTOR 2662.D 4,02F Q2 0.0 Bong“Pl £.C5F 03 1.96E=05% f.O4F T4 6.U5F "4 l.8RE OO 8-39E_61 0. 1,428 01 5.5CF 02 S.80F 02 2.015=-C2 2072;“02 DN 5.53£-~01 6.94F 27 2.87E r2 2L,A0F=-02 N.0 Q,3T7E 74 7.92E 04 0.0 T6TE-D5 1.172 8 35525,D0 2.53F Q0 0.0 3.93E-"1 1.82 O 1.96E-C5 h.56F D72 ASTE 2 1.88€ 0C 1.87F D 2.0 1.47F 01 D.0 O, HE-Co 226-91 CE 00 EE-(P ACE-02 oW D T O AR D O NTE D ) ,07F 2 QO WM s " 8 % " e s s & 0> NOOT OoONWOO DO * . . - . & » * N TR NNO O T 365252.D 0.0 0.0 2,89F~-01 0.0 1.96F=05 IQSQE‘Gé 1,5%0F-06 1.88F 00 1.88F 00 . 2 242F 01 . & - . b * v OF-~01 6F=-15 PO O T Pp2ODDRO LN N D N DO 0T - ® & & & @ 8 & & 4 ¢ b & 9 8 @ WODODOQDDODDONDBDODDDH TDD N o NUCLIDF THERMAL POWER, H KR RB SR SR Y Y IR NB NB TC 3 = 8¢ 89 or ag o1 o5 osm o 09 PPICY RHID2IM RUCH RH1 06 AG1liOM AGLID AGI ] CD11eM SN11am SN1272M SBl12a SNY 28 SBY2S TEY268M SR12¢& TE127M™ TE127 TFY2eM TEY29 Iy XE121 M £S134 CS13é6 €s137 BA12TM BA140 LAI4C CFl141 PRI< CEl4Ld PR14L ND1&47 PM147 PM148M PM148 SM151 EV152 GD153 FUl sS4 FUY158 EUYSE TBl160 GD162 TB162M SUBTOY TOTALS Table 3.11. A0S0 MW/MT , BURNUP= CHARGE g.0 a o . * L ] [ . . & MOOMTIONDNDOODDVOONODDIODDDOTNOD D MM IO I NICDIIOCOIIDIOONIOOD o - - - » [ ] . L ] . - . * a & . . . - . * * * *® . OODOODAODNOODTOONCODOOII00 INDO - OO IJIDDOTHIOONDCOIDNOTDTIOODODD s o 8 & ¢ 0 & & 3 . > o 3-21 Calculated Thermal Power of Significant Fission Product Nuclides Present in Spent Diablo Canyon Reference LWR Fuel and in the Wastes Generated by the Reprocessing of This Fuel NIABLO CANYCN REFERENCE LWR - WASTE DECAY TIMES (PRDCESSED 90 DAYS) POWERS= 90.0 2448F=-022 1.828 21 B.,CBFE=~02 TA9E D2 1.00E N2 4,60F 02 1.238 03 2,748 03 1.,5E 01 4,178 D3 9.,62E-0C3 8.,21% 02 A.,04F 01 2.72E 01 4,4L4F 03 4,97 00 ?oqlE-Ql 2.22E-02 4.7SE-01 £,81F-073 1.74F Q0 2.33F 00 10”!E“C1 2.R4F 01 2+.85E €O 0,25E~-N3 4,Q99E 09 1.6 01 4,508 01 £,29€ 1 1,57 00 1.03E-01 2.36E N3 T.89F 00 1,73E 02 3,928 02 3.74E O1 2.12F 02 4,N2€ 32 3.13€ 01 T.AR2E 02 6.63F N3 6.05E O 535 01 1.24E 01 T.21E-01 2.01E 00 1.415-01 3og3E-02 6.,44E 01 5.71E 00 2,70 01 4,508 00 6.24F=-01 1.258 CO 2.562F 04 2.62E D4 WATTS / 150.D 2.465-02 1.8%E 01 8-69?'03 3., 45F 02 a,a8e 1 4,388 02 6.04F Q2 1,458 C2 8,17 C¢C 2.48F 032 S.,62E=-02 2+.91F Q2 2.12E 01 2.4%F N1 3.96E 03 4,22 00 24TE-C] Q-GTE‘GS 1.80€E-01 E,77E=-3 1.25F QC 1.17¢ 00 1.216=-03 2+.73F O1 2.82F 00 6.11E-C2 3,40F OO 9.95¢ CO 1.33F 1} 1.56F 21 8-94?'03 3.18E-02 2.24E C3 2.22€-01 1.72E 02 3.91€ A2 1.45€ 00O 8.,21€ 0C 1.128 22 1.51F CC 6.76E 02 S.73F Q2 1.43%-01 5.13F 01 4,98 CC ZQSBE-Ol 2.01E 00 ICAOE-GI 3.22E—02 6.39E 01 5.36E DO 2.31€ 00 2.53E €0 5.66E-01 1.11€ 00 1.93€ 04 1.92€ 04 33000, 3265,0 2.3RF=-02 1.72 21 3.,05E~06 1,87 M 9,8aF M 4,32% 02 4,7°F 01 1.46F N2 8, 25E~01 2. 88 02 9062E-03 6. 75E 00 4 ,90F-01 l1.62E 1} 2¢ 66F 03 2+36F 00 1037E‘fi! ZOIBE-IB Be B4F=032 3.185“03 307°E‘01 gc?BE-OZ 1:585-18 2.34F 01 2.4RFE CC 5.99E~03 8. 68F-01 2+54E 00 1065E‘Gl !oqag-fil R, 15E-11 1.08E-0R 1.982E D3 3,23E~06 1.70FE Q2 3.85F C2 1|27E-05 1,62F 0C 1. 128 0D 2eB4E-(5 4, 00F 02 2,398 §3 2011E-07 4,39t 01 1.43E-01 T.425-02 2.,00E 00 1.35£-01 1.74E-02 6,.23E 01 4,288 OO0 1.12E-04 32.20E-01 3.76E-fl1 7.41E“01 1.0CE 04 1.00E D46 MWD/MT, 3652,D 1,423E=-02 3,71F OO 0.0 1.,R4F=~18 7.88F 01 2,46F 02 7T.15€E~16 8,79F=14 4,96F~-16 F3.05E=-08 9,62€F-113 0.0 0.0 3026E*62 5.21F 00 2.87E-04 1.68E-G5 0.0 D.G 3,49F=-07 4,A0E-09 1,15E=-18 0.0 2433 00 2.47TE=01 5.99E-N2 7.25%=-10 2.12E’09 3 I OO PO OO0D WO DI I o m Q o ? ® ® 8 ¢ & % 8w & & ¢ & & & 8 ¢ @ ROBE-D2 1.42E-06 4,228 C1 1.36E‘91 0.0 6.39E-15 7036E“04 1045E’QB 1.03E 03 1.03E N3 ARE25,.D BR,90E-I% 2.95E-O? 0.0 DN 8.56F 0O +T6E . wd oD 616~G3 4F-10C 8E-11 9E~-C3 DN NMNDOOTOODTQIOIO OO0 W ON = IOODOODODIOQ ® &€ & & & # ¢ & & & & & & & 3 & & b s 3E-12 - O PDOO0 DOO ) E 01 E 21 0 . & & ¢ 8 9 & & = e O 9 @ & & % 9 @ ¢ & ® DO DOPANIIPOOITBO ITOIDOD -~ m i Q e OSSO0 NP LVODI~O0N0O0QOWHONIODOQ . *® & ¢ ¢ 9 4 02 FLUX= 2,91F 12 N/CMx%2-SEC METRIC TON FUEL CHARGED TO REACTOR 348252 ,D 0L 0.0 0.0 el 1.95E-09 R.ETE~09 0.0 h_g - no oy S8F-03 - ¢ 4 & @ .« @ . HDONDIOODDOO HIQOTIOD 5E-03 1E-J8 LE~-C8 OWmO O DOODIOC N0 DO IO O DD 8 & 4 ® & 8 & % & % 4 & & @ * e & 0 & & & w & PODDOO OO VODIODIDNODADIOD TE-04 DOOROLWOOO MDD 30E~18 ® & & & & 9 @ e DO D 2€-02 1.77E-02 Table 3,12, 3-22 Calculated Thermal Power (Total) of Significant Fission Product Elements Present in Spent Diablo Canyon Reference LWR Fuel and in the Wastes Generated by the Reprocessing of This Fuel NTABRLD CANYON REFERFNCFE LWR -~ WASTE DECAY TIMES {PROCESSED 9f POWER= ELFMENT THFRMAL POWFER, H GA SE KR RA SR Y IR N8 MO TC RU RH AG €O IN SN SR TE ! X E s BA LA ce PR NP PM SM FU Gn T8 DY HO 23,00 MW/MT, CHARGE 228D DD . DDV OMDIDD IDOMTTVONNIDO ITIIAOIIDTIDDNOD IDAQOD - S 0300 2 DIDOIDONDD IO 2EmO R - QDO DD DIDO D TOTALS r,0C an,D 24BF-32 BFTE-15 1,49F-04 1.R2% 1 ?,0BF=-0? R.AQF 02 1.67E 73 2.74% 03 4,19 03 1.36E-r6 10005-02 P.FQc 02 4,5NE 012 5.28% 00 4,75€E-01 2.19E-C7 1.F5¢ CO 2.08E 01 1.208 02 1.7 CO 1.16F=-01 24555 N3 4.2NE 02 2.12F C2 1.8 ¢13 b.66F 3 6.0%F 00 6.77F €1 ?2.71E N 1.07% 02 6.,73E-01 S5.75¢ (N 2 QOAOE-‘A 8.31E~-C7 2.62E C4 BURNUP = 150.D 2,46FE-02 0.0 1,49E~C4 1.,81nF 0 R,69F=012 4.45F 02 ].,24E N2 1.45F N2 249F (03 4,59F~17 .62F=-02 2.15F 02 3,98 r2 4 46F OO 1.91€E-01 Q,5NE~-CHR 1.26€ 0OC 2.84E 01 45N MY R.ORE=-(07 2,19F=0N72 2.410 02 2,92 02 R.21E 0O TLRTE D F+T3E C3 1s42F=01 5.65E 01 2.,N1F °0 7,17% 01 5.98E'O] E4E I'T 4,57E-19 R.30E-0Q7 1.93E na& 3000, 26%.0 2.38E-"2 c.0 10‘?5-04 .78 1} 3.07F=08 1.,1RE D2 4. ACE 02 1.46F 02 2. 95F 0? f.0 D, 42F=07 2.79F 01 2.64F 3 2.48F OC . T0F=-02 6,82%-9 3.R2F-01 2 .35 01 6.25E €O 1.6AE-05 1.056~08 2.00E M2 3. R5F Q2 1. 62F 00 4,015 02 3.39F 02 2.11E=-C7 4 4E MY ?.NCE ON 6.£675 01 2,04F-mM 1.0€F O C.0 8.32CE-07 1.6CF na 26€2,D 1.43F=-02 0.0 1.,49F=N4 Q.TIE P 1.2%E=01R 7.28F 01 3.46F 02 2.23E=~04 1.4CQE=04 . 9,62E-03 3.26F=02 5.21€ on 3.03F¢04 2 ,59E-05 d.0 ?.95E~C4 2,238 N0 2 4HTE-DT 1.645-05 0.0 2.26F 02 3.13E 02 7.0 1.21F=01 1.11% 00 0.0 L, 06F T 1.86E 00 4.724F 01 36825,.N SOQQE-05 0.0 1-49F-56 2.35F=07? 1.28¢c=Q¢" B8.,56F N1 3.76F (1 2-23F-04 3,33F=-C4 a.0 9-6]F-03 DAYS) MWD/MT, FLUX= 2.91F 13 N/CM%%2=-SEC WATTS / METRIC TON FUEL CHARGED TO REACTOR 265297 .D CoC D.C 1.48F=06 C.C 10285-08 1.9%€E~09 B.STE-Q9 2.22E-04 2,38FE=04 2 o M i < W 2O 0 TIOW;m . L I . &+ ® *» » » 4% =08 F-Q4 F-18 POITOOLPNOODOWKD=TIIANGTTIODIDO DD o~ ODODPOD OO PO TDIND - - @ ¢ o ® @ 6F-07 ol -~ - im { 4,02€-02 I 1.00€8 C1 1,149"03 5E.35E=012 5.61€-01 0.0 2.,21E 02 1.08€-01 8.50F O 4.41% 02 4.,41F 02 FLUX= 2,91F 12 N/CM*%¥2-SEC / METRIC TON FUEL CHARGED TO REACTOR 13956.0 2+46E-04 5e21F=-1D 2.03F~-08 ? «35E-0T7 1.50F=-12 0.0 2,26F-09 tL0 2,58€-17 C.0 a0 4, T75E=04 A.765-07 1.025=-23 4,02F-02 N 2.07%8 01 1.13F-03 5.33E-03 f.61F-01 0.0 9.98F8 QO 1.39E-01 7.88F 01 2433E 32 2.23E 032 Calculated Thermal Power of Important Actinide Elements Present in Spent Diablo Canyon Reference IWR Fuel NTABLO CANYON REFERENCE LWR -~ FUFL DECAY TIMES 7.30E-05 4.995-C5 4.6TE (4 5.83E 04 6.,20E 02 4,82 02 1.12€ 03 ns 330 90.D 3.455-05 G, D4E=-E 3004E‘01 F.16E-02 1.228 22 6,30 0O T+99E Q2 9,28 02 0C. 120.0 3., 85E-05% 1. 89E-06 2-885‘02 5.16E-02 1.22E 02 &, 7TRE 00 T.14E 02 8.42€ Q2 MWD /MT, 15C.D 4,285-05 8.888-07 1062E'02 5+16E=-02 1.23€ 02 T.25E 00 6.38F Q2 T.69F 0?2 365.D 7.98€-05 6 .89E~09 1,58F=-02 5.15E~02 1.24E C2 1.06E 01 3,06€ Q2 4,41F 02 FLUX= 2.,91E 13 N/CM*¥*2=SE(C WATTS /7 METRIC TON FUFL CHARGED TO RFACTOR DISCHARGE 1096,D 2.46E-04 3,26E-09 1.65E-02 5.16E~-02 1.23E 02 2.12F 01 R,89t 01 2.33E 92 Table 3.19. 3-27 Masses of Actinide Isotopes Calculated to Be Present in Wastes Generated by the Processing of Spent Diablo Canyon Reference LWR Fuel DIABLO CANYON REFERENCE LWR - WASTE DECAY TIMFS (PROCESSED SC DAYS) POWER= NUCLIDFE CONCENTRATIONS, TH222 TH229 TH?23C TH221 TH222 TH233 PA23] PA22? PA227 PA234M PA234 U232 yz222 U234 u23¢ U226 uz237 U238 U239 NP236 NP237 NP238 NpP22¢e PU235 PU238 PU230 PU240 Puz241 PU242 PU243 AM?24] AM242M AM242 AM2472 AM244 CM242 CM2472 CM244 SuUsTOoT TOTALS POWER= 20,00 MW/MT, BURNUP= 33000, MWD/MT, FLUX= 2,91F 13 N/CM**2-SEC ELEMENT CONCENTRATIONS, GRAMS / METRIC TON FUEL CHARGED TO REACTOR CHARGE 90.0 150,0 365,00 3652.D0 36525.D 365250.D IH 000 2056E-04 2.56E‘04 2056E-04 20675'04 8-496‘04 1.675“02 PA N0 2.11E-06 2,.05E=06 2,N3E-06 2,04FE-06 2,07E-C6 2.3TE-06 U 1.00F 06 4,78E 03 4,78 03 4,78E 03 4,78€ 02 4,78F 032 4,79F 03 NP 0.0 T.62E 02 T.62FE 02 7.62F 02 T.62F 02 T.70E 02 B.06E 02 PU 0.0 4.56E N1 4,69E 0} 5,0CE 01 5.81F 01 7.28¢ 01 T7.30F 01 AM 0.0 144 02 1.,44E 02 1.44F 02 1.45E 02 1,40F 02 9,54F 01 CM 0.0 3.68E 01 2,53E 0! 3.19E 01 2,13F 0! 6,88F-01 1.04E-05 YOTALS 1.,C0E 06 ©.76E 03 5,76E 03 5,76 03 5,76F 03 5,76 03 5,.76E 03 20,00 MW/MY, BURNUP= CHARGE 3D . . - ¢ & ¢ & o 8 s 8 8 ¢ & & IDODODIDODIPIOINOOIOAODIODIDOODTOD * o . . L ] * & @ s a [ A ] DD DDAL IODLDONAD DO IDODIOD QD . . QL . MDD O MO DO DTIOODODODO CCE 06 1.CCE N6 Table 3.20. 0.0 1.27E-Cé6 6.06F~-08 6.80E~-C6 3.22-10 2.48F-04 r.r 2.03F~06 0.0 7.78F-08 0.0 0,0 1-!7E-66 R, A4FE-06 1.10E-02 3,98 01 2.04F 01 8.22E=-06 4,72F 03 "0 0.0 Te62E 2 3.,52F-13 T.48E=-05 4,91E-C6 B.26E-C1 2.59E C1 1.08F 01 S.O7E 00 1.75E 00 0.0 5.30E C1 4,14E-01 LOQ?E-O6 9.04E C1 0.0 5.83E 0O 8.765-02 3,09 01 5.76F 03 S.76E C3 150.D 1.20E-06 6-06E‘08 6.BLE=-NE 1.,61FE-12 2.48E-04 C.0 2.03F=-06 c.C 1.71E=-08 0.0 C.0 1.25E-G6 8-705—06 1.29€E-02 3.98E 01 2.04F 01 1.73E-08R 4., T2E O3 0.0 DQO T.62E D2 Bt81E-22 T+48FE-05 4,72E-06 2.12E 2 +69F 1.10F 5.03E 00 1.75E 0O 0.0 5.30F 01 4014E-01 4 ,97E=-06 Q.04FE D1 0.0 4.52E 00 B, 72E-02 2,7 01 S« THE 03 S.76E 02 Masses of Actinide 32000. GRAMS / METRIC TON 365,.0 9, 80E-07 6-Q6E-68 6 LRLE-DE 1.61E-12 2 4RE-06 .0 2.03E-06 0.0 T,41E-11 c.0 0.0 10°6E-06 B. 7T2E-06 2.92E-02 3.98E (1 2.04F 01 Ao‘?E-IB 4, T2E 03 C.0 0.0 T.62E 02 0.0C 7.48E‘05 4, 0QE~-D6 4, T6E 0D 2.+ €F 01 1.17e M 4,87 QC 1.75F 00 0.0 5.31t C1 4.,13E-01 4.95E“06 Q,04F M 0.0 1.81F 00 8.61E-02 3, 00E 01 5. T&E 03 5. 76E 03 MWD/MT, FLUX= 2,91E 12 N/CM*%2~SEC FUEL CHARGED TO REACTOR 3652.D 1.58E-e7 6.,00F=NR 1.28E=-25 1.61E~12 2.54E-N4 S5.43E 01 3.96E-01 4,T5E-06 9.,23E €1 0.0 9.62E-064 T.NQE-G2 2.128 01 5.76€ 03 5.T6E 03 36525.0 6.,35E=-08 6,378 Se43E-T4 1e61E-12 2L,06E-004 ~ n i} eis 2007&'06 0.0 N0 0.0 0.0 2278006 8.7T1E=-0¢ 2.,48F GC 3,98 M 2.07 0} 0.0 4.72E 03 JFE 0?2 O -~ OO O ~NOO Te41E=-05 1.43E-16 3,07 Q€ 2.7T7E 01 «L22F 21 2.48F-02 1.77F 0OC 0.2 5.,C00F C1 2.F3F=-01 3.185E-06 8.96E Ol 0.0 6.32E=-04 IQGIE‘GZ 6.77E-01 S.T6E 02 5.76F 03 365250.D 10105'11 9,00FE-C8 1'.59F=-02 1.64F-12 B.B5FE=D4 Z.C 2.37E-06 1.82E 00 G0 1.28E 01 4,23E-03 5-205-08 8.26F 01 0.0 1.04E-05 2.45E=-11 T.32E-16 5.7T6E 03 5.76E 03 Elements Calculated to Be Present in Wastes Generated by the Processing of Spent Diablo Canyon Reference LWR Fuel DIABLO CANYON REFERENCE LWR - WASTE DECAY TIMES (PROCESSED 90 DAYS) POWER= NUCLIDE RACIOACTIVITY, TH2?28 TH2?29 TH23D TH23? TH232 TH?2% PA2Z2] PA22? PA223 PA?24M PA?34 yz22 U221 U234 U23% U23£ uz3? uz23e yz2o NP22¢ NP2RT7 NP22R NP22Q PU23& PU23r PU?39 py24nr pPUzs? PyYz242 P24 AM2 4 AMP47M AM247? AM247 AMP 44 £M247 CM243 CM244 SURTOT TOTALS POWFR= FLEMENT RADIDACTIVITY, TH PA U NP Py AM CM TNATALS Table 3.21. 3-28 Wastes Generated by the Processing of Spent Diablo Canyon Reference LWR Fuel DIABLO CANYON REFERENCF LWR - WASTE DECAY TIMES (PROCESSED 9U DAYS) 2CLO0 MW/MT, BURNUP= CHARGFE . * ~ D D * 2 @ *® & 0 . & = -0 TE-04 O OO O 27E=01 - [ - . . L] . . . *» & & . - MO MOOCOOITIMNTIOANODNOD DO MION W Y AO TGO I0O0 0000 W e * a4 ® 8 e & o ¢ e D 3T oo » o 3 ¢ - ¢ 3,22F=-1 2,23E-01 Table 3.22. 3C.D 1.,08E~C3 1.30E-08 1.32E-07 1-7QE-04 2.715-11 c.C 6QE-(8 6.,7T1E-641 !nS?E‘O? 0.0 gt f.37E=-C1 Q. 19E=-0R 1.74F M1 2061E_03 1.29F 01 1.65F o0 2.39E (O 5.79F (2 6.21E-013 ".0 1.72F 4,02¢ 4.02E 1.74F N, 1.,92E 4.03F 2.50% 2.2 6F c?2 a0 no 01 nNa& 00 23 Db 2.26F 04 150.0 QORSE-OA 1.20E~08 1.325-67 B,S2E-N7 2.T1E=-11 0.0 Q.h9E~CR C.O B, L49E-04 Cad i * a - - 2.9CE-05 B.24E-0%8 T.99E-C5 R.E2E=-C7 1.295'0? 1,42E-072 1.S7E”03 G.0 Celd 5.37€-01 2.30F-1¢6 1.748 01 2+51E=-072 2,578 01 1,48 0 ?.43F 00 F.TUE 02 6.831F-002 0.0 1.72E 0?2 4,328 o0 4,228 0O 1,748 O C.0 1.5CE N4 4,31F 00 2.4Q9F (2 1.828 24 1.83E Q& 33Gfi50 CURIES / METRIC TON 265.D B, 05F=-C4 1030E“08 1.32e-07 R, 52E-07 2.72F-11 0.0C Q, AQE-NE N, n liq?E-Ofi Col a0 4, 20E-05 R, 26E=-NR ¢ !.F!E-OA 8.52E'07 1.20£-02 2, 65E-13 1.57E-03 £.81E-C72 C. 0 1.72E 02 4,1 ac 4.01FE OC 1.748 01 flog b NCE G2 2, 96E OO 2.42F G2 Q,20t 72 9,?7%E 02 MWD /MY, FLUX= 2.91E 13 N/CM*x2-SEC Calculated Radicactivity of Actinide Isotopes Present in FUEL CHARGED TO REACTOR 3652.0 1.29E=-04 1.30E-0R 2 44QE-07 8.53E~07 2.77F=11 0.0 S9.71E-C38 1029?“03 a0 1.57€=-02 0.0 el 5.2R8F=01 C.C 1.74E 21 2.&45—0“ 1.02FE 02 le66F 2N 4,48F 00 3.44F Q2 6.82E‘g% Ve 1.76E 02 3.,RA5F NN 3.85F 0O 1.74F OV Nelt 3,18 00O 3,26F 00 1.72¢ D73 Z2.lE 3 2+6GF 03 35652%,.,D 5.21E-25 1-36E-08 1.05€-0F goS#E“D’ 3.35E-11 N m | S o J¥ m t v) : N ¢ & @ . - [} ¢« & @& ~4 iT | o ODOP HTO RN IR AMIODIDDODO w ' - m ! o) vl . * & s @ @ MO N0 4D e BNVBDANIIOD OO ~ AV} ! e | e 7.60E=14 5.19% 01 1,77°% OC 8.87E CC 2 .84F CC 6.92E“:? 0.0 1.628 02 2.55E mn 2.55% 0OC 1.72 01 N,7 2.09F ©C 4,564F=D1 5.49E 01 YI.25E 22 3.25F G2 Wastes Generated by the Processing of Spent Diablo Canyon Reference ILWR Fuel DIABLO CANYON REFERFNCFE LWR =~ WASTF DECAY TIMES (PROCESSED SC DAYS) T G0 MW/MT , BURNUP = CHARGE Y. Y is) - 2,22F-01 N.0 C.0 2.0 2,23F="1 onN.D IQZBE-t? 1059E'?3 6.’4F-O! 1.79 1 E.Q7F 72 1.97€ ¢? 2.78F 04 226 (4 150.0 G,30%E=-N4 31495'@4 &-3QF‘Q? 1.79 01 6,14 02 1.97¢ C2 1.74E D4 1.83F ra 3zCEOo CURIES / METRIC TON 265,D R.NEE=D4 1, 61F=06 2, 00E-02 1, 75F 0 £.4YF 02 1.98F 02 8,43E N2 C,20E 73 MWD /MT, 365253.D 9,0 7F-D09 1,93E~-08 3.08E-04 B.T1E-0T7 Q.£8E-11 0.0 1.13E'07 R.45E 01 R.45E (1 Calculated Radioactivity of Actinide Elements Present in FLUX= 2.91F 13 N/CM*%2~SE(C FUEL CHARGFD TO REACTOR 15652.D0 1.31F=-004 QQ7IE-OQ 5.78E-03 1.78 M 4.,52F 02 2.,01F D2 1,728 03 2.40F 03 16528,.N 6.35E=-0F 9.85E=08 2.45F=02 1,787 21 6.52E M V.8&4E Q2 S5.T4E M 3.28F 07 265250.D 3.09€E-04 1.13E-07 4,45F=02 1.65F (1 1.04€ 01 5.75F Q1 2.,46E8-0572 R,e5F (1 Table 3.23. 3-29 Calculated Thermal Power of Actinide Isotopes Present in Wastes Generated by the Processing of Spent Diablo Canyon Reference LWR Fuel DIABLO CANYON REFERENCE LWR = WASTE DECAY TIMES (PROCESSED 99 POWER= NUCLIDE THERMAL POWER, TH22e TH229 TH230 TH231 TH232 TH232 PA? 3] PA232 PA232 PA234M PA214 U232 U233 U234 uz3s U236 ur37 Y238 y23e NP?34 NP227 NP?38 NP?239 PU276 PU228 PU230 PY240 PU241 PyU24? PU2432 AM241 AM242M AM242 AM243 AMP 44 CM242 CM242 CMo4L SUBTOT TOTALS 20 .00 MW/MT, BURNUP= CHARGE 6E-05 5e-03 ¢« & = * & ¢ % @ . e & & 9 * @ s @ . . @ IONOODOOTE0NO0OTO00NNS=TIN 00320030000 IOOD & & * o o & @ - . PO LSOO IODDODD TFOOTOO0D Do) I=ADOGODDIODOOTDOD = 9.i7F—D? 90.D 3.43F=-005 3.92E-10 3-735-09 2.35E=-C7 6.56F-13 0.0 2.96E=-009 1.976-06 2.376-C8 3-SIE‘G5 ! 1044E-03 3.°35-05 a0 N 0.0 4,T2E-10 5¢16E=02 9,.09E-05 4.62F-01 B.l13E~C2 T.456-02 2.40E-C2 2.01E-04 0.0 5.73E (0D 1.156=-C3 5.,27E-C3 5.61-01 £.0 7,11 ¢2 1.47-01 3.75€ (1 R.06E 02 150.D 3.24E~C5 3.92E-10 3.T74E-09 l.18E=-00 5«36E-C2 £ 5.61E-01 C.fl 5.51F 02 1.47E-01 B.70E N1 £ abE 02 2 2.17F-C3 R.06FE 02 €E.46FE 02 Table 3.2L. Calculated Thermal Reference LWR Fuel 332000. 36%.D 2e 64E-CH 3.92E-10 2. 75E-09 1.18E~CQ 6.5TE-12 c.0C 3 2. G6E-0Q C.C 3.845-09 r.0 0.0 !.35E-06 0 2 .4C0E~-CC . 22E=C6 2 .37E=-08 £ 2.51E-CF MWD/MT, 3652.D 4.24E-06 3,.,94E5-10 7.03E~09 1.18E=N9 6.T1E=13 OOG 2.96E-0Qg 0.¢ 0.0 0.0 0.0 A.55E=06 2.4$E-Qq 3.07E“05 237E~0OR 3.,51E-08 1.39F=01 1.43E-02 2.01E=-N4 O‘O 5.87E 00 2 1,10€E~-23 2,188 02 5.135”03 5.60E-01 Co 1,178-01 A.02E N1 7.06F 01 T.D8E 01 36525.0D 1. 71E=-Ds 4,12E-10 2.98€-07 IOIBE‘DQ R.10E~12 0.0 3.01E-G2 B e28E~02 2.76E-01 1.18€=9%4 2.04E-C6 J.0 Be41F 0O T.27E=N4 31,41F=02 5.56E-01 N0 T72E=02 Ioqu-O? 1,928 27 1.01FE 01 1.01e 0} DAYSY) FLUX= 2.91E 13 N/CM*%2-SfC WATTS / METRIC TON FUEL CHARGED TO REACTOR 3165250.0 2.976—10 5.825-10 8071E-06 1.23E-09 2-34E-12 0.0 5e-09 2.N7F=-15 2.27E QC Power of Actinide Elements Present in Wastes Generated by the Processing of Spent Diablo Canyon NIABLD CANYON REFERENCE LWR - WASTE DECAY TIMFS (PROCESSED or DAYS) AC,00 MW/MT, BURNUP= POWER= FLEMENT THFRMAL CHARGE TH ~on Pa 0.0 u 2.17E-03 NP fan PU 0.0 AM 0.0 CM N TOTALS 8,17E=-03 POWER, an,.Dn 3,L65F=05 4.,04E=-06 1.528-03 S1hE=-N2 6.!25-01 €.20% €0 T.90F 2 R,O06F 02 157.0 2,24E=05 B, RRE=-(7 9.11?-05 Ce16E-2 1.23F 00 £,30F (0 £ 3ARFE 07 EL46E 02 22000, 265.D 24 BLF=[5 8, 14F-08 S-lGE‘F? 2.RB2F 0O £.31F 00 A, 06F £? 3.15E @2 2652.0 4,25E-028 ?.9£E=09 I.KQEuOQ S.18E=-072 3.%8E 00 b.43F 0O ANBF 01 7.06€ 01 3652%,D 2.,01E=06 BQOIE*DQ 5.97E=C4 5.11E=r2 2.0RE QOQ 5.97F QO 2.71F 0O 1.01F 01 MWD/MT, FLUX= 2.,91F 13 N/CM**2~SEC WATTS / METRIC TCN FUEL CHARGED TO RFACTODR 365252 .D B.TIE=-D6 3,45E-09 IOZTF-O3 4,71F=0G2 3025E-01 1.9CE QO 1,276-013 2.27€ 00 3-30 Table 3.25. Masses of Activation-Product Nuclides Calculated to Be Present in the Zircaloy-l Cladding and Inconel Spacers of Irradiated Diablo Canyon Fuel Assemblies DIABLO CANYON REFERENCE LWR -- CLADDING ACTIVATION POWFR= 20,00 MW/MT, BURNUP= 3300C. MWD/MT, FLUX= 5,82E 12 N/CM*%x2-SEC NUCLIDE CONCENTRATIONS, GRAMS / METRIC TON FUEL CHARGED TO REACTOR CHARGE DISCHARGE 20.D 150.0 365.0 1096.0 10958.D C 12 1.8CE 01 1.,80F C1 1.B0F €1 1.8MF Q1 1,80E 21 1.80E 1 1,8DF 01 AL 27 1.08F 02 1.08F 02 1.08c 02 1.CfE 02 1.08F 02 1.08F 02 1.08E 02 ST 28 1,708 01 1,.71F O1 1.71€ O1 1.71F 01 1.71F 01 1,71F Q01 1,71F Q1 TI 46 1,40F 1 1,40FE N1 1,40F 01 1.40F C1 1,4CE 01 1.4CE 01 1,40F 01 TI 47 1.30F 01 1.30F 0! 1,.30F C1 1.30F 01 1.30F 01 1.30F 01 1.20F 01 TI 48 1,23FE 02 1.31FE 02 1.31F 02 1.21F 02 1.21F 02 1.31F 02 1.31F 02 TE 49 1.00F 01 1,186 O1 1.18€ 01 1.18E 21 1.18F 01 1.18E 01 1.1RF D1 TY s 1.0CE 0O 1.00F 01 1.00E O1 1.,0C0F O1 1.00FE C1 1.Q00E 0! 1.00E 01 v 51 0.0 7.19€ 00 7.47E CC 7.5CF 00 7.51E 00 T7.51F 00 7.51% 00 CR 50 2,52F 02 2.45F 02 2.45E 02 2.48E 02 2.,45F 02 2.,45F 02 2.45% Q2 CR 52 2,99 03 2,98F 03 2,9RE 02 2,9RFE 03 2,.9RF 03 2,9RF 02 2,98F 03 CR £33 3,40F C2 2.,34F 02 3,34FE C2 3,34F 02 3,24F 02 3,34E 07 3.34%F 02 CR 54 B,5CF 01 9,58 N1 9,58F C1 9.58FE 0} 9,58F N1 9,58F D1 9,58F 01 MN 55 1,8CE 201 1,77 01 1.78F &1 1.72E 01 1,79E 01 1,.82F 01 1,R5E Q1 FE 54 2,18F 02 2.17€ C2 2.17F 02 2.17F 02 2,17F 02 2.17E 0? 2.17F 02 FE S6 2,42F 03 3,408 07 2.4NF 03 3.40FE 02 3,40F 03 3.40E N2 3,40F (3 FE 87 R,20F N)] Q,68F 01 9,6RE 01 O, 68F N] 2,68F N1 G, 468F 01 9,6RF 01 FE 58 1,80F 01 2.09f€ 01 2.11F 01 2.,11F Q1 2.12E 01 2.12F 01 2.,12F Q1 CO 59 S.,40F 0 4,83F 0] 4,83 0 4,82E Q) 4,.,83E 01 &4.83FE 01 4.83F {1 CO &0t 0.0 S+H6TE (0 B5,48F 00 E,37F L0 4,97TE £ 2,82E 20 1.09F-01 NI SR £.40F 03 6.43F 03 A,43F 02 6.43F 02 $,43F 03 6,.,42F 02 6,43F 03 N1 %9 C.0 5.06FE C1 5.06F 01 5.06E 01 5.,06€ 01 S5.06F D1 5.06F 01 NI 6% 2,81E 03 2,50F 03 2,50F 03 2,8CF 03 2,50F 03 2,50E N2 2,50F 03 NT &1 1,14F Q2 1.25E 02 1.25F 02 1.2%F 02 1.25E 02 1.25% 02 1.25F 02 N1 52 3,50F 02 3,41E 02 2.,41F 02 3,4YF 02 3,41F 02 3,41F €2 3.41F {2 NI 62 C,0 9.12E 20 9.10F 0C 9,09F Q0 Q,N8F Ni §9,92F pN 7,27 €D NT 64 1.03F 02 1.03F 02 1.03F 02 1.C3FE 02 1,03F 02 1,03F 02 1,03 Q2 ZR 20 1.28F 0% 1.28E 05 1.28F 05 1,.,28F 05 1.28%F 05 1.28E 0% 1.28F (05 IR 91 2,79E 04 2,1N0E C&4 2,10F N4 2.,10F €4 2,10F 54 2,10F 04 2,1NE 04 IR Q2 4,25F 064 4,95F 04 4,95F C&4 &4,95F 04 4,95F 04 4 ,.95F 0¢ 4,95E Q4 IR 92 (C.0 3.978 Q1 3.97F 01 2,97 01 3,.,97% Q1 3.97Ff 01 2.,97F Q1 IR G4 4,32FE 04 4.32F D4 4,32F T4 4,32F 04 4,22F 04 4,328 04 4,22E G4 IR 96 H.96F 03 £.9hF 03 €.96F 03 6£.096F 02 6,96F 03 K,96F 02 6,25F 03 N8 @3 9,54F 02 9,52 02 0,52F 02 9.52F 02 9.52E 02 9.52F 02 9.52F 02 NB 94 1,0 1.83F 09 1,.,83F CC 1,83F N0 1,82F N0 1,83 00 1,83F N5 MO 9?2 8,8CF 1 8.R0F 01 P.80F 0O R, E0F 0! R8.80F (¢l R.80F 01 B.80F 01 MO 94 S,00E 01 S,00F 01 £,00F C1 5.00E 01 S.00F 0! 5.0CF QY S.COF 01 MO 9% 8.,8CF 01 R,RGE 01 9.1A0F 71 Q.,N4F OY Q.M9E N 9,09FE ) Q,N9F 1) MO 96 ©,20°F N1 1,72F 722 1,228 02 1.C28 02 1.02E 02 1.02% 02 1.C2F 02 MO Q7 6.30F N1 5.29F 01 S5.29F C1 5,209F (€1 5.2°F 01 5.29F ¢! S5,29F Q1 MO 98 1.33F 02 1.32F ©£2 1.32F 02 1.32E 02 1,32F 02 1,32F 02 1,32F {2 MOLIOE B ,40F 7] 5,278 (1 S5,37F 01 S5.37F 01 5,27E 01 5.37€ 01 5.37F 01 SN1l4 2,.5CE 01 2.50fF 01 2,50F QY 2,5CF Q! 2,5CFE 01 2,5CF Q1 2.%8C0F 01 SNI1S 1,.,3C0F 0 1,30F 01 1.30F €1 1.3CE 01 1,323€ 01 1,3RF 01 1,.20F 31 SNI1A S5.43F 02 5,36FE 02 5.36F (02 S.36F N2 5.26F 02 5,36F 02 5,.36F (2 SN117 2.89F 0?2 1.83F 02 1.83F 02 1.83F 02 1.R3E 02 1,82F% 02 1,83F 02 SN118 ©,12F 02 1.03F C3 1.03F €2 1.03F 03 1.03E 03 1,0%F 02 1,73E 03 SN119 2,26E 02 2.09E 92 2.09E 02 2.09F 02 2,.C9FE 02 2.09F Q2 2,.09% Q2 SNI20 1.25E 03 1.36F 03 1.36F 02 1.36F 03 1,.36F 03 1,36F 02 1,.26F 03 SN122 1.,76F 02 1,795 €2 1,.79€ ©2 1.79€ 02 1,79F 02 1,79E N2 1,79¢ g2 SNT24 2.26F 02 2.26F 02 2.26FE 02 2.265 02 2,26F 02 2,265 D2 2.26F 02 SUBTOT 2,71F 06 2.71E C5 2.71F 05 2.71F C5 2.,71F Q5 2,71t 0% 2,71€E 05 TOTALS 2.71E 08 2,718 £5 2,71F N6 2, TIE ©5 2,71€ N5 2,71F N% 2,.71F G5 RIABLG CANYON REFERENCE LWR POWER= FLEMFNT CONCENTRATIONS, H HE L1 BE noZmw NE NA MG AL SI p S CL AR K CA SC TI v ce MN FE co NI Cu SR Y IR NB M0 TC RU o SN S8 TE TOTALS Table 3.26. A0 L0 MW/MT, BURNUP= CHARGF 00 0.0 0.0 ~ o L Y - ) m 01 * o MmN o0 DD e e (DD 3 DD T D * & 3 8 8 & ® @ DR DIOODIIOTI DD T . 03 c1 £ 01 03 05 R4E 02 02 n3 2.71E 2 3-31 Masses of Activation-Product Elements Calculated to Be Present in the Zircaloy-lLi Cladding and Inconel Spacers of Irradiated Diablo Canyon Fuel Assemblies DISCHARGF 1.57F=09 2e26E~02 &'-’["BC-OQ 1.56E~16 1.,80F 01 1.42F=-14 T.73E-12 Te14F~19 2,828-09 3L,32E-C7 4,3BE~Q4 1.08 22 1.80F 01 4,27€E~-08 6.68E-17 1.045-09 3.49F=11 L, 6565F=04 1.808 02 7,228 20 3.66F% 03 1.78F 01 2,74 D13 5.43F 01 Q,56F 013 1.97€-01 5.91F=-03 2.49F €5 9.55¢F 02 B.68F 02 8.12E=-001 2.46E-01 1.73E-07 2,76E 03 8.345"'01 2.,76E=-02 2.7T1E €5 20.D 1.54E-09 2+26E-02 4 ,48F-00 3,326E-005 2.10E-16 1.808 01 1.80F=14 T.73F=1? 7.14F=-10 2,82E-00 2+30E-10 4,39E-04 1.08E 02 1.80F 01 4.°7E-OQ 7.136-12 1.04E-00 3.36F=11 4,66E~( 4 6. 57E-O5 1.89F C©2 T7.51F OO0 2,66FE 03 1.78¢ 01 3,74 72 5.39F 01 9.55F 03 3.88g-01 1.14E-03 2.49F OS5 9.54F 02 5.69E 02 R.12€-11] 2.46F-01 1 cqu-C-’ 3.THE (2 8.32€=-01 2,09E=-02 2.T1F C5 33000, GRAMS / METRIC TON 160.0 2+ 26F=02 4, 4R8F =00 A, BRE=NE l‘o 1&;—16 1.80E Q01 2.05E=14 T. RE-12 Tel4E=10C 3, B2E-00° 2.3CE-1C 4,39F-Ch 1.0RE 02 1.8CE 01 4.97E-08 3,52E-172 1.04FE-009 3.36E-11 4, 66[:-0‘* 4,27E-0% 1. 8CE 02 T. 528 NC 3.66E 032 1.78F M 3, T4F 03 5.3 01 9. 56F 03 3, 99F-1 1.95E-01 5.62F=-04 2449F DS 0,54F 02 5. 70 02 R, 12F=01 2.408F=01 1.69E-07 3. 76E (3 g, 31£-01 3,27E-02 2.T1E 05 -= CLADDING ACTIVATIIN MWD /MT, FLUX= S.82F 12 N/CM%%2-SE( FUEL CHARGED TO REACTOR 265,D 1.48F=~"Q 226E=02 t‘l“QF—Oq 2,26E=-05 T.REF=154 1.80F OV 2.96E=14 T«T3F=12 T l4E=20Q 3.,R2E-C° 2.20E-10C AOBQE“'OI‘ 1.088 N2 1.R0F 01 4.,97F=08 7.1?F-12 1.04F=09 3.36E-11 4,65F=04 1.,32E-05 1.80F 772 T.84F DO 2.66F 03 1,798 A1 3.74F 03 f.33¢ 01 .88 Q13 6.,45F-N1 1,05¢-01 4-’46':‘05 2+49E DS 3.54F Q2 5. TOF 02 R.12E-M1 2 4EE-01 1.698~-07 3,76E 03 R26E-01 2.81F=02 2.7T1E 05 1536,D 1.32E-(0 2 026:-9? 4, 48FE=-0C 2L,246E=LF5 2 08F=1% 1.80E 0! 6.05FE-14 T.72E=-12 T.16F=-10 3.82E-09 2.,30F=-10 4,39E-06 1.Nn8F 02 1.20F ) 7013&.-! 2 1 QOQE‘OQ 3.26F=-11 4,65F-00 T L4FE-06 1,80 N2 T.54F 0OC 2.665 Q% 1.2 01 3,74F N2 S.21% 01 .56 N2 5.75F=-(1 1.95€E-D1 9,54E-00 2 .49C (5 9.54E 02 5.7CF 02 R.12E=01 2.46E-01 1.69F-07 3.T6RE (3 R.13F~=01 5.,12E-0? 2.71E S 1C9%58,D 2up°E-10 2.265=02 4 L8F=00G 3026!:—’}5 1.01F=-14 1.R0E 31 4,T76E~-13 TaT3F=12 701‘?F-19 2,82E-09 2.30%-10 4,39E=04 1.0RE 02 1.80F 01 4,97F=(8 7.13E=-12 3.,52E~13 1.04F-09 3.36E-11 46504 T.45F-06 l.0F 22 7.54% 00 2.66F G2 1.85F 2.T74F 4 ,85F 3.56F 2.22F 1.95€=-01 T.56E-1D 2+49F G5 9.54F 02 E.70F 02 g.,12E-01 2.46E-01 1 .69E-Q7 3.76F 03 7.94F=-01 T7.07€-02 2.71F {5 Table 3,27. 3-32 Calculated Radioactivity of Activation-Product Isotopes Present in the Zircaloy-l Cladding and Inconel Spacers of Irradiated Diablo Canyon Fuel Assemblies DIABLO CANYON REFERENCE LWR —-— CLADDING ACTIVATICN PDWER = NUCLINE RADIDACTIVITY, CHARGE DISCHARGE 90.D 150.0 3265,D 1096.,D 1095R8,D SC 46 (0.0 4,24F Q0 2.01F 00 1.23F OC ?.08E~-D]1 4,95FE~-04 (0.0 CR %Y .0 2.92F N4 R,10F 02 £,95F N2 3,2T7F N0 3,97E=-0R (0 MN 54 1,0 2.48F N2 2,01E 02 1,76F 02 1.07F 02 2.02F 01 3,23E-Q09 FE 85 0,0 1.99¢ 03 1.,8AF 02 1,78BF 03 1,52F 02 8,94F 02 &K.68F-(G1 CO 58 A0 Q,27E C3 3,87 Q3 2,16FE 02 2 ,67F 02 2,19F=01 0.,C €0 60 0.0 £.42F 03 6,22E 03 6.,CRFE 02 5,63F 03 4,32F 02 1,.23F 02 NI 8¢ 0.0 ALR3IE 00 2,83F £C 2,82F CC 3,83F 00 3,82 (" 2,83F 0D NT A3 .0 E.62F N2 §,62F 02 5,.61E 02 5,59F 22 5.50F (2 &,4%F (2 SR Ra 02,0 4,74F 01 1,31F 01 S5.87F 00 3.,34FE-01 1,96E-0% 0.0 ¥ 91 0.0 R,O5F DY 2.79F D1 1.37F 01 1.79F 0O 1.97E~-4 2,0 IR 95 .0 2.87F D4 1,1NE N4 S5,80C N2 5,86E 02 2.41F-(1 C."° NB 95 0.0 2.78F 04 1.7RE Q4 1,06F (04 1,22% N2 5,12F-01 0.0 SN1aMm 2.0 2.61F 01 2.02F 01 1.72E Q) 9.47F OO 1,25k 0 1 ,66F-12 SB124 0,0 129F 01 &4,93F 0 2,47E OO0 2,06E=-"1 4 ,42F=0C 0,0 S$812% 0.0 4,4AF 01 4,19F Q1 4,02F DY 2,45F 0y 2,07 0OY 2,.01F=-02 TE125M 0.0 2.028 01 1.80F Y 1,69F 01 1.423E D1 B,56% 00 R,35E-03 SUBRTOT 7,7 1.05F 05 4,48F N4 2,8CF N4 Q,98F N3 §,R2F N3 §,77F N2 TOTALS 0.0 1.24FE 05 &4, 48F Q04 2,80F 04 9,058F 03 5,82F 02 5,77F Q2 Table 3.28. Calculated Radioactivity of Activation-Product Elements Present in the Zircaloy-l Cladding and Inconel Spacers of Irradiated Diable Canyon Fuel Assemblies DIABLO CANYON REFERENCE {WR —- CLADDING ACTIVATION POWER= 20,00 MW/MT, BURNUP= 320C00. MWD/MT, FLUX= 5,82F 13 N/CM%%2-SE(C ELEMENY RADIOACTIVITY, CURTIES / METRIC TON FUEL CHARGED TO REACTOR CHARGE DISCHARGE 9%.0 150.D 3685.,0 1996.,D 10958.D H 0.0 1.526-05 1.50E-0% 1.48E-0F5 1.43F=05 1.28FE-05 2.80F=06 BE C.0 B8,95E-04 3.26E-11 3,26E~-11 3,26€=-11 3,26E-11 3,26F~-11 C CoC Se69F=10 S5,869E=10 S,E9FE=1C 5,69F-10N 5,69E=-1] S5.,67F-10 P 0.0 1.29E=07 1,£4F=-00 B,9tE~11 2,67E-15 0.0 0.0 S 0.0 S5.T4F=C9 4 ,60E-19 2.87E=19 5.27E-20 1.66E=-22 (0.0 cL 1.0 1o4685=12 2.,42F=26 2,42F=26 2.,62E-26 2.42E=-2t 2.42E=26 AR 0.0 3.,30FE-07 4.33E=16 4,22F=16 4,32FE~16 &4,30F=-16 4,.01FE-16 K 0.0 4,97E-C6 1.10F-28 1,10F-2R 1.10F-?8 1.10E-2R 1,10F~-28 CA Cal 2TE-C2 2.16E=02 1,6PE=N2 6,80F=03 3,15F=04 2,31F=25 SC D 1.23% 01 2.01F 0O 1.23€ 00 2.08E-Q1 &,95E-D& 0,0 CRr 0.0 2.95E 04 3,10E 02 €,95F 02 3,27f 00 3.97k-CR (0.0 MN N0 3.11F 03 2,Q1E D2 1.76FE 02 1.07F 22 2.02F N1 3,23E=09 FE CL.C 2.21F 03 1,92 02 1.80F C3 1.52F 03 B.,94E (? 6.68F=(1 co 0.0 2.178 04 1,01F 04 8,24F 03 5,90F 03 4.32€ 02 1,23F 02 NI Q.0 1.41E €3 5,66F N2 5,65F G2 5,62E 02 5.54F N2 4,53E 02 Y 0.0 1.5YF 03 2.79F 01 1.37E Q1 1.,09F 00 9,98E-C4 4,11F-04 IR 0.0 2iB7E 04 1.10E 04 S.80E 03 5.,86F {2 2,43F=-01 1.02F-D1 NB A 2.79F 04 1,78F 04 1,06F 04 1,22F 03 §,36F-N"1 9,52F=02 MO 0.0 2+26% 03 2.008-C2 2.00E-02 2.008-02 2.00E-02 1,99E-02 TC 0.0 2.03F 03 1.45F=-02 1.40F=02 1.40F=-N2 1.40F=02 1,40E=D2 cn Q.0 2.83E-03 1,39E-15 Q.0 D.0 0.0 PL.C SN 0.0 2.49F 03 2,08F 01 1.75FE 0Y 9.62F 00 1.33F 00 6.56F=-02 S8 0.0 Bet64F 01 4,.70E 01 4,26FE 01 3,47€ 01 2.07¢8 01 2,ClE-D2 TE C.0 - 2.028 01 1,80 0 {.69F 01 1.43F D1 B8.56F 00 8,35F=03 TOTALS 0.0 1.24F 0S5 4.48FE 04 2.80F 04 9,96F 03 5.82F 03 S§,.77F 02 30,00 MW/MT, BURNUP= 32000. CURIES / METRIC TON MWD /MT, FLUX= S5.,82E 13 N/CM%x%2-SEC FUEL CHARGED T0O REACTOR 11 i Table 3,.29. 3-33 Present in the Zircaloy-L Cladding and Inconel Spacers of Irradiated Diablo Canyon Fuel Assemblies DIABLO CANYON REFERENCE LWR —- CLADDING ACTIVATION POWER= NUCLIDE THERMAL POWER, CHARGE SC 46 CR 51 MN 54 FE 55 FE 59 Co 58 co &n NI 63 SR 89 Yy o1 IR 95 NB 92 NB 95 SN119M SN123M SB124 $B125 $B126 TE125M SUBTOT 2 & @ ® ¢ & 8 @ 8 & P & 9 8 B 8 s O OVOO0ODOO0ODO0OVAOODOONTIOD » o ODODOIOO0 300200 QQOTIDOD TOTALS Table 3.30. 3@000 Hfl?HT, BURNUP= 5.95E-02 1.30F 02 2.00E DO 2.59E N 1.75¢ 00 1.7T4E 02 1.00E ©2 9.01E-02 1.56E=~01 3, D6FE-01 1.50F 02 7.B7E“02 1,33 22 10378—02 1.51€-03 1.8Ge-01 1.50E-01 1099E*01 1.74E=02 6.95€ 02 T.93F €2 S0.D 2.83E-02 1.38E 01 1,62 °C 2.43E CO 4.37E’01 T.24E (01 9,69E 01 8.,99E-02 4,TCE-02 1.C6F=01 5. T6E 01 1-69E-04 8,52 N1 1.07E-02 9,18E-04 6.675‘&2 1.41€E-01 1.35€-03 1.54E*G2 3.31F 02 3,318 02 15Q¢.D 1.72E=-C2 2,08E CO 1.42E €0 2.32E OC 1.72e=-01 4.,04F 01 0,48 0 B, 9RE~02 2.11E-02 5.23E‘Q2 3.04F 01 2« 83E~06 5. (9F 1 9,C7e~-02 6.58E-04 2.33E-02 1.356-01 4, 86E-05 10465-02 2+ 24F Q2 2.24F 02 265.0 2+92E-03 1.458-02 B.66E-C1 1.99€ 00 6.22E=03 5.,00F D0 8,776 01 8,94F-02 1.20E-013 4.,15FE=-N2 2,07 0O 1,21€-12 5.85F OO0 5.00€8-02 2.00E-04 2.7T8E-D3 1.16E-01 3.236-10 1.235‘“2 1.05F 02 1.C5¢% 02 1096,.D 6.95€E-06 1.76E-10 1.63E—01 1.17€ 00 8 .09FE-0¢ 4,10F-01 6.74E 01 B.81F=-02 7.fi5E~fl8 7053E”07 1.26£-03 0.0 246E~02 6.58E=-04 3.,4T7FE=-06 5.98E-C7 H.,92E-02 0,0 7.36E-07 6.89F (1 6,89 O1 Irradiated Diablo Canyon Fuel Assemblies DIABLD CANYON REFERENCE LWR == CLADDING ACTIVATION 1@@6.0 4.56F-10 1.69€-13 0.0 4 ,73E-26 AOSIE—ZQ 5.°6E-1Q 1.92E-07 6.95E~06 1.76E-10 1.63E-0? 1.17F 0O 6. T4t 01 8,81F=C2 1.11E-1¢ S e33E-CH 1.27E-073 2.46FE~013 4 ,97E-CH 2., 40F=05 0.0 T.51E=-D4 h.93E-C2 T26F=C2 Calculated Thermal Power of Activation-Produgt Isotopes 33000. MWD/MT, FLUX= S.82E 13 N/CM&*x2-SEC WATTS / METRIC TON FUEL CHARGED TO REACTOR DISCHARGE 10958.0 C.C 0.0 2.61E-11 2.,71E-04 Calculated Thermal Power of Activation-Product Elements ‘Present in the Zircaloy-l Cladding and Inconel Spacers of FLUX= 5.,R2E 13 N/CM**2-SE( 10958.0D 9.95€c-11 1.68F=-13 0.0 0.0 4,%1F=29 5.56€E-19 S.84F=-28 2.0 0.0 2.61E-11 2,71E=-04 1.92F 00 T.18E-02 5'36E‘07 2.35F=06 1.21€£-05 2.09F=05 4 ,Q8FE=(5 2 .40F-C5 0.0 6.,99E-35 £, T6E=05 7017E“06 pDNERn 30.00 HW’MT, BURNUP= 22000, MHD,MT, FLEMENT THERMAL POWER, WATYS / METRIC TON FUEL CHARGED TO REACTOR CHARGE DISCHARGE 90.D 150,D 3e5,D H 0.0 5.40F=-10 $,32F-10 5.27%-10 5.10E-10 C N 1.69F~13 1,69E=-12 1,69F-13 1,£9E~-113 P 7.0 5.31FE=-10 6.,76E=-12 3,69F-12 1,10E-17 S 0.0 1.21F=-10 1.21€=-22 8,156-22 1,50FE-23 cL o 2.66E=-14 4,51€~29 4,51F~29 4,5]1E-29 AR D 5.07F=09 6£,00F-19 ¢,00E=-19 5,99E-190 CA 0.0 2.09F=05 1.,32E~-05 1,C2FE~-05 4.15E~06 SC C.0 1.35E-C1 2.83E-N2 1,726-02 2.92€E~0C3 CR G.G 1.325 QZ 1.385 93 3.095 OO l.hSE-QZ MN n,0 4,78 01 1.62 CC 1.42F 00 B.65E-01 FE C.0 4.34F CQ 2,876 00 2.5CFE OF 1.99E 3D cn C0 2.TTE 02 1.h9E N2 1,235 £2 9.,27F 31 NI 0.0 6,17 CO 8,99F-02 8,9RE-02 B,94FE~02 SR ¢.0 2.305~01 4,70F=-02 2,11E-02 1,20F~03 Y (U ¢ BJ4TE NQ 1, NEE=-N1 5 23E-0C2 4,15FE-02 IR N.0 1.50FE 0?2 5,76F 01 3.04F 01 2.07E DO NB .0 Y 24&F 02 ..EZE 01 5. (OF 01 5,85%F NN MO £ 1.38F 1 4,97E=-0% 4,97E~"8 4 ,07E~DE TC .0 3.71F 00 2.45F=-0% 2.40F-08 2.,40E-05 ch 0-0 1-48E-05 é.b?E-lB 0'0 0.0 SN C.r 2.54F 0N 1,20F=02 9,83F=-02 5,29F-013 S8 0.0 Bo44F-01 2.09F=01 1, £4RE-01 1.,19F-01 TF C.0 1e74FE~02 1,545-02 ) ,46F-02 1.23E-02 TOTALS .7 T93E 02 3,31F N2 2,.24F (2 1.,N8E N2 6.89E 1 1.99F 00 Al REFERENCE OXIDE LMFBR — WASTE DECAY TIMES BURNUP= POWER SE 80 B8R B8] SE 82 KR 83 KR B84 Table 3.31. 3-3L Masses of Fission Product ILsotopes Calculated te Be Present in Spent IMFBR (AT Reference Oxide) Fuel and in the Wastes Generated by the Reprocessing of This Fuel (PROCESSED AT 30 DAYS) = 58.23 MW/MT, NUCLIDE CONCENTRATIONS, 20.D CHARGE ® % & 9 & ¢ ¢ ¢ 4 ¢ ¢ & ¢t & ° & % 8 S € & B P B St ° B+ " & w ¢ G s O O 9 b P "V s s e * e VOO0ODO0OD0VO0D0OOCUDVO0OODVOCOODIRROVODOOVOCDVOHOIOROO0DOOCOOCO Q0000000 VUO0LOO00LAOONAALVARIAOOOLDLOLODOCOO0OCON0ODLONDNADOTODTOIDOC0 * & & & 5 & & & v 8 e 1.34E 2.23E 4.29E 6.54F 1.14E 2.61E 4.41F 1.08E 1.34F 2.04E 2.26E 2.23F 3.07€ 6.65E 3.77€ 3,09E 4.49F 5.56% 6.30E 9.92E 6.78E 6.,17E 7+ 34E 1.30€ 9.84F 8.56E 8.7T4E 1.01€ 2.04F 8+34E 9.36E 5.49E B.19E 8. TCE 5.61F 3.84% 2.94F %4.96F 3.11F 3.63E 3.95E 1.84F 6.13E 3,715E 3.49E 4,01E 2.03€ 1.05€E 4.16F 1.48F 2.26E l.47€ 1.39€ 1.24F 1.61E 1.52€ 1.63E 1.85€ 3, 66F 5.99E 365.0 1.34E CO 2.23E 0C 4.29E 00 6.54E 0Ol 1.14E Q2 2+46F 01 4.56E 01 1.08E 02 1.34F (2 2.04E 02 2.60E-01 2.45E 02 3.00FE 02 1.35E 01 T.27E-C1 3.46E C2 4.49E 02 5.56F G2 6.3CE 02 2.79E 00 3.19E 00 T.78E 02 T.34E 02 1.30E 01 9.84E 02 8.56E 02 8.74F 02 1.01E 03 2.04F 01 8.34E G2 9.36E 02 1.56E-01 8.74E 02 8.70E 02 5.61E 02 2.04E 02 4.T74E 02 4.96E 02 3.11E G2 3.63E Q2 3,95 00 1.86E 01 6.13E 01 3.75E C1 3.49E QO 4.01E 0O 2.04E 00 1.05€ 00 4.,16E 00 1.48E 00 2.26E 00 l.47€ €O 1.38E 00 1.24E 01 l.61E 00 1.59e 00 1.63€ 00O 1.47€ 01 7.65E 00 5.99€t 01 32977. GRAMS / METRIC TON 1096.D 1.34E 00 2+23E 02 4.29E 09 6.54E 01 1.14F 02 2.16F 01 4,85F 01 1.08€ 02 1.34E 02 2.04E 02 1.52E-05 2.46E Q2 2.86E 02 2.80E 01 1 l32£"04 3.47E 02 4.49E 02 5.56E 92 5.30E 02 1.156-03 1.315“'03 7.84E 02 T.34E 02 1.30E 01 3.84E 02 8.56E 02 8.74E 02 1.01E 03 2.04E 01 B.34E 02 9.34E 02 4.3ATE-Q7 B.T4E 02 8.,70FE 02 5.61E 02 5.13E 01 6.27E 02 4.96E 02 3.11E 02 3.63F Q2 3.95E 09 1.87E 01 6.13E 01 3.75E 01 3.49E 00 4.01E 0O 2.04F 00 1.05E 00 4.16E 00 1.48E 00 2.27E 09 1.47E 00 1.35E 00 1,24E 01 1.61€ 00 1.60F 00 1.63E 00 8.77E 00 1.37€ 01 5.99E 01 MWD /MT, FLUX= 2.65E 15 N/CM**2-SEC FUEL CHARGED TO REACTOR 3652.D 1.34E 00 2.23E 20 4.29E G0 6.54E 01 1.14E C2 1.38c 01 5.64E 01 1.08 02 1.34E 02 2.04E 02 2+43E-20 2.46E 02 2.41E 02 7.33E C1 1.08E~-17 3.47E 02 4.49E 02 5.56E 02 6.30E 02 1.67E~-15 1.91E~15 T.84E 02 T.34E 02 1.30E 01 9.84E 02 B.56F 02 8.T4E 02 1.01E 03 2.04E 01 8.34E 02 9.36E (2 0.0 8.7T4E 02 8.70E 02 5.61E Q2 4.11E-01 6.77E 02 4.96E 02 3.11€ 02 3.63E Q2 3.95E OO 1.88E 01 6.13E 01 3.75E 01 3.49€ 00 4.0l 00 2.04E 00 1.05€ 00 4.16E 00 1.48E 00 2.27E 00 1.47E 00 1.27E 00 1.25E 01 lL.61E 0O 1.60E 00 1.63E 00 1.45E 00 2.12E 01 5.99E 01 36525.D 1.34E 00 2.23E 00 4,29€ GG 6.54% 01 1.14E 02 4.18E-02 7.01F 01 1.08€E Q2 1.34% 02 2,045 02 0.0 2.46E 02 2.61E 01 2.88E 02 0.0 3.47E 02 4.49E G2 5.56¢ Q2 6.30F 02 0.0 0.0 7.84E 02 T.34E 02 1.30€ 01 9.84E 02 8.56F 02 8.T4E 02 1.01€ 03 2.04F 01 B.34E 02 9.36E 02 0.0 8.74E 02 8.70E 02 5.61E 02 0.0 6.,77TE 02 4.96E 02 3.11€ 02 3.63F 02 3.95E 00 1.88¢ 01 6.13E 01 3.75€ 01 3,49 00 4.0l 00 2.04E GO 1.05€E 00 4.16E 00 1,48E 00 2.27E 00 l1.47E 00 5.59E-~01 1.32E 01 1.61E 0O 1.50F 0C 1.63E 00 1.35E-10 2.26F 01 5.98¢€ 01 365250.D 1.34E 0C 2.23E 0C 4,29 CC 6.54E (1 1.14E 02 c.0 T.02E 01 1.08€ 02 1.34F 02 2.04E 02 0.0 2.46E 02 5.96€E-09 3.14E 02 0.0 3.47TE 02 4.49E 02 5.56E 02 6.30E 02 0.0 C.0 T.84E 02 T.34t 02 1.30€ C1 9.84F Q2 8.56E G2 B.71E 02 1.01€ 03 2.04€ 01 B.34E 02 9.36E 02 0.0 8.74E 02 8.70E 02 S.61E G2 G.0 6.77E 02 4.96F C2 3.11E 02 3.63E Q2 3.95E €O 1.88E 01 6.13E 01 3.75E 01 3.49E 00 4.01€ 00 2.04E 00 1.05E 00 4.16€ 00 1.48E 00 2.27E 00 1.47E 00 1.53E-04 1.38E 01 1.61E 00 1.60E 00 1.63€ 00 0.0 2.26E C1 5.95E 01 3-35 Table 3.3l (Continued) Al REFERENCE OXIDE LMFBR — WASTE DECAY TIMES (PROCESSED AT 30 DAYS) POWER= 58.23 MW/MT, BURNUP= 32977. MWD/MT, FLUX= 2.65E 15 N/CM*%2-SEC NUCLIDE CONCENTRATIONS, GRAMS / METRIC TON FUEL CHARGED TC REACTOR CHARGE 30.D0 365.0D 1096.0 3652.0 36525.0 365250,0D TE126 0.0 1.60E 00 1.61E 00 1.62E 00 1.62E 00 1.66E 00 2.03F OC TEL127M 0.0 6.4TE 00 7.69E~01 T.36F-03 6.44E-10 0.0 0.0 1127 0.0 1.15E 02 1.21E 02 1.22E 02 1.22E D2 1.22E 02 1.22E 0z TE128 0.0 1.61€ 02 1.61E 02 1.61E 02 1.61E 02 1.61F 02 1.61E 02 XE128 0.0 3.T6E 00 3.76E 00 3.76E 00 3.76E G0 3.76E 00 3.76E 00 TE129M 0,0 6.08E 00 6.58F-03 2.22E-09 0.0 0.0 0.0 1129 0.0 3.27E 02 3.33E 02 3.33E 02 3.33FE 02 3.33E 02 3.33F C2 TE130 0.0 3.65F 02 3.65E 02 3.65E 02 3.65E €2 3.65E 02 3.65E (02 XE130 G.C S.T77E 00 9.77E 00 9., 7T7E Q0 9.77TE 00 9.77E 00 9.77E 00 XE131 0.0 6480E 02 6.81E 02 6.81E 02 6.8lE 02 6.81FE 02 6.81F 02 XE132 0.0 9,78E 02 9.78E 02 9.78E 02 9.78E 02 9.78BE 02 9.78F 02 CS133 0.0 1.24E 03 1.24F 03 1.24EF D3 1.24F 03 1.24E 03 1l.24E 02 XE1l34 0.0 1.37€ 03 1.37E 03 1.37F 03 1.37E 03 1.37E 03 1.37F 03 CS134 0.0 2.22E 01 1.63E Ol 8.29E 00 7.78E~01 4.96E~-14 0.0 BAl34 0.0 4,61E 00 1.05E 01 1.86F 01 2.61F 01 2.68E 01 2.68F 01 CS135 0.0 1.33E 03 1.33E 03 1.33E 03 1.33E 03 1.33FE 03 1.33E 03 XE136 0.0 1.22E 03 1.22€E 03 1.22F 03 1.22E 03 1.22E 03 1.22E 03 BAl136 0.0 3.79 01 3.82E 01 3.82E 01 3.82E 01 3.82F 01 3.82€F 01 CS137 0.0 1.25F 03 1.23E 03 1.17E 03 9.97E 02 1.25E 02 1.16E-07 BA137 0.0 2.45F 01 5.08E Ol 1.06E 02 2.81EF 02 1.15E 03 1.28E 03 BAl138 0.0 1.22E 03 1.22E 03 1.22E 03 1.22E 03 1.22E 03 1.22€ ©3 LA139 0.9 1.14E 03 1.14E 03 1.14E 03 1.14E 0G3 1.14E 03 1.14F 03 BAl140 0.0 7.17€ 00 9.34E-08 0.0 0.C C.0 0.0 CE140 0.0 1.24F 03 1.25E 03 1.25E 03 1.25f£ 03 1.25€E 03 1.25E 03 CEl41 0.0 5.18E 01 4.00E-C2 6.45E-09 0.0 0.0 0.0 PR141 C.O 1.08% 03 1.13E 03 1.13F 03 1.13E 03 1.13E 03 1.13E 03 CEl42 0.0 1.01F 03 1.01€E 03 1.01E 03 1.01E 03 1.01E 03 1.01E G3 ND142 0.0 4,78E 00 4.7BE 00 4,78E 00 4.78E 00 4.78E 00 4,78E 0C PR143 0.0 9.66E 00 4.15E-07 3.60E-23 0.0 0.0 ¢.0 ND143 0.0 1.03E 03 1.04F 03 1.04F 03 1.04FE 03 1.04E 03 1.04F 03 CEl44 0.0 4.,00F 02 1.77E 02 2.96E 01 5.79E-02 0.0 0.0 ND144 C€.0 4.46F 02 6.69E 02 8.16F 02 8.46F 02 B.46E 02 B.46E 02 ND145 0.0 6.,65E 02 6.65F 02 6.65E 02 6.65E 02 6.65E 02 6.65E 02 ND146 0,0 6.17E 02 6.17E 02 6.17E 02 6.1TE 02 6.17E 02 6.17E 02 ND147 0.0 2.31E 00 1.90E-09 0.0 0.0 0.0 0.0 PM14T 0.0 3.80E 02 3.00F 02 1.77E 02 2.77E 01 1.26E-C9 0.0 SM147 0.0 8.71E 01 1.,70E 02 2.93E 02 4.42F 02 4,.70E 02 4.70F 02 ND148 0.0 3.83E 02 3.83E 02 3.83F 02 3.83E 02 3.83E 02 3,.83F 02 PM148M 0.0 1.97E 00 7.83E-03 4.55E-98 0.0 0.0 0.0 SM148 0.0 3.44E Ol 3.64E 01 3.64FE 01 3.64E 0l 3.64E 01 3.64E 0Ol SM149 0.0 2.83E 02 2.83FE 02 2.83E 02 2.83E 02 2.83E 02 2.83E 02 ND150 0.0 2.26E 02 2.26E 02 2.26E 02 2.26E 02 2.26E 02 2.26E 02 SM150 0.0 2,07 01 2.07E 01 2.07€ 01 2.07E O1 2.07E Ol 2.07E Ol SM1S51 0.0 1.72E 02 1.71E 02 1.68E 02 1.59E 02 7.78E 01 5.97E-02 EU151 0.0 1.09€ 00 2.34E 00 5.05€ 00 1.42E 01 9,.57E 01 1.73E 02 SM152 0.0 1.53F 02 1.53F 02 1.53E 02 1.53E 02 1.54E 02 1.54E 02 EUls53 0.0 T.48E Ol 7.48E 01 T.48E 01 T7.48F Ol 7.48FE 01 7.48E Gl SM154 0.0 6.04E 01 6.04E 01 6.04E 01 6.04E 01 6.04F Q01 6.C4E 01 EUl54 0.0 6.T2E 00 6.46E 00 5.93E 00 4.328E 00 8.87E-02 1.03E~-18 EUl55 0.0 6.23FE 01 4.38E Ol 2.04E Ol 1.40E 00 1.51E~15 0.0 GD155 0.0 2.08E Ol 3.93E 01 6.27E 01 8.,17E Ol 8.31E 01 8.31€ 01 GD156 0.0 4,80E 01 4.85E 01 4.85E O1 4.B5E 01 4.85F 0l 4.85E 01 GD157 0.0 2.76E 01 2.76€E 01 2.76E 01 2.76FE 01 2.76E 01 2.76E 01 GD158 0.0 TeS54E 01 T.54E 01 T.54E 01 7.54E 01 7.54E Q01 7.54t 01 TBl59 0.0 3.75€ 01 3,75E 01 3,75E Ol 3.7S5E 01 3.75E Ol 3.75E 01 GD160 0.0 3.12€ 00 3,.12E 00 3,12E 00 3.12E 00 3.12E OC 3.12E 00 DY1I60 0.0 2.72E 00 3.52E 00 3.55E 00 3.55E 00 3.55E 00 3.55& 00 DY161 0.0 T.68E 00 7.69E 0D 7.69E 00 T7.69E 00 7.69E 00 7.69E 00 GD162 0.0 2.00E 00 1.06E 00 2.65E-01 2.07E-03 0.0 0.0 DY162 0.0 1.79E 00 2.73E 00 3,53E 00 3.79E 00 3.79E 00 3.79E 00 oY163 0.0 1.46E 00 1.46E OC 1.46E 00 1.46E 00 1.46E 00 1.46E 00 SUsTOT 0.0 3,49 04 3.49E 04 3.49E 04 3.49E 04 3.49F 04 3.49E 04 TOTALS 0.0 3.49E 04 3.49E 04 3.49E 04 3.49E 04 3.49E 04 3.49F 04 A1 REFERENCE OXIDE LMFBR - WASTE DECAY TIMES POWER= 58,23 MW/MT, BURNUP= Table 3.32. 3-36 Masses of Fission Product Elements Calculated to Be Present in Spent IMFBR (AI Reference Oxide)} Fuel and in the Wastes Generated by the Reprocessing of This Fuel {PROCESSED AT 30 DAYS) ELEMENT CONCENTRATIONS, CHARGE 30.0 9.,61E-02 1.20E-08 5.24E-09 1.63E-01 1.86E-02 6.5TE CO 2.23E 3.13E 1.78E 5.34E 2.61F 2. T8E 6.7T9E 3.48E 8+ T4E 3.10E 8.19E 1.67E 3.64E 1.26F 2.C4E T.42E 3.24F 5.45E 4.43F 4.26F 3.85E 1.29¢€ 1.14E 2.T0E 1.09E 3.38E 3,.82E B.11E 1.45E 1.77E 3.83E 1.43E 2.21E-01 1.565-01 3.49E 04 32977. GRAMS / METRIC TON 365.D 9.13E-02 0.0 0.0 1.63E-Cl 1.86E-02 6,57 0OC 2.23E 00 3.11E 02 1.79€ 02 5.05E 02 2.46E 02 2.73E 3.19E 3.64E 8. 74E 2.86E 8, 74¢ 1.85F 3.63E 1.27E 2.08E T.41E 2.86E 5.37E 4.54E 4.26E 03 2.82E 1.32E l.14E 2.43E l1.13E 3.61E 3. 00E 8.95E 1.27E 1.95E 3.75E 1.61E 2.21E-01 1.56E-01 3.49E 04 1096.D 8.16E-02 0.0 0.0 1.63E-01 1.86E-02 6.,57€ 00 2.23E 3.08E 1.82E 4.90E 2.46E 2.74E 2.28E-03 3.64E 03 8.74E 02 2.7T1E 03 8.74E 02 2.00E 03 3.63E 02 1.27E 02 2.13E 00 T.40E 01} 2.28E 01 5.42F 02 4,55E 02 4.,26E 03 3.75E 03 1.38E 03 l.14E 03 2.28E 03 1.13E 03 3.7%E 03 1.77€ 02 1.02F 02 l.06F C2 2.19€ 02 3.75€ 01 1.69E 01 2.21E-01 1-56E-01 3.49E 04 MWD/MT, FLUX= 2.65E 15 N/CM**2-SEC FUEL CHARGED TO REACTOR 3652.D 5.50E-02 0.0 0.0 1.63E-01 1.86E-02 6.5TE 2.23E 3.01€ 1.90E 4 J45E 2.46E 2.79E 2.T6E-Q3 3.64E 8. T4E 2.66E 8.74E 2.05E 3.63E 1.26F 2.27E T.39E 1.55E 5.50E 4.55E 4.26E 3.57E 1.56E la.1l4E 2+25E 1.13E 3.78F 2.77E 1.16E 9.48E 2.39E 3.75E 1.71E 01 2.21E-01 1.56E-01 3.49E 04 36525.D 3.45E-04 C.0 C.0 1.63E-C1 1.86E-02 6.57TE QO 2.23E 0C 2.87E 02 2.04E Q2 2.31F 02 2.46E 02 3.00E 03 2.59E-02 3.64E 03 B.74E Q2 2.66E 03 B.74E 02 2.05E 03 3.63E 02 l.26E 02 2.60E 00 T.32E 01 1.48E 01 5.51E 02 4,55 02 4.26E 03 2.70E 03 2.44E 03 l.14E 03 2.25E 03 1.13E GC3 3.78F €3 1.26E-09 1.10E 03 1.71E 02 2.45F Q2 3.75E 01 1.71E Q1 2.21E-01 1.56E‘01 3.49E 04 365250.D G.0 0.0 0.0 1.63E-C1 1.86E-C2 6.56E 0OC 2.24E 00 2.87E 02 2,04E 02 2.04F 02 2.46E 02 3.03 03 2.57E-01 3.64E (3 8.71E 02 2.66E 03 8.74E 02 2.05€ (3 3.63E 02 1.26E €2 2.61F 00 T.23E 01 1.54F 01 5.51E 02 4.55E8 02 4.26F 03 2.57E 03 2.56E 03 1.14E 2.25E 1.13E 3.78E 0.0 1.02E 2.48E 2+45F 3.75E 1.71E 2.21E-01 1.56E~01 3.49E Q4 3-37 Table 3.33. Calculated Radioactivity of Fission Product Nuclides Present in Spent IMFBR (AT Reference Oxide) Fuel and in the Wastes Generated by the Reprocessing of This Fuel A1 REFERENCE OXIDE LMFBR -~ WASTE DECAY TIMES POWER= 58,23 MW/MT, BURNUP= 32977. MWD/MT, (PROCESSED AT 30 DAYS) FLUX= 2.65FE 15 N/CM**2-SEC NUCLIDE RADIOACTIVITY, CURIES / METRIC TON FUEL CHARGED TO REACTOR o CHARGE 20.D 365,D 1096.D 3652.D 36525.D 365250.D H 3 C.0 9.32E 02 8,85E 02 7.91E 02 5.33F 02 3.,34E C2 0.0 KR 85 0,0 1.02E 04 9,63E 03 B.46F 02 5,39E 03 1.64% Cl1 0.C RB 86 0.C 1,028 03 4.03E-03 6.49E-15 0.0 0.0 0.0 SR 89 C.0 6.37E 05 7.33E 03 4.295-0]1 6.85E-156 0.0 0.0 SR 90 0.0 4,34E 04 4,25E 04 4.,04F 04 3.40F 04 3.69E 02 8,.,43E-07 Y 90 0.0 4.35FE Q4 4.25E 04 4.04F 04 3.40E 04 3.70FE 03 B.42E-07 Y 91 C.0 9,21E 05 1.78E 04 3.,21E 00 2.64E-13 C.0 0.0 IR 93 Q.0 1e43F 00 1.43E 00 1.43FE 00 1.43E 00 1.43FE CO0 1l.42E 0OC IR 95 .0 2.10E 06 5.,B9F 04 2.43E C1 3.53E-11 0.0 0.0 NB 95M (,C 4.45E 04 1.25E 03 5,15E-01 7.49E-13 0,0 0.C NB 95 0.0 2466F 05 1.25F 0F 5,15F (01 7.49E-11 D.0 0.0 MO 99 0.0 1.81E 03 0,0 0.0 0.0 0.0 0.0 TC 99™ (0.0 l1.73E 03 0,0 0.0 G.C C.C 0.C TC 99 Q.0 1.49E 01 1.49E 01 1.49E Q1 1.49E 01 1.49E 01 1.48% 01 RUI03 0.0 1.76E 06 5.00FE 03 1.40E-02 0.0 0.0 0.0 RH103M G.0 1.76E 06 5.00E 03 1.4CE-02 0.0 0.0 0.0 RUl106 0.0 1.29E GC6 6.85E 05 1.,72E 05 1.38E 03 C.D 0.C RH106 O0.C 1.,29E 06 6.85E 05 1.,72% 05 1.38E 03 0.G 0.0 AGl10M 0.0 1.59E 03 6.34E 02 B.55E 01 7.78E-02 0.0 0.C AG110 0.0 2.06E 02 8.,24% 01 1.11FE ¢l 1.01E-02 0.0 0.0 AGlll Q.0 1.26E 04 4.50E-10 0.0 0.0 0.0 C.0 CD113M 0.0 1.26E 02 1.20F 02 1.09t 02 7.72E 01 8,95£-01 3.98E-2C IN114M 0.0 1.43F CO0 1.3BE-02 5.45E-07 0.0 0.0 0.0 IN114 C.0O 1.38E 00 1.33E-02 5.26E~07 0.0 0.0 0.0 CD115M 0.0 2469F 02 1.22E 00 9.33E-06 0.0 0.0 0.0 SN119M G.0 2.1GE 01 B8.2BE 00 1.09E 00 9.12E-04 0.0 0.0 SN121M 0.0 5.41F 01 5.,36E 01 5.26F 01 4.94E 01 2.,17E 01 S5,.96F-02 SN123M 0.0 6.86E 02 1.07F 02 1.86E 00 1.29E-06 0.0 C.C TE123M 0.0 2.91E 00 4,0CF-01 5.27TE-03 1.39E-09 0.0 0.0 SN125 9.0 6.,72F 03 1,26E-07 0.0 0.0 0.0 0.0 SB125 0.0 1.96E 04 1.55E 04 9.29E 03 1.54F 03 1.43E-C7 0.0 TE125™ 0,0 6.86FE 03 6.41E 03 3.86E 03 6.39E 02 5.91F£-08 0.0 SN126 0.0 1.708 00 1.70E OO0 1.70F 00 1.70E OO0 1.70E CO 1.6%E 0O SB126M 0.0 1.7CE 00 1.70€ 00 1.70E 00 1.70E 00 1.70E OC 1.69E OC SBl12é6 9.0 9.35E 02 1.68E 00 1.68E 00 1.68E OC 1.88E 00 1.67E 0O $B127 C.0 1.60E 03 0.0 C.0 0.0 0.0 0.0 TE12TM 0.0 6.11FE 04 T7.26E 03 6.95F 01 6.08E-06 0,0 C.C TE12T 0.0 6.,18E C4 7.17€E 03 6.87E Q01 6.01lE-CH6 0.0 0.0 TE129M C.0 1,81F 05 1.96E 02 6.81E~05 0.0 0.0 0.0 - TE129 (C.0C 1.16E 05 1.,26E 02 4.23E-05 0.0 0.0 0.0 1131 Q.0 1.39E C5 4.12E-08 0.0 0,0 0.0 0.0 XE131M Q.0 6.19E 03 2.44E-05 0.0 0.0 0.0 0.0 TE132 0.0 4,17€ 03 0.0 0.0 0.0 0.0 0.0 " 1132 0.0 4.30E 03 Q.0 0.0 0.0 0.0 0.0 XE1l33 0,0 T+44E 04 5.44E-15 0.0 0.0 0.0 N.0 CS134 (.0 2.90E 04 2.,12F 04 1,08E 04 1.,01E 03 6.46E~11 C.0 CS135 0.0 1,176 00 1.17E 00 1.17E 00 1.17E 00 1.17E 00 1.17€ 00 CS136 0.0 2.88E 04 5,01E-04 G.0 0.0 0.0 0.0 Al REFERENCE OXIDE LMFBR - WASTE DECAY TIMES 3-38 Table 3.33 (Continued) POWER= 58,23 MW/MT, BURNUP= 32977, NUCLIDE RAODIDACTIVITY, CURIES / METRIC TON CHARGE 30.D 365.0 1096.D €S137 0.C 1.09E 05 1.07€ 05 1.02E 05 BALI37M 0.0 1.02% 05 9.99E 04 9.53E 0% BA140 0.0 5.23F 05 6.81E-03 0.0 LA14C 0.0 6,01 05 7.8B3E-03 0.C CEl4l 0.C 1.48F 06 1.14E 03 1.85E-04 PR143 0.0 6.44F 05 2.,7TE~Q02 2.4CE-18 CEl4e4 0.0 1.28E 06 5.64E 05 9.47E 04 PR144 O0.C 1.28E 06 5.,64E 05 9,47 04 ND147 0O.C 1.85E 05 1.52E-04 0.0 PM147 0.0 3.53E 05 2.79E 05 1.64E 05 PM148M 0.0 4.155 Q4 1.65E 02 9.56F-04% PM148 0.0 4.93E 03 1.32E €1 7,68BE-05 PM149 0.0 6.15E 01 0.0 0.0 SM151 0.0 4.69E 03 4.66F (03 4.59E 03 EUl152 0.0 1.05E 01 9.93E OC 8,84F 0D EUl54 0.0 9.76E 02 9.38E 02 8.6CE 02 EU155 0.0 T«94FE 04 5.59FE 04 2.60F 04 gEUlsSe 0.0 3,06E 04 5.BC0E-03 1.24E-17 TB160 G.0 3.46E 03 3.78F 02 3,.35E-01 18161 C.0 9.08E 02 2.,20FE-12 0.0 GD162 0.0 4.42F 03 2.,34E 03 5,.84FE 02 TB162M C.0 .42 03 2.24F 03 5.,84EF 02 SUBTOT 0.0 2.01E Q07 3.43F G6 1.04F 06 TGTALS 0.0 2.01F 07 3.43FE Q6 1.04E 06 {PROCESSED AT 30 DAYS) MWD /MY, FLUX= 2.65E 15 N/CMx%2-SEC FUEL CHARGED TC REACTOR 3652.0 8.67E 04 8.11E 04 0.0 1.85E 02 1.85E G2 8% 04 2.81E 05 2.81€ 05 36525.D 1.08E 04 1.01E 04 0.0 0.C . o TE-06 2E 03 6E-02 SE Ol 3E-12 ¢ & & & * o & & & o @ QOO DONNMFHOOOGMMOOODOQ WOOOOOrMPNDIOOD=OOIO0OOQ 06E C4 3,06E C4 365250.D 1.01E~05 9.44E-Q6 & & & & o € * NDOCOODOMROEODOMNDODOONDOODO HFO000O0OVMOROO0OOND00O0O0OOOO ® & & 8 & & & & 0o+ o 2 1€ 01 2.56F 01 Table 3.3k, Al REFERENCE POWER= Calculated Radioactivit in Spent IMFBR (AT Reference Oxide 3-39 by the Reprocessing of This Fuel BURNUP= 32977. OXIDE LMFBR — WASTE DECAY TIMES 58.23 MW/MT, ELEMENT RADIOACTIVITY, CURIES / METRIC TON H IN GA AS SE BR KR RB SR TOTALS CHARGE c.0 6.0 c.0 0.0 .0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 30.0 9.32€ 02 1.136-02 1.62E-02 8.92E-C4 3.92E-02 2.06E-04 1.02E 04 1.02€ 03 6.81E 05 9.64E 05 2.10€ 06 2.71E 06 1.81€ 03 1.74E 03 3.05€ 06 3.05E 06 5.03E-~06 1.44E 04 3.96E 02 3.40E 00 T.48E 03 2.22E 04 4.31F 05 1.43E 05 B.06E 04 1.67E 05 6.25E 05 6.01E 05 2.76E 06 1.92E 06 1.85€ 05 3.99E 05 4.70E 03 1.11E 05 4.42E 03 1.48E 04 6.13E-01 9.14E-01 2.01E 07 365,0 8.85E 02 02 1.22€E 02 2.70E-02 l1.71€ 02 1.55€ 04 2.12€ 04 S«43E~02 2044E‘05 1.28E 05 9.99E 04 T.83E-03 5.65E 05 S¢64E 05 1.52E-04 2. T9E 05 4.66E 03 5.68E 04 2.34E 03 2.T2E 03 0.0 0.0 3,43E 06 1096.D T«91E 02 0.0 o0 OWwWwoOo «0 «92E-02 «0 8.46F 03 1.105’05 4.04E 4.04E 2.57E 5.23E 0.0 1.49E 1.72€ 1.72E 0.0 9.86E 01 1.09€ 02 1.,07e-06 5.T3E 01 9.29E 03 3.99E G3 5.43E-02 0.0 1.13E 9,53E 0.0 9.47E 9.47E 0.0 1.64E 4.59E 2.68E 5.84E 5.85E 0.0 0.0 05 04 04 04 1.04E 06 MWD/MT, of Fission Product Elements Present Fuel and in the Wastes Generated (PROCESSED AT 30 DAYS) FLUX= 2.65E 15 N/CM&®%2-SEC FUEL CHARGED TC REACTOR 3652.D 5+33E 02 0.C «0 CWwWwoo 0 092E-02 «0 5.39E 03 1.10E-05 3.40E 04 3.40E 04 1.43E 00 6.01E-01 C.0 1.49E 01 1.38€E 02 1.38€E 03 0.0 80795‘02 7.T2E 01 0.0 5.11E 01 l.54E 03 6.39E 02 5.43E-02 0.0 8.78E 8.11F 0.C 1.85E 1.85E 0.0 2.58E 4.34F 2.42E 4.57E 4.57E 0.0 0.0 04 04 02 c2 04 c3 03 00 00 2.81E 05 36525.D 3.34F 00 G.0 0.0 0.0 3.92E-02 0.0 1.64E 01 1.10E-05 3.69E 03 3.70E 03 l1.43E 0O 1.42E 0C 0.0 1.49€ 01 1.01E 04 0.0 0.0 0.0 0.0 1.17E-06 2.12E 03 1.29E 01 0.0 0.0 0.0 0.0 3.06E 04 3652%0.0 0.0 8E-02 DO WDHOO OQOO®DHOO 1.10E-05 B+43E-C7 8.43E-C7 1.42E 00 1.42E 00 o . o 1.48E 01 OO0V OOQOOPmOPOWTPDOLOOOO 8E-20 € 00 E 00 o 0 3E-02 TE 00 4E-0Q6 3E 00 OE-~-16 N OO0 OHMOODOOOVMFROVOWMOWOOOD Al REFERENCE OXIDE LMFBR - WASTE DECAY TIMES POWER= 58,23 MW/MT, NUCLIDE THERMAL POWER, 172 < Pt b O = QOUVDOOOOOLOOLOOQLDNMYVOOLROOOOOOCORDAOLOOOOOLDOoLDDOODODOOOC *« & % & © 5 B 2 9 & @ Table 3.35. CHARGE # & 2 9 % 8 ¢ 9 & & ¢ & & & & o & & o & e % & » o . OO0OQOODOO0OOQCOO0OLOO0DO0VO0O0VUVOOOOOO0O0OOOCOOOLODOODOOOODOOOODOO 3-L0 Calculated Thermal Power of Fission Product Nuclides Present in Spent IMFBR (AI Reference Oxide) Fuel and in the Wastes Generated by the Reprocessing of This Fuel 30.D 3.31E-02 1.64E 01 4.81E 00 2.29E 03 S5.67TE 01 2+.49E Q2 3.50E 03 1.10E 04 6.20€ 01 1.28E 04 8.11E 00 1.46E 00 1.01E-02 5.75E 03 4.18E 02 T.65E 0Ol 1.25E 04 2.56E 01 1.50E 00 3.01E 01 1.67E-01 2.03E-03 6+49E-03 1.10E 00 1.11E-02 5067E_02 2+34E 00 4-285-03 1.04E 00 4.,06E 01 6.57TE 01 5.89E 00 1.02E 01 8.81E 00 3.37€ 01 1.01E 02 3.59E Q2 4.21E Q2 5.71E 02 6.02E 00 5.695 00 6.88E 01 8.03E 01 3.,04F 02 4.45E 02 1.77e 02 4.00F 02 1.76E 03 9,98 03 BURNUP= 365.0 3.15E-02 1.55€ 01 1.90E-CS 2.64E 01 5.54E 01 2.43E 02 6.T6E 01 3,08E 02 1.74E 00 6.01E 02 0.0 0.0 1,01E-02 1.63E 01 1.19¢ 00 4.06E 01 b.62E 03 1.02E 01 5.98E-01 1.08E-12 1.59€E-01 1.96E-05 6.25&‘05 4.95E-03 4.375-03 5.62E-02 3.65E-01 5.88E-04 2.16E-02 T.50E-10 5.21E 01 5.51€ 00 1.85E-G2 0.0 4.0CE 00 1.17E 01 3.88E-01 4.565*01 1.7CE~10 2.3T7E-08 0.0 0.0 5.87E-18 2.23E 02 7075E’06 1.73E 02 3,92E 02 2.,30E-05 1.30€E-Q4 32917. 1096.D ZOBIE-OZ 1.36E 01 3.05E-17 1.54E'03 5.27E 01 2.31E 02 1.22E-02 1.27E-01 T.17E-04 2047E-01 0.0 0.0 1.01E-02 4,57E-05 3032E“05 1.02€ 01 1.66E 03 1.38F 09 8.06E’02 0.0 1.44E-01 T.7T6E-10 2.48E-09 3.80E-08 5.75E-04 5.52E-02 6.33E-03 T.74E-056 4.65E-06 0.0 3.12E 01 3.31E 09 IQBSE-OZ 0.0 3.83E-02 l.12E-01 1.31-07 1.54E-07 0. 3E 02 5€ 02 4E 02 CSCOWHOOOOD O OOYdOO-HDODOOO0O MWD/MT, (PROCESSED AT 30 DAYS} FLUX= 3652.0 1.90E-02 B.66E 00 0.0 2.4TE-18 4.44E O1 1.95E 02 1.00E~15 1.85E~13 1.0§E-15 3.595‘13 0 0. 0. 1.01E-02 0. C. OO0 8.18E=~02 1.33E 01 1025E‘03 7 .34E-05 0.0 10025-01 D.0 0.0 C.0 4.815-07 5.18E'02 4.405-09 2.04E-12 6.97€E-19 0.0 5.17€ 00O 5.49E~-01 1.84E-02 0.0 3.355‘09 9,80E-09 0.0 0.0 0.0 0.0 0.0 0.0 0.0 1.06E 01 0.0 l1.40E 02 3.18F 02 0.0 0.0 2.65E 15 N/CM**2-SEC 36525.0 1.19E-04 20635‘02 0.0 0.0 4.82E 00 2.11€ 01 0.0 C.0 0.0 BE-03 ¢ & & & ¢ & & 2 P e > 8E-02 8E-10 BE-11 4E-02 WATTS / METRIC TON FUEL CHARGED TO REACTOR 365250.0 0.0 0.0 E-09 E-09 N QO - ® QOODDOOL~OOD a & & 9 @ 0E-02 QO NYOOOOOLOOLOOUODLPWOOOOOODNOCOOONOOOODOOO 6E-23 s 8 & & & 5 2 & @& » 3E-02 E-08 E-08 £ W OCOWMRMOCOOOOO0OOOCOOAMIODIO0OOTVOOOVOOOLOOLOOOO0D00L,=O0O e & & 8 ¢ o e & & S & - 0 s P s 0 = 3-41 Table 3.35 (Continued) AT 30 DAYS) 2.65E 15 N/CM*#%2-SEC 0.0 3E-03 1E~-18 000 ODOP,POH®DOODDOOO00 AT REFERENCE OXIDE LMFBR - WASTE DECAY TIMES (PROCESSED POWFR= 58,23 MW/MT, BURNUP= 32977. MWD/MT, FLUX= NUCLIDE THERMAL POWER, WATTS / METRIC TON FUEL CHARGED TO REACTOR CHARGE 30.D 365.D 1096.D 3652.D 36525.D 365250.0 CEl41 0.0 2.91E 03 2.25E 00 3.63E-07 0.0 0.0 PR143 0.0 1.40E 03 6.01E-05 5.20E-21 0.0 0.0 CEl144 0.0 1.125 03 4.95E 02 8.31E 01 1.62E5-01 0.0 PR144 0.0 9.50E 03 4.20E 03 7.05€E 02 1.38E 00 0.C ND147 0.0 5.18F 02 4.,26E-C7 0.0 0.0 0.0 PM147 0.0 1.82E 02 1.44E 02 8.46E 0Ol 1.33E 01 6.05E-10 PM148M 0.0 5.26F 02 2,09E 00 1.21E-05 0.0 0.0 PM148 (0.0 4.03E 01 1.08E-C1 6.28E-07 0.0 c.0 PM149 0.0 1.54E~-C1 0.0 0.0 0.0 c.0 SM151 0.0 8.18E 00 8.,12E 00 7.99€ 09 7.56E 00 3,69E 00 EVl52 0.0 1.286-01 1.21E~01 1.C8E~01 7.21E-02 3,98E-04 EUl54 0.0 9,14E 00 B,79E 00 8.06E 00 5.95E 00 1.21E-01 EU155 0.0 6.68E 01 4.70E 01 2.19E Ol 1.50E 00 1.62E-15 EUlS56 0.0 3.,23F 02 6.12E-05 1,.,31£-19 0.0 C.0 TBl60 O.C 7.96E 01 3,18E 00 2,82E-03 6.00E-14 0,0 T8161 0.0 1.48E OG0 3.59E-15 0.0 0.0 0.0 GD162 0.0 1.50E O1 7.96E 00 1.99E 00 1.55E-02 0.0 T8162M 0.0 2.96F 01 1.57E Ol 3.91E 00 3.06E-02 0.0 SUBTOT 0.0 B+.03E C4 1.38E 04 3.58BE 03 7.66E 02 8,72FE 01 TOTALS 0.0 8.03€ 04 1.38E C4 3.58BE 03 7.66F 02 8,72E 01 Al REFERENCE OXIDE LMFBR - WASTE DECAY TIMES Table 3.36. 3-42 Calculated Thermal Power of Fission Product Elements Present in Spent IMFBR (AI Reference Oxide) Fuel and in the Wasites Generated by the Reprocessing of This Fuel (PROCESSED AT 30 DAYS) 2.65E 15 N/CM**2-5EC 3652.0 36525.D 365250.D 0.0 0.0 G.0 1.49E-05 0.0 2+63E-02 7.18E-09 4.82E 00 2.11F 01 1.69E-04 2.52E-04 2.37E-02 1.94E-02 5.08E-11 0.0 1.75E 01 3.98 921 0.0 . (o] 5£-1C 9E 00 1E-01 DOOCDmWwrOOO * 8 8 2 o & o * @ COQONITOOCO E-03 E-18 SOO00=NOOOOO ® & & 4 ® 8 8 & 0 0 - OCOODPrOIIDO0O POWER= 58,23 MW/MT, BURNUP= 32977, MWD/MT, FLUX= ELEMENT THERMAL POWER, WATTS /7 METRIC TON FUEL CHARGED TO REACTOR CHARGE 30.D 365.D 1096.0 H 0.0 3.31E-02 3.15E-02 2.81E-02 1.90E-02 1.19E-04 0.0 IN 000 1-64E"05 000 0‘0 000 AS 0.0 1.24E-06 0.0 0.0 0.0 SE 0.0 1.49E~0C5 1.49E-05 1.49E-CG5 1.49E-05 BR 0.0 394DE"'°6 0.0 0.0 0.0 KR 0.0 1.648 01 1.55E 01 1.36E 01 B.66E 0O R8 0.0 4,81FE 00 1.90E-05 7.18BE-09 7.18E-09 SR 0.0 2.35E 03 B8.18E 01 5.27E 01 4.44E 01 Y 0.0 3.75E 03 3.11E 02 2.31FE C2 1.95E @2 ZR 0.0 1.10E 04 3.08E 02 1.27E-C1 1.69E-04 N8 0.0 1.28% 04 6.03F 02 2.48E-01 1.07E-04 MO 0.0 8.11E 00 0.0 2.0 0.0 TC 0.0 1.47E Q0 1.01E-02 1.01E-02 1.01E-02 RU 0,0 S«82E 03 5.,69E 01 1.02E 01 8.18E-02 RH 0.0 1.29% 04 6.62E 03 1.66E 03 1.33E 01 PD 0.0 2.86E-09 0.0 0.0 0.0 AG c.0 S« 72 01 1.08E 01 1.46E 00 1.33E-03 cD 0.0 1.27€ 00 1.64E-01 1.44E-01 1.02E-01 IN 0.0 9.71E-03 B8.2GE-05 3.25E-09 0.0 SN 0.0 4.31E 01 4.,26E-01 6.30E-02 5.27E-02 SB 0.0 8.58E 01 5.21E 01 3.12fE C1 5.19% 0O TE 0.0 9.26E 02 2.20E Ol 3.46E 0O 5.49E-01 1 0.0 6.40E 02 2.35E~-05 2.33E-05 2.35E-0Q5 XE C.0 8,63 01 2.37€E~C8 0.C C.0 cS 0.0 9.25E 02 3.96E 02 2.78F 02 1.51F D2 BA 0.0 2.16E 03 3,92E 02 3,74E 02 3.18E D2 CE 0.0 4,03E 03 4,97F 02 B8,31E 01 1.62E-~G1 PR 0.9 1.09E 04 4.20F 03 7.05E 02 1.38€E 00 ND 0.0 5¢1BE 02 4.26E~CT7 0.0 C.0 PM 0.0 T.49F 02 1.46F 02 8.46E 01 1.33c 01 SM .0 B.19E 00 B.12E 00 7.99E 00 7.56E 090 EU C.?D 3.99E 02 5.59% 01 3.0CE 01 7.52& 00 GD 0.0 1.505 01 7.96E CO 1.99F 00 1.55E-02 T8 0.0 1.11E C2 1.88E 01 3.,92F 00 3.06£-02 DY 0.0 5.19E-04 0,0 0.C C.0 HO 0.0 4.01€-03 0.0 0.0 G0 TOTALS 0.0 8.03E 04 1.38E 04 3,585 C3 T7.66F 02 8.72E 01 3.41E-02 3-L3 and thermal power, respectively, for individual isotopes; the same data, summed for each fission-product element, are given in Tables 3.32, 3.3, and 3.36 respectively. Actinides. - Tables 3.37 through 3.l2 present the calculated masses, radioactivity, and thermal power of important actinides in the fuel, com- piled for each isotope and summed for each element; Tables 3.42 through 3.48 present comparable data for the waste. In making the calculations for the waste, we assumed that reprocessing occurs 30 days after the fuel is discharged from the reactor and that 0.5% of the uranium and 0.5% of the plutonium in the spent fuel appear in the waste. Cladding. - Tables 3.49 through 3.5L present the calculated masses, radioactivity, and thermal power of neutron-induced activation products of the oxygen, stainless steel, and sodium in the cladding of the mixed core and blankets., These data include only the cladding in the zones exposed to neutrons, and assume a 0,00l-in,-thick layer of sodium at the fuel-cladding interface. In addition to the neutron-induced 1sotopes calculated here, ORNL hot-cell experience indicates that about 0.03% of the plutonium in the core and blankets may be associated with the cladding.7 3.L4. Shipments of Spent Fuel Spent nuclear fuel will continue to be shipped, as at present, from reactors to a processing plant in large shielded casks. Cask sizes will tend to become larger to permmit higher payload ratios and to minimize the shipping cost. Sizes in the 100- to 120-ton range are anticipated. Most shipments will be carried by rail; barges and trucks will be used to a lesser extent. Although transportation by barge is economical, water routes between reactor and processing plant sites are not always available. Shipments by truck are relatively expensive because the cask weight in this case is limited to about 30 tons. The anticipated growth of the shipping industry from 1970 to 2020 is indicated in Table 3.55. The figures are based on Phase 3, Case L2 of the SATF study, which assumes an LWR-LMFBR nuclear economy, and are, POWER= Table 3.37 58.23 MW/MT, 3-Lk Spent IMFBR (AL Reference Oxide) Fuel Al REFERENCE OXIDE LMFBR - FUEL DECAY TIMES BURNUP= 32977. NUCLIDE CONCENTRATIONS, GRAMS / METRIC TON TH228 TH229 TH230 TH231 TH232 TH233 PA231 PA232 PA233 PA234M PA234 U232 U233 U234 U235 U236 U237 U238 U239 NP236 NP237 NP238 NP239 PU236 PU238 PU239 PU240 PU241 PUR242 PU243 AM241 AM242M AM242 AMZ243 AM244 CM242 CM243 CM244 susTOY TOTALS CHARGE *« & & @ ¢ & & » @ . 05 Q m CO00OO0OVOOMODOO00000OODO0000O . COOO0OOONODOPHPOOO0OCOOCOOO0O0O0OO0O L . + Q m o N O - P b w0 mmm Q H 3.17E 03 OO0 O0ODOOO * ¢ o % o 8 v @ COO0O0OO0OO0QOOoO0 1.00E Cé6 1.00E 06 Table 3.38. IMFBR (AX DISCRARGE 4 915'07 1.17€-08 1. 71E-05 4.99E-09 IOOGE‘Oa 1. 76E~-12 5.89E-07 4.11E-10 2.81E~Q9 2.T6E~16 9. 12E"1‘" le24E-04 8.048-04 8.04E 00 l.42E 03 3.75t 01 2.28F 00 B.77E Q5 1.48E 00 1.24E 02 9.06E-02 2.13E 02 T7+84E-04 6.62E 02 5.73E 06 1.93E 04 5.28E 03 3.26E 03 1 0685‘01 4,61E 02 8.92E 00 1.76E~-01 2.57E 02 1 . 78E‘03 2.23E 01 8.42E-01 1.53E 01 9.66E 05 9.66E 05 30.0 5 . 7BE"0 7 1.20€E-08 1.89E-05 5. T4E~11 1.16E-06 0.0 5.94E-07 5.92E-17 1.32E-09 0.0 0.0 1 .‘I‘OE‘O‘Q B8.04E-04 8.46E 00 1.42E 03 3.77E 01 1.05€E-01 8.77€E 05 0.0 T.05€8-16 1.26E 02 4 DH4E~06 3.10E-02 T« T2E-04 6.64E 02 5.75E 04 1.93E 04 5.25E 03 3,26E Q3 0.0 4.84E 02 8.91E 00 1.0TE~04% 2.58E G2 0.0 1.98€ 01 8.40E-01 1.53€ 01 9.66E 05 9.66F 05 60.D S 1’35-07 1.22E-08 2.09E-05 5.7T4E-11 1,25E-06 0.0 5.95E-07 0.0 6.16E-10 0.0 0.0 1 - 55E-04 8.04E-04 8.88E 00 1.42E 03 3.79E 01 4.81E-03 B.77TE 0S5 0.C 0.0 1.27E 02 2432E~10 20 1 8E'04 T.56E-04 6.,66E 02 5.75E 0% 1.93€E 04 5.23E 03 3.256E 03 0.0 5.C7TE 02 8.91E 0D 1.07E~04 2.58E 02 0.0 1.74E 01 8.395-01 1.53E 01 9.66E 05 9.66E 05 MWD/ MT, FLUX= Masses of Actinide Isotopes Calculated to Be Present in 2.65E 15 N/CM**2-SEC FUEL CHARGED TO REACTOR 90.D T.7T8E-07 1.25E-08 2.29 E-OS S.T4E-11 1.34E~-05 3.80FE 01 2.21E"04 B.77TE 05 0.0 0.0 1.,27E 02 1.19E-14 2.13E-04 7.41E-04 6.68E 02 5.75E 04 1.93E 04 5.21E 03 3.26E 03 0.0 5.29E 02 8.91E 00 1.07E-04 2.58BE 02 0.0 1.53€ 01 B«+37E~01 1.52€ 01 Q9.66E 05 9.66E 05 150.0 1,01E-00 1.,31E-08 2+73E-05 5.7T4E-11 1 ® 52 E-08 0.0 5.99E~07 0.0 6.,33E-11 0.0 0.0 1.97TE-04 8.04E-04 1.01E 01 1.42E 03 3.83E 01 4 4+65E~07 8.7T7TE 05 0.0 0.0 1.27E 02 2.98BE-23 2.13FE-04 T.12E-04 6.T1E 02 5.T5E 04 1.93E 04 5.16F 03 3.26E 03 0.0 5.75E 02 8.90E 00 1.07E-04 2.58E 02 0.0 1.19€ 01} 8..34E-01 1.51E 01 9.66E 05 9.66E 05 365.D 2-05E"06 4.63E-05 5.7T5E-11 2.17E-06 B.77E 05 ¢.0 0.0 1.27€ 02 0.0 2.13E-04 6.1TE~C4 6.T4E 02 5.75E 04 1.93E 04 5.00€ 03 3,26E 03 0.0 T.34E 02 8.88E 00 1.07E-04 2.58E 02 0.0 4.78E OC B.24E-01 1.48E 01 9.66E 05 9.66E 05 Magses of Actinide Elements Calculated to Be Present in Reference Oxide) Fuel and in the Wastes Generated by the Reprocessing of This Fuel Al REFERENCE OXIDE LMFBR - FUEL DECAY TIMES 2.65E 15 N/CM**2-SEC FUEL CHARGED TO REACTOR POWER= 58.23 MW/MT, BURNUP= 32977, MWD/MT, FLUX= ELEMENT CONCENTRATIONS, GRAMS / METRIC TON CHARGE DISCHARGE 30.D 60.D 90,.D TH 0.0 1.86E-05 2.07E-05 2.28E-05 2.51E-05 PA 0.0 5.92E-07 5.96E-07 5.96E-07 5.97E-07 u 9.21E 05 B8.79E 05 B8.79E 05 8.,79E 05 8,.79E 05 NP 0.0 3.38€ 02 1.26E 02 1.27TE 02 1.27E 02 PU 7.93E 04 8.58E 04 8.60E 04 B8.60E D4 8.60E 04 AM 0.0 T2TE 02 7.50E 02 7.73E 02 7.96E 02 CM™ 0.0 3.85F 01 3.59t 01 3.35E 01 3.14E 0Ol TOTALS 1.00E 06 9.66F 05 9.66E 05 9.66F G5 9,.66E 05 15C.D 2.99E-05 5.99E-07 8.79% 05 1.27€ Q2 8.59€ 04 8.41E 02 2,78 01 9.66E 05 365.0 5.05E-05 6.07TE-07 8.79E 0% 1.27€ G2 8.58F 04 . 1.00F 03 2.04E 01 9.656E 05 Table 3.39. 3-L5 Calculated Radicactivity of Actinide Isotopes Present in Spent IMFBR (AI Reference Oxide) Fuel Al REFERENCE OXIDE LMFBR - FUEL DECAY TIMES POWER= 58.23 MW/MT, BURNUP= 32977, NUCLIDE RADIOACTIVITY, CURIES /7 METRIC TON TH228 TH229 TH230 TH231 TH232 TH233 PA23] PA232 PA233 PA234M PA234 U232 U233 U234 U235 U236 v237 U238 U239 NP236 NP237 NP238 NP239 PU236 PuU238 PU239 pU240 PU241 PUZ42 PU243 AM241] AM242M AM242 AM243 AM244 CM242 CM243 CM244 SUBTOT TOTALS CHARGE . MODDOOOOODOFOOOO000O000LOODO 2E-05 6E-01 04 ® & 9 & 8 ¢ 2 ° & B e . = (D » @© mm 03 06 0l = OO0OO0OOLOQOHMHFIAPNFIOOODOODOWOOWOOODOODO0OOODOOODDOO . o x® m * ¢ ® & 4 9 & o 2w @ 06 1.10E 06 Table 3.40. DISCHARGE 4903E'04 2.50E~-09 3.31E-07 2.65E-03 1.18E-13 6.45E-05 2.80E-08 1.75E-04 5.715E-05 1.90E-07 1081E‘07 2.665‘03 T.62E-06 4098&”02 3.04E-05 2.38E~03 1.B86E 05 2.92£-01 4,97 07 3.03€ €O 8.75E-02 2.37E 04 4.96E 07 1,12F 3,52E 4,26F 6.,02F 1.27E 4436E 1.49E B.6TE l.42€ 4,95E 5.28€ T+.40E 3.87E l.24F 1,01E 1,01 08 30.D 4, T4E-04% 2.56E-09 3,6BE-C7 3.04E~05 1.27€-13 0.0 2.83E-08 2.52E-11 2.69E-05 0.0 0.0 2.99E-03 T.62E-06 5¢23E-02 3.04E-05 2.39E-03 8.54E 03 2.92E-01 0.0 4.26E~-10 8.91E-02 1.19€ €O 7.22E 03 4.10E-01 1.12E 3.53E 4.26E 6.00F CF 1.27€ 0.0 1.57E 8.6TE B.6TE 4 .S6E c.0 6.55E 3.86E 1.24E T.03E 7.03E 05 60.D 5.53E-04 2.62E-09 4.06E-07 3.,04E-05 1.37E-13 2¢40E'03 3.92E 02 2.925*01 C.0 0.0 8.92E-02 6.07E~-QS 5.06E 01 4.02€-01 1.13E 04 3.53E 03 4,26E 03 5.97E (5 1.27€ 01 0.0 1.64E 8.66E 8,64 4.96E 0.0 5.77E 3.86E 1l.24E 6. T7E 03 01 01 01 6. T7TE 05 MWD /NT, FLUX= 2.65E 15 N/CM**2-SEC FUEL CHARGED TO REACTOR 90.D 60395‘04 2:68E-09 4 46E-07 3.05€E-05 1.47E-13 0.0 2084E‘08 0.0 5.90E-06 0.0 0.0 3.,62E-03 T«.62E-06 5. T5E-02 3,05E-05 2.41E~03 1.80E 01 2.92E-01 0.0 0.0 8.93E-02 3.11E-09 4,96E C1 3.945‘01 1.13E 04 3.53E 03 4,26E 03 5,95E 05 1,27E 01 .0 1.72E 8.66E B.66E 4 .96E 0.0 5.08E 3.85E 1.23F 6.,68E 03 01 01 01 6.68E 05 150.D 8.,29E~04 2+.80E-09 5.32E-07 3.05E-05 1.66E~-13 0.0 20855-08 0.0 1.295-06 0.0 0.0 4.,23E-03 TH2E-06 6.27E-02 3.C5E-05 2.43E-03 3.80E-02 2.92E-01 0.0 0.0 B8.94F-02 T.79E-18 4.,96F 01 3079E-01 1.13E 04 3.53E 03 4.26E 03 5.89E 05 1.27E 01 0.0 1.86E 8.65E B.45E 4 .96F 0.0 3.94E 3,84% 1.22E 6451F 03 o1 o1 01 6.51E 365.D 1.68E-03 30225-09 8.99E-07 3.05E-05 2+38E-13 0.0 20895-08 0.0 5.626'09 0.0 0.0 6.20E-03 T.62E-06 8.15E~02 3.05E-05 2.50E-03 9.80E-12 2.92E-01 c.0 0.0 8.98E-02 ¢.0 4.96E 01 3.288-01 l.14E 04 3.53E 03 4.26E 03 5.7T1E G5 1.27€ 01 0.0 2.38E 8.63E 8.63E 4 .96E 0.0 1.58E 3.79F 1.20E 6.10F 03 01 c1 01 6.10E Calcul ated Radioactivity of Actinide Elements Present in Spent IMFBR (AI Reference Oxide) Fuel Al REFERENCE OXIDE LMFBR - FUEL DECAY TIMES POWER= 58,23 MW/MT, BURNUP= 32977, ELEMENT RADICACTIVITY, CURIES / METRIC TON TH PA CHARGE DISCHARGE 30125‘03 2.33E-04 4,99F 07 4,97 C7 1.06E 06 1.97¢ €5 7.52F 04 TOTALS 1,1CE C6 1.01% ©8 36.D 5.05E~-0¢4 2.705“05 B.54E 03 T.22E 03 6.19E 05 1.79€ 03 6.68E 04 T.03E 05 6C.D 5.34E-04 1.26E-05 3,93E 02 5.07€ 01 6.,16F 05 1.86E 03 5.90E 04 6., TTE 05 MWD/ MT, FLUX= 2.65E 15 N/CM*%2-SEC FUEL CHARGED TO REACTOR 90.D 6.70E-04 5.93E-06 1.84F (1 4.,97E 0Ol 6.14E 05 1.94E 03 5.21E 04 6.,68E 05 150.0 8.60E-04 1'32E-O6 4.00E‘01 4,97 01 6.0BE 05 2.08E 03 4.06E 04 6.51t 05 345.0 1.71E-03 3.#5E‘08 3.83E-0C1 4,96E 01 5.50E 05 2.6CE 03 1.7T1E 04 6.10E 05 POWER= Table 3.41. 56423 MW/MT, NUCLIDE THERMAL TH228 TH229 THZ230 TH231 TH232 TH233 PA231 PA232 PA233 PAZ234M PA234 U232 U233 U234 uz23s uz23é U237 U238 uz239 NP23s NP237 NP238 NP239 PU236 PuU23s8 PU2393 PU240 PU241 PU242 PU243 AM241 AM242M AM242 AM243 AM244 CM242 LM243 CM244 SUBTOT TOTALS CHARGE 7E=-07 OO TPOO0OOOOOOOOCOOCO0 75£=-03 * & & % & & & ¥ ¥ & & 5 4 S B & 0 s & s @ 2 QOOCOO0~NOOWOCOOLOOOOOOLOOOCOO OO0 O0OQOO0 Table 3.42. 3-L6 Calculated Thermal Power of Actinide Isotopes Present in Spent IMFBR (AI Reference Oxide) Fuel Al REFERENCE NOXIDE LMFBR - FUEL CECAY TIMES BURNUP= 32877. MWD/MT, FLUX= 2+.65E 15 N/CM*%x2-5EC POWER, WATTS / METRIC TON FUEL CHARGED TO REACTOR DISCHARGE 1.326-08% Te56FE=11 9.37E-06 3. 6£5-06 24 345 -15 1-@1E'07 Be26E-10 5.825'07 1.46E-07 QCTSE-IO l. 33E-09 8.55F=05 2.22c-07 l.43E~0C3 5.455-07 b.45E=05 3.96F 02 Te39E-03 1.18t 05 g.54c=-03 0.0 1.228 02 1.47E 0% le48E=02 3., 70E 02 1.09E 02 1.33E 02 2.50E 01 30765-01 €.21E Q2 4.982 01 247E-02 1.90E Q2 1.6GE 00 l.68E 02 2.73t C3 le41E 00 4,35E (1 2. T0E 05 270z 05 30.0 IQSEE-CS Te74t=-11 1.04E‘08 Loa20E-Q8 3.08fF=-15 0.0 8.64E=-10 B.38E-14 be84E-C3 2.0 0.0 3.60E-05 2.22E=-07 1-51E-03 Be4S5E~-CT 6.,48E-05 1.63F (C1 T.29E-Q03 0.0 1l.20E-12 0.0 6.13E-03 2a14E 01 l.43E-C2 3.72E C2 1108 C2 l.33F C2 2.49E 01 3.76E-01 0.0 5.23E 01 2447E-02 lel6E-CL 1.€0£ 0OC 0.0 2.42E 03 1.41F 0O 4,34 01 3.19F C3 3.19E 03 60.0D 1.81E-05 Te92E-11 1.15E-08 4.21E-C8 3.31E-15 0.0 8.66E~-10 0.0 3.,20E-08 0.0 0. 0 1. 06E-04 2+,22E~07 1-58E-03 8.45E“O7 €« 51E=05 8042E-01 Te33E-03 0.0 0.0 0.0 3.14E-07 1.50E-01 1,40E-02 3. 73 02 l1.,10E Q2 l.33E 02 2.48E 01 3.765-01 0.0 5.48E 01 20475'02 ls 16E-01 l.60E Q0 0.0 2e 13E l.41E 4e32E 2.8T7E 2.8TE 90.0 2.09E-0% 3.10E-11 l.25E~-C8 L,21E-C8 3.55E-15 1.60E~11 l1.47E-C1 1.37E-02 3.74E 1.10E 1.33E 2.47E 3.76E~01 0.0 5.72E C1 2446E-0Q2 l.16E-01 l.60E 00 0.0 1.87E le41lE 4.31E 2.62FE 2.62E 150.0 24 72E~05 Be46E-11 1.,50E-08 4.21E~-08 4,02E-15 0.0 3.71E~10 0.0 3.,28E-09 0.0 0.0 le 36E-04 2.22E=-07 l.31£~03 Ba45E=-07 6.595-05 8.15E~-05 7e39E-03 0.0 G.0C 0.0 4.02E-20 1047E-01 1l.32E-02 3.75E @2 1.10E 02 1.33E 02 2+45E 01 3.76E-01 0.0 6.21FE 01 2e46E~-02 1.15E-01 1,60 €O .0 l1.45E 032 1.40E QO 4.28E 01 2.20E 03 2.20E 03 365.D 5.51E=-05 GeT4E-11 2.54E-08 Calculated Thermal Power of Actinide Elements Present in Spent IMFBR (AI Reference Oxide) Fuel Al REFERENCE OXIDE LMFBR - FUEL LCECAY TIMES POWER= 58,23 MW/MT, BURNUP= 32977, MWD /MT, FLUX= 2,65F 15 N/CM**2-SEC ELEMENT THERMAL POWER, WATTS / METRIC TON FUEL CHARGED TO REACTOR TH PA TOTALS 7.89E 02 2.70E CHARGE 0.0 0.0 T.75E-03 0.0 T+.89E 02 0.0 0.0 DISCHARGE 1.70E-05 T+ 31E-07 1.18¢E le47E 1.26E 4, 09E 2. TTE 30.C 1.5¢E-05 6.52E-08 1.83¢ 2.14E 6639E 5.40¢ 2¢46E 3.1GE 60.0 1.82E-05 3.29E-08 8451E-Cl 1.50E-01 6.40E 02 5.65E 0l 2.1TE 03 2.87E 03 S0.D 2.10E-05 1.59E~C8 4o T9E-C2 1.47E-01 6es41lE 02 5.90E 01 l.92E 03 2.62E 03 150.D 2. T2E~-05 4416E-09 9.48E-03 1.47E-01 be42E 02 6+3%9E Ol 1.50E 03 2.20E 03 365.D 5.52E-05 8.97E-10 1.00€-02 l1.47E-01 6.43E Q2 8.11E 01 6.26E 02 1.35E 03 A1 REFERENCE OXIDE LMFBR - WASTE DECAY TIMES 58,23 MW/MT, BURNUP= POWER= Table 3.43. 3-L7 Masses of Actinide Isotopes Calculated to Be Present in Wastes Generated by the Reprocessing of Spent IMFBR (AL Reference Oxide) Fuel 32977, MWD/ MT, FLUX= (PROCESSED AT 30 DAYS) 2.65E 15 N/CM**2-SE( NUCLIDE CONCENTRATIONS, GRAMS / METRIC TON FUEL CHARGED TO REACTOR CHARGE TH228 TH22% TH230 TH231 TH232 TH233 PA231 PA232 PA233 PA234M PA234 U232 U233 U234 uz23s Uy236 U237 uz23s U239 NP 236 NP237 NP238 NP239 PU236 PU238 PU239 PU240 PU241 PU242 PU243 AM24] AMZ242M AM242 AM243 AM244 CM242 CM243 CM244 SUBTOY TOTALS Al POWER= o m QOOVDO0OOVOOODOOOOODOOQODOO o m s & ¢ 8 & 0 & * b " ¥ B s 6 4 " 0 s O s e e COO0OO0OOONDCOCPOOOOODODOODOODOOOOC Ot 1.00E 03 05 02 04 04 03 03 06 06 30.D 5.78£-07 1.20E-08 1.89€~05 5.74E-11 1.16E-06 0.0 5.94E-0T 5.92€-17 1.,32E-09 0.0 0.0 6.98E-07 4-025-06 4.23E-02 7.10E 00 1.88E-01 5.23E-04 4.39€ 03 0.0 7.05E-16 l.26E 02 4.54E-06 3.108-02 3.32F 00 2.88E 02 3.65E 01 2.63E 01 1.63E 01 0.0 4.84F Q2 8.91€ 00 1.07E-04 2.58E 02 0.0 1.98€ 01 8.40E-01 1.53F 01 S.T4E 03 5.74E 03 365.D 4.226-07 1.2CE-Q8 1.91€-05 2.87E-13 1.17E-06 2.0 5.955‘07 0.0 2.7T5E-13 0.0 0.0 1045E-°6 4.,02E-06 1.,29E-01 T.11E 00 1.,97E~01 6.00E-19 4.39E (3 0.0 0.0 1.27€ 02 0.0 2.13E-04 3.09E-CH 1.8CE 01 2.88E 02 9.70E 01 2.50E 01 1.63€ 01 0.0 4.84E 02 8.88E 00 10075‘04 2.58E 02 0.0 4.78E 00 8.24E-01 1.48E 01 S5.T4E 03 5.74E 03 1096.D 2.335-07 1.20E-08 2.07E-05 2.88E~13 1.18E-06 4,02E~-06 4.53E-01 T.13E 09 2.17E-01 C.0 4.39¢ 03 0.0 0.0 1.28E 02 0.0 2.13E-04% 1.90E-056 2.22E 01 2.88E 02 9.81E 01 2.25€ 01 1.63E C1 0.0 4.85E 02 8.80F 00 1.06E-04 2.58E Q2 0.0 2+34E-01 7.89E-01 1.37€E 01 5.74E 03 5.74E 03 3652.D 1.06E-07 1.21E-C8 4009E-05 2.90E'13 1.23E-06 C.0 5.95E-07 4.02FE-06 1.63E 00 T.18E 00 2.87TE-01 0.0 4.39E 03 4E 02 QO OO0 CWwWwoo 2.13E-04 3.46E-07 2.15€ 01 2.88E 02 1.01E G2 1.55€ 01 l.64E 01 .0 4.87E 02 8.52E 00 1.02E-04 2.57E 02 0.0 2.06E-02 6.78E-01 1.05E 01 5.74t 03 5.T4E 03 36525.0 4.76E-08 10355-08 1.98E-03 3.2CE-13 3.,21E-06 0.0 6.,01E~07 OO0 0 - » * * OO0 1.78E-06 4.02E-06 1.29E 01 7.91t 00 1.27E 00O 0.0 4.39E 03 0.0 0.0 1.96E 02 0.0 2.11E‘04 lcOBE-lfi 1.23€E 01 2.90E Q2 1.10F 02 1.28€-01 1.69F 01 C.0 4.39F 02 5.65E 00 6.79E~-05 2.55E 02 0.0 1.36E-02 9.65E~-02 3.33E-01 S«7T4E 03 S.74E 03 365250.D B.27E~-12 2.70E-08 €.64E~-C2 6.21E-13 1.61E-04 0.0 9E-07 2.35E 02 0.0 2.25E-04 3029E'10 3.,60E~-16 S.74E 03 5.74E 03 Table 3.hli. Masses of Actinide Elements Calculated to Be Present in Wastes Generated by the Reprocessing of Spent IMFBR (AI Reference Oxide) Fuel (PROCESSED AT 30 DAYS) 32977, REFERENCE OXIDE LMFBR - WASTE DECAY TIMES 58,23 MW/MT, BURNUP= ELEMENT CONCENTRATIONS, GRAMS / METRIC TON CHARGE ™ PA U NP PU AM cM 0.0 0.0 30.D 2.075-05 5.96E~07 9.21E 05 4.39E 03 0.0 1.26E 02 T<.93E 04 4.30E 02 0.0 0.0 T.50€ 02 3.59E 01 TOTALS 1.00E 06 S5.74E 03 365.0 2.07€-05 5.95E-°7 4.39E 03 1.27E 02 4.44E 02 T.51E 02 2.04FE 01 5.74E 03 1096.D0 2.21E-05 5.95E-07 4.39E 03 1.28E 02 %4.47TE 02 T.52E 02 l.47E 01 S«T4E 03 MWD /MT, FLUX= 2.65E 15 N/CM*%2-SEC FUEL CHARGED TO REACTOR 3652.0 4.22E-05 5.95E-07 4.40E 03 1.34E 02 4.43E 02 T.53E 02 1.12E 01 5.T4E 03 36525.D 1.98E~03 6.01E-07 4.41E 03 1.96E 02 4,29E 02 7.00E 02 4.43E-01 5.74E 03 365250.D 6.655'02 6.89€-07 4.44E 03 5.17E 02 4,21F 02 3.48E 02 2.25E-04 5.74E 03 Table 3.45. Al REFERENCE OXIDE LMFBR - WASTE DECAY TIMES Table 3.46. Al REFERENCE OXIDE LMFBR - WASTE DECAY TIMES 3-L8 Calculated Radiocactivity of Actinide Isotopes Present in Wastes Generated by the Reprocessing of Spent LMFBR (AL Reference Oxide) Fuel (PROCESSED AT 30 DAYS) FLUX= 2,65E 15 N/CM*%2-Sg( FUEL CHARGED TO REACTOR POWER= 58,23 MW/MT, BURNUP= 32977. MWD/MT, NUCLIDE RADIOACTIVITY, CURIES / METRIC TON CHARGE 30.0 365.0 1096.D 3652,0 TH228 0.0 4, T4E-04 3.4TE-04 1.91E-04 B.T1E-0O5 TH229 0.0 2.56E-09 2.56E~09 2.57E-09 2.59E-09 TH230 0.0 3.68E-07 3,71E-Q7 4.02E-07 T7.94E-07 TH231 0.0 3.04E-05 1.52E-07 1.53E~07 1.54E~07 TH232 0.0 1.27E-13 1.28E-13 1.29E-13 1.35E-13 TH233 0.0 0.0 0.0 0.0 0.0 PA231 0.0 2.83E-08 2.83E-08 2.83E-08 2.84E-08 PA232 0.0 2.52E-11 0.0 0.0 0.0 PA233 0.0 2.69E~-05 5.,62E-09 5.23E-17 0.0 PA234M 0.0 0.0 0.0 0.0 6.0 PA234 0.0 0.0 0.0 0.0 G.0 U232 0.0 1.50E-05 3.10E-05 5.52E~05 8.29E-05 u233 0,0 3.81E~-08 3.81lE-08 3.B1lE-08 3,.81E-08 U234 0.0 2.62E-04 B.OCE-04 2,80E-03 1.01E-02 U235 3,12E-05 1.52E-07 1.52E-07 1.53E-07 1.54E-07 U236 0.0 1,20E-05 1.25E-05 1.38E-05 1.82E-05 U237 0.0 4.27E 01 4.90E~-14 0.0 0.0 U238 3.06E-0l 1.46E-03 1.46E-03 1.46FE-03 1.46E-03 U239 0.0 0.0 0.0 0.0 0.0 NP236 0.0 4426E~-10 0.0 0.0 0.0 NP237 0.0 8.91E~-02 B.96E-02 9.,06E-02 9.42E-02 NP238 0.0 1.19E 00 0.0 0.0 0.0 NP239 0.0 T«22€ 03 4.96E 01 4.95E 01 4.95Z 01l PU236 0.0 2.05E-03 1.64E-03 1.01E-03 1.84E-04 PU238 1,59E 04 5.61E 01 3,04E 02 3.75E 02 3,62E 02 PU239 2.88E 03 1.76E 01 1.77E 01 1.77E 01 1.77E 0Ol PU24C 4.14E 03 2.13FE 01 2.14E 01 2.16E Ol 2.23E 01 - PU241 1.08E 06 3.00E 03 2.86F 03 2.57€ 03 1.77E 03 PU242 1.24E 01 6.36E-02 6.36E-02 6.37E-02 6.39E-02 PU243 0.0 0.0 0.0 0.0 0.0 AM24]1 0.0 1.57€ 03 1.57€ 03 1.57¢ 03 1.58E 03 AM242M 0.0 B8.67E 01 8.63E 01 B.55F 01 8.,28E 01l AM242 0.0 8.6TF 01l 8.63E 01 8.55E 01 8.28E 01 AM243 0.0 4.96E 01 4.96E 01 4.95F 01 4.95E 01 AM244 0.0 0.0 0.C 0.0 0.0 CM242 0.0 6.55E 04 1.58F 04 T.74FE 02 6.82E 01 CM243 0.0 3.86E Cl 3.79E 01 3.63t 01 3.12E 01 CM244 0.0 1.24E 03 1.20F 03 1.11E 03 B.48E €2 SUBTOT 1.10E 06 T7.90E 04 2.21E 04 6.74E 03 4,.96E 03 TOTALS 1.10E 06 T.90F 04 2.21E C4 6.T4F 03 4.96E 03 36525.0 3.91E-05 2.89E-09 3.85E-05 1.70E-07 3.51E-13 0.0 2.86E-08 0.0 0.0 0.0 0.0 3.80E-05 3.81E-08 8.01E-02 1.70E-07 8.04E-05 0.0 1.46E-03 0.0 0.0 1.38E-01 0.0 4.,91F 01 5.7T3E~-14 2.07E 02 1.78E 01 2.43E 01 1.46F 01 6.59E-02 0.0 1.42E 03 5.50E 01 5.50E 01 4.,91E 01 0.0 4.51E 01 4.44E Q0 2.T0E 01 1.97F 03 1.97€ 03 365250.D 60805-09 5.77E-09 10295-03 3.,29E-07 1.76E~-11 3.795-08 1.81E-01 3.29E-C7 6.88BE-04 0-0 1.46E-03 ZOIOE-ZO 6.,97TE-02 0.0 3.64E 02 9-07E’01 9.07€-Q1 4.53F 01 0.0 7.44E‘01 1.52E-C8 2.916-14 5.01E 02 5.01E 02 Calculated Radioactivity of Actinide Elements Present in Wastes Generated by the Reprocessing of Spent IMFBR (AT Reference Oxide) Fuel POWER= 58,23 MW/MT, BURNUP= 32977. ELEMENT RADIOACTIVITY, CURIES / METRIC TON CHARGE 30.D 365.D 1096.D TH 0.0 5.05E-04 3,4TE-C4 1.92E-04 PA 0.0 2.T0E-05 3.40E-08 2.83E-08 U 3.06E-01 4.27E 01 2.30E-03 4.33E-03 NP .0 T+22E 03 4.96FE 01 4.96E 01 PU 1.10E 06 3,09E 03 3.20E 03 2.98F 03 AM c.0 1.79E 03 1.79E 03 1.79E 03 CM™ 0.0 6.68F 04 1.71E 04 1.92E 03 TOTALS 1.10E 06 7.90E 04 2.21E 04 6.74E 03 MWD/MT, FLUX= (PROCESSED AY 30 DAYS) 2:65E 15 N/CM*%2-SEC FUEL CHARGED TO REACTOR 3652.0 8.81E-05 2.84E-08 1.16E-02 4.96E 01 2.17€ 03 1.79E 03 9.47TE 02 4.96E 03 36525.D 7.77E-05 2.865‘08 8.16E~02 4.93€ 01 2.64F 02 1.58E 03 T7.65E 01 1.97E 03 365250.0 1.29E-02 3.28E-08 1.83E-01 4.56E Cl 4,27t 01 4.11E 02 T«44E-Cl 5.01E 02 Al REFERENCE OXIDE LMFBR - WASTE LCECAY TIMES 58.23 MW/MT, BURNUP= POWER= Table 3.L7. NUCLIDE THERMAL TH228 TH229 TH230 TH231 TH232 TH233 PA231 PA232 PA233 PA234M PAZ34 U232 U233 U234 y23s U236 V237 U238 U233 NP236 NP237 NP238 NP239 PU236 PU238 PU239 PU240 PU241 PU242 PU243 AM24 1 AM242M AM242 AM243 AM244 CM242 LM243 CM244% SUBTOT TOTALS Al REFERENCE OXIDE LMFBR - WASTE DECAY TIMES 58,23 MW/MT, POWER= CHARGE 7E-07 5E-03 MOOOoODOQQO~-1 QO WO OO0 OoOOCOoOOOO000 COOOOO~NCOCOOOTOOOO0OOO0OODOO0O0OO0C0O & # & & & 9 & B * ¢ " 4 & 8 8 6 & b » b s 8 s 2 3.65E-01 0«0 0.0 «0 0 0 3.0 0.0 0.0 0.0 040 T.89E 02 T+83E 02 Table 3.48. 3-L9 Calculated Thermal Power of Actinide Isotopes Present in Wastes Generated by the Reprocessing of Spent IMFBR (AT Reference Oxide) Fuel (PROCESSED AT 20 DAYS) 32977. MWD/MT, FLUX= 2,65E 15 N/CM*%2-SE(C POWER,y WATTS /7 METRIC TON FUEL CHARGED 7O REACTOR 3C.D l.556=05 70745‘11 le O4E-08 4e 2CE-08 3.08E-15 0.0 Be £4E~-10 8s 38E~-14 e 84E-08 0.0 0.0 4. 30E-07 1, 11E-09% Te53c=0¢€ 4e 22£-06 3. 24E-07 9.17£-02 3. 70E-05 0.0 le 2CE=12 0.0 bel32E~03 2+.14F 01 T« L4E~-05 1.86% 00 5.48E~01 6. 63E-01 1,24E-01 1.88E£-03 0.0 £.23t 01 20“75—02 1,16£-C1 l.6CE QO 0.0 2.%2E 03 l1.,«1% 00O 4,34F 01 254 03 2054 C3 36540 lel4E~C3 7075E*11 1.05E~C8 2+10€E~10 3.09c~15 0.0 8065&”10 0.0 le43E-11 0.0 0.0 9.C4E-C7 1.11E-09 230E-05 4.,23E-09 3.3%E-07 1.05E~16 3.70E-0% OOOO 0. O. 0. O 1l.47E-01 5.71E-0% 1.01F C1 5.48t-01 6067&'01 1.188-C1 10885“03 0.0 5248 Gl 2.46E~02 l.15E~01 1.608 00 0.0 5.83F C2 1.38E 0O 4.19F (1 b.52E C2 £.G2E (2 1096.D 6.27E-06 T« 77E~11 le14E-08 2.11&-10 3.12E~15 0.0 Be 65E-10 0.0 1.33E-19 0.0 0.0 1. 77E-06 l1.11E-09 8. 07E-05 4424E-09 3. T3E-C7 0.0 3.70E-05 0.0 0.0 0.0 0.0 1.47E£-01 3.51€E-05 l.24& 01 5.42E-01 6. 74E-01 1.06E-01 l. £8E-0C3 0.0 £.25E Cl Ze43E-02 lel4E~0Ol 1.60& 00 0.0 2.85E 01 1.32€E ©O 3.88E 0Ol l.37E 02 1.37E 02 3652.0 2.85E=06 7.84E-11 2+25E~C8 2.13E-10 3-265-15 0.0 8.65E-1C 0.0 0.0 0.0 0.0 2+.66E-06 l1s11E-09 2+30E-04 4,27E-03 4494E-07 0.0 3.70E~C5 l.4TE-01 6+ 40E~06 1.20E 01 5.49E-01 6.95E~01 T«33E~C2 1,83E~C3 0.0 527TE €1 2436E~02 1.10&‘01 l.60E 0O 0.0 2.51% 00 1.14E GO 2.97€ 01 1,01 C2 1.0l C2 36525.D 1.,28E-06 8.75E-11 1. 09E-06 2e34€£~10 8,50E~15 0.0 8+.73E-10 0.0 0.0 0.0 0.0 1.22E~06 1.11E~-09 2+ 30E~-03 4,70E~09 2.1BE=06 0.0 BOTOE-OE Qe O 0.0 0.C 0.0 10465‘01 1,99€~-]15 6.86E 00 5.53E-01 7.57E‘01 54 04E=-04 1-95E-03 0.0 4+ T4E 01 1,5¢E-02 7+33E-02 1.58E 00 0.0 1«69E 00O lléZE-Ol Q. 45E-01 be 02E Ol £+ 02E 01 365250.D 2.23E-10 1.74E-10 3.65€-05 4,55E-10 4+25E-13 OE-09 1.21£-03 l.%6E QO 0.0 2-74E‘02 5«33E-10 1.02E-15 1.51F 01 1.51F 01 Calculated Thermal Power of Actinide Elements Present in Wastes Generated by the Reprocessing of Spent IMFBR (AT Reference Oxide) Fuel (PROCESSED AT 30 DAYS) BURNUP= 32377, MWD /MT, FLUX= Z.65E 15 N/CM*k2-5EC ELEMENT THERMAL POWER, WATTS / METRIC TON FUEL CHARGED TO REACTOR TH PA U NP Py AM CM CHARGE 30.0 1.565-05 6.932E-08 93.17F=02 2.148 01 3,20t 00 £.40E 01 2.46t Q3 TOTALS 7.89E 02 2.54f 03 3é5.0 ls14E-C5 Be73t-10 6.13E’05 1.¢7F-01 1.14£ 01 /0*1& 01 002&2 02 5.92¢8 G2 109640 6-28E-06 8. 565E~10 1.20E“O4 1.47E-01 1.33€ 01 5.42F 01 €.86F 01 1.37€ 02 3652.0D 20885’06 30655-10 3.30E~-04 1.47E-0C1 1.33E 01 Se44E 01 3.33€ 0Ol 1.01E Q2 36525.D 2.37E=06 €.73E-10 2¢35E-03 1-46E“01 B8s18E 0O 4.91E 01 2.80E CO 5., 02E 01 365250.0 3.65E£~-05 1-005'09 5.26E'03 1.34E-01 1.33E 00 l1.36E 01 2eT4E~-C2 1.51E 01 3-50 Table 3.h9 Masses of Activation-Product Isotopes Calculated to Be Present in Irradiated LMFBR Fuel Cladding Al REFERENCE OXIDE LMFBR - CLADDING ACTIVATION POWER= €8,23 MW/MT, BURNUP= 32077, MWD/MT, FLUX= 2.,65E 15 N/CM*¥2-SEC NUCLINE CONCENTRATIONS, GRAMS / MFTRIC TON FUEL CHARGED TC REACTOR CHARGE DISCHARGE 37,0 az,n 15C.D 1096,C 10958,D HE &4 0,0 2,45E CQ 3.45F 00 3.45F Q0 3,45F D0 2.45F QC 2,45 0C C 12 1,728 02 1.72% 22 1.72F 02 1.72FE 02 1.,72E 02 1.72% 02 1.72F {2 C 13 2.,82F 00 Y,31F 21 1,318 ©1 1.23F 1 1,31F 21 1.31F 21 1,31 3} C 16 1.34F 05 1,24F 5 1,24F 0S5 1,24F 05 1,34F Q05 1.,34F 0F 1,34F Q5 C 17 S,T71FE DY 6.94FE 01 £.94% 01 A.9F 01 K,94F 01 6,.94F N1 £,04F Q1 IR 3,328 N2 2,338 02 3,23F 02 3,23F 02 3,330 02 3,33 22 2,228 (2 NA 22 1,74F 02 1.74F 03 1.74E 03 1,74F 02 1,74F 02 1.74F 02 1,74F 03 ST 22 1,80CF 03 1.50F 03 1.50F N3 1,%CE 02 1.50% 093 1.,50F 02 1,S0F 03 S1 29 7.,97F N1 7,875 01 T.976E 71 7,67 N1 T7,97E 71 T.07F "1 T.,97F (1 ST 37 B,41F 01 5.41F C1 £.41F8 G1 5.41E Q1 5.41F 01 S,41¢% Q) 5,41F 01 P 31 8,72F 01 R,73F 01 8.,73F 01 R.72F (0! AR,73F 01 9,72¢ () R2,73E 01 S 32 AJ19F T1 6L18F 01 AL19F M1 6419 €Y 4,18F 1 6.,19F 21 6,19E 31 S b 2.,97F D0 2,928 10 2,928 OC 2,92F QC ?2,92E 00 2.92F 0C 2.92F 00 v 51 0.0 1.C6F CO 1.10F CC 1,12 0C 1.14F 00 1.14F 00 1,.14F OO0 CR S50 1.,62F N2 1.62F N2 1.62FE N2 1,62F 02 1.,62F N2 1.62F 02 1.62F 32 CR 52 7,28F 74 3,28F 24 2,28F 4 2,2RFE 04 2,28F 04 3,28BF D& B,28BF Q4 FR 53 2,81F 03 3,728 03 2,72E 02 2,72F 02 2,728 03 3,72F (03 2.72F 03 CR F& O,7CF 02 1.C8E €3 1,7°3F 72 1,08F 03 1,08F N3 1,78F 02 1,08F 23 MN 54 .0 BL,80E 10 8.,22F TC T,16F O £.25F 70 TL,17E-021 1.15€-10 MN 58 4,36F 02 4,35F 03 4.35F 02 4,3%E (02 4,25F% 02 4,36F 02 &4,.354F 03 Fc 54 R,70F 02 B,66F 03 A,AKFE 02 B, 66FE O3 8.66F O3 B 6AF D2 B,E4F )3 FE 85 0.0 1.45F 1 14428 01 1,26 01 1.30E 01 6,51F D0 4.86FE~D3 FE 54 1.42F 0F 1,42FE 05 1442F 08 1,42F 05 1,42% 05 1,42F 0% 1,42F 05 FE B 3,446F 03 2,6AF 03 3,66 02 3,066E 023 3,66E 03 3,66FE D2 3,66F 33 FE B8 6§,26F 02 B,TNE N2 B.7IF ©£2 E,73F N2 §5,74F 02 5,T4F D2 S.,76F 32 co s/ 0,¢ 5.90F 00 4.48F CC 2.50E 00 1.3G6F 00 1.46F~04 0LC C0 59 2,18F N2 2.18F 02? 2.18F 02 ?2,1PE 02 2,18F 02 2.18€ 02 2.1RE 32 NI 52 1.17E D4 1.16F €& 1.16E 6 1,16F D4 1,165 N& ] ,16F N6 1,16FE 34 NI 82 Q.0 1.36C 01 1.26E 01 Y.36%F 01 1.36F 01 1.36F 0OV 1,26F )1 NI &N 4,6TE 03 4 ,ABFE Q3 4,A6F 02 4L,6HF D3 4, 66F 23 4 ,66F 02 4,.66F 03 NT 61 2,148 £2 2,188 N2 2,18 €2 2.,1°PF 02 2.18F {i2 2.18F 32 2.18F 72 NI &2 6.72F 02 6.73E 02 £,.73E 02 6.72F 02 6.72F 02 6.73F Q2 6.T72F 02 NI A4 2.06F 02 2.36E 02 2.06F 02 2.06F 02 2.06F 02 2.06F 02 2.06F Q2 SUBTOT 3.53FE 05 3,53 €05 3.53F (5 3,53k 05 2,53F 25 3,53 05 3,53F (5 TOTALS 3.532F 0S5 3.53E 05 3,53f 0% 2,52F 0% 3,53F 05 3,53F 05 2,83F 05 Al REFERFNCF OXIDE LMFRR POWER= H HFE Ll RE R T2 t NF NA MG AL S1 p S CL AR K Cha SC T1 v ce MN Fo cn NT cy Table 3.50. 3-51 in Irradiated IMFBR Fuel Cladding = CLADDING ACTIVATION 2.65F 18 N/CM%%2-SEC Masses of Activation-Product Eloments Calculated to Be Present FUEL CHARGED TO REACTOR 58,23 MW/MT, BURNUP= 132977, MWN/MT, FLUX= FLEMFENT CONCENTRATIONS, GRAMS / METRIC TON CHARGE DISCHARGE 3D Q0.0 150,0 TGt 1,20F-N09 1 ,29F-0Q 1 ,2RE=-"C 1 ,27E="0C n.0 2,458 D R,485F 00 2,45F QL 2,45k Q0 N.0 S5e6TE=QT S ,6TE=0T B,67TE=-0T7 S5,575=-07 CLr 1e22F-02 1,32FR=-12 1 ,32E-02 1,228-03 ¢, 2.,05E-07 ?,08F=07 2,055-07 2.05F=-07 1.75F N2 1,85k 02 1,8B8F 2 1.,8%F 02 1.REF 97 D et 4 ,S85F=07 5, 08FE="T7 £,2F="T7 T ,53E-"7 Yo24F A5 1 ,34% 08 1,34F (0% 1.34F G5 1,34F 05 2.0 2,08F=02 3,08F=02 2,0%F=02 3,05F=02? g f 1,0RE-02 1.96F="2 1.9€E=-"2 1,067 1748 22 1 ,74F (03 1,74FE 02 1,74F 02 1.74E 03 0.0 €.852F=-N1 B,83F=01 €,81E-01 §.,58%F-N1 oL 1.276=04 1,27C=04 1,276=04 1, 27TE-"4 1.64F D3 Y A48 03 1 p4F 022 1, 66F 02 1 ,64F (03 8,72F 01 R,73F C1 8,73F 0NY R, T2F 01 ],73F O} 6,565 01 6,83F M) £,52F 71 £,52E 7)1 4,528 M) it et 1.45F="8 1 S4F=CR 1,67F-08 1 ,75FE~-0R N.0 A ENE-1T 1,008=-17 11,0017 VL.0025=-17 Q;O E.TQE‘Yfl g.?OE-la 8.7PF'14 3.7“F-14 el 4,785=07 4, 7T7E=0T &L, TTE-"T 4. TT7E-0T Z.0 BeNAFE=0R 1 ,16F=10 H,TIE=17 7,E0E=-12 0.0 2.2C0F=C1 2,29F=0) 2,29F=-01 2,2%E-N1 ™ e 1,12€ °A 1,188 1 1,21 7 1,22E °C 2L,T78F 04 2,TRE 04 3,78E 04 2,70F N4 3,TARE D4 L,2¢F 02 &4 ,36F 03 4,36F D2 4 ,2RF 02 4,36F N3 1,8E68F N8 1 ,68F 8 ], 85FE £F 1,88 R 1 ,86F 05 2JIRE Q2 2.,25F Q2 2,.23F 02 2.21F 02 2.,20F 02 TLTE4F 24 1,74F Q4 1,745 06 1, 74E D& 1 [ T4E Q4 S 1¢2TE=AT 1, 27E-001 1 ,2RFE3Y 7 ,28F="1 05 2,53F 05 23,83F 0S5 3,53F 05 3,53F 05 TOTALS 2.°52¢ Al POWERS= p P CR MN FE £c co co NT N1 SURTOT REFERENCE OXIDF LMFRR 22 23 |1 54 55 59 5 A0 50 £ Table 3.51. 10¢6.D lolfiF—gQ 3.45F OC 5.57TE=-07 1-325-53 2 .08E=-07 1.95F C? 2.68F*36 1.34E DF 3,085-(02 1QFRE-G2 .74t 02 5¢53F=01 10?75-24 1,64F D72 a,73F Q1 6.538 21 1.395-05 1.00E-17 2,T"E~14 4, TTF=07 !.10E‘1} 2 .29F-01 1.22% 2,78F 4 ,3ARF 1.55F 2.10E D2 1.T4F D& 1.4GF-"1 o0& 3 ke 102e58.D 2.60€C=10 3.65¢ QG 5.675“07 1,32E-213 ?2.08E~07 1.8B5%E 02 227515 1.74F 05 1,056-02 Y.06E-D2 1.,74F (2 S.F2E=D1 1.27E~34 1.64F G3 2,73 01 $.53F 11 1,P9E~DA 1.00E-17 P.TLE~14 at7?F-O7 1.10E~11 2.29F=31 1.22E 0 2.79F€ Q04 03 s 02 1.74F 04 2ebbsE=(1 31.52F 05 Calculated Radicactivity of Activation-Product Isotopes Present in Irradiated ILMFBR Fuel Cladding — CLADDING ACYIVATION FLUX= 2,65E 1% N/CM*%2-5SEC FUEL CHARGFED TO REACTOR ER,23 MW/MT, BURNUP= 32077, MWD/MT, NUCLIDFE RADIOACTIVITY, CURIES / METRIC TON CHARGE DISCHARGE 3.0 An LN 157.0 ol VT 73 2,49F 02 1 ,26E 01 7,641F-01 2.0 A,7YE 00 2,70F 00 S.12E-C1 G9.71F=02 D0 T.9AE Q3 3,77E 02 8,44F 52 1,R9F 72 {ei’ 7.0 14 6,54F N4 B TPF L4 4 ,00F T4 C.0 1,A2F 04 3,54F 04 2,208 04 3,25F (4 .0 4,71F 03 2,97 02 1.1RE Q3 4,.68F 72 T ar 1,898 8 1,41F 5 7,808 4 4,41F N4 C.0 6ALTE 02 £,40F 02 £,26F 02 A,13F 02 D.0 1.02F €0 1.03F 00 1.03E CC 1.03F 00 Tl 2,785 €1 2,27F 01 3,27F 11 3,278 71 C.C 3,10F 08 2,5CF 08 1.,72F 05 1.28%F 06 E.A67E 08 2,50F 08 Y1,.728 05 1.28% 05 TOTALS 0.0 129¢6.D 0.0 3.92%-172 1.08F=018 5.73E (2 1.628 04 2,20E-04 4 ,63F 07 4 ,3RE 02 1.03F 0D 3,206 21 2.25%E 06 2 ,25F 10958.0 -, i3 Ty e PRI OO Table 3.52. 3-52 Calculated Radioactivity of Activation-Product Elements Present in Irradiated IMFBR Fuel Cladding Al REFERENCE OXIDF LMFBR -~ CLADDING ACTIVATION 2.65F 15 N/CM**2-SE( 10958.D 2.22F~06 4.355-06 2,T3E-02 Dl L. 0.0 1.62?'39 e S.62FE-25 3.795'25 D0 2.35E‘13 0.0 9.17E-07 01 0l 01 1.22F 1.24F 2.72F Cc.C 5.18F 01 POWER= 58,23 MW/MT, BURNUP= 132977. MWD/MT, FLUX= ELEMENT RADIOACTIVITY, CURIES /7 METRIC TONTFUEL CHARGED TO REACTOR CHARGE DISCHARGE 290.0 g0 .0 150.0 1096,.D H 0.0 1.2HFE«05 1.25F=05 1 .,24F=05 1,23E=-05 1 ,06E-05 BE .0 Z2eSIE~N1 4 ,35F=06 4,25F=06 4,35FE-0'6 4 ,35FE=0% T 0.0 2o TAE=02 2.T4E=02 2,74E=02 2,T4E=-02 2.74F-02 NA 0.0 T.13F 03 2.83F~-11 0.0 0.0 0.N p n.n 1.G7F €3 2.52F 02 1,41E 01 8.38F=01 3,92E-13 S 0.0 1.85E=~C4 1.,46E=-04 9,12E-Q5 5,6RFE=05 3.,30F-0°f cL 0.0 T.?20E~14 1,.,62E-20 1,62€E-30 1,82F-3Q 1.562F=30 AR DL 2.34E“n9 3.64E*21 lcllE’ZJ 3-385‘22 G-Q K 0.0 6£.81F=09 5,862E=~25 5,82FE=25 5,462F=25 5 ,42FE=25 CA C.0 2.43E=-04 2,57E-06 R,2ATE~-(8 &,48F=08 1,22E=-0° SC CoN 6.99E=-02 R,94FE-"% 1,5FfF=-C8 R, 7TNE-T9 3 ,53E-12 v 0.0 1.17E 05 2.1R8E-03 1,92E-03 1,69FE=03 2,32F=04 CRrR 0.0 0,48F N3 3,7T7E 03 8,44F 02 1,89F 02 1.08F=-0R% MN QL 1.95F 05 6,56F N4 §,7T2E 4 4 ,99F 04 §,73F )2 FE n.0 4,00 Q4 3,84 04 2,51F 04 2.,29F 04 1,.63F Q4 co D.0 1.90F 056 1.42F 05 T.9rF 04 4,47F 04 4 ,40F D2 NI CL.n T6RE 72 2,38F N1 3,376 €1 2A,37F 01 3.31F 7} CU Gofl 2.828 CO IOOIE“IR 0.0 Q.O 0.0 TOTALS 0.0 S.6TE 05 2.50E 0% 1.73F 05 1.28F 05 2,25F D4 Table 3.53. Calculated Thermal Power of Activation-Product Isotopes Present in Irradiated LMFBR Fuel Cladding A1 REFERENCE OXINE LMFBR - CLADDING ACTIVATION 1096.D 0.0 0.0 1.77€-1¢ 4.79F-11 L.62F N1 2,125 01 1.70E=-06 B,67C=07 6.79F 00 5,13F-02 7.62F N} FLUX= 2.65E 15 N/CM*%%2-SEC 1C958.D POWFER= ©®8,23 MW/MT, BURNUP= 132977, MWD/MT, NUCLIDE THERMAL POWER, WATTS /7 METRIC TON FUEL CHARGED TO REACTOR CHARGE DISCHARGE 30.0 Qf.D 15¢.D NA 24 0,0 2.C0F 02 7.0RF=~12 0,0 0.0 P 22 ", 4,298 N0 1,028 CC §,50F=02 2 NEF-13 P 32 N0 2.80F=072 1 ,22E=03 2.31E=04 &4 ,2TE-NSK CR s1 0.0 3.54% 01 1.A7E 01 3,75F 00 B,40DF-D1 MN sS4 0,0 Se66F (12 B5,29E 12 4 ,81F (32 &4, N2F D Fe 58 N .0 4,T72F 01 4,62F NY 4L 42F ) 4 ,23F 01 FF 89 N0 2,65F 01 2.30F Q1 Q.12E QC 3.,62F 00 cn 52 9.5 F.55F 03 2.85E N3 1,4RE 03 R 28EF N2 co &7 N0 1.01F ©1 Q,98FE NO Q,77F N a 854F N NT £ 0,0 B,24F=N3 £ 24F-0N2 & ,23F-02 5,23F-02 SURTOT 2,9 L,48F 03 2,28F D2 2,01F 02 1,28F M1 TOTALS ¢ F,2AC 73 2,28 2 2 718 02 | 20F ~1 2.14F=01 3-53 Table 3.5L. Calculated Thermal Power of Activation-Product Elements Present in Irradiated IMFBR Fuel Cladding Al RFFFRENCE OXIDE LMFRR - CLADDING ACTIVATION POWER= SR,23 MW/MT, BURNUP= 32977, MWD/MT, FLUX= 2.6%E 15 N/CM*xx2-SEC FLEMENT THFRMAL POWER, WATTS / MFTRIC TON FUFL CHARGED TO RFACTOR CHARGE DISCHARGE An,D 97 ,D 182.D 1nQé,0 1MA58,D H e 4.48E-10 4,46E=-11 4 ,42F-10 4,3R8E=-10 3,7RF-]1C B,26E~11 C Ne0 R,130-C6 R,13E-06 R,I2E-CE *,13F-06 B,12E~0L B,103F~06 NA 0.0 2,008 02 T,58E=-12 0,0 .0 fL0 Coh P Sl 4,408 0N 1 ,02F 00 5, £2F-N2 R,L1INE=03 1,77E-18 G.0 S C.0 Be2TE-QR 4,1£C=08 2,F0F=0R 1,672%=-08 2,40E=12 C.0 CL D0 1.32E-18 2,01€E-22 2 ,ME-22 3 ,M1E=33 3,01€6-32 2 ,01E-33 AR Y 2,59F~-11 1,75F-23 S,3FF=24 1,63F-24 0,2 0.0 K 0.0 !o?BF-IO OOD 0.0 3.0 O'O 0.0 Ca C.0 2+10E-06 2,14E-CR 5,31E-11 3,96F~-11 7,44F-12 O, 53E-28 SC el Te?28E=06 T,756-07 2, 13E=-10 1,22€-10 4,94E-14 0,0 v Ca? 1.80F 02 R,01F=06 T,06FE=0F £,23F-06 B,53F-07 B,H62F-16 CcRr 0.0 4,645 01 1,A7%8 01 3,75F CO 8.40F~-C1 4,70F-]11 2,0 MN e 2.57S 03 BL,29F N2 &4, A1F N2 4,025 02 4,52 31 7.39E-)9 Fe 2.C £,2TE Q1 A,92E 01 K.34F C1 &4,ACF 01 2.,12F CY 1,.89F=-02 cn D.0 2,54F 032 2,665 02 1.40F 03 8,35F 02 6,88 00 1.94F-C1 NI el R,27F N B, ,24F=-N"2 E ,D23F-N2 §,23E-03 5 ,13E=-"2 4 ,18F~-C3 cuy 0.0 1,0926-02 6£.195=-21 0.0 0.0 0.0 C.0 TOTALS 2.0 £.3AF £3 2,28F 02 2,00F 03 1.28% 03 7.,43E 21 2,14FE-01 Table 3.55. Anticipated Growth of the Spent-Fuel Shipping Industry from 1970 to 2020 Tnstalled Fuel Load Number of Number-of Loadgd Bstimated System Growth Pattern Year Nuclear Capacity (metric Casks Shipped Casks in Ivansit Number Shipping Loaded Casks Ending [1000 Mw (electrical)] tons/year)® per year 500 Miles 1000 Miles of Plants~ Distance® in Transitd 1970 1k 9k 30 1 1 700 1 1975 6l 1,400 470 6 9 3 600 6 1980 153 3,500 1,200 1L 23 L 500 1L 1985 250 7,500 2,700 30 52 5 450 27 1990 368 13,500 6,800 75 130 6 400 60 2000 735 15,000 9,500 105 181 6 L00 85 2020 2210 33,000 20,000 220 382 6 ele 170 #Metric tons of total heavy metal per year. Source: Systems Analysis Task Force, Phase 3, Case 42. bEstimated number of fuel processing plants in operation, CApproximate average distance in miles, assuming roughly uniform geographical distribution of plants. dThis is the average number of loaded casks that might be expected to be in transit on any given day of the year, hs-€ 3-55 of course, very approximate. The number of casks shipped annually was calculated on the basis of average loads of 3 and 1.2 metric tons per cask for LWR and LMFBR fuels respectively. The average load, in the case of the LMFBR fuel, includes both core and blanket material. The radial blanket material has a relatively low radicactivity level because of its low exposure, and can be carried in loads of 3 or l metric tons per cask. Loads of core--axial blanket fuel, however, are limited to about 0.5 to 1 metric ton per cask because of the heat-removal problem under accident conditions, The number of loaded casks in transit at a given time depends on the average length of the trip from the reactor to the reprocessing plant, Table 3.55 shows estimates for average distances of 500 and 1000 miles, using one-way transit times of Ly days and 7 days respectively. The last three columns in the table are based on calculations using current estimates of the growth pattern of the fuel reprocessing industry. These estimates were made in connection with the work of the AEC Fuel Recycle Task Force. The average shipping distance is based on an approxi- mately uniform geographical distribution of plants, which is equivalent to the assumption that suitable sites can be found in most areas of the country. No attempt was made to weight the average by locating plants close to large load centers, although this factor is always considered when choosing actual plant locations. 3.5 Waste Management Projections Estimates were made of waste management conditions anticipated for the period 1970 to 2020, These estimates were based on the SATF Phase 3, Case L2 projections of nuclear power growth in the United States (Fig. 3.1). In making these estimates, the LWR and the IMFBR were considered separately, and the results were combined to obtain composites reflecting the overall economy, 3.5.1 High-Level Wastes Light-Water Reactors. — In the case of IWR's (Table 3.56), it is assumed that the fuel has been continuously irradiated to a burnup of Table 3.56. (Aqueous processing of all fuels) 3-56 Projected Wastes from LWR Reactor Fuels Calendar Year Ending 1970 1980 1990 2000 2020 Installed capacity, 10° Mw (electrical)® 1k 153 223 209 5h1 Volume of waste generated, as liquidb Annvally, lO6 gal /year 0.017 0.97 1.98 1.58 .62 Accumulated, 10° gal 0.0L7 L.Lo 21.L 39.2 87.4 Volume of waste generated, as solid® Annually, 10° £t3/year 0.17 9.73 19.8 15.8 L6.2 Accumulated, 103 f£t3 0.17 W.o 21k 392 876 Accumulated radioisotopesd Total weight, metric tons 1.75 L51 2180 1,000 8960 Total activity, megacuries 210 18,900 54,500 62,550 1L2,700 Total heat-generation rate, Mw 0.91 81.6 226 2l 571 9OSr, megacuries 3.98 962 L340 7085 13,900 137¢s, megacuries 5.27 1280 5800 9530 18,900 1297 curies 1.85 176 2320 L4250 9510 85Kr, megacuries 0.56 12l 501 701 1280 3H, megacuries 0.033 7.29 30.2 h3.h 80 238Pu, megacuries 0,002 1.20 6.3 11.6 2h.5 3%y, megacuries® 0.00009 0.022 0.107 0.196 0.438 2hop,, megacuries® 0.00013 0,0L09 0,239 0.53 1.37 2m‘Pu, negacuries® 0.0295 6.63 27.7 L0.3 7h.1 2thu, curies® 0.354 91 L1 807 1806 2hlAm, megacuries® 0.0089 2,31 11.3 20.8 L6.6 2h3Am, megacuries 0.0009 0,232 1,13 2.07 L .62 2lhion, megacuries 0.128 29.9 130 200 379 2h2Cm, megacuries 0. 725 L3.2 90 72 211 ®Data from Phase 3, Case L?, Systems Analysis Task Force (April 11, 1948). bAssumes that wastes are concentrated to 100 gal per th Mwd (thermal) and that there is a delay of 2 years between power generation and waste generation, Assumes 1 ft° of solidified waste per 104 Mg (thermal). dAssumes that fuel was continuously irradiated at 30 Mw/metric ton to a burnup of 33,000 Mwd/metric ton, and that fuel is processed 90 days after being discharged from reactor, ®Assumes that 0.5% of the plutonium in the spent fuel is lost to waste. 3-57 33,000 Mwd/metric ton at an average specific power of 30 Mw/metric ton, that reprocessing is done 90 days after the fuel has been discharged from the reactor, and that there is a two-year delay between power generation and waste generation., Assuming that the fuel is reprocessed by aqueous methods and that the resulting waste is concentrated to 100 gal per 10,000 Mwd (thermal) burnup, the volume of waste generated annually will increase from 17,000 gal in 1970 to 1.58 million gal in 2000, If the waste is stored as a liguid, 39.2 million gal will accumulate by the year 2000, On the other hand, if it is converted to a solid form, waste volumes may be reduced by a factor of about 13. The weight, radioactivity, and heat-generation rate of all the fission products, and the accumulated activities of each significant fission-product and actinide isotopes (t > 10 years), is also shown in Table 3,56, 1/2 Fast Breeder Reactors. — For IMFBR's (Table 3.57), it is assumed that the core is continuously irradiated at a specific power of 148 Mw/metric ton to a burnup of 80,000 Mwd/metric ton, the axial blanket is irradiated at .6 Mw/metric ton to a burnup of 2500 Mwd/metric ton, and the radial blanket is irradiated at 8.L Mw/metric ton to a burnup of 8100 Mwd/metric ton. In addition, it is assumed that the fuel is reprocessed 30 days after it has been discharged from the reactor, and that a two-year delay occurs between power production and waste generation, With aqueous proc- essing of the spent fuels, it is estimated that 20,9 million gal of liquid waste, concentrated to a volume of 100 gal/10,000 Mwd (thermal) will accumulate by the year 2000. If this waste is converted to solids, 209,000 ft3 will accumulate by 2000. The levels of accumulated fission products and actinides in these wastes are also given in Table 3.57. Total Nuclear Economy. — The projected annual and accumulated volumes of wastes for the total U. S. nuclear economy are given in Table 3.58. Data for the principal radioisotopes in the wastes from spent LWR and IMFBR fuel processing are also given. The total accumulated radioactivity and thermal power of these radioisotopes are shown in Figs. 3.6 and 3.7 respectively. It should be pointed out that the total quantities of actinides in the wastes would be less than is estimated here if there is a significant contribution from thorium-fueled reactors in the nuclear economy . 3-58 Table 3.57. Projected Wastes from IMFBR Reactor Fuels (Aqueous processing of all fuels) Calendar Year Ending 1985 1990 2000 2020 Installed capacity, 10° Mw (electrical)? 28 15 shé 1669 Volume of waste generated, as liquid® Annually, lO6 gal/year 0.118 0. 7L 3.02 9.08 Accumulated, 100 gal 0.248 2.1 20.9 150.6 Volume of waste generated, as solid® Annually, 10° £t°/year 118 7.1 30.2 90.8 Accumilated, 10° ft° 2.8 2), 209 1504 Accumulated radioisotopesd Total weight, metric tons 25 260 2200 15,640 Total activity, megacuries li, 388 30,000 146,450 523,300 Total heat-generation rate, megawatts 17.4 117 563 1949 POy, megacuries 31.8 300 2165 15,500 es, megacuries 78.3 %0 6070 38,600 1297 curies 39.1 380 3300 22,690 85Kr, megacuries 7.2 66 L89 2620 3H, megacuries 0.653 6.0 L6.1 252 238Pu, megacuries® 0.18 1.98 9.1 1.5 2394, megacuries® 0.013 0.128 1.11h 8.0L 2hOPu, megacuries® 0.0161 0.156 1.38 10,0 2b’lPu, megacuries® 2.12 19.5 150.7 835 2h2Pu, curies® L8 W69 14063 29,09 2l megacuries® 1.18 1L 100 716 2b3ym, megacuries 0,037 0.36 3.12 22,1, 2thmd megacuries 0.73 7 55 321 2u20m, megacuries 1.5 95 L5 1279 ®Data from Phase 3, Case 42, Systems Analysis Task Force (April 11, 1968). pssumes that wastes are concentrated to 100 gal per 10h Mwd (thermal) and that there is a delay of 2 years between power generation and waste generation, Assumes 1 ft3 of solidified waste per 10h Mwd (thermal). dAssumes that the core was continuously irradiated at 148 Mw/metric ton to a burnup of 80,000 Mwd/metric ton, the axial blanket was irradiated to 2500 Med/metric ton at 4.6 Mw/metric ton, and that the radial blanket was irradi- ated to 8100 Mwd/metric ton at 8.4 Mw/metric ton. the fuel was processed 30 days after discharge from the reactor. ®Assumes that 0.5% of the plutonium in the spent fuel is lost to waste. It was also assumed that 3-59 Table 3.58. Projected Fuel Processing Wastes from Total U. 5. Nuclear Power Economy (Aqueous processing of all fuels) Calendar Year Ending 1970 1980 1990 2000 2020 Tnstalled capacity, 10° Mw (electrical)® 1L 153 368 735 2210 Volume of waste generated, as liquidb Annually, 106 gal/year 0,017 0.97 2.69 .60 13.7 Accumulated, 10° gal 0.0L7 4.40 23,8 60.1 238 Volume of waste generated, as solid® Annually, 10° £t°/year 0.17 9.73 26.9 L6.0 137 Accumulated, 10° £t° 0.17 1.0 238 600 2380 Accumulated radioisotopesd Total weight, metric tons 1.75 L5l 2410 6200 2ly,600 Total activity, megacuries 210 18,900 84,500 209,000 666,000 Total heat-generation rate, megawatts 0.91 81.6 343 807 2520 9OSr, megacuries 3,98 962 4640 9550 29,400 13705, megacuries 5.27 1280 6540 15,600 57,500 1291, curies 1.85 L76 2700 7550 32,200 85Kr, megacuries 0.56 124 567 1190 3500 3H, megacuries 0.033 7.29 36.2 89.5 332 2385, megacuries® 0.00? 1.20 8.28 30.7 166 239py, megacuries® 0.00009 0.022 0.235 1.3l 8.45 2LOp, megacuries® 0.00013 0.0409 0.395 1.91 11.4 2blp,, megacuries® 0.0295 6.63 17.2 191 909 Ethu, curies® 0.354 91 910 14870 30,900 2blyn, megacuries® 0.0089 2.31 22,7 121 763 2b3yn, megacuries 0.0009 0.232 1.49 5.19 27.0 2Uiny megacuries 0.128 29.9 137 255 700 2LL2Cm, megacuries 0.725 h3.2 185 L87 1490 %Data from Phase 3, Case 12, Systems Analysis Task Force (April 11, 1968), bAssumes that wastes are concentrated to 100 gal per 10)‘L Mwd {(thermal) and that there is a delay of 2 years between power generation and waste generation. Chosumes 1 £43 of solidified waste per 10% Med (thermal). dAssumes that IWR fuel is continuously irradiated at a specific power of 30 Mw/metric ton to a burnup of 33,000 Mwd/metric ton, and that the fuel is processed 90 days after discharge from reactor; LMFBR core continuously irradiated to 80,000 Mwd/metric ton at 148 Mwd/metric ton, axial blanket to 2500 Mwd/metric ton at h.6 Mw/metric ton, and radial blanket to 8100 Mwd/metric ton at 8.4 Mw/metric ton, and that fuel is processed 30 days after discharge. ®Assumes that 0.5% of the plutonium in the spent fuel is lost to waste. 3-60 ORNL—DWG 69 — 7869R {(¢) TOTAL WASTE SYSTEMS {#) FAST BREEDER FUEL WASTES ISCTOPIC ACTIVITY IN WASTE SYSTEMS (meqacuries) (¢} LIGHT — WATER FUEL WASTES 1970 1980 1990 2000 2040 2020 CALENDAR YEAR ENDING Fig. 3.6. Total Accumulated Radiocactivity in Wastes Generated by Reprocessing Spent ILWR and IMFBR Fuels. 3-61 ORNL-DWG 69 —6787R {¢) TOTAL WASTE SYSTEMS 5 (6) FAST BREEDER FUEL WASTES {¢) LIGHT —WATER FUEL WASTES ISOTOPIC POWER IN WASTE SYSTEMS (megawatts) 1970 1980 1990 2000 2010 2020 CALENDAR YEAR ENDING Fig. 3.7. Total Accumulated Thermal Power in Wastes Generated by Reprocessing Spent LWR and IMFBR Fuels. 3-62 Annual Generation Rates. — The estimated annual generation of key fission-product and actinide isotopes is presented in Table 3.59. 3.5.2 Management of Solidified High-Level Wastes Estimates were made of the conditions that would exist if high-level liquid wastes from fuel reprocessing were solidified and then shipped to a salt mine repository for permanent storage. Table 3.60 presents the volumes in storage, the required storage canal capacities, the number of waste shipments, and the total and accumulated mine space needed if the wastes were solidified immediately and shipped after 5 years of interim storage on-site. Table 3.6l gives analogous data for the case of storing the solidified wastes on-site for 10 years. 3.5.3 Intermediate- and Low-Level Liquid Wastes The projected generation of so-called intermediate- and low-level liquid wastes as the result of LWR and LMFBR fuel reprocessing is given in Table 3.62. Volumes of these wastes were calculated on the basis that about 200 gal and 10,000 gal, respectively, are produced per metric ton of fuel reprocessed. This corresponds rougnly to present practice; how- ever, 1t can be anticipated that, in the future, intermediate-level wastes will be combined with high-level wastes and converted to a solid form. In addition to the volumes of low-level wastes shown in Table 3,62, about 3 million gal of low-level waste is generated annually at each reprocessing plant from sources such as cell drainage, equipment decontamination flushes, and laboratory sinks. 3.5.4 Solid Wastes If mechanical decladding, as exemplified by a shear-leach head-end step, is used in reprocessing, the cladding hulls and associated fuel- assembly hardware containing neutron-induced radioisotopes (as well as some of the actinides) will constitute an important source of solid waste. Calculated levels of these isotopes for Zircaloy cladding and for stain- less steel cladding are given in Tables 3.25-3.30 and Tables 3.49-3.5L respectively. Annual and accumulated volumes of cladding wastes, assuming compaction to 70% of their theoretical densities, are given in Table 3.63. Wit W Table 3.59. Projected Annual Generation of Key Fission Product agd Actinide Isctopes in Wastes from IWR and IMFBR Fuel Processing®s Light-Water Reactor Fuels Fast-Breeder Reactor Fuels Total Calendar Year Ending 1970 1980 1985 1990 2000 2020 1985 1990 2000 2020 1.970 1980 1.990 2000 2020 Psr, megacuries/year 3.99 227 110 162 367 1080 15.5 93.5 398 1180 3.99 227 556 765 2280 L3705, megacuries/year 5.55 316 571 643 511 1500 38,8 233 1000 3010 5.55 316 878 1510 Ls10 1297 curies/year 1.96 112 202 320 181 531 19,0 118 189 1WT0 1.9% 12 135 670 2000 ke, megacuries/year 0.59 33 60 63 5 158 3.63 22 93.6 282 0.59 33 90 WE ko 34, megacuries/year 0.036 2.1 3.7 L.? 3.3 9.8 0.33 2.0 8.6 25.8 0.036 2.1 6.2 11.9 35.6 238Pu, megacuries/year® 0.00072 0.041 0.07L 0,084 0,066 0.195 0.02 0.121 0.515 1.55 0.00072 0.041 0.205 0.58 1.75 239Pu, megacuries/year" 0.000085 0.00L9 0.0088 0.0099 0.0079 0,023 0.0063 ©.0379 0.162 0.lL86 0.000085 0.0049 G.0LT8 0,17 0.509 2hOPu, megacuries/year® 0.000123 0.00705 0,0L27 0,014k ©.Cllh 0.0335 0.0076 0.046 0.196 0.589 0,000123 0©,007L 0,0604 0.207 0.623 Eh1Pu, megacuries/year® 0.03 1.71L 3.09 3.u8 2.7 8.11 1.07 6.6 27.5 82.9 0.03 1.7 2.9 30.3 91.0 2h2Pu, curies/year® 0.35 20,1 36.3 40.9 32.5 95.4 22.6 137 58l 1760 0.35 20.1 178 617 1850 thAm, megacuries/year 0.009 0.51 0,92 1.03 0.821 2.1 0.56 3.38 1. L3.b 0.009 0.51 L. 15.2 45.8 2h3Am, megacuries/year 0.00021L 0,012 0.021 0,024 0.0192 0.056 0.0177 0.107 0.L6 1.37 0.00021 0,012 0.131 O.47h 1.43 2h2Cm, megacuries/year 0,99 56.9 103 116 92,1 270 23.3 141 601 1810 0.99 56.9 257 693 2080 thCm, megacuries/year 0.129 7.37 13.3 15.0 11.9 35 0.l 2.67 11.5 34.3 0.129 7.37 17.7 23.3 69.3 %Based on Systems Analysis Task Force, Phase 3, Case 42 (April 11, 1968), and assumes a 2-year lag in waste generation after power producticn. bAssumes that the LWR Mel ig continuously irradiated to a burnup of 33,000 Mwd/metric ton at a specific power of 30 Mw/metric ton; also assumes that the LMFBR core is rontinuously irradiated to a burnup of 80,000 Mwd/metric ton at 148 Mw/metric ton, the axial planket is continuously irradiated to 2500 Mwd/metric ton at h Mw/metric ton, and the radial blanket is contlnuously irradiated to 8100 Mwd/metric ton at o.i4 Mw/metrlc ton, Cpssumes that 0.5% of the plutonium in the spent fuel is lost to waste. 3-6L Table 3.60. Waste Management Data for Conversion-to-Solids Concept (5-year interim solid storage) Calendar Year Ending 1980 1990 2000 Solid waste generation, ftB/yeara 9,730 26,900 L,6,000 S-year interim solid storage Volume in storage, ft° 3l,400 117,500 206,000 Length of 2L-ft-wide canals, ft 690 2,340 4,100 1000-mile shipment to salt minesb Number of shipments per year 62 332 61l Number of casks in transit® 2 7 12 Disposal in salt mines Area required, acres/year 17 83 157 Accumulated area used, acres L3 540 1780 b ®One cubic foot of solid waste per 10~ Mwd (thermal) irradiation, PFach shipment consists of thirty-six 6-in,-diam pots containing 8 megacuries of radioactivity and generating 100,000 Btu/hr. ®One-way transit time is 7 days. p 3-65 Table 3.61. Waste Management Data for Conversion-to-Solids Concept (10-year interim solid storage) Calendar Year Ending 1980 1990 2000 (20L0)? Solid waste generation, ft3/yearb 9,730 26,900 L6,000 (-=) 10-year interim solid storage Volume in storage, ft° 13,800 194,000 363,000 (--) Length of 2l-ft-wide canals, ft 870 3,860 7,230 (--) 1000-mile shipment to salt mines® Number of shipments per year 3 172 L77 (81L) Number of casks in transit® 1 L 10 (16) Disposal in salt mines Area required, acres/year 0.7 1O 113 (197) Accumulated area used, acres 0.7 186 1010 (2560) 2Ccommitments made in the year 2000. bOne cubic foot of solid waste per 10,000 Mwd (thermal) fuel exposure. ®Each shipment consists of thirty-six 6-in,-diam pots containing 5 megacuries of radioactivity and generating 56,000 Btu/hr. dOne-way transit time is 7 days. 3-66 Table 3.62, Estimated Volumes of Low- and Intermediate- Level Liquid Wastes® Calendar Year Accumulated Ending Gallons per Year Gallons® Intermediate-Level Waste® 1970 31,000 31,000 1980 777,000 3.5 x 10° 1990 2.6 x 10° 2.0 x 10 2000 3.2 x 10° L.9 x 107 Low-Level Wasted 1970 1.6 x 10° 1980 3.9 x 107 1990 1.5 x 10° 2000 1.h x 108 ®Based on fuel processing projections of Phase 3, Case }2, Systems Analysis Task Force (April 1968). bIn the future, these wastes will probably be combined with high-level wastes and solidified, ®Based on the generation of 200 gal of intermediate-level waste per metric ton of fuel processed, dBased on the generation of 10,000 gal of low-level waste per metric ton of fuel processed. These wastes are decontaminated of radioisotopes to required levels and discharged to the environment. et wub 3-67 Table 3.63. Solid Wastes from Spent LWR and ILMFBR Fuel Processing™ Calendar Year Ending 1970 1980 1990 2000 Volume of cladding wasteb Annual, 10° f£12 0.3 8.3 L1 87 Accumulated, 10° £t 0.3 37 320 1030 Total volume of solid waste® annual, 100 £t 0.03 0.8 2,0 3.2 Accumulated, 10° £t 0.03 3.5 16 19 Burial ground aread Annual, acres 0.6 16 L N Accumulated, acres 0.6 70 320 980 *Based on fuel processing projections of Phase 3, Case 1j2, Systems Analysis Task Force (April 1968). bBased on 2.1 ft3 of cladding hulls per ton of LWR fuel processed, and 8.7 £t3 of cladding hardware per ton of IMFBR mixed core and blankets processed. ®Based on an average volume of 200 £13 of solid wastes per ton of fuel processed, dBased on burial of 50,000 £t3 of solid waste per acre of burial ground. 3-68 Other solid wastes that are generated as a result of routine reproces- sing plant operation vary widely in size and characteristics. Annual and accumilated volumes were estimated (see Table 3.63) by using 200 ftB/metric ton as the average volume of all solid wastes produced in fuel reprocessing. The land area needed for the ultimate disposal of all these solid wastes, assuming that the burial of a SO,OOO—ft3 volume requires one acre, is also shown in the table. 3.6 References United States Federal Power Commission, National Power Survey, Parts 1 and 2, U, S, Government Printing Office, Washington, D.C. (196l). Forecast of Growth of Nuclear Power, USAEC, Division of Operations Enalysis and Forecasting, WASH-108l, (December 1967). M. Searl, USAEC, memorandum ("Electricity Requirements and Generating Capacity by Areas for the Year 2000") to J. Vallance, March 1, 1967. Jackson & Moreland and S, M, Stoller Associates, Current Status and Future Technical and Economic Potential of Light Water Reactors, WASH-1082 (March 1968). K. Buttrey, O, R, Hillig, P. M. Magee, and E. H. Ottewitte, Liquid Metal Fast Breeder Reactor Task Force Fuel Cycle Study, NAA-SR-MEMO- 1260 (January 1968). Chem, Technol. Div. Ann, Progr. Rept. May 31, 1969, ORNL-LL22, pp. 89-91. J. H. Goode, ORNL, personal communication, January 8, 1969, L-1 ;. TECHNICAL CONSIDERATIONS li.7 Design of Fuel Reprocessing Plants Fuel reprocessing plants are characterized by their complexity. Typically, a fuel recovery process entails shearing the fuel (to rupture the corrosion-resistant sheath and expose the fuel), dissolution of the fuel in nitric acid, separation and purification of the uranium and plu- tonium by solvent extraction and jion exchange, and conversion of the product nitrates to oxides suitable for refabrication into fuel elements. In addition to the primary process, there are many auxiliary operations: treatment of the solvent to provide for its reuse, recovery of nitric acid from the agueous streams, management of the gaseous, liquid, and solid waste effluents, and the speclalized techniques and equipment re- quired for process control and personnel protection. The spent fuel is transported from the reactor to the reprocessing plant in heavy, shielded casks. The cask is unloaded in a water-filled pool, and the fuel is stored under water, which serves both as a trans- parent radiation shield and as a coolant. The fuel elements to be proc- essed are transferred to a head-end cell and sheared intoc 2-in. lengths to expose the inner core, which is then leached with nitric acid in batch dissolving tanks. The leached hulls constitute a solid waste that is ultimately disposed of by land burial. The nitric acid solution of the fuel, containing the uranium, plutonium, and nearly all of the fission products, is the feed solution for the solvent extraction process. Solvent extraction processes exploit the wide difference in concen- tration distribution between two immiscible phases — the organic and the aqueous. Nearly all major fuel reprocessing facilities employ some form of the Purex process,1 which makes use of the organic complexing compound, tributyl phosphate (TBP), in an inert hydrocarbon diluent. When this organic mixture is brought into countercurrent contact with the aqueous feed solution, the TBP extracts both the uranium and the plutonium into the organic phase, leaving the fission and corrosion products behind in the aqueous phase. L-2 The TBP-organic solution of uranium and plutonium is stripped or back-extracted with dilute nitric acid. The back-extraction of plutonium can be vastly enhanced if the plutonium is reduced to the trivalent form, usually with ferrous sulfamate; this makes it possible to back-extract selectively first the plutonium and then the uranium. The two aqueous solutions are usually further purified by a second extraction cycle or by ilon exchange. The ferrous sulfamate that is used for reducing plutonium is, itself, oxidized to ferric sulfate in the reduction process, and thus contributes to the waste and interferes in the chemistry of subsequent plutonium purification steps. Uranium in the tetravalent state has been successfully used by the Europeans for this purpose; also, plutonium reduction can be effected with hydrogen.2 Neither of these reductants contributes spurious chemicals to the process. The uranium and plutonium may be precipitated from dilute nitric acid solution with oxalic acid. These oxalate precipitates are then removed by filtration, and the filter cakes are thermally decomposed to produce ura- nium and plutonium oxides. The oxides are sintered and ground, or extruded into pellets for fabrication into new fuel elements. 3 In the recently developed sol-gel process,” the nitric acid is removed from the agueous solutions of plutonium or wranium by extracting the acid with an amine solvent. As the acid extraction proceeds, a stable, colloidally dispersed suspension of uranium oxide is formed. This is a "sol," which can be handled like a true solution. Progressive removal of water by evaporation or by extraction with a hygroscopic solvent converts the sol to a plastic gel. The sol can be formed into gel micro- spheres of controlled size by adding the sol dropwise into a stream of the hygroscopic solvent. When fired to about 1200°C, the gel attains a density near the theoretical density and is suitable for fabrication into reactor elements, Sols of plutonium and uranium can be combined and gelled to form "mixed" oxide microspheres in which the two elements are homogeneously dispersed. Radioactive gaseous wastes from these operations are treated chemi- cally, as well as by filtration, sorption, and scrubbing in order to reduce h-3 their radioisotope content to levels that can be discharged to the atmo- sphere. The agueous radioactive wastes that contain essentially all the fission products are generally concentrated by evaporation and stored on an interim basis in underground tanks. The evaporator overheads are sufficiently decontaminated of radioisotopes to permit their discharge to the environment under existing regulations. ,.1.1 Preventive Measures and Containment Criteria Criticality. — Criticality is normally prevented by a combination of the following: limiting the concentration or quantity of material "in-process" by administrative means; imposing dimensional limitations on the process equipment; and adding parasitic neutron absorbers, either soluble or fixed to the process tanks. (The latter are usually in the form of raschig rings or parallel spaced plates.) Administrative control is usually arranged so that the positive, simultaneous action of two responsible operators is required to add criti- cal material to an "in-plant" inventory and to transfer material within the plant. A visual display of the fissionable material inventory status in each area is maintained, and transfer valves are kept locked with the keys in the immediate control of supervision. "Double-batching'" is thus prevented, and the plant or discrete portion of the plant is made safe by limiting its in-process inventory to less than the minimum critical quantity. Neutron absorbers, such as boron and cadmium added directly to the dissolver as soluble salts, are effective for criticality control in the L dissolution and feed adjustment steps. These absorbers remain with the aqueous waste. Tanks of large volume packed with borosilicate-glass raschig rings may be used for the storage of fissile product solutions. Parallel plates of boronated stainless steel have been used in the bell-shaped end sections of pulsed columns. Radiation. — Airborne radiocactivity in the cell ventilation system, and in personnel operating areas, is usually detected by radiation-sensing instruments focused upon a filter through which a constant volume of air is drawn. Some designs use continuous filters and are set to alarm at L=k certain radiation levels; others employ a fixed filter but are set to alarm at a given rate-of-rise of the filter activity. Radiation fields in operating areas are monitored by ion chambers with level alarms. In order to avoid the hazard of spurious alarms, signals from two out of three instruments are required before an alarm sounds. The inadvertent entry of personnel into shielded process areas having high radiation fields 1s prevented by securely locking these areas, with access in the immediate control of supervision or other properly designated authority. Containment Systems. — Processing plants are designed to ensure con- tainment of airborne radicactivity by providing increasing levels of vacuum in three successive envelopes so that all air leakage flows from areas of low to those of high contamination potential. The building forms the outermost envelope, and it is operated at a pressure approximately 0.3 in. Hy0 lower than atmospheric pressure. This is a higher vacuum than a 30-mph wind could be expected to produce on the lee side of a rectangular building. All openings in the building communicate either with uncontaminated persormel areas, or with two doors in series, only one of which can be open at any time. The vestibule formed by the space between the doors is maintained as an uncontaminated area. The shielded process cells form the second envelope of containment and are operated with a vacuum of about 0.7 in. H,0 with respect to the building. The cell exhaust system has a rated capacity of approximately 0.1 cell volume per minute (to accommodate explosions or fires without pressurizing the cell), and the cell in-leakage rate is limited to approxi- mately 107° cell volume per minute at a 2-in. H,0 differential pressure. Seals that are used to close cracks and crevices are designed to withstand a minimum pressure of 10 in. H.,0. The cell structure and its closures are designed to withstand the pressure that could be generated by any credible accident. Finally, the process equipment in the cell is operated at a negative pressure, with respect to the cell, of about 10 in. HZO, Similarly, the direction of air flow through personnel areas in the building is controlled by introducing a positive air flow into offices that exhaust into corridors. From the corridors, the air flows succes- sively toward operating areas, to limited access areas, and to hot labora- tories, from which the air is exhausted through filters to the atmosphere. L-5 Off-Gas Treatment. — Because of its higher radioactivity and chemi- cal fumes content, the dissolver off-gas in most radiochemical plants is treated, in turn, for nitric acid recovery, for iodine removal, and for removal of residual acid fumes before being blended with the off-gas from the vessels in the balance of the plant. The vessel off-gas is usually scrubbed with caustic, dried, and filtered through one roughing and two high-efficiency filters (HEPA, asbestos-glass fiber paper, 99.97% DOP efficiency). Todine in most of its chemical states is removed from gas streams by reaction with AgNO, impregnated on ceramic packing and by scrubbing with Hg(NO5),-HNO; or caustic solutions. However, organic iodides, par- ticularly methyl iodide, can be removed most efficiently by catalytic decomposition and sorption on silver, copper, or iodine-impregnated char- coal. The efficiency of iodine removal units is sharply dependent upon the concentration of the iodine, but 99.5% is a commonly quoted design efficiency in cases where organic iodides are not removed.5 The efficiency of charcoal impregnated with potassium ilodide has been quoted at 99.99%.6 All off-gas streams from the plant are blended with the cell ventila- tion streams and passed through a sand filter,7 a deep-bed fiber-glass filter,8 or a bank of high-efficiency particulate air filters before being monitored and discharged up a stack. }.1.2 Probable Trend of Plant Design The principal concern in chemical plant design is safety, but economy is a necessary parallel objective. The size of processing plants will in- crease to take advantage of the lower unit processing costs associated with higher plant capacity. Newer reactor fuels, the IMFBR fuels in particular, will contain higher quantities of fissionable material. The high value of this fissionable material will supply an economic inducement for minimizing out-of-reactor processing time; thus fuel may be processed with as little as 30 days preprocessing decay time. This short decay period, combined with the increasing specific power and high burnups of future reactor de- signs, will exaggerate many fuel processing problems; for example, there will be more decay heat to dissipate, more radiocactive off-gas to contend L6 with, more extensive disintegration of process reagents due to radiation, and more severe plutonium criticality considerations. In addition, the product-finishing end of the plant must be shielded owing to the presence of certain isotopes in recycled plutonium and uranium (238pPu, <° b GAS TO CLEAN-UP WASTE _-Lgp_._(_.-J ATOMIZING GAS | L. - 1tk ! ° t CALCINER CALCINER o > STORAGE POT |° ° L ; fo | ' °| STET7T CONTINUOUS MELTER TO RECEIVER POT ¥ a - Pot Solidification b - Spray Solidification GAS TO - “ WASTE (== GAS YO CLEAN-UP CLEAN-UP A - GLASS FORMING CHEMICALS SLURRY ' EVAPORATOR -+ GAS TO CLEAN-UP - | B «=FLUIDIZING GAS e | D MELTER oI o ‘ ' ! ' , SR - o , T B ¥ TO RECEIVER POT ¢ - Phosphate Glass Solidification d - TFluidized Bed Solidification Fig, 4.3. Primary Solidification Techniques in the United States. (Drawing by courtesy of Battelle-Northwest) Eid L-29 Table L.3. Summary of Research and Development on the Solidification of High-Level Waste' Process Pilot Plant and Lab Scale b Capacity Chemical Status Sites Time Span Radiocactivity Radiocactivity (liters/hr) Product Additives of Work Pot Calcination ORNL 1958-1965 None None 25 Calcine Calcium, sulfate Completed BNwW 1959-1962 None None 10 Calcine Sulfate Completed 1962 to date H H 20 Calcine Sulfate, calcium In progress Spray BiW 1959 to date H H 20 Ceramic, Phosphate, In progress glass borophosphate USSR ~1961 to date ? ? 20 Calcine, Borosilicate In progress glass Phosphate Glass BNL 1960 to date None None 20 Glass Phosphate In progress BNW 166l to date H H 20 Glass Phosphate In progress Fluidized Bed ANL 1955-1959 None L 6 Granules None Completed INC 1955 to date L I 300 Granules None In progress BNW 1959-1961 No work None 20 Granules None Completed USSR ~1962 to date 7 ? 30 Glass, Boresilicate In progress granules Pot Glass AERE 1959-1966 None H 6 Glass Borosilicate Completed FAR 1962 to date H L 20 Glass Boroalumino- In progress silicate Phosphosilicate CpP 1969 startup No work H 20 Glass Boroalumino- In progress silicate ORNL 1961-1966 Note None 3 Semiglass Phosphate, Completed Borophesphate Rotary Kiln BNL 1955-1963 None None 20 Powder None Completed FAR 1960 to date None None 6 Glass Phosphosilicate, In progress Borosilicate Ceramic Sponge LASL 1959-198) None L by Ceramic None Completed balls ork is also being done in Canada, Germany, Denmark, India, Japan, bValues are based upon ref }5: H is » 70 Ci/kg of solid; T is 0.07 o 70 Ci/kg of solid; L is < 0.07 Ci/kg of solid. Abbreviation SWMATY 5 ORNL USSR AERE . Do~ OFEFwrn — —_ LASL Oak Ridge National Iaboratory, Oak Ridge, Tennessee BNW Battelle-Northwest, Richland, Washington Union of Soviet Socialist Republics BNL Brookhaven National Laboratory, Upton, Long Island, New York ANL Argonne National Laboratory, Argonne, Illinois INC Idaho Nuclear Corporation, Idaho Falls, Idahe Atomic Energy Research Establishment, Harwell, Berks, England FAR Center for Nuclear Studies, Fontenay-aux-Roses, France CPP Center for Plutonium Production, Marcoule, France Los Alamos Scientific laboratory, Ios Alamos. New Mexico and Czechoslovakia. L-30 impurities when compared with the chemical fission product content of the wastes. The variations usually have marked impact on the solidification process conditions and on the nature of the final solidified waste. A1l processes for solidifying high-level waste generate additional waste streams that contain intermediate levels of radiocactivity. These are the vapor or condensate streams from the solidifier that have been decontaminated by factors of 10 te 1000. From this point, decontamination requirements of the effluents are comparable to those from the high-level liquid waste handling system of the fuel reprocessing plant. Processing of these effluent streams would logically and readily be done by recycle routing to the existing high-level liquid waste concentration and proces- sing equipment. Only a modest increase in capacity (on the order of 10%) of the liquid waste processing capacity of the reprocessing plant would be required. The first part of Table l.lL describes five waste compositions that bracket the ranges of nonfission product compositions of wastes expected from fuel reprocessing by solvent extraction. All compositions are shown at a volume of 378 liters per metric ton of uranium fuel (100 gal/metric ton) to provide a common basis, although concentrations greater than about L N in total metallic lons will generally result in excessive precipitation 28,26,L6 which is unmanageable for extended storage. The compositions shown in Table L.l assume that the fuel cladding is not dissolved with the fuel; consequently, the fuel cladding constituents are not present in the high- level waste. Waste composition No. 1 is typified by a very high content of iron and a low content of other constituents. This waste has been generated by one reprocessorh7 by dissolving an iron fuel container with the fuel. Composition No. 2 is a moderately "dirty" waste from first-cycle waste combined with second-cycle waste that contains sulfate (which comes from a reductant in the uranium-plutonium partitioning step). Waste No. 3 is the same as waste No. 2 except that it has been neutralized prior to stor- age. Waste No. L is a "clean" waste, which would come from the first- solvent extraction cycle if reasonable care is taken to maintain a flow- sheet reasonably free of nonradicactive chemicals. This is expected to L-31 Table l.l,. Range of Chemical Compositions of High-Level Liquid Wastes Concentration (M at 378 liters/metric ton) for Waste Composition Constituent No. 1 No. 2 No. 3 No. It No. & A, General Chemical Composition of Inert Materials Na Low High High Low Tow e High Medium Medium Low Iow A% 0 0 0 0 High 30, 0 High High 0 O B. Actual Chemical Composition of Inert Materials H 3.7 3.93 (-)o0.0 6.29 .25 Fe 0.93 0.4h5 0.Lh5 0.05 0.05 Cr 0.012 0.02) 0.024 0.012 0,012 Ni 0.0085 0,010 0.010 0.008 0.008 Al 0.001 (0,001 0.001 0.001 0.65 Na 0,138 0,93 3.67 0.10 0.10 U 0.010 0.010 0.010 0,010 0,010 Heg <0.,001 <0,001 <0.001 <0,00" <0 ,001 Nog 7.5 5.37 2.0 6.66 6.5 50, - 0.87 0.87 - - PO, 0.003 0.006 0.006 0.003 0.003 Si04 0,010 0,010 0.010 0.010 0.010 F <0.001 <0 .001 <0.,001 <(,001 <0,001 + chhem (ref. a) 3.03 2.48 5.22 0.365 2,31 kg oxide/metric ton 31.7 28.1b 60b b.6 17.2 C. Chemical Composition of Major Materials from Nuclear Fission Fuel Exposure in Thermal Reactors 20,000 Mwd/metric ton at 15,000 Mwd/metric ton at 15 Mw/metric ton 30 Mw/metric ton Mo 0,065 0,130 Te 0.01L 0,031 Sr 0.0155 0.036 Ba 0,0195 0.ok1 Cs 0,035 0.078 Bb 0,007 0,01y Y+RE® 0.12 0,27 Zr 0,065 0,143 Ru 0.032 0,082 Rh 0.0074L 0,013 Pd 0,017 0,043 Ag 0.0008 0.0016 Cd 0., 0008 0.0025 Te 0.006) 0.014 zMI’:p (ref. a) 0.91 2.11 kg oxide/metric ton 22 L9 ®M" is metal equivalents, or normality of metal ions (does not include acid). bDoes not include the sulfate. If sulfate is not volatilized, approximately 27 kg of additional oxides per metric ton are formed. CRE is rare earth elements. L-32 be a fairly typical waste in the near future. Waste No. 5 is a waste that is generated in a TBP-25 process in which aluminum nitrate is used for the salting agent in the solvent extraction process. Chemical adjustment of waste No. 3 is required before solidifica- tion since direct calcination will form unstable, hygroscopic Na 0. Upon acidification, the composition of this waste then approaches that of waste No, 2. The TBP-25 process offers no known major advantages over the Purex process for commercial plants; therefore, waste No., 5 is believed to have only minor importance in the future. Consequently, wastes having composition Nos. 1, 2, and li bracket the range of expected high-level liquid waste compositions. The third part of Table L.}y shows the amounts of fission product elements resulting from fissioning in thermal reactors with moderate— and high-exposure histories. The 20,000-Mwd/metric ton exposure is typical of current reactors, and the 45,000 Mwd/metric ton exposure represents probable maximum exposures in future thermal reactors. 8 It is obvious that, unless intermediate-level wastes from fuel reprocessing are mixed with the high-level wastes, the chemical content of fission produces will be significant in essentially all fuel reprocessing schemes. In fact, with moderate attempts to minimize the inert contaminants in the waste, the chemical equivalents of fission products will exceed those of the non- fission products, and the chemistry of the fission products will be the controlling factor in the waste treatment steps. Another point of interest is that the absolute minimum weight of solidified waste (that of fission product oxides alone) is about 1.1 kg/ 1000 Mwd thermal exposure. Contributions from inert chemicals in the compositions shown in Table L.l can increase that volume by a factor up to about L. Additional chemical additives are often needed to perform chemical functions during solidification. These additives are based on the total composition of chemicals present, and can increase the waste volumes by as much as a factor of 2. Solidification processes that form melts require significant chemi- cal modification of almost any waste composition. Compositions for waste L-33 solidification generally require at least 70 mole % of inert chemicals to incorporate the fission products into materials that are meltable at reasonably low temperatures (i.e., at less than about 1000°C). (A more typical value for inert chemical content in melts is 85%). Melts have L9-51 57 been developed in which the major melt-making fluxes are phosphates, 52-54 55,56 52,55,56 In most cases, workable chemical composition ranges have been defined. borophosphates, silicates, borosilicates, and borates. However, general correlations for chemical compositions are somewhat difficult to define because of the complex interaction of all the consti- tuents in the wastes. Therefore, each waste composition encountered usually requires at least some laboratory investigation of melt-forming composition. Similar studies are usually necessary to predict the occur- rence of special problems, such as ruthenium or sulfate volatility, foam- ing, stickiness, etc., for all solidification processes. Three of the inert chemical constituents listed in Table 4.4 are sufficiently troublesome during solidification to merit efforts to keep them out of high-level wastes. These constituents are sulfate, fluoride, and mercuric ions. Sulfate ion is generally unstable chemically at the higher range of temperatures reached in solidification (700°C and higher) and tends to volatilize. Retention of sulfate in the solidified waste at temperatures above 700°C requires chemical additives (usually calcium); for melts, it becomes very difficult above 950 to 1OOO°C.h9 The volatil- ization of sulfate results in added corrosion problems in the off-gas system recycle and in increased sulfate concentrations in the liquid waste for cases of partial volatilization; in cases of complete volatilization, another medium-to-high-level waste stream requiring special treatment and disposal is produced. Sulfate also causes severe precipitation and resul- tant solution handling problems from sodium — rare earth sulfates when the 58 latter are present at concentrations of approximately 0.5 M or greater. Fluoride is retained with difficulty (by using calcium) during solidi- 59 fied waste processing up to temperatures of about 600°C,”” and is nearly 0 impossible to retain significantly at higher tem.peratures.6 If it cannot be retained, it must be disposed of by another means (e.g., via the plant stack or discharge in a separate lower-level waste stream). A fluoride L-34 content much greater than about 0.001 M will significantly increase corro- sion of stainless steel and titanium (used generally in waste processing systems), although this corrosion can be partly overcome by use of com- plexing agents (aluminum, zirconium, etc.). Mercury cannot be retained in solidified waste that is processed at temperatures above 100 to 500°C. When volatilized, the mercury and its oxides condense at temperatures of about 350°C and provide relatively serious potential plugging problems. A means for pretreating the waste for removal of mercury has been developed in the 1aboratory.61 Ruthenium is just as troublesome in waste solidification as it is in fuel reprocessing. Its removal from the off-gas stream is more diffi- cult than that of nonvolatile materials. One to eighty percent of the ruthenium will usuvally oxidize and volatilize during solidification. Additions of certain chemicals are sometimes required to minimize oxida- tion to the volatile RuO, form. Even then, volatilization of at least 1% is usually encountered. Pot Calcination. — Pot calcination, which was developed at ORNL, is a batch process that has been developed to a state of readiness for com- mercial radioactive use. It is presently being demonstrated with full- level radioactivity on a pilot-plant scale. Its advantages are that it is a simple process and is adaptable to a wide variety of feed compositions. Its disadvantages are: (1) a stainless steel pot is required, (2) the amount of heat that can be incorporated into a pot is limited, (3) the capacity of a system using this process must be increased by multiple-pot lines, and (L) the solidified waste is more leachable than glassy solids. Pot calcination is a batch process in which the principal processing vessel, the pot, is also the final container for the solidified waste. In pot calcination, liquid waste 1s added to a pot that is heated in a multiple-zone heating and cooling furnace, The waste is sufficiently con- centrated at a constant volume that scale (salt cake) forms on the walls of the pot. As calcination continues, the scale grows in thickness and reduces the capacity for heat transfer from the pot wall to the boiling sludge; therefore, the feed rate must be reduced proportionately. When the feed rate is reduced to an "unprofitable" rate (about 5 liters/hr), L-35 the feed supply is shut off. At this point, the scale has grown inward from the pot wall and upward from the bottom of the pot to fill the pot, except for a thin-cone-shaped liquid-containing void in the upper 2 to 3 ft of the salt cake. Heating is then continued until the liquid is boiled to dryness and all of the waste in the pot has been calcined and has reached the temperature of 850 to 900°C. The pot is then cooled in the furnace, removed, sealed, and sent to storage. The product from pot calcination (i.e., the solidified waste) is a mixture of the oxides (and sulfates, if sulfate is present in the waste) of the metallic constituents in the original liquid waste. The product is a porous, friable calcine with & low thermal conductivity and a rela- tively high solubility in aqueous solutions. The basic items of equipment required for pot calcination are: (1) a multiple-zone furnace for heating and cooling the calcine, (2) a pot for calcining the waste, and (3) an off-gas line from the pot to the first process condenser which can be washed down continually. The successful performance of pot calcination equipment fulfilling these requirements has been demonstrated, using full-level wastes in the Waste Solidification Engineering Prototypes. Because the pots serve as the processing vessels, they are exposed to severe corrosion conditions during clacination; therefore, they must be made of corrosion-resistant material. Corrosion of type 30LL stainless steel was found to be negligible during processing (< 0.0003 in./day). The pots must be equipped with liquid-level and temperature-measurement devices. Liquid level may be measured with a standard gas-purged dip tube or with an interhal temperature sensor located near the top of the pot. In demonstration tests, temperature measurements were taken, with in-place thermocouples, at the center line and at the pot walls in each zone. Because of the significant cost of thermocouples, an incentive exists for either reducing the required number of these devices or for making them reusable. Internal heat from the decay of radioactive constituents reguires slight modifications of operating techniques. When internal heat is L-36 present, the pot wall must be cooled before the material at the center of the pot has reached its final maximum temperature; if cooling is not available, the center temperature will exceed that desired. (Higher temperatures result in severe corrosion and potentially undesirable vola- tility of some constituents.) Control of this temperature has been success- fully demonstrated by using a simple three-step reduction of furnace tem- 6 peratures, based on pot wall and center temperatures. 2 The pot calcination cycle may be divided into three major periods: (1) feeding and concentrating the waste at a constant feed rate, (2) pot wall scaling and calcining, which cause a gradual reduction of feed rate, and (3) calcining and cooling when the feed supply is turned off and the calcine is heated to 850 to 900°C and then cooled in preparation for re- moving the pot from the furnace. Typical time requirements for the steps are summarized in Table }.5. Since the diameter of the pot has a rela- tively small effect on the overall processing capacity, an increase in capacity must be obtained by effecting changes in pot geometry (e.g., by use of annular pots) or by multiple pot lines. Table 1,.5. Time Cycles and Capacities for Pot™ Calcination62 Pot diameter, in. 8 12 Pot height of fill, ft 6 6 Volume of calcine, liters 60 120 Feed volume, liters 500 1000 Initial feed rate, liters/hrb 30 60 Time at initial feed rate, hr’ 10 10 Time at reduced feed rate, hr 20 30 Calcining and cooling time after feed is turned off, hr 10 30° Total time cycle, hr L0 70 Overall cycle capacity, liters/hr 12 14 Equivalent waste processing capacity,d metric tons/day with Feed Concentration = 378 liters/tonne 0.75 0.9 ®For pots with an internal heat of 5 kw. bFor feeds relatively free of foaming tendencies. CEstimated; exact data not available. dWith a feed concentration of 378 liters/metric ton. h-37 Longer pots provide a slight increase in capacity because there is no marked increase in calcining and cooling times for such pots. Maximum boilup rate is limited by entrainment in the upper part of the pot, or by the cross sectional area of the pot. Some feeds may contain signifi- cant amounts of foam-making constituents (e.g., dibutyl phosphate from the reprocessing plant). If foaming is present, feed rates during the initial boiling period must be reduced from those shown in Table [.5. During the pot calcination of Purex wastes, ruthenium is volatilized to the extent of about 5% and 10 to 30% for low-sulfate wastes and high- sulfate wastes respectively.62 Lower volatility can be effected by the 63 addition of chemical reductants, such as nitric oxide or phosphites. When sulfate is present in the waste, less than 2% of it will be volatilized from the calcine if the chemical composition of the feed is adjusted in such a manner that the chemical equivalent of alkali or alkaline-earth metallic ions is present. In practice, sodium and/or calcium nitrates are usually used. The volatility of cesium and rubidium, which are always present as fission products, can be virtually eliminated by adding enough sulfate or phosphate ions to the feed to be chemically equivalent to the total amount of alkali metals present. When the pot calciner is operated on a reasonably conservative basis, entrainment from it corresponds to approximately 0..% of the total feed.62 Spray Solidification. — Spray solidification is a continuous proc- ess that has been extensively developed and is approaching readiness for commercial use. It is currently being demonstrated with full-level radio- activity on a pilot-plant scale. The spray solidification process was developed at Battelle-Northwest. Its advantages are: (1) it is a con- tinuous process with low hold-up volumes, (2) it is adaptable to a mod- erately wide variety of feed compositions, and (3) it produces a variety of good-quality solids. Its disadvantages are: (1) it is a moderately complicated system, (2) it requires good flow control of sometimes difficult-to-handle feed solutions, (3) its performance requires high- quality atomization, and (/) at present, it requires the use of a rela- tively expensive platinum melter., Results obtained from current pilot L-38 plant tests of melting the calcined powder in the receiver pot, rather than using an expensive platinum melter, may eliminate one of the dis- advantages. In the spray calciner (see Fig. 4.3), liquid waste (which contains some or all of the melt-making additives) is fed through a pneumatic atomizing nozzle into the top of a heated cylindrical tower. The atom- ized waste is sequentially evaporated, dried, and calcined to a powder as it falls into a continuous melter (below the calciner), where it is melted at temperatures of 800 to 1200°C. Process gases from calcination flow into the adjacent filter chamber, carrying much of the calcined powder as dust. The dust collects on the porous metal filters as the gas passes through. The dust deposits are periodically blown off the filters by sudden pulses of high-pressure steam or air that is directed backward through the filters by small nozzles. The dislodged dust falls into the melter with the main stream of powder. The molten calcine flows through an overflow weir or a freeze valve into the receiver-storage pot below. After the pot is filled, it is cooled in the furnace, sealed, and sent to storage. The product from spray solidification is & monolithic solid that is formed after the melt is cooled. The solid is a tough, microcrystalline, rock-like material having a good thermal conductivity and a moderately low solubility in aqueous solutions. (Glassy solids have also been pre- pared in the spray solidifier, but primary emphasis has been on micro- crystalline materials.) The basic items of equipment required for spray solidification are: (1) a pneumatic atomizing nozzle and a spray tower for atomizing and drying-calcining the feed, (2) a multiple-zone furnace for heating the spray tower, (3) an off-gas cleaning system near the spray tower to remove the bulk of the entrained calcine dust from the off-gases, (L) a continuous melter for melting the powdered calcine, (5) a furnace for heating the melter, (6) a pot for receiving the molten waste, and (7) a multiple-zoned "furnace" for cooling (and possibly heating) the receiver pot. L-39 In the continuous melter, the small amounts of residual nitrate and water present in the calcine are volatilized, and the calcine is melted. To date, platinum is the only reliable metallic material of construction that has been found to withstand the environment of corrosion and high temperature. The capacity of a platinum melter that is 10 in. in diameter 57 Platinum has been used extensively at temperatures up to 1250°C. A spe- and has a 1L-in.-high heated section is 1.7 liters of melt per hour. cial alloy of 50% chromium — 50% nickel is generally satisfactory at temperatures up to 1000°C; also, steels with high chromium and nickel contents, as well as some alloys with a high nickel content, are satis- factory at temperatures up to 900°C, The discharge of melt from the melter has been adequately demonstrated both on a continuous basis, using overflow weirs, and batchwise, using straight-tube freeze valves in which a plug of melt about 2 in. long is melted or frozen to provide on-off flow control. The pot for receiving the molten waste may be made of mild steel if the pot is to be filled with melt by large, rapid, batchwise "dumps" from the melter, or if the pot is to be filled with a melt having a low melting point (less than about 700°C). Mild steel is acceptable since the pots must be heated under most conditions only to the point where the melt will slump; this ensures complete filling of the pots, without forma- tion of stalagmites or voids. (Mild steel pots can acceptably resist temperatures up to about 650°C for several-day periods.) Corrosion of mild steel or stainless steel pots by phosphate melts at temperatures of about 700°C or lower is negligible.éh_66 The spray solidifier concept requires that the sintering point of the calcined feed in the spray tower be higher than the temperature of the walls of the spray tower. Adherence to this limit will prevent gross sticking of calcine to the tower walls. In addition, the melting point of the final powder must be no more than about 900°C (see above). The chemical composition of the feed is then adjusted to fit these limitations. Some or all of the melt-making flux can be added, as a solid, directly 1o the melter to further widen flowsheet and operational flexibility. L-L0 The capacity of a spray tower increases significantly with (1) wall temperature, (2) degree of atomization, or spray drop size, (3) decreased stickiness of the feed, and (L) length and diameter of the tower. The capacity increases by about 30% for each 100°C increase in wall tempera- ture in the normal operating range of 500 to 750°C. Atomizing quality can affect capacity by a factor of 2. The drying capacity during the calcination of nonmelting calcines is about 30% less than that with water; the capacity for the calcination of "melting" feeds is approximately a factor of 2 lower than that for nonmelting calcines. The calcine capacity increases approximately linearly with diameter up to about 2 ft and with length up to about 10 ft. Scale-up factors beyond these size limits are not yet well-defined. The typical capacity for a melting feed in a spray calciner of the size used in Waste Solidification Engineering Prototypes (13 in. in diameter by 6 ft long) is 20 liters of liquid waste per hour. Most of the flowsheets used for spray solidification at Battelle- Northwest produce alkali metal — phosphate solids. These are used pri- marily because (1) they offer a relatively large latitude in chemical composition, (2) they have generally low melbing points (700 to 900°C), (3) they produce melts with reasonably low viscosities (less than 50 poises) at operating temperatures, and (L) the chemically adjusted feed solutions are easier to handle than those of other flowsheets and generally produce homogenous melts, The primary disadvantage of phosphate melts is the associated corrosion rate which is higher than that for other melts such as silicates, borates, etc. With the typical phosphate melts, microcrys- talline solids are formed in spray solidification by adding enough phos- phate to approach the composition of orthophosphate melts (total normality of cations/phosphorus = 2.5 to 3.0). Sufficient alkali metals are added to reduce the melting point to 700-%00°C., Although the flowsheets are not always compatible with the spray solidifier, glassy solids are formed by adding more phosphate to the range of metaphosphate of hypophosphate melts (total normality of cations/phosphorus = 1.0 to 2.0). Calcium is added in excess to melts containing sulfate; the calcium combines chemically with the sulfate and retains it. In some cases, a small amount of aluminum is added to increase the sintering temperature to L=l achieve more efficient operation in the spray calciner. Then, enough alkali-metal and phosphate ions are added to reduce the melting point to about 700°C. Some of these ions are added, in the form of solids, directly to the melter to permit operation of the spray calciner with a melt of a chemical composition having a higher melting point than that of the final melt. With the conditions used in the spray sclidifier, 95% of the sulfate is retained in the final solid. Up to 75% of the ruthenium can be volatilized from the spray calcina- tion step (not during melting) with the phosphate flowsheets.67 This volatility can be reduced by eliminating the melt-making flux from the feed and adding it to the melter, 6 or by reducing the oxidizing potential in the calciner. The volatilities of cesium and rubidium have not been significant in spray solidification flowsheets. Phosphate Glass Solidification. — Phosphate glass solidification is a continuous process that has been extensively developed at Brookhaven National Iaboratory (BNL) and is approaching readiness for commercial use. It is being demonstrated with full-level radiocactivity on a pilot-plant scale. Its advantages are that it is a continuous process and it yields a good-quality glass product. Its disadvantages are: (1) it is a moder- ately complicated system, (2) it requires operation with slurries that are difficult to handle, (3) it cannot retain sulfate in the final solid, and (i) at present, it requires the use of a relatively expensive platinum melter. In phosphate glass solidification, liquid waste that contains all of the melt-making additives is first fed to the evaporator, where it is con- centrated and denitrated, by factors of 2 to 10, to a thick, syrupy, aqueous phosphate slurry.68 The slurry is fed to the continuous melter, where final volatilization of the water, nitrates, and other volatile con- stituents is accomplished; then the resulting material is heated to 1000 to 1200°C to form a molten glass. The molten glass flows through an over- flow weir or a freeze valve into the receiver-storage pot below. After the pot is filled, it is cooled in the furnace, sealed, and sent to storage. L-h2 The product from the phosphate glass process is a monolithic, moder- ately brittle glass that is formed after the melt has cooled. This glass has a fairly good thermal conductivity and a low solubility in aqueous solutions. The basic items of equipment required for phosphate glass solidifi- cation are: (1) a continuous evaporator to concentrate the feed to a syrupy consistency, (2) a means to provide controlled feeding of the syrupy concentrate to the melter, (3) a continuous melter for final evapo- ration and melting of the waste, (L) a furnace for heating the melter, (5) a pot for receiving the molten waste, and (6) a multiple-zoned fur- nace for heating and cooling the receiver pot. The last four needs are essentially identical to those in the spray solidifier. The requirements for the continuous melter are essentially the same as those discussed previously for the spray solidifier. Exceptions are that, in the phosphate glass melter, the net heat transfer requirements are 50 to 100% higher (primarily because of the added evaporation load) and the desired freeboard requirements above the melt level are somewhat 69 The capacity of higher because of the foaming tendency in the melter, a platinum melter that is 10 in. in diameter and has a 1l in.-high heated section is 1.2 liters of glass per hour, or about 3 liters of slurry feed 43,70 The vapor stream from the melter is hot (100 to 600°C) and 0,71 71051 and must be routed through platinum piping until the tem- per hour. corrosive, perature is reduced to about 120°C. The pot for receiving the molten glass is similar to that for the spray solidification process. The low slump point (600 to 700°C) and the continuous viscosity-temperature relationship for the phosphate glasses permit the filling of pots by the slow continuous dripping of the melt while the pot is heated only to 500—600°C.71 Mild steel can tolerate these conditions during the filling of one pot. The phosphate glass process can readily solidify high-level waste solutions that contain sulfate, but the sulfate is completely volatilized from the melter. In this case, the vapor stream from the melter forms a separate stream of intermediate-level waste. This stream contains all the L L-L3 sulfate, and normally about 30% of the nitrate, 5 to 10% of the radio- ruthenium, and less than 0.5% of all other radioactivity that was origi- 70 nally in the liquid waste stream. Because the sulfate cannot be reused, it requires special treatment for final disposal. When sulfate is not present, the condensate from the melter contains only nitrates and can be combined with the condensate from the denitrator-evaporator; alternatively, it can be condensed separately and recycled to the denitrator-evaporator to reduce the overall off-gas activity from the solidifier to less than about 1% for radioruthenium and to 0.5% or less for all other radionuclides. The chemical adjustments required for the phosphate glass process consist mainly of adding phosphoric acid to the feed to obtain a meta- phosphate melt (total normality of metal ions/phosphorus = 1). The con- centration (mole %) of the oxides of the alkali metals is maintained at about one-half of that of the total metal oxides in the melt in order to obtain a glass that forms at a reasonable temperature (850 to 1000°C), melts at a low temperature (650 to 700°C), and has good handling properties. The solids in the chemically adjusted feed to the denitrator-evaporator are gelatinous and are readily suspended. Concentration in the denitrator- evaporator sometimes progresses through stages of foaming or heavy crystal- line deposits at lower than, as well as higher than, normal concentration factors.72 These conditions must be defined for each flowsheet. Fluidized-Bed Solidification. — The fluidized-bed solidification proc- ess that has been extensively developed for use with aluminum nitrate and zirconium fluoride - aluminum nitrate wastes. Development of this process was initiated at ANL, and has been extensively demonstrated by Idaho Nuclear Corporation., It has been extensively demonstrated with moderate radio- activity levels in production-scale equipment since 1963, and is now ready for commercial application. Development with the more complex Purex wastes has been limited. The advantages of the fluidized-bed process are that it is a continuous process with a relatively high capacity for a given equip- ment size, and the solidified waste product is readily transportable by pneumatic means. Its disadvantage is that 1t is a moderately complicated system. L-Lh In fluidized-bed solidification, liquid waste 1s continuously con- verted to granular solids by being heated in a fluidized bed of the solids, and the solids are continuously withdrawn from the calciner to storage bins (or the solids may be further converted to monolithic forms). The liquid waste is injected through pneumatic atomizing nozzles into the side of a heated (LOO to 600°C) bed of granular solids. This bed is continuously agitated (fluidized) by sparging gas upward through the fluidized-bed reactor. Contact of the waste with the hot, granular bed results in evapo- ration and calcination of the feed as coatings of the bed particles. The calcine that is entrained with the process gases from the calciner is re- moved from the gas stream by cyclone separators or filters, and is then returned to the main stream of particles. The main stream of particles is continuously removed from the reactor and transported to storage bins, The product from fluidized-bed solidification is granular, with a mean particle diameter of about 500 pm. The granules may be composed of crystals or amorphous solids. The granules are generally spherically shaped, and are moderately soft and friable. The thermal conductivity of the bulk calcine is relatively low. The basic items of equipment required for fluidized-bed calcination are: (1) an atomizing nozzle and a reactor for atomizing and calcining the feed, (2) a means for heating the bed of calcine in the reactor, (3) an off-gas cleaning system located immediately downstream of the fluidized-bed reactor to remove the bulk of the entrained calcine dust 73 from the off-gases,'~ and (L) a storage container for the calcined solids. The heat for calcining must be provided in such a manner that the maximum temperature of the heat-transfer surface is less than the sinter- ing point of the calcine, and the heat must be distributed in such a manner that is can be absorbed by the needs of the reactor. For small reactors (less than about 12 in. in diameter), the heat has been provided solely through the walls of the reactor, using conventional electric heat- ing sys’t:ems.m_76 For larger reactors (and for some smaller reactors), additional heat has been added through heat-transfer surfaces inside the reactor bed to provide better heat distribution. Liquid NaK has been satisfactorily demonstrated as a heat-transfer fluid at the Waste r L-L5 Calcination Facility (WCF), located at Idaho Falls, Idaho, and the com- bustion of gases is being investigated as an alternative heating method.77 For high-level wastes having high rates of self-heat generation, the fluidized-bed system requires a means for cooling the contents of the bed or for dumping the bed during shutdown periods. Such provisions will eliminate the potential for self-overheating of the bed when the flow of feed to the bed is terminated. Containers for fluidigzed-bed calcine may be individual pots, as dis- cussed previously, or they may be large slab or annular containers, as demonstrated at the WCF.78 by air or water circulating around the outsides of the concentric annuli The latter geometry provides for heat removal between the concentric storage bins. Thus far, containers for storing fluidized-bed calcine have been made of stainless steel; however, mild steel could possibly be used if air cooling were provided. The fluidized-bed process has been amply demonstrated in the WCF with aluminum nitrate and aluminum nitrate — zirconium fluoride wastes 77,79,80 having moderately high radioactivity levels. The relatively limited development with Purex wastes indicates that the calcination of such wastes by fluidized-bed calcination is expected to be successful.TS’Bq Purex wastes are less amenable to processing than aluminum wastes because of their greater solubility in the feed solution, their relatively high decomposition temperature, and the low melting point of the sodium nitrate in the wastes. Although these characteristics cause increased agglomera- tion of particles and increased formation of lumps around the nozzle, the formation of agglomerates can be controlled by impingement air-jet grinding, variations in fluidizing gas rates, and simple modifications to commercial 75,81 atomizing nozzles. The volatility of ruthenium from aluminum nitrate wastes varies from less than 1% at 550°C to greater than 90% at 350"0,75 the WCF during operation at hOO"C.80 The addition of chemical reductants and averages 0% in greatly reduces the volatility of ruthenium. For example, the volatility from Purex wastes at 500°C was about 70%, but was reduced to about 1% when 75 sugar (a chemical reductant) was added to the feed. L-L6 Sulfate is retained (greater than 99%) in the fluidized-bed calcina- 75 Corrosion is controlled in the fluidized-bed calcination of zirconium tion of Purex waste. Fluoride is also retained (99%) with the calcine. fluoride — aluminum nitrate wastes by adding calcium in stoichiometric 59 equivalence to the amount of fluoride present. Characteristics of Solidified Waste. — The three conditions that will determine the desirable characteristics of solidified waste are: (1) in- terim storage, (2) transportation to long-term storage, and (3) long-term storage. The basic criterion is that radicactivity beyond safe limits is not permitted to enter the human environment. The desired character- istics of solidified waste with primary importance are: (1) high thermal conductivity, (2) low leachability by water (or possibly air), (3) good chemical stability and radiation resistance, ();) mechanical ruggedness, (5) noncorrosiveness to container, (6) minimum volume, and (7) minimum cost., The net effect of high thermal conductivity is to increase the allow- able heat-generation rate in a pot. This characteristic also reduces the amount of time that liquid waste must be stored before solidification and permits possible reductions in the volumes for solidified wastes. These effects are summarigzed in Fig. L.L for values that are typical for wastes from thermal reactors. The heat-generation rates of wastes from processing the mixed core and blanket fuels of future fast reactors will not be sig- nificantly different from those in Fig. L.l after the first half-year of decay. Low leachability of the solidified products is desired in order to minimize the amount of contamination resulting in any water that might contact a breached container of solidified waste. The leachability and other characteristics of wastes solidified by the processes developed in the United States are shown in Table l;.6. In the case of the best solidi- fied waste materials produced to date, less than one-millionth of the radionuclides are leached per unit specific surface per day. On the first contact of melt-solidified waste with water, the leach- ability is relatively high; then, over a period of 10 to 50 dags, it decreases by about a factor of 10 to a relatively steady rate. i " L-L7 ORNL DWG 68-5840R1 N o — | 1 1 I 1 { | i3 | Bases: Solidified Woste Volume =1 ft>/ton. Thermal Conductivity Units for _ | Solidified Waste are Btu hr 'ft °F > - | o T Maximum Power —emmin 6-in.-diam Pot Maximum Power in "THERMAL POWER IN WASTE FROM 1 METRIC TON FUEL (kw) S5 {2-in.-diam Pot : 4t 3 45,000 Mwd /ton at 30 Mw/ton 2 _—~ 20,000 Mwd/ton at 15 Mw/ton — | ’/ — 0.l | L1 | | 1 ] | 1T 1 | | "0 l 2 3 4 5 6 7 8 9 [0 i 12 13 DECAY TIME( years) Fig. 4.Lh. Effect of Thermal Conductivity upon Maximum Thermal Power Which May Be Stored in Containers of Solidified Wastes. Table 4.6, Characteristics of Solidified High-Level Waste Pot Spray Phosphate Fluidized- Calcine Melt Glass Bed Calcine Form Calcine cake Monclithic Monolithice Granular Description Scale Microcrystallinea Glass Amorphousb Chemical composition, mole % Figsion product oxides 15 to ~ 80 S to 30 5 to 25 g to 50° Inert metal oxides 10 to 50 Lo to 50 10 to 30 10 to > 0 Sulfur oxides (if in waste) 0 to L0 0 to L0 0 0 to L0 Phosphorous oxides ~ 0 25 to 40 ~ 60 ~ 0 Bulk density, g/ml 1.1 to 1.5 2.7 to 3.3 2.7 to 3.0 1.0 to 1.7 Thermal conductivity, 0.15 to 0.25 0.4 to 1.0 0.4 to 1.0 0.10 to 0.25 Btu/hr~* f471 °F-* Maximum heat, w/liter 85 205 190 70 soligd Leachability in cold water, 1.0 to 1077 1079 to 10-8 10°% o0 1077 1.0 to 1677 g/em™® day~* Hardness Soft Bard Very hard Moderate Friability Crumbly Tough Brittle Moderate Residual nitrate, wt % < 0.05 < 0.005 < 0.005 < 4.0 of product Volume, liters/1000 Mwd (thermal) Tt 2.5 1.2 to 3 1.5 %0 § 1.5t 5 Maximum stable temperature, °C ~ 900 Phase separation Devitrifies ~ 600 at ~ 500 at ~ 500 Container material Stainless steel Mild steel or stainless steel Mild steel or stainless steel Mild steel or stainless steel aGlassyproducts can also be made with some difficulty. bMicrocrystalline products can also be made. cComposition ranges for fluidized bed are also for Purex waste and are estimated, dApproximate values for storage in air in 8 in,-diam cylindrical pots to maintain pot center- line temperatures at less than 900°C and pot wall temperatures at less than }25°C. Average k values were used. 811 L-L9 The chemical stability and radiation resistance of solidified waste are important for two reasons. First, they ensure that gases, which may significantly affect the integrity of the product (or container, if pres- ent), are not generated during storage. Second, they ensure that the basic structure and properties of the solidified waste are known. Expe- rience to date indicates that the formation of gas from solidified waste in enclosed containers is generally not significant if the storage tem- 78,82,85-87 perature does not approach processing temperature. However, a few exceptions have been indicated for calcine prepared from feeds with a high sodium nitrate content (nitrogen oxide volatility)86 and for some phosphate-sulfate melts (sulfur oxide volatility).88 Some "nonvolatile" constituents have been found to volatilize at temperatures above process- ing temperatures. For example, at 800°C, significant volatilization of cesium and ruthenium occurs from alumina solids prepared by the fluidized- bed process;78 at 1200°C or higher, boron is volatilized from borosilicate glasses, and some phosphate is volatiligzed from phosphate m_elts.82 The basic structure and chemical properties of solidified waste will change with time because about 15% of the fission products present after 6 months out of the reactor will eventually decay to other chemical ele- ments. For calcines, this 15% represents up to 10% of the oxides present in the total waste; for melts, it represents up to 5% of the oxides present. A clear definition of these changes with regard to properties and their effects is not well known. Some glasses will devitrify to microcrystalline structures if held at L0OO to 800°C for days or weel{s;51’82’83’85’89’9O calcined alumina granules change from amorphous to crystalline form;78 some volatile constituents migrate from thermally hot locations and con- dense at cooler locations,78 and phosphates and other glasses sometimes 85,89 exude liquids. Mechanical ruggedness of the solidified waste package is desirable, primarily during transportation. In the event that the container is breached, the ruggedness of the solidified waste is important in terms of its tendency to be dispersed. A waste that has low leachability, but is very brittle or easily scattered, may contaminate the environs to the same degree that a physically rugged waste with a higher leachability would. L-50 The corrosiveness of the solidified waste to the container deter- mines, in part, the life of the container. Corrosion of containers by solidified wastes has indicated no problem areas in limited measurements to date;éh_66 however, very long-term effects have not been evaluated. The useful life of the containers is expected to be much longer than the 15 to U0 years for containers for liquid wastes.” 272 The minimum volume of the solidified waste is important, primarily, for economic reasons. In general, reducing the volume will reduce the size and cost of containers, container storage areas, shipping equipment, and land to be used for storage areas. Minimizing cost, without affecting quality, is an obvious merit. Near Future Technology. — The technology of solidification has prog- ressed to the point that three of the major processes in the United States are being demonstrated with full-activity-level wastes in engineering-scale equipment, and the fourth major process has been demonstrated with lower- activity-level wastes in large-scale equipment for six years. Most of the basic technology has been obtained; nonradioactive development work is nearly completed; and fully radioactive tests are in progress. The status of the ragéoactive demonstration program at the WSEP has been summarized recently, Table U.7. and experimental results from that program are presented in The modest amount of nonradiocactive development work on solidification processes now in progress in the United States is expected to be completed within the next two years, unless new applications arise. A small amount of laboratory-scale flowsheet work for special problems may continue beyond that time. Current work includes that on fluidized-bed calcination at ICPP, phosphate glass solidification at BNL, and spray solidification at BNW. On completion of these studies, basic process and equipment technology will have been developed for general use. Also, during 1970, demonstration of three processes (pot, spray, and phosphate glass solidification) with Purex wastes will be completed in the WSEP. The processes will have been demonstrated using fully radiocactive wastes with thermal power and fission-product contents equivalent to the P & L-51 maximum expected from advanced light—water reactors and fast-breeder reactors. The operation of the fluidized-bed calcination facility at ICPP will continue to convert aluminum nitrate and zirconium fluoride wastes to granular calcine having a thermal power up to about 1 w/liter. Table };.7. Overall Status of Radiocactive Demonstrations at the WSEP as of February 1970 Solidification Method Phosphate Pot Spray Glass Total Runs completed 6 10 1 27 Megacuries solidified L.o 17.5 19.3 N Equivalent metric ton processed 11.3 1.6 12.8 39 Mwd (electrical) repre- sented by waste® 75,000 98,000 106,000 279,000 Metric ton/day rate 0.6-1.0 0.5-0.9 0.3-0.7 - Maximum kw in one pot 5.1 12.7 11.8 153° Maximum w/liter in b 8-in.-diam pot 85 205 195 - Maximum center-line temperature in pot, °C 940 930 8L0 - Iiters of solid/metric ton L0-50 30-65 50-100 - Runs to complete 6 3 0 9 aAssuming 33% thermal efficiency for 20,000 Mwd/metric ton and 45,000 Mwd/metric ton. Prna 6-in.-diam pot, 315 w/liter has been attained. “Potal kilowatts encapsulated to date. In about two years, the fluidized-bed scolidification system at the Midwest Fuel Recovery Plant will be converting Purex wastes, diluted with aluminum nitrate, to solidified, granular calcine having power densities up to about 200 w per liter of solid. L-52 During the next two to four years, the technology obtained from this testing program will be as complete as reasonably possible. The character- istics of the solids generated in the WSEP program will have been measured oL : This is the time period during which the solid is at its highest temperature and and evaluated for the first few years following solidification. about one-half of the total radiation dose is obtained. Measurements will be made on core-drilled specimens from actual solidified wastes. Charac- teristics of solids generated in the ICPP and stored at higher temperatures (about 700°C) will also have been investigated.73 Current datth’92’9S on economics of waste solidification and its management will be updated and well defined. At least a small amount of developmental effort on any process in the nuclear fuel reprocessing industry, including waste solidification, will be required for any specific application that has not been previously demon- strated. The developmental requirements may be limited to laboratory tests; however, because of the high degree of reliability needed in the nuclear fuel reprocessing industry, a short demonstration program in pilot plants is frequently warranted. ;.3.3 Interim Storage of Solidified Wastes Conceptual designs and cost estimates have been made for the storage of solidified, encapsulated power-reactor-fuel wastes for periods up to 30 years in water-filled canals, in air-cooled annular bins, and in air- cooled concrete vaults. While the bins and vaults were not characterized specifically as facilities for either "interim storage" or "permanent disposal,' the systems, as conceived, should probably be considered suit- able for storage over decades rather than centuries. The interim storage of solid wastes can be accomplished safely in much less complex and less expensive systems than those required for the storage of the corresponding liquid wastes, although equivalent amounts of decay heat must be dissipated in all facilities having comparable inventories of radionuclides. An essential safeguard to be supplied is a thoroughly reliable and independent backup method for removing this heat in the event of a failure of the primary systemn. L-53 Water-Filled Canals. — Canals were chosen for the interim storage of solids in ORNL waste-management evaluations because of their better heat-transfer environment and because canals present simpler mechanical problems in handling and transfer of the packaged 1;Jas*tes.96 The storage facility (Fig. l.5) consists of a central-facility canal, storage canals, and a service area containing water-cooling and purification equipment. The central facility was designed for receiving cylinders of waste from the solidification plant, and for routing them to the proper canal for storage. It was equipped with bridge cranes of 100- and S5-ton capacities mounted on tracks overhead. The containers of solidified waste were stored upright in a series of 2l-ft-wide canals adjoining the central facility. The depth of the canals, as determined by the thickness of water needed for shielding and by the depth regquired to maintain the cylinders in an upright position, varied from 23 to 28 ft. As an aid in locating defective cyl- inders during storage, aluminum partitions were provided, spaced 8 ft apart, along the lengths of the canals. These partitions would channel the water for purposes of monitoring. The canal water was recycled for demineraliza- tion and cooling, and a structure was provided to house the area. About 500 ft of 2h-ft-wide canals would be required to store the solidified acid wastes that would accumulate over a 10-year period from an installed 23,500- Mw nuclear economy. In the year 2000, it is projected that 7230 ft of ?2li-ft-wide canals will be required for 10-years interim storage of all so- 1idified high-level fuel reprocessing wastes (see Table 3.61). Air-Cooled Annular Bins. — A conceptual design for the storage of granular solids obtained from the fluidized-bed solidification of power- reactor fuel-reprocessing wastes was patterned after the originmal solids- storage facilities at the ICPP.97 The solids are pneumatically transported to nested annular, vented, air-cooled, stainless steel bins contained in underground concrete vaults. The thickness of the bins is dependent upon the volumetric heat-generation rate and the thermal properties of the solids; heat is removed by air that is circulated by forced convection through 2-in.- wide passages separating the annular sections. Air-Cooled Concrete Vaults. — Five 150-ft-long, 28-ft-wide, and 18-ft- high concrete vaults, ventilated through a single stack, were proposed by CYLINDERS IN STORAGE 5-TON BRIDGE CRANE (NEAR WATER LEVEL) 5-TON BRIDGE CRANE (OVERHEAD) ORNL-LR-DWG 75604R1 5-TON BRIDGE CRANE (NEAR WATER LEVEL) - T T ¥ .9 P O ¥ g W v - - L T8 VN v_ . - ’ . . \ * bzl q y = g STORAGE CANAL 1 ' | L LT A b L o | SRS B >ALUMINUM PARTITIONS ' 10000000 CANAL 4 i A == - ,._O_.. T s Fleob—r— .;' CENTRAL e T v ¥ ¥ v d v by ‘.‘ ; » B . & - & - — 4 . -V . ' CANAL "- i v - v . v ¥ - e W © v ‘ . : % ‘ \ " v 9 fi e — |e}, BULKHEAD IN PLACE 1] - 4 FOR DRAINAGE OF - ' | Ik - @ ol I P o Trd;;]fl——,;rj—u——;r‘ = » = ¥ =% ---L-x..,..;u.' / 5-TON BRIDGE CRANE 5-TON BRIDGE CRANE 100-TON BRIDGE CRANE (NEAR WATER LEVEL) (NEAR WATER LEVEL) (OVERHEAD) y =l \ Ve 2 HEAT EXCHANGER FILTER 4 COOQLING TOWER Q DEMINERALIZATION COOLING SYSTEM SYSTEM Fig. 4.5. Concept of Interim Solids Storage Facility. S L-55 the British for storing waste that had been converted into cylinders of glass by the Fingal process.98 In this concept, the vaults are con- structed above grade and equipped with a mild-steel liner, which is surrounded by thermal insulation to maintain the concrete at ambient temperature. Air is circulated by fans to remove decay heat during the early years of storage, with the expectation that natural-draft ventila- tion would suffice thereafter. A facility of the size considered in this concept would contain the solidified wastes that would accumulate from reprocessing 1500 metric tons of Magnox fuel annually over about a 28-year period. To maintain safe storage conditions, ventilation would be required for 200 years. L,.3.4, Disposal of Solidified Wastes in Bedded Salt Formations Background. — In September 1955, at the request of the AEC, a com- mittee of geologists and geophysicists was established by the National Academy of Sciences - National Research Council (NAS - NRC) to consider the disposal of high-level radioactive wasies in geologic structures within the continental United States. This committee proposed storage in natural salt formations as the most promising method for the near future.99 As a result of the recommendations of this committee, a study of the problems of disposing of high-level radioactive waste in salt was begun. Some of the advantages of natural salt formations as repositories for radioactive wastes are: 1. Salt is essentially impermeable due to its plastic properties. 2. Salt is widely distributed and abundant, underlying about 500,000 square miles in the United States (see Fig. L.6) and 1 1 with known reserves greater than 6 x 101° tons. 00,10 3. The cost of developing space is relatively low as compared with other rock types. i. The heat-transfer properties of salt are good as compared with other rock types (k = 2.5 Btu hr-* ft~1 °F~2 at 200"}?).102 5. Salt formations in the United States are located in areas of low selsmicity. ORNL-DWG 69-7130 — F— 7 . /’T’;‘( / /// LA /] ROCK SALT DEPOSITS IN THE UNITED STATES (SEE PIERCE AND RICH, U.S.G.S. BULL. 1148) ~ SALT DEPOSITS HAVING THICKNESS OF AT LEAST 200 ft AND LYING WITHIN 2000 fi OF LAND SURFACE Fig, L4.6. Salt Deposits in the United States, and Those Portions Suitable for Initial Waste Disposal Facility. 95-1 -Hild L] ik L-57 6. The compressive strength of salt is similar to that of concrete, or about 3000 psi. Barly investigations in the laboratory and in the field were aimed at the disposal of liquid wastes. - O This approach was prompted by the fact that, while processes for converting aqueous fuel-reprocessing wastes to solids had been proposed, they were, at that time, in a very early stage of development. In December 1961, the NAS-NRC committee met at AEC's Savannah River Plant to discuss progress made since the 1955 meeting and to make recom- mendations regarding future work. The conclusions and recommendations of 108 the committee at this meeting were: "that experience both in the field and in the laboratory on disposal of wastes in salt have been very productive, well- conceived; and that plans for the future are very promising. The Committee noted that the interpretations relating to disposal in salt are by the very nature of salt deposits capable of being extrapolated to a considerable degree from one deposit to another...;" {and] "that the effect of storing dry packaged radioactive wastes in a salt deposit be tested, and urges the Atomic Energy Commission to consider using, at an early date, Federally controlled land in the Hutchinson area." Following this meeting, the AEC requested that ORNL consider the possibility of testing or demonstrating the disposal of high-level radio- active solids in a salt mine, using whatever sources might be available. Consequently, several radiation sources were examined, with the final choice being fuel assemblies from the Engineering Test Reactor (ETR) at the National Reactor Testing Station (NRTS). In 1962, a preliminary study indicated that it was feasible to use irradiated fuel elements to establish the practicality of using salt for waste disposal. In this way, it would be possible to demonstrate the practicability of the disposal-in-salt concept before significant gquanti- ties of solidified waste would be produced. Early in 1963, discussions between ORNL and the AEC led to a decision to extend considerably the scope of the demonstration as conceived in the L-58 preliminary study. This extension in scope made it possible to obtain a vast amount of additional information on the deformational properties of salt at elevated temperatures, which would be valuable in the design of an actual disposal facility, and to demonstrate the use of prototype waste-handling equipment. Major Conclusions from Studies of the Disposal in Natural Salt Formations. — The operation of Project Salt Vault (an experimental dis- posal of high-level radiocactive waste solids in a bedded salt mine at Lyons, Kansas, using Engineering Test Reactor fuel assemblies in lieu of actual solidified wastes) successfully demonstrated waste-handling equipment and techniques similar to those required in an actual disposal operation.109 A total of about L million Ci of fission products in 21 containers, each containing an average of about 200,000 Ci, was trans- ferred to the disposal facility in the mine and then returned to the NRTS at the end of the test. No hot cells were used at the mine; and, even under these conditions, the maximum personnel exposure was only about 200 mrads to the hands and head. In an actual disposal facility, hot cells would be required since the waste containers will offer only single con- tainment. (The fuel cladding and the sealed canister was considered as double containment.) During the 19-month operation of the radiocactive phase of the demon- stration, the average dose to the salt over the depth of the fuel assembly container holes was about 8 x 10° rads, and the peak dose was about 10° rads. The dose decreased very rapidly with distance out into the salt; for example, the dose at 6 in. into the salt was only about 102 rads. As anticipated from the laboratory studies, no significant radiation effects were detected. Theoretical studies indicate that some free chlorine should be radiolytically produced within the salt structure; however, as predicted from laboratory studies, no detectable quantities of chlorine were re- leased. Small quantities of what is believed to be a radiolytically pro- duced organic peroxide were detected when the salt temperature exceeded 175°C, but this is expected to be of no consequence in an actual disposal operation. Although ultimate doses to the salt by wastes of the future L-59 may exceed 10'° rads, the mass of salt involved will still be small and no detrimental effects are anticipated. Both theoretical and experimental results indicate that rock salt is approximately equivalent to concrete as an absorber of gamma radia- tion.11o If this is true, approximately 5 ft of solid salt or 7-1/2 ft of crushed salt (assuming one-third to be composed of voids) will give adequate biological shielding to allow unlimited access to a salt mine room whose floor is filled with the most radioactive waste containers of the future. The containers would be located in backfilled holes in the floor, with the tops of the containers at the proper depth and with con- tainer spacing based on heat dissipation calculations. Field tests have indicated that the heat-transfer properties of salt are sufficiently close to the values determined in the laboratory that confidence can be placed on theoretical heat-transfer calcu]_ations.1lIO Calculations to date have generally been approximate and on the conserva- tive side, but more precise calculations are being made by using more sophisticated heat-transfer models. At the beginning of the study of the use of salt for waste disposal, very little was known about the effects of heat on the behavior of salt in mines. It soon became apparent that, due to the unusual quasi-plastic properties of rock salt, there was little hope of developing exact theo- retical solutions for the effects of stress, temperature, and other vari- ables on the behavior of salt in mines. Consequently, the use of model salt pillars was investigated and found to be applicable. The behavior of model pillars at ambient temperatures was found to correlate with observed phenomena under actual mine ccnditions.”1 It was thus concluded that the behavior at elevated temperatures could be extrapolated to mine conditions. This conclusion has been borne out by the field tests. The most significant finding in the field tests regarding the effects of heat on salt behavior is that the insertion of heat sources in the floor of a mine room produces a thermal stress whose effects are instantaneously transmitted around the opening (to the pillars and roof).109 These stresses produce increased plastic flow rates in the salt, and could possibly cause 14-60 - mine stability problems if the roof of the room is very near a shale - layer (a plane of weakness). In the demonstration area, such a shale layer existed at about 2 ft above the ceiling; however, it was found that conventional roof-bolting techniques were adequate to handle the problem. In an actual disposal operation, it is anticipated that rooms would be filled with waste and then backfilled with crushed salt rapidly enough that roof bolts would probably not be required. The combined field and laboratory tests have provided sufficient information on the deformation characteristics of the salt to allow the development of both general and specific empirical criteria for design of a disposal facility in almost any bedded salt deposit. In the course of these tests, it was discovered that small brine- filled cavities (in general, roughly cubic in shape, with sizes ranging from a few millimeters to microscopic) migrate toward a heat source.109 - A typical bedded salt deposit might contain about 1/2% water by volume. Calculations based on theoretical models and laboratory tests of the migration rates, as a function of temperature, were in reasonable agree- ment.112 Based on theoretical calculations, one might expect a total inflow of 2 to 10 liters of brine per waste container hole, which would take place over a period of 20 to 30 years after burial. The peak inflow rate of 200 ml to 1 liter per year per hole would occur about 1 year after burial. This brine inflow rate would be expected to taper off and approach zero after 20 to 30 years. Inflow rates similar to these were observed -~ in the demonstration. The field tests indicated that, once the migrating brine reaches the - crushed salt backfilling the hole, the moisture moves upward and condenses - in the colder regions above the waste containers. Since the upper regions - of the waste containers may not be full of waste, the upper ends of the containers may be located in the condensation zone under some conditions. If this is the case, stress-corrosion-cracking of these portions of con- tainers made of stainless steel may be anticipated. However, container failure would not be anticipated during the relatively short period of operation in an individual room (typically, about 1 month). If the con- tainers are made of mild steel, then only generalized rusting may be L-61 expected, and container integrity should be maintained for an indefinite period of years. Even if some containers do fail, this should not produce any problem since there should be no gas pressure in the containers. In the event that a cylinder becomes pressurized and then ruptures, the 7 to 8 ft of crushed salt above the containers would be expected to act as a filter and absorber for any material released. If some material should manage to escape the hole in spite of the crushed salt, the anticipated operating procedure will prevent ventilating air from coming in contact with personnel after it passes a waste storage room. Provision will also be made to route the ventilating air through an air-cleaning system and up a stack in the event of an activity release. Generalized Concept of a Disposal-in-Salt Facility. — A generalized conceptT13 of a mine facility to dispose of containers of solidified high- level radioactive wastes has evolved over a period of more than 10 years of research at ORNL. The facility is discussed here primarily to intro- duce the mode of operation, the basic elements, and the various require- ments. Such a disposal facility could be located at any suitable place where an area underlain by bedded salt of appropriate thickness and depth is available. One quadrant of this postulated area is shown in Fig. 4.7. Each quadrant or sector around a central shaft complex would be developed and utilized in sequence. Initial development requires outlining the sector with dual corridors in order to maintain a dual ventilation system throughout the operations. Also, one row of rooms would have to be exca- vated before disposal could commence, One ventilation system serves the salt excavation activities, while the other isolated system provides fresh air to the waste-disposal operations. Operations would be conducted in such a manner that the fresh air never passes the front of a previously filled room before reaching areas of active waste-disposal operations. The top of the waste shaft, which is used solely for lowering waste containers into the mine, is contained within a topside hot cell. The shipping casks, containing a number of waste containers, are unloaded in this facility; the containers are inspected, recanned or decontaminated if required, and lowered, successively, into the mine., A second hot cell, located at the bottom of the waste shaft, serves primarily as a radiation ly-62 ORNL-DWG 64-3902R MINING SHAFT 1 WASTE SHAFT e Lze ol lz T ST T oot s ....l.lr.tl..r..rr... 3 .r.IIJ1AI.I.L.‘- . DISPOSAL ROOM: 50 X300 ftiiii TEnanas < St e W Ll Ll Lo o e s e LT rEfi[LFIFILIIlI:lllI:III-y 2600 ft Quadrant of Mine Based on the 1-Square-Mile Concept. Fig. L.T7. iR vy 1i-63 shield and a containment shell. Each waste container is lowered into this hot cell and into an underground waste transporter, which might be similar to that shown in Fig. L;.8. The waste container, enclosed in the lead-shielded transporter, is carried to the currently active waste- disposal room and deposited in a hole drilled in the floor; then the hole is backfilled with crushed salt via remote methods. When the entire floor area of a single room has been filled with waste containers, in a spacing pattern dictated by the heat-generation rate of the waste, the room itself is backfilled with crushed salt obtained from the excavation of the next row of rooms. In this general concept, the corridors connecting the filled waste rooms would also be backfilled, allowing the deformed solid salt to reconsolidate and, in time, the crushed salt, to recrystallize. Recrystallization of the crushed salt at the elevated pressure and tempera- tures would completely isolate the waste materials and thus prevent any possible contact with the environment. Concept of an Initial Repository. — A study was made of the feasi- bility of establishing a repository in salt to serve the nation's needs for the next two to three decades. The availability of solidified wastes from the nuclear power industry and the cumulative space in salt required for their burial after l; years and 10 years of aging are summarized in Table l;.8. These two particular aging periods were chosen as a basis for the study because burial prior to L years after their generation would entail a considerable cost premium, whereas there would be relatively little cost reduction for wastes aged more than about 10 years. In actual practice, wastes may become available at some intermediate age or, what is more probable, in a mixture of ages and container sizes. It was con- cluded from Table /.8 that, for li-year-old wastes, the quantity available for burial will be large enough and will increase so rapidly that a dis- posal facility should be ready to accept wastes near the end of calendar year 1975. In the case of wastes that have been aged for 10 years, the first facility should be ready by 1981. Although the wastes could be back- logged at their source for several additional years before they are buried in salt, it is considered essential to inaugurate a new facility of this nature slowly and to gain experience with waste-handling equipment and ORNL DWG 65— TT9CRI CENTERING FUNNEL '9 LD. WASTE SHAFT ‘ /| s S P S L ’ / - / s / - s /// //////// ///// //////,‘/'/ ) / ,'/ ’/’, ’,//,/ // WINCH FOR LOWERING AND REMOVING ) CANISTER FROM STORAGE HOLE g::’EEL'; B’;g’;g”“““ LEAD SHIELD RINGS CANISTER POSITION INDICATOR TRANSPORTER SHIELD GUIDE TUBE TRANSPORTER SHIELD ~ N’ /——‘—CAMSTER POSITION INDICATOR LOWER TRANSPORTER SHIELD DOORS Fig., L.8. Underground Transporter. o~ sl 4 L-65 Table 4.8. Estimated Quantities of Soliditied Wastes and Uumilative Salt Space Requirements for Nuclear Power Industry® Number of Waste Containers Cumulative Acres of Salt Buried During Year Space Used by End of Year Calendar Reprocessing Wastes Aged For: Reprocessing Wastes Aged For: Tear L, years 10 years L, years 10 years 1975 170 - 1.2 - 1976 270 - 3.2 - 1978 660 - 11 - 1980 110 - 28 - 1981 2000 170 43 1.7 1982 2700 270 65 3 1984 L600 660 120 10 1986 6100 1410 205 25 1938 8300 2700 315 56 1990 10,300 L600 L50 110 1992 12,200 6L00 530 185 1994 14,300 8300 815 285 1996 16,700 10,300 10L0 110 1998 - 12,200 - 560 2000 - 14,300 - 735 2001 - 16,700 - 835 ®Based on an installed nuclear electrical capacity of 11,000 Mw in 1970, 145,000 Mw in 1980, and 735,000 Mw in 200C; a delay of 3 years between power generation and fuel processing was assumed, bAssumes 1 £1% of solidified waste per 10* Mwd (thermal), and that wastes are enclosed in -in,-diam by 10-ft-long containers. L -66 procedures before undertaking high-volume operations. If 20 years is assumed to be a reasonable operating life, based on economics and obso- lescence, a gross mine area of 835 acres is indicated. In the consideration of siting requirements for an initial reposi- tory, salt domes were eliminated because of inadequate knowledge of their possible behavior (some domes are believed to be undergoing movement), and because some are in contact with circulating ground water, which is known to have caused flooding in at least one mine (see Sect. L.3.2). A suitable bedded salt deposit should be at least 200 ft thick and lie between 500 and 2000 ft below the surface. To ensure long-term stability, a considerable thickness of shale or some other impermeable rock should overlie the salt, and the excavation itself should be located well within the salt deposit. The maximum depth is governed by mine stability condi- tions during the operating period and cost considerations (costs increase at greater depths due to shaft length and the increased amount of salt that must be left as support pillars)., For example, a disposal facility at a depth of 500 ft (about the minimum desirable depth) could be operated at a cost about 5 to 7% less than a similar facility at 1000 ft. At 1500 ft, the operating costs would be 15 to 18% more than at 1000 ft; at 2000 ft, the cost would be 25 to 33% greater than at 1000 ft. The four areas known to meet these criteria are shown in Fig. l;.6. The largest area (about 10,000 square miles) lies in central Kansas; two smaller areas are in Michigan; and one small area is in west-central New York. Other major siting requirements are: (1) the site must offer means for disposing of excess salt as a "backup" in the event that the salt can- not be routinely marketed, (2) it must not be adjacent to large population centers and related high land values, (3) it must be accessible by rail and highway, and (L) it must be acceptable to public officials and private citizens of the area. For purposes of cost estimation, a mine depth of 1000 ft was selected and disposal costs were calculated for l- and 10-year old wastes over mine operating periods of 1975-1995 and 1981-2002, respectively. Two possibili- ties were considered: (1) that a new mine was developed especially for this purpose, and (2) that initial phases of the operation began in an L-67 existing, inactive mine. 1In all cases, costs were escalated to 1971 levels, and 5% was used as the cost of money. For a new mine, the initial capital outlay required before start of operations was about $17.5 million; in contrast, the initial outlay for a facility starting in an existing mine was only about $70.5 million. The total costs over the entire period ranged from $91 - $95 million, for use of an existing mine, to $101 - $106 million for a new mine. The corresponding costs, in terms of kilowatt-hours of electricity produced, ranged from 0.0055 to 0.0067 mill/kwhr. 4L.3.5 Disposal of Solidified Wastes in Rock Types Other than Bedded Salt The widespread occurrence of rock salt throughout the United States has been commonly accepted as one of the principal advantages for the use of these rocks as storage sites for radioactive waste materials. Indeed, salt deposits do underlie portions of 2L of the 50 states; however, from recent laboratory and field studies on the flowage of rock salt at ele- vated temperatures and high overburden loads, it is apparent that many of these deposits are unsuitable for disposal sites.“LL At present, there are perhaps three principal areas in the United States where dis- posal in salt would appear to be highly desirable. These areas are: (1) the Silurian salt deposits of the Northeast, which underlie parts of New York, Pennsylvania, West Virginia, Ohio, and Michigan; (2) the Permian basin salts, which underlie parts of Kansas, Oklahoma, Texas, and New Mexico; and (3) the Gulf Coast Embayment salts, which underlie parts of louisiana, Texas, Arkansas, and Mississippi (see Fig. ;.6). The first two areas are bedded deposits, while the latter contains only salt domes. Most of the other deposits throughout the United States are less suitable because of their great depths below the surface, their numerous inclusions of other rock types, or a general lack of knowledge concerning their extents, depths, etc. In general, mine workings at great depths in salt are initially expensive to open and, due to the greater overburden loads, accelerated deformation of the salt occurs. The presence of other rock types with the salt beds may further accelerate the deformation of L-68 the salt. For instance, in the Williston basin, which covers a part of North Dakota, the minimum depth to salt is 3600 ft; also, the bed is only about 20 ft thick. Thicker deposits occur, but they lie between 4300 and 2000 ft below the surface. Salt beds are also present in Florida, but they are only about 30 ft thick and occur at depths of 10,000 to 12,000 ft. Even in the Permian basin, much of the salt is located at great depths below the surface and contains numerous inclusions of other rocks. Perhaps the principal concern in the disposal of waste in bedded deposits is the stability of the structure at elevated temperatures and stresses. This has been found to be especially significant when shale beds occur interbedded with the salt. These shale beds are usually absent in dome deposits; thus, in this respect, domes may be favored over bedded deposits for waste disposal sites. From recent laboratory and field tests, it appears that efficient and safe operations in bedded rock salt can be designed; however, there are several problems, unique to salt dome deposits, that require investigation before a similar operation can be designed for these structures. More than 300 salt domes are now known to be present in the Gulf Coast Embayment. There are no shale beds overlying these formations, the salt is of a higher purity than that found in bedded deposits, and the domes often occur relatively near the land surface, Many lie between 500 and 1000 ft below the surface; and, of course, the salt extends to depths of many thousands of feet. A large part of the available mined-out space in salt deposits also exists in salt domes. Approximately LO%Z of the total space vacated by rock salt mining each year results from workings in the domes of the Gulf Coast region. The principal technical concern in the disposal of waste to salt domes is in ensuring that migrating waters do not reach the stored waste., The recent flooding of the Winnfield, Iouisiana, dome mine may serve to illus- trate the concern. Also, in Germany, where domal-type salt structures have been mined for many years, at least 20 mines have been reported to have been flooded by groundwaters. At present, apparently 1little is known about the movement of groundwater in the viecinity of salt domes; thus, prior to utilization of these domes for radioactive disposal media, investigations cnof L -69 would have to be initiated to ascertain the geohydrological factors or other parameters that were, or appeared to be, instrumental in the flood- ing at Winnfield and other salt domes. It is obvious that the distance between mine workings and aguifers is important, but it is not possible to state at this time, for example, what the minimum distance would be under specified conditions of mine depth, structural and stratigraphic conditions of the intruded native rocks, etc. Once the important para- meters that bear most directly on mine flooding are identified, laboratory and/or field investigations, if necessary, would have to be initiated to demonstrate that safe and efficient disposals can be made in salt domes. Even though salt is believed to be the most suitable environment for the ultimate disposal of high-level waste, and it is widely distributed throughout the country, it does not underlie any of the major AEC labora- tories and plants that are currently engaged in fuel reprocessing or waste disposal. 1In addition, of the six geologic basins recently dis- cussed by the American Association of Petroleum Geologists for radiocactive waste-disposal potential, only three contain salt deposits.115 Thus, even though high-level wastes could be shipped to areas where salt structures are available for ultimate disposal, they could probably also be safely and economically stored in other rock types that may be available at a given site or areas adjacent to 1it. Dry mine workings are probably not as commonplace as are wet mines in the United States. However, thick and relatively undisturbed beds of limestone and shale (and even granite and other crystalline rocks) exist, many of which are essentially free of circulating water. For instance, 116 in Barberton, Ohio, a dry 2000-ft-deep limestone mine is in operation. Also, mined caverns in chalk near Demopolis, Alabama, have been found to 17 Excavations in thick shale beds in Tllinois have remained dry since they were opened.99 It is reported that be relatively free of Water.’l 2 mine in crystalline rocks in Ontario, Canada, has remained free of water even though the mine is situated directly beneath a large 1ake.99 Loess deposits offer another possibility for the disposal of high-level waste in some areas above the water table.99 Evaporite deposits other than rock salt (e.g., potash, trona, anhydrite, gypsum, etc.) may also be suitable. 4-70 For hard rock, such as limestone and granite, it is expected that mined cavities will remain stable under loads up to several thousand psi and temperatures up to a few hundred degrees centigrade. Recent model pillar tests on samples of dolomite from a local (ORNL) quarry show that there are negligible amounts of deformation in the rock up to loads of 10,000 psi and temperatures as high as 200°C. In comparison, it is of interest that, in similar pillar model tests for rock salt at tempera- tures of 200°C and 6000 psi, pillar deformation had exceeded 35% after ohly 1 hr., Thus it appears that the structural integrity of the exca- vated openings in these rocks, due to the superincumbent load, will not be of primary concern in the event that these rocks should be used as storage media; however, it is likely that such factors as ensuring the isolation of these excavations from migrating groundwaters and the geo- graphic location of suitable deposits and their vertical and lateral extents, along with possible radiation and heat effects on the rocks, would be critical. On a regional basis, it appears that the most promising areas of rock deposits suitable for radicactive waste storage would include rela- tively tectonically undisturbed areas such as the mid-continent region of the United States. Other areas, such as the Colorado Plateau, would also appear to be highly desirable. In parts of the arid west, where there is no groundwater recharge from rainfall and where site locations in rock exist above the water table, suitable excavations may also be practicable., In many localities within these areas, it is likely that horizontal shaft-type or tunneling operations may be feasible. This method of excavation is preferable, in many respects, to vertical shaft mining since it is generally agreed that mining costs are lower and the openings are more accessible. Tunneling into the faces of hills, escarp- ments, or other topographic features of high relief is a common method for mining limestone in many areas where horizontal bedding prevails. Many mines of this type in Middle Tennessee havé been found to be struc- turally stable; and, except for some leakage at the mine entrances, they are entirely free of circulating groundwater. On a larger scale, under- ground excavations in limestone near Kansas City, Kansas are currently ) L-71 being used as refrigerated cold storage bins, Here, facilities have been provided to accommodate the storage of entire rail cars and their refrigerated products in tunneled-out, dry cavities. In summary, it is apparent that dry openings that could be utilized for the storage of radiocactive wastes can be excavated in rocks other than salt; however, investigations are needed to define more precisely such factors as the geohydrological and geotopographical conditions that determine the usefulness of local sites within the most desirable geo- graphic regions and the effects of heat and radiation on the enclosed rock media. li.li Waste Management Technology: Intermediate- and Low-Level Wastes The volumes of intermediate-level wastes obtained from evaporating second- and third-cycle raffinates, product concentration, cell and equip- ment decontamination, solvent cleanup, and off-gas scrubbers, range from 200 to 500 gal per metric ton of fuel processed. They are principally nitrate solutions of sodium, potassium, aluminum, and iron, and often contain sulfate, fluoride, and phosphate in addition. Their activity levels are generally several tenths of a curie per gallon. In the United States, they are stored in underground tanks, sometimes mixed with clad- ding wastes. In the United Kingdom, they are discharged to coastal waters under carefully monitored conditions after suitable periods of decay.H8 At Marcoule, they are partially decontaminated by coprecipitation and coagulat%gn at a pH of about 11.5, using lime, NaH,PO,, A1,(S0,),, and tannin.1 The resulting sludges are mixed with asphalt, packaged in barrels, and stored in protected areas; the decontaminated effluents are discharged to the Rhone River, In addition, a plant may discharge several tens of thousands of gal- lons of contaminated organic solvent wastes annually. These are usuvally either burned or stored in tanks. The low-level liquid wastes from fuel processing are not greatly different, chemically, from natural waters. They contain only very small amounts of inert chemicals and radionuclides in addition to those chemicals L-72 that contribute to natural hardness. The radionuclides of greatest routine importance in these wastes are °°Sr, *°7Cs, *°°Ru, and °H; however, under unusual circumstances of accidental contamination, other fission products, as well as °°Co, U, Pu, and Th, may also be present. These wastes are very large in volume. Evaporator condensates alone may average 10,000 gal per metric ton of fuel processed, and the total generation from all sources within a plant may average several hundred thousands of gallons per day. Because of their great volumes and low concentrations of radionuclides, these wastes have been suitable for environmental disposal. At AEC production sites, where processing plants are located on large tracts of land, ground disposal via seepage basins, cribs, trenches, etc., has been practiced. In these cases, the sorptive capacity of the soils is such that the majority of the isotopes are retained and, in turn, contamination of the ground water is reduced. Each year during the past 5 years, almost 1 billion gal of low-~level waste, containing an average of 35 to U415 kilocuries of radionuclides, has been safely discharged in this manner. In Europe, such wastes receive appropriate treatment for decay or decontamination and are then released to the sea or to rivers. The release of limited amounts of radionuclides to the environment has played an important part in waste management practices to date. It has not been uncommon for low-level liquid wastes to be discharged di- rectly to environmental waters without treatment, depending upon large dilution factors to reduce potential radiation exposure of populations to acceptable levels. A review of these practices in North AmericamO (sum- marized in Tables .9 and [;.10) shows that the quantities of isotopes released have been controlled so that the exposure of people from this source has been considerably less than the limits recommended by the ICRP and other authoritative bodies, The trend, however, is toward relatively less dependence on the environmental disposal of radicactive wastes. This reflects an awareness of the projected greatly increased production and application of radio- isotopes, and of the realization that the pressures of an expanding popu- lation and nuclear industry will make it difficult for "remoteness" to L-73 Table 4.9. Quantities of Low-Level Radioactive Wastes Added to Streams (curies/day) Origin of Wastes Nuclear Power Nuclide Half-Life Chalk R. Hanford Qak Ridge Savannah R. Stations Activation Products a4 a b c Cu 13 h - 2007 -1000 - - - R4Na 15 h 910 200-1000 - - - 7é4s 26 h g 50-300 - - - 238N 2.34d - 200-1000 - 2® - 32p 14 d 0.01-0.1 20-T0 - - - Bl 28 d - 600-2000 - 3.5 Td Fe 15 @ - ¢ - - 74 S804 71 4d - Td - - Td 383 87 d 0.00" 4 - 0.l - &87n 250 4 0.002 30-100 - 0.09 - €00, 5.3y %_0.08 1-2 0.0L-0.2 .01 ¢ 4 12 ¥ 0.1-20 T - 208 - Fission Products 1311 8 d - 1-3 0.001-0.01 0.1 Td 140p, 13 4 - 74 - 1 4 95}, 35 d - - 0.002-0.2 - - 895y £0 d - Td - 0.09 - o5z, 65 a - ¢ 0.001-0.1 0.1 - 1240 285 d 0.005-0.015 - 0.003-0.1 0.1 - L08Ry 1y 0.002 - 1-5 0.03 - 903 28 v 0.001-0.005 0.1 0.02-0.2 0.03 £ 137Cs 30y 0.002-0,02 7 0.01-0.2 0.3 - Total beta {Exclusive of °H) 0.05-15 2000 1-6 (0.5 in river) 1075-0.01 Recelving Stream Ottawa R. Columbia R. Clinch R. Savannah R. Various Flow, Range 3-13 10-75 0.6-2 1.L-8 1-10 109 liters/day avg, 6 27 1 2.5 Measurement point Ottawa R.; Golumbia R. White Qak Storage- Various process sewer; {Pasco) Creek basin waste Perch lake discharge streams a_bWhere a substantial variation is reported, both the low a and high b values are listed. ©(.) Indicates the nuclide is not reported. other nuclides encountered. dT = trace. “Where a yearly average is reported, or there is little variation, only one value is listed. Trallout contributed from 0.1 to 1 curie of €98r per day to large rivers of North America in 1963. It may be present, but in amounts that are trivial in relation to Table L.10. Significance of Exposure from Various Sources Nuclides of Mode Type of Person Percent Greatest of Critical Receiving of a Reference Site Interest Exposure Organ Greatest Exposure Limit Year Limit Chalk R. °03r Drinking water Bone Pembroke resident < Tb ICRP, populaticn at large 32p Fish Bone Fisherman < 0.1 1963 ICRP, Group B (c)°© Hanford 8=p Fish and Bone Fisherman, farmer < 10 ICRP, Group B (c) irrigated crops 76As + 2%¥Np Drinking water G.I. tract Pasco resident < § 1963 ICRP, population + 81gp at large 18171 Drinking water Thyreid Pasco child < 6 FRC,d exposed population QOak Ridge °Cgr Fish Bone Fisherman < BOe ICRP, Group B (c) 205y Drinking water Bone Clinch R. resident® <5 1961 ICRP, Group B {c¢) 10&py Drinking water G.I. tract Clinch R. residentf <5 ICRP, Group B (c) Savannah R. “H Drinking water Whole body Savannah R. residentf < b ICRP, genetic apportionment 1831 Drinking water Thyroid Savannah R. resident’ < 0.3 1963 ICRP, population at large ®08p Fish + water Bone Fisherman < 1P ICRP, population at large Power 89%0 Fish G.I. tract Fisherman < 0.07 ICRP, Group B (¢} #Excludes atmospheric pathways, but includes contributions to the same organ from other man-made isotopes present in the water,. bMost of the ®°Sr contributing to this exposure was from fallout, and not from plant operations. ®Recommendation adopted September 9, 1958. Group B (c) is "members of the public living in the neighborhood of controlled areas.” dFederal Radiation Council {(US) Recommendations (September 1967). ®Assumes that whole fish, including bones, is eaten. fA hypothetical person who drinks untreated river water. No such person has been found. If only the flesh is eaten, the estimate is < 6 per cent. il nL-% L-75 provide the necessary safety factor between the point of waste discharge and the point of population exposure. Present regulations, 10 CFR 20, encourage a minimum of dependence on environmental dispersion and contain a standard clause requiring reduction of the radioactivity in effluents Yo 10% of the continuous occupational MPC before the effluents are dis- charged to unrestricted areas; however, amendment of licenses to permit higher limits is possible if the licensee makes a "reasonable effort" to minimize radioactive discharges and if the resulting exposure of indi- viduals in nearby areas is not likely to exceed 10% of the continuous occupational MPC. Iy.i.1 Treatment of ILiquid Wastes Evaporation, ion exchange, and coprecipitation and coagulation proc- esses are most frequently used for removing radionuclides from low-level wastes; the choice of treatment depends on factors such as the degree of decontamination required, the volume of waste to be treated, and the con- siderations of cost that pertain at the installation in question. Al- though evaporation generally yields the highest decontamination factors (i.e., ratios of the activity in the feed to that in condensates of 10* to 10° are routinely obtained), the cost is in the range of several cents per gallon. Single-stage coprecipitation processes typically remove from 60 to 90% of the radioactivity at a cost of $0.25 to $1.00 per thousand gallons. Ion exchange with either natural minerals or organic resins is frequently used in conjunction with precipitation for additional decon- tamination at extra cost. In addition to partially decontaminated waste water (which can be released to surface waters), each process produces a sludge, a slurry, or a solution containing the separated isotopes. This material is usually packaged and may have to be shipped off-site for burial. There has been an increasing emphasis on research and development aimed at treatment processes that will provide high decontamination fac- tors for the bulk of the waste volume. Such processes will permit envi- ronmental disposal at or near MPC levels, and allow concentration of the bulk of the radionuclides into a relatively small volume, which can be =76 stored or converted to an essentially insoluble solid suitable for dis- posal by burial. Improved scavenging-precipitation methods have been studied, both alone and in combination with ion exchange and other sorp- tion processes. Attention has been given to incorporating the precipita- tion sludges, organic wastes, the ion exchange regenerants, the ion exchange media, and ashes from the incineration of combustible waste mate- rials into low-solubility solid bodies for disposal by burial. In addi- tion, a method, based on the hydraulic fracturing of shale, has been developed for disposing of liquid wastes. Scavenging-Precipitation. — The treatment of low-level liquid waste has usually involved a scavenging-precipitation step, either alone or as the first step in a series. This step includes: (1) formation of a bulk precipitate that contains some of the trace-level radioactive species; or (2) precipitation of a flocculating agent such as ferric hydroxide or aluminum hydroxide to promote separation of suspended solids, precipitates, and colloidal species in the waste; or both (1) and (2). Common examples of coprecipitation include strontium with calcium carbonate or calcium phosphate, and cesium with copper or nickel ferrocyanide.121 Single-stage scavenging-precipitation processes do not usually give high decontamina- tion factors (they are typically 2 to 10, and rarely as high as 100). The actual value obtained depends on the radioactive species, the chemis- try of the precipitation step, and the efficiency of the clarification method. Recent work on improving clarification efficiency includes the use of zeta-potential control to optimize flocculation conditions,122 especially with regard to radiocolloid removal, and the use of an optimum arrangement of filter coal and sand in a polishing filter after floccula- 123 tion and clarification. Inorganic Ion Exchange. — The use of inorganic exchange materials in waste treatment has received considerable attention. This attention can be attributed to: (1) studies of exchange reactions of minerals that have been made in connection with ground disposal of wastes, (2) a desire to use inexpensive natural sorbents that can be disposed of as solid wastes instead of more-expensive synthetic materials, which usually must be re- generated and reused, and (3) an attempt to find sorbents that are highly L-77 selective for particular waste components. The use of vermiculite col- umns by the British to provide additional waste decontamination, espe- cially for cesium, after one or two scavenging-precipitation steps is the "classical! exaTgie of the application of natural exchange materials rial to be developed for sorption columm application is clinoptilolite, 125,126 The addition of to waste treatment. The most promising natural mineral exchange mate- which has been studied extensively at Hanford. grundite clay in scavenging-precipitation steps to improve cesium decon- tamination is another example of the use of natural exchange materials in waste treatment.122 The use of an activated alumina bed to remove phosphate, which otherwise would interfere with the precipitation of calcium carbonate from low-level waste, is an interesting application of a synthetic inorganic sorbent.122 Organic Ion Exchange. — The application of inorganic ion exchange resins to radicactive waste treatment has received considerable study, beginning early in the atomic energy program. However, the use of organic ion exchange in actual low-level waste treatment has not been widely prac- ticed because its cost is typically higher than a standard single-stage scavenging-precipitation process and because the potentially higher decon- tamination factors have not been considered necessary. As a rule, ion exchange resins are too expensive to discard as a solid waste after a single use; hence they are normally regenerated, and the regenerant waste is subsequently treated as an intermediate- or high-level liquid waste. Most ion exchange resins are not highly selective; that is, calcium and magnesium must generally be removed with strontium, sodium must be removed with cesium, etc. The high decontamination factors possible with ion exchange processes are usually based not so much on selective sorption as on the fact that performance corresponding to a large number of transfer units or theoretical stages can be obtained with a single piece of equip- ment . An exception to the low-selectivity rule is the preference shown for cesium over sodium by phenolic-base cation exchangers at pH values high enough to ionize a significant fraction of the phenolic groups. The cesium- sodium separation factor for a resin containing only phenolic exchange L-78 groups is about 160; however, despite somewhat lower cesium-sodium sepa- ration factors, polyfunctional resins such as phenolic-sulfonic and phenolic-carboxylic have more useful capacities for treating wastes con- taining calcium and magnesium. A several-month series of pilot-plant tests of an integrated scavenging-precipitation, phenolic-ion exchange process were conducted at a 10-gal/min scale with ORNL low-level waste. In its final form, the flowsheet included a fluidized-bed alumina column to prevent the interference of phosphate with calcium-magnesium-strontium precipitation. It also included a provision for the recycle of ion ex- change regenerate waste to the scavenging-precipitation step with grundite clay addition. All of the removed radionuclides are concentrated in the clarifier slu.dge.122 The overall decontamination factors varied from 1200 o 12,000 for strontium, 100 to 3000 for cesium, 20 to 700 for rare earths, 10 to 150 for zirconium-niobium, and 1.5 to 8 for ruthenium; the radioactivity of the effluent was reduced to less than 2% of the continuous occupational MPC., Cost estimates for a 750,000-gal/day plant waste treat- ment rate were 60 to BOZ per thousand gallons for this process under vari- ous conditions. Demineralization and Waste-Water Recycle. — High-decontamination- factor processes such as demineralization may yield treated water that is of higher quality than the normal water supply of the waste-producing nu- clear facility. This raises the question of whether the waste water should be reused instead of being discharged to the environment. Burns and Glue- kauf considered three possible alternative schemes and concluded that lim- ited reuse for certain purposes could be justified economically, but that complete demineralization and general reuse were more expensive, at least under the assumed Harwell conditions;jzh however, work concerning the ion exchange and electrodeionization of waste water has been continued on labo- ratory and pilot-plant scales at Harwell. Work at ORNL on a "drinking water" process gave decontamination factors of greater than 1000, 300, 1300, 200, 600, and 25 for Sr, Cs, Co, Ru, Ce, and Zr-Nb, respectively, with all the activities being reduced to analytical background levels when low-level waste was treated successively by: (1) alum coagulation under optimum zeta- potential conditions, (2) ion exchange demineralization, and (3) passage through activated carbon.122 L-79 Insolubilization of Waste Concentrates, — The immobilization of wastes by incorporation into relatively inert solid materials prior to storage or disposal can be advantageous for safety and economic reasons. Liquid wastes, such as ion-exchange regenerant wastes, and solid wastes, such as scavenging-precipitation sludges and incinerator ashes, have been mixed with cement or concrete to give moderately insoluble solid blocks. A 1:1 mixture, by volume, of expanded vermiculite and cement gives a stronger solid, with a lower leaching rate, than is obtained when vermicu- lite 1s not included.121 A substantial volume increase occurs during the conversion of liquid and solid wastes to concretes because of the rela- tively large amounts of cement (and vermiculite) required. A promising recent development is the use of bituminous material to solidify and in- solubilize waste concentrates. This technology originated in Europe and currently is in widespread use there on an industrial scale. A process designed to incorporate all types of organic and alkaline aqueous wastes or slurries in asphalt or polyethylene is being developed at ORNL. This process appears to offer greater versatility and economy than any others developed thus far.122’127 L.L.2 Problems with Tritium Tritium is produced in the fission of *2°U and *°°Pu, with yields of about 0.01% and 0.02% reSpectively.128 It merits special consideration from the standpoint of its management in fuel reprocessing because it is unresponsive to separation and concentration by conventional procedures 129,130 for treating waste. In fuel reprocessing, as much as 25% of the tritium may be released as a gas during the dissolution of metallic fuels, but apparently less than 1% can be expected to volatilize during the 137 There is experimental evidence that tri- 132 dissolution of oxide fuels. tium tends to escape from oxide fuels during reactor operation; however, the tritium remaining with the fuel can be expected to appear as tritiated water in the reprocessing plant evaporator condensates. Based on the projections of Sect. 3.5, the annual generation of fission-product tritium from the Civilian Power Program may be expected to increase from about 36,000 curies in 1970 to about 12 megacuries in the L4 -80 year 2000, Allowing for natural decay, the accumulated quantity should increase from about 36,000 curies in 1970 to about 90 megacuries in 2000. The subsequent discussion is based on the assumption that all of the tri- tium will be present in the fuel at the time of reprocessing. If this tritium could be uniformly dispersed throughout the environ- ment, the resulting increase in background would be of little signifi- cance.133 In the actual case, however, a fuel-processing plant will have only its immediate environs available for dispersion, and the capacity of these environs to accept tritium will depend on the rate that the latter is released, as well as on the many environmental factors that pertain to the particular site. Two immediately available possibilities for the release of tritium- bearing wastes under existing regulations are: (1) dilution and release directly to surface waters, and (2) distillation into the plant off-gas system and subsequent release up the stack. The quantity of tritium that can be released to surface waters can be computed within the limitations that the concentration shall not exceed the permissible concentration in water under 10 CFR 20, or 3 x 107°% yc/cc at the boundary of the controlled zone, and that the concentration shall subsequently not exceed 1 x 107° juc/cc for the general population. If the controlled zone borders a stream of any significant size, the first of these restrictions is controlling. A ton of fuel irradiated to a burnup of 33,000 Mwd contains about 700 curies of tritium, which would require dilution in water to the extent of about 63 million gallons before it could be released from the controlled zone at the permissible concentration of 3 x 107% yc/cc. The total aqueous effluent from a plant operating with a Purex process flowsheet may be as much as 10° gal per ton of fuel processed, but this is far short of the requirements for tritium dilution. The most practical means of achieving the on-site dilution requirement would be to have available, for this purpose, a stream flowing through the controlled area., To meet the speci- fication for use by the general population, this stream would have to flow into a larger body of water to achieve additional dilution by a factor of 3 or more. " 1181 Tt is desirable that a plant be situated adjacent to a large, pref- erably navigable, river for other (and possibly more important) reasons than tritium disposal; however, it is much less obvious that acceptable sites should be limited to those which, in addition, encompass a stream of the size useful for dilution. Therefore, we conclude that, with restrictions as presently interpreted, the alternative of release to sur- face waters is of very limited applicability as a general case. Distillation into the plant off-gas provides a more effective means of releasing tritium. Calculations presented in Sect. 8 indicate that plants having spent-fuel capacities up to 20 metric tons/day and site boundaries two to three km distant can release their tritium in this manner under existing regulations. This is not to imply, however, that attempts should not be made to develop methods for removing tritium, before it becomes greatly diluted with air or process streams, and encapsulating it for long-term storage. L,.;.3 Disposal by Hydraulic Fracturing The study of a method for disposing of intermediate-level radioactive wastes, based on the oil-field technique of hydraulic fracturing, was ini- tiated at ORNL in 1959.13LL 1966. To date (February 1970), 540,000 gal of concentrated intermediate- The first actual waste was injected in December level waste containing almost 340,000 curies of fission products has been disposed of at depths of 360 to 900 ft, well below the zone of circulating water. The method consists of mixing the aqueous wastes with preblended dry solids containing principally cement, and then pumping the resulting slurry down a well and out into a conformable, nearly horizontal fracture in a thick shale formation at the desired depth (Fig. L.9). The cased well is prepared for the injection by perforating the casing at the desired depth and pressurizing the well with water. This induces a fracture in the rocks, which is further extended as the slurry is pumped into it. After the pumping phase is completed, the cement slurry is allowed to harden under pressure, thereby forming a thin, horizontal grout sheet. This pro- cedure can be repeated successively up the well, creating a stack of hori- zontal grout sheets. 1-82 ORNL-DWG 63-3830 DRY SOLIDS STORAGE BINS PUMP HOUSE VALVE PIT EMERGENCY WASTE TRENCH WASYE STORAGE TANKS ) . 2 WELLHEAD CELL Mg “. HIGH.PRESSURE PUMP . S 400 GRAY SHALE s _: 600 900 tIMESTONE BED RED SHALE wm OBSERVATION WELL Fig. L4.9. ORNL Fracturing Disposal Pilot Plant, SCALE IN FEET L-83 The successful application of this method required research and development in three main areas: (1) design, construction, and testing of the plant and equipment, including tanks, bins, mixers, and pumps capable of safely handling the materials; (2) chemical development of mix formulations providing, at minimum cost, a pumpable slurry and a grout offering maximum radionuclide retention; (3) development of an understand- ing of the mechanical behavior of the host rock under the influence of repeated injections and suitable instruments and techniques for monitoring that behavior. The Plant and Its Operation. — Immediately prior to a waste injec- tion, the dry solids are blended and temporarily stored at the site along with the waste solutions. After the equipment has been checked and the well has been prepared, the dry solids and the liquid wastes are vigor- ously mixed, at a constant flow rate, by the jet mixer. The slurry is then pumped to the wellhead, down the well, and out into the prefractured shale. At ORNL this usually requires a pumping pressure ranging from about 1500 to 2500 psi. The jet mixer, the high-pressure injection pump, and the wellhead are enclosed in individual concrete cells to provide shielding and to facilitate decontamination. Development of the Mix, — The cost of the dry solids to be mixed with the waste solutions represents one of the larger fixed expenses of dis- posal by the hydraulic fracturing method. The development of the slurry formulation was, therefore, mainly a search for less-expensive materials and the establishment of the minimum required quantities of these materials. Specifications that had to be met with regard to the slurry were: (1) the slurry should have a viscosity and a thickening time such that the slurry could be pumped and would remain fluid during the entire injection phase, which might last up to 8 hr, (2) the slurry should harden into a grout having at least some physical integrity within a reasonable period, (3) all of the fluid should be taken up during the setting process so that there would be no phase separation, and (L) the radionuclides should be firmly retained in the grout in a reasonably unleachable state. These requirements were met by developing a solids blend, based on Portland cement, which provided the hardening and strength characteristics -8l of the grout sheet. The cement also combines chemically with the radio- strontium in the waste, providing satisfactory retention of that nuclide, Since a high-strength grout was not necessary, the quantity of cement used was approximately 5 1lb/gal, about one-third the usuval concentration. Attapulgite clay was used to prevent any possible phase separation of the slurry as the result of this low quantity of cement. Adeguate pump- ing time was assured by the addition of a small quantity of commercial organic retarder (a sugar, delta gluconolactone). Radiocesium, the major radionuclide in the waste, was retained by the addition of illite (Grundite) clay. Finally, it was discovered that highly siliceous possolanic mate- rials, such as fly ash, could be substituted for part of the cement (up to 2.5 1b/gal) with a further reduction in cost and the added dividend of an improved strontium retention capacity. The formula for the mix was usually modified slightly for each injec- tion because of small differences in the composition and concentration of the waste, but, in general, it met the slurry specifications and provided for about 99% retention of all radionuclides as measured by water-leaching tests. Monitoring. — It was realized from the beginning of the developmental program that the behavior of the shale near the injected grout sheets and the rocks making up the rest of the system would exercise a controlling influence on the general applicability of the method. The rocks overlying the injections provide both shielding and an isolation barrier, the integ- rity of which must be maintained if the method is to be successful. Obvi- ously, it is not possible to continue to inject grout sheets indefinitely, one on top of another, with each injection adding an increment of rock deformation and surface uplift, Monitoring of the operation at Oak Ridge is carried out in several ways. The injection pressures are carefully noted during the progress of each injection. Any departure from the normal pattern would require shut- ting down the operation until a survey could be made. A number of small- diameter cased wells, which extend below the deepest fracture, are logged with a gamma-sensitive probe after each disposal operation; new peaks of activity show where the latest injection has intersected each of these L-85 wells. In this way, the location and extent of each grout sheet may be determined. At intervals, core drilling is used to confirm the informa- tion derived from the logging and to obtain samples of the grout sheet. The continued integrity of the rock cover is tested by periodically attempting to pump water down each of a number of wells that are uncased for an interval of about 100 ft, a little above the depth of the shallowest fracture. At present, each of these wells will accept only a few gallons of water before the pressure reaches the limit (75 psi) of the test pump. Any marked increase in the volume of water that can be injected in this manner would indicate an increase in the permeability of the rock cover. The elevation of each of a widespread network of bench marks in the dis- posal area is determined periodically with high-precision equipment. The normal response of the land surface is to arch up very slightly with each injection, the uplift forming a smooth dome without any marked steps or discontinuities. If the cover rock fails in shear, there should be irregu- larities in the surface uplift. These several methods of monitoring pro- vide a high degree of assurance that the disposal operation is proceeding as planned and that no hazardous conditions are being created. The cost of disposing of intermediate-level waste by hydraulic frac- turing has been estimated, based on the limited experience with the ORNL plant (which was, of course, originally an experimental facility). The total unit cost, including capital investment charges, for a plant of approximately the same size and similar design as the one now in operation at ORNL, disposing of approximately ;00,000 gal/year in 150,000-gal batches, would be expected to be in the range of $0.30 to $0.35/gal. Summary. — Although hydraulic fracturing has been an extremely satis- factory disposal method at the Oak Ridge site, it is not yet possible to consider it without reservation for any other site. Further work is re- quired in two main areas: (1) further development of the understanding of the mechanisms of fracture propagation and the disturbance created in the host rock, and (2) determination of site-testing procedures and accep- tance criteria. O0il-field experience suggests that vertical fracturing is more common than the (near) horizontal fractures required for waste disposal. Since the orientation of hydraulically induced fractures is L-86 influenced by many factors, some of which (e.g., the state-of-stress in the ground at the site) cannot be predicted in advance, it will be neces- sary to conduct site tests prior to adopting this method of waste disposal. The development of improved site-testing procedures, especially with a view toward reducing their cost, is currently in progress at ORNL., Also, a research program to understand and predict the underground behavior of the injected grout sheets is belng continued. ;.5 Waste Management Technology: Solid Wastes This section is limited to considerations of the solid wastes from fuel reprocessing operations other than the solidified high-level raffi- nates from the solvent extraction processes that are discussed in Sect. 4.3. L.5.17 Land Burial Much of the information summarized below was taken from a report that was prepared primarily for those who may be involved in the evaluation and 135 approval of proposed waste burial operations. It contains current in- formation and recommendations regarding commercial waste burial practice. Waste solids that may be radioactive are produced in practically all operations involving the production or utilization of nuclear materials. The low-level solid wastes of greatest volume, for which land disposal is most suitable and advantageous, are designated as "low-hazard potential! and consist typically of paper trash, packing material, broken glassware, clothing, experimental animal carcasses, and contaminated equipment or building material. Table l;.17 shows the volumes of solid waste buried at AEC sites begin- ning with fiscal year 1961. Total volumes of solid waste buried at the commercial burial grounds beginning in 1962 are shown in Table h.12.136 Burial charges have ranged from $1.50 to about $0.70/ft®. Based on current average charges of about $1.00/ft®, this table is a reasonable indication of the size of the market for burial service. The practice of burying solid wastes at selected land sites began very early in the Manhattan District and AEC programs. The possibility that the L,-87 buried radionuclides might be leached, with resulting contamination of groundwater (and possibly of surface water), prompted extensive studies of various types of soils and of burial techniques. DMuch has been learned scientifically and technically of the proper procedures for disposing of solid wastes with maximum safety in various situations. Table L.11. Volumes~ of Solid Radjoactive Waste Buried at AEC Sites'3 From AEC and From Other Fiscal AFC Contractor Government From Year® Operations Agencies® Licensees" Total 1961 2,892,600 20,600 74,400 2,987,600 1962 2,268,200 21,800 68,900 2,358,900 1963 1,698,900 2,500 77,700 1,801,100 1964 1,697,400 2, 700% 15,300 1,715,400 1965 1,454,300 1,454,300 1966 1,413,000 1,413,000 1967 1,800,000 1,800,000 %Values are given in cubic feet. bFiscal Year is from July 1 to June 30. CBuried at Oak Ridge and National Reactor Testing Station (Idaho) under AEC Interim Burial Program. dBuried during the period July-August 1963. ),-88 Table );.12. Volumes® of Solid Radicactive Waste Buried at Commercial Sites Year Jan,-June July-Dec. Annual Total 1962 36,281 36,281 1963 119,069 95,821 214,890 196k 241,660 205,434 LL7,094 1965 258,997 230,982 189,979 1966 261,800 238,172 502,972 1967 380,584 393,266 773,850 1968 321,940 341,630 666,570 1969 306,522 a . . . Values are given in cubic feet. As the nuclear industry developed, certain AEC installations, which had established facilities for the burial of their own wastes, made their burial grounds available for the disposal of solid wastes from industrial users of radioisotopes and from other AEC installations. In 1960 the AEC announced that regional sites for the permanent dis- posal of solid low-level packaged radioactive wastes would be established on land owned by the state or federal government, and sites were desig- nated for this purpose at ORNL and at NRTS. The AEC continued to furnish this service until 1963, when commercial service became available at two locations (Beatty, Nevada, and Morehead, Kentucky) from one company. In late 1969, service was available from two companies operating burial grounds at five sites (Fig. L.10). On-site burial facilities are main- tained by the AEC at Oak Ridge National Iaboratory, the Savannah River Plant, the National Reactor Test Station (Idaho), Hanford, and Los Alamos Scientific Laboratory. ORNL DWG 70-2773 /) RICHLAND,WAS ® NUCLEAR FUEL SERVICES WEST VALLEY,NY, | \ _______,r———-—i NUCLEAR ENGINEERING CO. [ - ! SHEFFIELD, - [ —— \ ’ o jLL. \ - NUCLEAR ENG!NEE ING BEATTY NEV \ —CO. fl / ! \ OREHEAD mill/kwhr, for interim liquid storage as a function of storage time, in years. Interim liquid storage costs are based on the same tank-farm design and capital and operating costs that were used for perpetual liquid storage. The wastes were stored at a concentration corresponding to 100 gal per 10,000 Mwd (thermal), and a tank lifetime of 50 years was assumed. The tank size was optimized for each storage period, and provi- sions were made for the reuse of tanks when possible. Costs for interim 3 mill/kwhr for storage liquid storage ranged from 13 x 107> to 27 x 10~ periods of 1 to 30 years respectively. There is very little difference in cost between a 20- and a 30-year storage period because the same total storage capacity is required and none of the tanks can be reused. Pot Calcination and Shipment. - Figure 6.6 presents pot-calcination and waste-shipment costs as a function of the age of the waste at the time of calcination and shipment. Costs were calculated for calcination in 6-, 12-, and 2li-in.-diam cylinders for every case in which the center-line temperatures of the cylinders were permitted to remain at less than 1650°F when the cylinders were standing in air. The volume of calcined solids was taken to be 1 ft° per 10,000 Mwd (thermal) of fuel irradiation. Cal- cination costs were computed after the costs from the earlier studyl7 had been escalated as follows: permanent facilities, 50%; calcination pots, 30%; labor, increased to $10,000 per man-year; overhead, assumed to be 100% of the labor costs. The calculated costs ranged from 16.5 x 1073 mill/kwhr, for the calcination of l-year-old wastes in 6-in,-diam pots, to 1.8 x 107> mill/kwhr for the calcination of 30-year-old wastes in 2l- in, -diam pots. Costs for 1000-mile shipment of the pots in lead-shielded casks weighing 50 to 90 tons, without forced convection cooling enroute, ranged from 2.3 x lO_3 mill /kwhr for shipment two years after fuel reproc- essing to less than 1 x 1073 mill/kwhr for shipment LO or more years later. In arriving at these estimateg, freight costs were escalated by 20% over those previously used;19 the purchase price of the casks was based on $1.25 per pound of welght; labor costs were increased to $100 per man-day, including overhead; the cost of the loading crane was escalated to $1200 per ton; and a period of 1} days was allowed for a 2000-mile round-trip shipment. Time, COST (10”2 mills /kwhr) 6-33 OCRNL-DWG 69-11851R 35 30 ot/ 15, 10 0 5 10 15 20 25 30 35 INTERIM LIQUID STORAGE TIME (years) Fig, 6.5. Cost of Interim Liquid Storage as a Function of Storage 6-3U ORNL-DWG 69-11852A 16\ 14 \ COST (102 mills/kwhr) a - POT CALCINATION IN \\ 6-in.-diam CYLINDERS \ 12-in. diam \ N 24-in. diam SHIPMENT O ! O 10 20 30 40 50 60 70 AGE OF WASTE (years}) ’/// Vil / Fig. 6.6. Present-Valued Costs of Pot Calcination and Shipment of Solidified Wastes as a Function of the Age of the Wastes. 6-35 Interim Solid Storage. - Present-valued costs for interim solid storage of the calcined wastes in water-filled canals are presented as a function of the age of the waste (specifically, interim liquid storage time) for 1, 3, 10, and 30 years of storage (Fig. 6.7). The costs range from 16 x 107> mill/kwhr (obtained by extrapolation) for 30-year storage of l-year-old waste in 6-in.-diam pots, to 0.5 x 107> mill/kwhr for l-year storage of 30-year-old waste in 6-, 12-, or 2lL-in.-diam pots, For these calculations, the costs that were escalated over those used in the previous study18 are: excavation, $S/yd3; concrete in place, $120/yd3; epoxy lining, $l.50/ft2; aluminum partitions, $lO/ft2; 5-ton crane, $7000; track, $37/ft; demineralizer system, $L50/gpm; Geiger tube detectors, $1200 each; scintil- lation detectors, $2300 each; service and office building, $35/ft2; and a building to house the storage canals, $lO/ft2. The costs of the aluminum stands to hold the cylinders were estimated at $25, $39, and $63 each for 6-, 12-, and 2)j-in,-diam pots respectively. Cooling system costs, which include cooling towers, heat exchangers, and pumps, were increased by 25%. The discontinuity in each of the cost curves of Figs. 6.6 and 6.7 at the 19- to 20-year marks is a characteristic of the particular economic model that is used. In this model, it is assumed that any investments made after the end of the 20-year income-recelving period would have to be financed out of the escrow fund, Investments made during the 20-year period, on the other hand, would be recovered from incomes received during that same 20-year period. Since amny investments outstanding during this period are expected to yield an annual interest of 12.L4%, the incomes used to retire these investments are, in effect, earning at this rate all through the payout period. An investment made in the twentieth year must be re- covered (reduced to zero) at the end of that year. This is done by making use of accumulated annual incomes that have been received during the previous 20 years and that have, in effect, been accumulating interest at 12.1% per year. If these incomes had been accumulating at an interest rate of only 5% per year, the amount of annual income required to retire the same investment would have had to be much larger. The discontinuity comes about because of the assumption that an investment made in the twentieth year can be retired by incomes that have been accumulating for 6-36 ORNL-DWG 69-11853A 13 ] CYLINDER DIAMETER (in.) 12 \ & —_——12 —_————24 " \\ 10 A = 8 £ z X ~ w 7 .é \ 0 IQ 6 |,_ 1) O O 5 4 \ X \ \ \\\\ \ D ,-',;\ INTERIM SOLID \ \k \\\sTORAGE(yecrs) ~ ~ 2 \\ S ] \,.:\ ‘-t:_::‘_:-:} 20 ~J -______ O 0 5 10 15 20 25 30 35 AGE AT TIME OF STORAGE (years) Fig. 6.7. Present-Valued Costs of Interim Storage of Solidified Wastes as a Function of Their Age at the Time of Burial. 6-37 20 years at 12.4% per year, while the same investment made in the twenty- first year must be retired by incomes that have been accumulating at only 5% per year. For example, in Fig. 6.6, a 19-year age for the waste at the time of calcination means that the investment for the calcination facility was made in year 20, while a 20-year age at the time of calcina-~ tion means that the investment was made in year 21. In constructing the model, there seemed to be no simple, valid method of eliminating this discontinuity, which never amounts to more than about 2% of the cost in any event, Disposal in Salt Mines. - The estimated costs for the disposal of solidified wastes in a salt mine are presented in Fig. 6.8. As in the previous study,Zl the cylinders are buried in vertical holes in the floor of a mine that is excavated 1000 ft below the surface., The pots are spaced in such a manner that the decay heat can be dissipated without increasing the temperature of the salt above 200°C. Costs that were developed in the study of an initial govermment-owned repository (de- scribed in Sect. L.3.L) were utilized in arriving at the estimates shown in Fig. 6.8. A cost of $381,000 per acre of the net mine area, including all capital and operating costs and 5% annual interest on money, was an average of several cases considered. To determine the costs of burying a can of waste, the required mine area was first calculated from consider- ations of the heat-generation rate and the age of the waste at the time of its burial. This area was multiplied by $381,000; then the product was converted to mills per kilowatt-hour of electricity originally gener- ated, and present-valued. Disposal costs lie in the range of 0.1 x 10—3 o0 10.9 x 107> mill/kwhr; they increase with pot diameter because the heat is dissipated easier from smaller vessels, thus permitting more efficient utilization of space in the mine, Total Costs of Management., - Minimum total costs for six cases repre- senting different schedules of waste management operations carried out sequentially are summarized in Table 6.8. In addition to pertinent descrip- tive data, the initial capital investment and present-valued unit cost are given for interim liquid storage, pot calcination, interim solid storage, and shipment of the solidified waste. For the salt-mine repository, only 14 12 10 mills / kwhr) COST (1073 N Fig. 6.8. Present-Valued Costs of Disposal in Salt as a Function of 6-38 ORNL-DWG 69-11854R \ 6-in-diam CYLINDERS \ \12-in. diam A \\ 10 20 30 40 20 60 AGE OF WASTE AT TIME OF BURIAL (years) the Age of the Waste at the Time of Its Burial. 70 B Il e 6-39 Table 6.8. Optimal Schedules and Costs of High-Level Waste Management Case No, 1 2 3 L 5 6 Interim liquid storage Storage time, years 20 10 5 0 0 0 Number of tanks, with spare 6 5 5 0 0 0 Tank size, 106 gal 1.05 0.88 0.55 0 0 O Initial capital cost, $lO6 13.2 12.2 10.0 O 0 0 Unit cosb, 107> mill/kwhr 27.7 25.9 22.9 O 0 0 Pot calcination Residual solids volume, £t3/10"* Mwd (thermal) 1.0 1.0 1.0 1.5 1.0 1.5 Solids conductivity, Btu hr-1 ft-1 °F- 0.26 0.26 0,26 0.26 1.h 0,2 Pot diameter, in, 9 12 12 6 9 6 Number of pot lines 5 5 5 10 22 10 Initial capital cost, $1o6 L.l .2 L.l 6.5 8.3 6.5 Unit cost, 107> mill/kwhr 3.7 L.8 7. 2hl 22,7 2h.l Interim solid storage Storage time, years 0 0 5 3 l 1 Initial capital cost, $lO6 0 0 2.8 3.8 3.5 2.2 Unit cost, 107> mill/kwhr 0 0 3.0 6.6 6.3 L.l Shipmenta of solidified waste Number of shipments per year L5 L L5 67 L0 67 Number of casks, with spare 3 3 3 L 3 L Initial capital cost, $1o6 0.4 0.5 0.5 0.8 0.6 0.9 Unit cost, 107> mill/kwhr 0.6 1.0 1.0 3.0 1.7 L.3 Salt mine repository Unit cost, 107> mill/kwhr 2,0 5.0 L.7 5.1 6. 1L.2 Total cost Maximum unrecovered capital, millions of dollars 13,2 12,2 13.9 10.4 11.9 8.7 Unit cost, 107> mill/kwhr 3,.0 36.7 39.0 38.8 37.1 L3.7 22000-mile round trip. 6-110 the unit cost of disposal is given since it is assumed that this is a national facility owned and operated by the govermment on a full cost- recovery, but nonprofit, basis. Finally, the minimum total unit cost, in mills per kilowatt-hour, is given for each case, as 1s the maximum amount of capital that remains to be recovered at any time during the 20 years that income is received. Cases 1 through 3 indicate that the total cost increases from 0.03lL mill/kwhr to 0,039 mill/kwhr as the interim liquid storage time is de- creased from 20 years to 5 years. In case 1, calcination in 9-in,-diam pots and shipment to the repository are carried out after a 20-year ' storage period; these operations are financed entirely by the fund which is established for that purpose. This is the least expensive of any of " the cases that were considered. Case 2 shows that calcination in 12-in,-diam pots, shipment, and disposal can be carried out, after only 10-years liquid storage, for 0.0367 mill/kwhr. However, if the waste is solidified after only 5-years storage, it can be seen from case 3 that an additional S-years storage of the solids prior to shipment and disposal is justified. Because of their high heat-generation rates, these wastes cannot be calcined without some prior storage as liquids unless the residual solids are either diluted with inert material or treated in some manner designed to increase their thermal conductivities. In case I, it is assumed that the volume of residual solids has been increased by 50% [i.e., from 1.0 to 1.5 ft3/10LL Mwd (thermal)] without affecting the thermal conductivity. " The wastes are calcined immediately in 6-in,-diam pots, stored in canals . for 3 years, and then shipped to a salt mine for permanent storage. The . total cost for this schedule of operations is 0,0388 mill/kwhr, about 14% more than for case 1. Results of laboratory studies show that our capability of producing dispersions of calcined solids in a sodium tetraborate matrix is potenti- ally good. These dispersions would have a thermal conductivity of about 1.4 Btu hr™t £yt (°F)-l, and their volume would be no greater than that | of the calcined solids.27 Case 5 assumes the formation of such a disper- , sion in 9-in,-diam pots without prior liquid storage., After l-years 6-41 storage in solid form, this waste is shipped to the repository; storage and shipment result in a combined cost of 0.0371 mill/kwhr. Case 6 represents the most accelerated schedule possible, as deter- mined by the maximum allowable heat-generation ratés of wastes buried in salt. Here, it is assumed that the wastes are solidified immediately [after dilution of the residual solids to 1.5 ftB/th Mwd (thermal)], and then shipped to the repository after only 1 year of interim solid storage. A total cost of 0,0437 mill/kwhr” is estimated for this case. In summary, these costs indicate that the least-expensive management consists of storing the waste in liquid form for the full 20 years that income is received before solidifying and shipping it to the salt-mine repository. Storage as a liquid for only 10 years, followed by solidifi- cation and shipment to the repository, increases the total cost about 8%; and immediate solidification, followed by 3 to L years on-site storage of the solids prior to shipment and disposal, costs 9 to 1L4% more, If for any reason, such as for enhanced safety, the schedule should need to be carried out with the least practical delay, we would expect to pay about 30% more, Effect of Scale., - A consideration of the major components of the costs in the management of high-level wastes indicates that these costs should be dependent on the guantity of fission products handled rather than on the mass of fuel reprocessed. The quantity of fission products can be represented by the burnup of the fuel, and, on an annual basis, it is equivalent to the total number of thermal megawatt-days represented by all the fuel reprocessed during that year. Accordingly, cases 2, L, and 5 were investigated from the standpoint of the total annual waste management cost over the burnup range 1,13 x lO7 to 1.82 x lO8 Mwd (thermal) /year. (These burnups are equivalent to about 340 and 5500 metric tons/year, respectively, of fuel that has been exposed to 33,000 Mwd/metric ton.) The total annual costs for case 2 are represented, within i;O%, by: 0.76 $/year = };.07 x 106 [MWd (theig%l)/yeaf} . 6-42 Case L is represented, within +10%, by: Mwd (thermal)/year}o'85 . $/year = L.05 x lO6 I 7 10 Case 5 is represented, within +10%, by: 0.86 $/year = 3.80 x 106 [MWd (therma%)/year] - 10 A1l costs calculated for the three cases can be represented, within +25%, by: Mwd (thermal)/year}o'8l $/year = 3.97 x lO6 [ 7 10 The 10' figure in the denominators of the above expressions is a normalizing factor, 6.3.4 Comparison of Salt with Concrete Vaults and Granite The choice of a permanent disposal site for solidified high-~level radioactive wastes must be made on the basis of both safety and economic considerations. Although the safety and cost requirements cannot be rigorously defined, the hazards associated with the wastes are of suffi- cient magnitude that provisions for contaimment outside the environment are required, virtually forever. It is implicit that this containmment be effected under conditions requiring a minimum of surveillance and at a cost commensurate with the costs of other items in the power reactor fuel cycle. In the United States, cavities mined in natural salt formations are believed to offer the best possibilities for the permanent disposal of high-level radioactive wastes, However, since a salt mine could be located at a considerable distance from a fuel reprocessing plant, ship- ment of the wastes would almost certainly be required. Suitable deposits of granite or shale might be more accessible to a plant., Also, it is conceivable that high-integrity concrete vaults could be constructed at the plant site to serve the purposes of permanent containment, 6-L3 If we have available, as a point of reference, the more detailed analysis of the cost factors in the disposal of calcined wastes in salt mines,21 a rather perfunctory analysis can show the relative costs for disposal of wastes in concrete vaults at the surface of the earth and in areas excavated from granite formations. In lieu of a formal safety analysis, a qualitative observation can be made that disposal in granite would, at best, be no safer than in salt. In addition, concrete vaults would be less safe because of the I1imited period of integrity of the concrete and the proximity of the waste to the biosphere. Therefore, in order for these alternative methods to be competitive, the costs of mining space in granite should be as low as the costs for salt, whereas the costs for vaults should be lower than those for salt. Costs were estimated for the permanent storage of calcined radio- active wastes in concrete vaults and in rooms mined out of granite forma- tions.22 The costs for concrete vaults were five to seven times as much as the previously estimated costs for storage in salt mines, whereas the costs for storage in granite were only about twice as much., Thus, its economic advantage, as well as the greater safety it is believed to offer, makes salt the preferred choice. While it is possible to design vaults of lower costs than those calculated in ref, 22, it seems unlikely that the costs could be reduced sufficiently to make them comparable to those for storage in salt formations, unless many safety features are sacrificed. Vaults for use in the storage of high-level solid wastes would be similar in their gross features to many of the tanks built for storing liquid radiocactive wastes, in that they would be underground structures of reinforced concrete with an earth cover, To make storage in vaults as safe as possible, we assumed that the vaults were sealed completely from the surface. Space requirements were calculated by assuming that the heat of radioactive decay was dissipated via conduction through the earth cover, Vaults of two types of concrete were considered: ordinary concrete (capable of withstanding temperatures of LOO to 500°F), and "high-temperature concrete" (capable of withstanding a temperature of 1000°F), 6-Ll Space requirements for storing calcined wastes in rooms mined out of granite formations are about the same as those for storage in salt forma- tions. However, mining costs are higher for granite because heavier equipment is required, drilling is more difficult and slower, and costs of explosives are higher, It must be borne in mind that, whereas the costs given in ref, 22 should be valid for comparative purposes, a more subtle interpretation may be placed on them when they are used to optimize the total costs of waste management. For example, disposal in concrete vaults has been estimated to be five to seven times as expensive as disposal in salt formations; however, a concrete vault located at the plant site would eliminate the need for waste shipment, Although the costs for shipping long-decayed waste in the largest-diameter cylinders would be 10%, or less, of disposal costs, shipment of the wastes in smaller cylinders after short periods of decay could amount to as much as 25 to 504 of the costs of disposal in vaults. Again, with a moderate increase in cost, the vaults could be equipped with forced-air convection cooling (i.e., to serve as interim storage facilities) until the fission product heat generation had decreased to a level that would allow the vault to be finally sealed. In this way, interim storage in canals prior to permanent disposal could be avoided. Of course, these same considerations would apply to granite or salt in the event that the fuel processing plant is situated adjacent to sultable deposits of either. As has been previously pointed out, a really meaningful optimization of waste management must include safety, as well as economic considerations. At present, it is believed that salt offers the greatest safety and that the costs for disposal in salt, even allowing for shipment under reasonable conditions, are less than for other alternatives. 6.3.5 Salt Mine Repository Charges A preliminary estimate of the charges for handling and emplacement of containers of solidified waste at a salt mine repository has been developed using the latest information available at the time this report was published. 6-U5 Subject to a minimum charge for handling, the cost for disposal of a container of waste is determined as the product of the cost per unit of floor area of the mine and the area that is required to provide for sufficient dissipation of heat from the container. The characteristic mine area required by a container is determined by the transient thermal power of the container and imposed conditions that are necessary to ensure thermal stability of the mine and acceptable temperatures near the land surface. A unit cost of $9.90 per £1° of floor area of the mine was de- rived on the basis that 245 net acres of burial space would be utilized over the 20-year life of the repository and that the total cost incurred in this period would be $106,000,000. This total cost (in 1970 dollars) provides for full recovery of all capital and operating expenses over the life of the repository, with 5% annual interest on outstanding debt and provisions for a $500,000 fund for "perpetual' surveillance after decommissioning. The following equation has been found to correlate the unit cost data that have been developed in comprehensive studies: * al®) C = 1.174 2 gy, VE 0 where _ C = unit cost for receipt, handling, and disposal of a container at the repository, dollars per container, a(t) = linear thermal power of the container, w/ft, and t = time since receipt of the container at the repository, years. For single radioisotopes (or mixtures of isotopes if proper considera- tion if given to radiocactive daughters), this equation may be expressed as: 6-L46 where q? = injitial linear thermal power of isotope i upon receipt at the i repository, w/ft, and Ty half-1ife for radicactive decay, years. These equations and/or the containers for emplacement in the mine are subject to the following restrictions: (1) Containers shall be right cylinders having outside diameters in the range of 6 to 2L in, and heights in the range of 2 to 10 ft. (2) The minimum cost per container shall be $300, (3) A container may not have transient power density such that the calculated cost is greater than $10,000, (L) The initial linear thermal power of a container shall not exceed 500 w/ft. (5) At any time greater than 5000 years after receipt of the container, the following equation shall be satisfied: q{t) < ¢/1000. Table 6.9 presents estimated costs, assuming that the waste consists of mixed fission products and actinides resulting from the processing of fuel from an LWR. The value of the integral in the cost equation was obtained by summing contributions from individual isotopes. 6.4 Site Costs In the NFS cost estimate, $500,000 was included for the 1300-acre 4 This is less than 2% of the estimate for the total capital investment. If the cost of the site plant site furnished by the state of New York. 6-147 Table 6,9, Estimated Costs for Receipt and Storage of Solidified Fission Product Waste from LWR Fuel at a Salt Repository as a Function of the Age of the Waste® Post-irradiation age of waste, years 1 2 3 5 7 10 15 Thermal power, w/metric ton of fuel 10,320 5200 3490 2130 1540 1100 826 l;%l& fv” alt) dt, dollars/(w/ft) 3.85 5.30 6.55 8.70 10.45 12,74 1h.1 a 0 Wb Initial linear thermal power, w/ft Cost, dollars/container 30 300 300 300 300 310 380 420 60 300 320 390 520 630 760 850 120 L60 540 790 1040 1250 1528 1690 180 690 950 1180 1570 1880 2290 2510 21,0 920 1270 1570 2090 2510 3060 3380 300 1160 1590 1970 2610 3140 3820 4230 360 1390 1910 2358 3130 3760 L4590 5080 120 1620 2230 2750 3650 4390 5350 5920 h80 1850 2540 3140 L4180 5020 6110 6770 500 1930 2650 3280 L350 5230 6370 7050 Cost, dollars/metric ton of fuel Waste in 10-ft-long containers 3970 2760 2290 1850 1610 14,00 1170 ®The fuel is assumed to have been irradiated at an average specific power of 30 Mw/metric ton to an exposure of 33,000 Mwd/metric ton. The waste consists of all fission products plus the actinides remaining after remval of 99.5% of the uranium and plutonium following a postirradia- tion decay period of 150 days. 6-48 were a function of the 0,70 power of the plant size, instead of the 0.35 power assumed for total capital investment, it would contribute about 5 to 6% of the total capital investment we have allowed for a LO-metric ton/day plant. However, unless the noble gases and iodine are removed from the off-gas to a much greater extent than at present, the cost of the reprocessing plant site area will be a function of greater than the 1.0 power of the plant area and, in turn, the site costs could become a substantial fraction of the total capital investment (see Sect, 8). As discussed in Sect. 8, the enhanced removal of noble gases and iodine, improved containment of internal explosions, and earthquake-resistant structures can reduce site size requirements to the point that considera- tions other than health and safety are controlling., The extra cost of these safety features might add 10% to the capital costs estimated in Sect. 6.1, thereby increasing the total reprocessing cost estimates (including waste disposal) by less than 10%. 6.5 Costs of Engineered Safeguards The word "safeguards” has been used to refer to engineered safety features designed to protect the public against the potential hazards of reactor or other nuclear facility operations or accidents. It has also been used to imply inspection procedures for ensuring that fissile materials are not diverted to unauthorized uses. In this report, we have used the term "contaimment systems! instead of "engineered safeguards." We have not independently estimated the costs of "political safeguards," but have assumed that these costs will be paid by the national or inter- national agencies responsible for the policing rather than being charged to the utilities whose fuel is being reprocessed. It has been reported that 29 people would be required to police NFS; on this basis, we can estimate that such antidiversion inspection could increase the total reprocessing cost estimates of Sect. 6.1 by less than 2.5%, assuming that present labor and overhead costs are involved. (No allowance is made for increased capital cost or reduced operating efficiency.) 6-L9 We have not made any recent estimates of the cost of noble gas removal. In an early cost analysis of the Idaho Chemical Processing Plant, the added cost of 99% noble gas removal was less than 3% of the total capital cost.28 New technology is under development, and new cost estimates are needed, particularly for the case of short-cooled FBR fuel; however, we still feel that the removal of noble gases will add only a few percent to plant capital cost. To date, we have not made any estimates of the cost of high-degree jodine removal (by factors of 108 or more in some cases discussed in Sect. 8). However, we plan to make some estimates in connection with our FBR reprocessing development program. Again, our tentative appraisal is that iodine removal will increase the capital cost of the plant by only a few percent. Examples of the incremental cost of earthquake-resistant structures have been summarized by Bell and Lomenick.29 Cost increases due to aug- mented earthquake survival capacity in structures can be divided into those for design and those for construction, Unfortunately, the art of cost-benefit analysis has not yet been established, even for conventional structures. For both conventional structures and nuclear plants, the burden of work on the structural design engineer is greatly increased as seismic loading is increased. It is estimated that the recent building code changes in Los Angeles, which permit high-rise reinforced concrete structures in cases where ductility requirements are met, will increase the proportion of cost allocated to structural engineering by 50%. The proportional change in the costs of construction and materials will not be as great. It was estimated that a conventional 12-story building would cost about L to 6% more if designed for 0.10 gravity than if earth- quake excitation were not considered. An inspection of the Preliminary Hazards Summary Report for the proposed Malibu reactor allows a rough estimate of the added cost of steel for earthquake reinforcement of the containment system to be made as no more than 10% of the containment cost, or less than 1% of the total plant cost. We have not made a detailed estimate of the added cost of protecting reprocessing plant and waste tanks against possible 100-psi internal 6-50 explosions. Such a detailed study would be desirable; however, we feel that this added contaimment will add only a small percent to our reference capital cost estimates. A1l in all, we feel that the possible additional capital cost for these "engineered safeguards’ or "contaimment systems" is on the order of 10% of the capital costs allowed for in Sect., 6.1 (with the uncertainty already being on the order of +30%). 6.6 References 1. Reactor Fuel Cycle Costs for Nuclear Power Evaluation, WASH-1099 (in press). 2. W, H, Farrow, Jr., Radiochemical Separations Plant Study, Part IL - Design and Cost Estimates, USAEC Report DP-566 (March 1961). 3. R. J. Christl, J. E. Crawley, and C. S, Otto, Radiochemical Separa- tions Plant Study, Limited Maintenance -~ Case VII, USAEC Report DP-%30 (September 196L). i. Chemical Reprocessing Plant (Hearing Before the Joint Committee on Atomic Energy, 0lst Congress of the United States, May 1, 1963), U.S. Govt, Printing Office, Washington (196L). 5. Atomic Energy Clearing House 13 (40), 8 (Oct. 2, 1967). 6. L. Thiriet, C. Oger, and P, de Vaumas, Long-Term Developments in Irradiated Natural Uranium Processing Costs - Optimal Size and Siting of Plants, French Report CEA-R-26L2 (August 196lL); also published as Paper A/Conf. 28/P/98, Third United Nations International Conference on the Peaceful Uses of Atomic Energy, Geneva, 196k, 7. L. Thiriet et al., "Capital and Operating Costs of Plants for the Reprocessing of Irradiated Natural Uranium," Proceedings of the European Atomic Energy Soclety Enlarged Symposium on Fuel Cycles for Nuclear Power Reactors, Baden-Baden, Germany, Sept. 9-1l, 1965; also n available as French Report CEA-R-2937 (January 1966). 8. Chem, Technol. Div. Ann, Progr. Rept. May 31, 1966, ORNL-3945. 9., Chem. Technol. Div. Ann, Progr. Rept. May 31, 1967, ORNL-L41L5. 10. M. W. Rosenthal et al., A Comparative Evaluation of Advanced - Converters, CRNL-3686 (January 1965). 11. R. Salmon, A Computer Code for Calculating the Cost of Shipping - Spent Reactor Fuels, ORNL-3648 (August 196lL). ) i 12, 13, 1h. 15. 16. 17. 18, 19. 20, 21, 22, 23. 6-51 R. Salmon, Estimation of Fuel-Shipping Costs for Nuclear Power Cost- Evaluation Purposes, ORNL-39L3 (March 1966). Code of Federal Regulations, Title 10, Part Tl, as published in the Federal Register 31 (141), (July 22, 1966). Tnterstate Cormerce Commission Order No. 70, Nov. 18, 1965, amended Apr. 1L, 1966, as reported in the Federal Register 31 (83), (Apr. 29, 1966). M. J. McNelly, Liquid Metal Fast Breeder Reactor Design Study (1000 Mwe UO,-PuQ, Fueled Plant), Vols. I and II, GEAP-4418 (January 196L). R, L. Bradshaw, J. J. Perona, J. T, Roberts, and J. 0. Blomeke, Evaluation of Ultimate Disposal Methods for Liquid and Solid Radio- active Wastes., 1, Interim Liquid Storage, ORNL-3128 (August 1961). J. J. Perona, R, L. Bradshaw, J. T. Roberts, and J. O. Blomeke, Evaluation of Ultimate Disposal Methods for Liquid and Solid Radio- active Wastes. II. Conversion to Solid by Pot Calcination, ORNL-3192 (September 1961). J, 0. Blomeke, J. J. Perona, H, O, Weeren, and R. L. Bradshaw, Evaluation of Ultimate Disposal Methods for Liguid and Solid Radio- active Wastes. III., Interim Storage of Calcined Solid Wastes, ORNL-3355 (October 1963). J. J. Perona, R. L, Bradshaw, J. O, Blomeke, and J. T, Roberts, Evaluation of Ultimate Disposal Methods for Liquid and Solid Radio- active Wastes. 1V. oShipment of Calcined Solids, (RNL-3356 (October 1962), J. J. Perona, J. O. Blomeke, R. L. Bradshaw, and J. T, Roberts, Evaluation of Ultimate Disposal Methods for Liguid and So0lid Radio- active Wastes. V. Effects of Fission Product Removal on Costs of Waste Management, ORNL-3357 (June 1963). R. L. Bradshaw, J., J. Perona, J. O, Blomeke, and W. J, Boegly, Jr., Evaluation of Ultimate Disposal Methods for Ligquid and Solid Radio- active Wastes., VI. Disposal of Solid Wastes in Salt Formations, ORNL-3358 (Rev.) (March 1969). J. J. Perona, R. L. Bradshaw, and J. O, Blomeke, Comparative Costs for Final Disposal of Radioactive Solids in Concrete Vaults, Granite, and Salt Formations, ORNL-TM-66lL (October 1963). J. 0. Blomeke, E., J. Frederick, R, Salmon, and E. D. Arnold, The Costs of Permanent Disposal of Power-Reactor Fuel-Processing Wastes in Tanks, ORNL-2873 (September 1965). 2l. 25. 26. 27. 28, 29. 6-52 J. 0, Blomeke, R, Salmon, J. T. Roberts, R, L. Bradshaw, and J. J. Perona, "Estimated Costs of High-Level Waste Management," pp. 830-39 in Proceedings of the Symposium on the Solidification and Long-Term Storage of Highly Radioactive Wastes, CONF-660208 (November 1966), R. Salmon, A Revision of Computer Code POWERCO (Cost of Electricity Produced by Nuclear Power Stations) to Include Breakdowns of Power Cost and Fixed Charge Rates, (RNL-4116 (August 1969). R. Salmon, A Procedure and a Computer Code (POWERCO) for Calculating the Cost of Electricity Produced by Nuclear Power Stations, ORNL-39L) (June 1966). Chem. Technol. Div, Ann, Progr, Rept. May 31, 1969, ORNL-LL22, pp. 136-41. P. L. Robertson and W. G. Stockdale, A Cost Analysis of the Idaho Chemical Processing Plant, (RNL-1792 (Jan. 3, 1955). C. G. Bell and T. F. Lomenick, Earthquakes and Power Reactor Design, ORNL-NSIC-28 (to be published). i 7-1 7. SLTING CONSIDERATIONS 7.1 Environmental Factors 7.1.1 Meteorology Meteorology and Atomic Energy,l AFCU-3066, provides a detailed dis- cussion of meteorology and its role in air quality from the standpoint of the nuclear industries. A referral to the Nuclear Safety Information Center (NSIC) will furnish up-to-date studies and data on specific related subjects. For convenience, the meteorological factors can be divided into two groups: (1) commercial considerations, and (2) health and safety aspects. Commercial Considerations, - Technology has obviated the importance of most meteorclogical parameters in the location and design of industrial facilities., Electrical power "outages," for example, have yielded to advanced circuitry, thereby reducing the importance of thunderstorm fre- quency to electrically dependent industries. The climate inside most new offices is controlled by decree, not by the outside environment; consequently, the same types of office bulldings may be found throughout all climatological regions. Rainfall and wet-bulb temperatures continue to be ilmportant criteria for manufacturing industries with large cooling requirements. Rainfall -- its frequency, duration, distribution with time, and reliability -- deter- mines the input parameters for hydrological considerations. Wet-bulb data are necessary when the water in the cooling facilities is to be recycled rather than immediately being discharged into nearby rivers or lakes. Wet-bulb frequency tables, as well as temperature and dew point tables, are obtainable from the Envirormental Science Service Administration (ESSA), Department of Commerce, in the United States, and from the national meteorology services of other nations. Subjective aspects of the use of cooling towers are the appearance of vapor clouds and the attendant hazards to motorists traveling nearby roads during cold weather. 7-2 Health and Safety Aspects. - From the moment a particle or a gas escapes to the outside enviromment, its fate is determined by the prevail- ing meteorological conditions. The wind and temperature fields traversed by the effluent are of primary importance. The following paragraphs out- line the variables that are likely to be important in the considerat ions of various sites. Wind Speed and Direction, - The significance of detailed wind data is obvious. The frequency of wind direction toward any given sector determines the actuarial experience or potential to the population or facilities within that sector from material emitted upwind, The wind speed directly affects the dilution rate of effluent material, Conse- quently, the first consideration of the meteorologist is to obtain or construct wind roses, Once wind vector data become avallable, calcula- tions of the concentration of material vs emission can be made, based on a number of references.l_h Where possible, "night" and "day" wind roses (lapse vs inversion data would be more useful, but they are rarely avail- able) are preferable to the 2l-hr averages. There are a number of locations where a simple wind rose is neither available nor can be readily estimated. For example, sites in deep valleys and sites near oceans or lakes have complex wind patterns that vary from night to day, The normal regional wind flows are superimposed on these local effects. Thus, although a few broad generalizations may be made, special observations over a period of time are necessary for detailed analysis. Calm winds (generally, wind speeds of less than 3 mph) require more than superficial analysis. Winds of 1 to 3 mph are considered "light air® on the Beaufort scale,5 and are described as follows: '"Direction of wind shown by smoke drift, but not by wind vanes." Here, the effects of local terrain, surface heating vs inversions, etc., on effluent behavior are most marked, Moreover, the effluent's own characteristics play an import- ant, perhaps dominant, role in the height to which the material will rise. The skillful analysis of "calm" wind data and potential effects on effluent material presents a challenge to the most experienced investigator. M 7-3 The persistence of wind vectors or calms is generally secondary to the wind roses themselves. If a potentially harmful release over a pro- tracted period is considered, the potential hazards to a sector or to a number of adjacent sectors are calculated by assuming a degree of contin- uous flow toward the areas of concern.6 The persistence of calms may become a limiting factor in areas of persistent anticyclonic circulation. Wind variation with height (or elevation) assumes increasing import- ance as the height of the stack is increased. Surface wind data are generally sufficient for estimating average downwind concentrations when stacks are 30 m or less in height and the effluent temperature 1s near ambient, However, if the terrain has a noticeable effect on lower-level winds, wind data obtained above the first 100 m or so must be considered. Also, tall stacks and high effluent temperatures introduce effluent to wind regimes that are different from those at ground level. Interpolation between wind observations made near the surface and routine information that has been gathered from the upper air network of ESSA can serve as a first approximation to the wind pattern at intermediate heights. The temperature variation with height was a parameter used primarily to determine the extent and the intensity of inversions, It is now being utilized, when available, for determining effective stack heights of fossil-fueled power plants and other facilities where large quantities of heat released to the atmosphere may be used to advantage. Rainfall, Rainout, and Washout. - Specific elements of interest are removed by rainfall at rates varying with the form of material and the intensity of the rainfall.7 The removal of effluents by rainfall poses the question of the amount of material deposited on the area involved. Where possible, a "rainfall wind rose' should be constructed. The signifi- cance of rainfall rates vs wind direction may then be applied to effluent removal models, Deposition. - The removal of material from the atmosphere via depo- sition is strongly dependent on the concentration profiles of the effluent in the vertical. The parameters of stability and wind speed, in combina-~ tion with a deposition model (such as Chamberlain's deposition velocity), can be statistically manipulated to estimate the depletion of material as 7-L a function of distance from its source and, as a corollary, the amount and pattern of deposition around the source. The terrain has been mentioned in relation to its contribution to wind behavior., In a slightly different manner, we may consider terrain as a parameter in source configuration or emission rates. For example, if an industrial site is located in a deep valley, the air in the valley has a capacitance effect; that is, it smooths the ocutflow of the effluent over a period of time. Aside from possible increased deposition to the sides of the valley, concentrations and deposition calculations over periocds of a day or longer are only slightly different from the estimates obtained from flat terrain. The "smoothing'" delays, but does not alter, the introduction of effluent into the surrounding wind pattern, However, a sudden increase in the rate at which the material is emitted will be mitigated by the capacitive effect. The nearby enviromment suffers a longer and a heavier effluent burden, whereas the distant terrain, although receiving the same total amount of material, experiences a lesser load over a greater period of time than would have been the case if the effluent moved freely from its source. When the amount of material that is released to a volume of air exceeds the inflow of "fresh'" air, stagnation begins, Stagnhation is not confined to valleys by any means. It is quite pronounced in arctic towns and cities in the wintertime, as well as in cities like London and Los Angeles in temperate climates. Nevertheless, in temperate climates, valleys are more prone to suffer from stagnation than open areas, A quick estimate based on the routine off-gas levels and the volume of surrounding air will give some idea of how susceptible a facility is to stagnation problems. Vegetation, particularly wooded areas, can be considered a meteorolog- ical factor that alters the wind behavior in the vertical and changes the deposition and concentrations accordingly. 7.1.2 Geology and Hydrology The geology and hydrology of the site of a nuclear fuel reprocessing plant can influence: (1) the foundations of the plant, (2) the emplace- ment of underground waste-storage tanks, (3) the water supply, (L) the routine disposal of liquid and solid radiocactive wastes, (5) the danger from earthquakes, and (6) the consequences of an accidental release of significant quantities of radicactive materials. Geologic conditions that would be favorable for one of these considerations might be unfavor- able for another; therefore, an ideal enviromment does not exist, and the selection of any actual site will require a compromise. Perhaps the only valid generalization is that all of these considerations will be easier to evaluate if the geology and hydrology of the site are simple (although determining what constitutes a "simple enviromment may frequently be difficult). Foundations., - Satisfactory foundations can be provided in almost any geologic environment, although water-saturated, poorly compacted clays and silts may require deep and expensive excavations or pile driving. Many limestones contain extensive and unpredictable networks of solution cavities, which are difficult to wash free of mud and fill with cement grout. Fault zones, particularly in basic igneous rock, weather more deeply than the adjacent unfractured material and may cause difficulties if they are not detected by preliminary test bores, These points may appear somewhat elementary; however, they bear repeating since several of them have, on occasion, been neglected in the siting of nuclear facilities both in this country and in Europe, Emplacement of Underground Waste Storage Tanks. - A nuclear fuel reprocessing plant will probably need waste-storage tanks, which may be located underground. If hard bedrock exists at a shallow depth, the emplacement of the tanks may be relatively expensive. On the other hand, some uncongolidated deposits may need to be shored up (i.e., to hold open excavations), which is both expensive and potentially dangerous. The most important requirement, however, is that the excavations in which the tanks are located be well drained and located safely above the water table. 7-6 (Partly filled tanks will float up out of the ground if surrounded by water.) If located near a river, the plant site must be well above any possible flood level, Water Supply. - If sufficient quantities are available, groundwater is generally preferable to surface water as a source of water supply, since surface water is nearly always more variable in temperature and chemical composition. This suggests that the plant site should be underlain by at least a moderately deep deposit of sand or other permeable material. Permeable limestone, which can yield an excellent supply of groundwater, may not be desirable (as has been mentioned), because of potential prob- lems with the plant foundations. Also, aquifers with an irregular and unpredictable permeability may seriously complicate the problems of routine low-level waste disposal or of remedial action following an accidental release of radiocactive materials. Routine Waste Management. - High-level radiocactive liquid wastes may have to be aged in tanks prior to ultimate disposal. At least 50 ft of easily excavated sediments, or deeply weathered rock of moderate permea- bility, and appreciable ion exchange capacity are required for satisfactory tank emplacement, After aging, the wastes can be solidified and then shipped to a permanent disposal site. As described in Sect. L.3.5, exca- vations in salt or other dry, underground workings are highly desirable for a permanent repository, and sites with these characteristics may be many miles distant from the reprocessing plant. Intermediate-level liquid wastes can best be handled by combining them with high-level wastes for eventual solidification and permanent storage outside the biosphere., A less desirable alternative is to solidify the intermediate-level wastes, perhaps by incorporation in asphalt or cement, and to ship them off-site to licensed burial grounds or special reposito- ries, Still another alternative may be the disposal of these wastes by hydraulic fracturing (Sect. 4.Lh.3). At least several hundred feet of flat- lying beds of shale at depths between 500 and 3000 ft are required for this method of disposal. The shale would have to be tested to make cer- tain that horizontal, rather than vertical or steeply dipping, fractures will be formed. =7 Low-level liquid waste and potentially contaminated cooling water are produced in such volumes that storage is impractical., This waste is treated, usually by evaporation or ion exchange, and then released to the ocean, to a river, or into a shallow or deep groundwater aquifer, Although release to the ocean has apparently been satisfactory at various sites that routinely use this type of disposal, it has required detailed, con- tinuing surveillance of fish, sea weed, and beaches in the affected area, Release to rivers has, perhaps, been somewhat less satisfactory, depending greatly on the particular circumstances; however, even under favorable conditions, it has entailed extensive, detailed envirommental monitoring, In cases where the circumstances have been even mildly unfavorable, severe restrictions have been placed on the concentration and the total quantities of radiocactive materials that could be discharged, necessitating careful and potentially expensive treatment of the low-level waste prior to dis- charge, Successful releases of waste into shallow (water-table) aquifers also depend greatly on local conditions. In sparsely inhabited areas, large, unused aquifers may be avallable to provide dilution and, more importantly, long holdup times for the waste. In instances where the aquifer that is used for disposal is also used in the same general area for water supply, extensive geohydrologic investigations will be required; even then, severe restrictions may be placed on waste discharge, The most favorable situation for disposal of low-level liquid waste would be pro- vided by a deep, permeable, and porous artesian aquifer that is not a source of water supply. This would be the case if equally good (or better) sources of water were available on the surface or at a shallow depth. Then a deep, moderately permeable aquifer could be developed in such a manner that it would receive all the low-level waste and cooling water from a large nuclear fuel reprocessing plant, However, to ensure that no hazard would exist, considerable geologic and hydrologic information would have to be assembled and analyzed. Such information has been collected for several areas in the United States by a committee of the American Assoclation of Petroleum Geologists.9 The areas described as favorable in this report would be particularly suitable sites for nuclear fuel reprocessing plants. 7-8 Consequences of the Accidental Release of Radioactive Material. - If large quantities of radiocactive materials were accidentally released and then quickly reached a stream, a river, or the ocean, with only a small proportion being held back at or near the plant site, the consequences could be serious. Rapid movement and low retention should be expected at a site where the earth or rock is sufficiently impermeable to allow released liquids to move rapidly over the surface and virtually no ion exchange with the soil to take place. On the other hand, in many perme- ' able terrains, the released radiocactive materials would travel through the soil (i.e., below the surface); and, if the plant were located at even a . short distance from the nearest stream or river, the travel time might be long enough to permit important remedial action to be taken. Also, during this time, appreciable guantities of many of the radionuclides might be fixed virtually permanently in the soil and thus be rendered effectively harmless, In connection with the consequences of accidental release, simplicity in the geologic enviromment is particularly desirable. Only in cases where the conditions can be analyzed in detail and with considerable confidence can predictions of the possible results of an accident be made. These predictions will allow proper precautions to be taken against such an eventuality, as well as suggest effective remedial measures in the event of an accident. A simple geologic and hydrologic environment also makes it possible to determine, with confidence, the most effective local methods for ultimate disposal, the maximum guantities of radiocactive material that may be released to the envirormment, and the best methods for monitoring the enviromment to make certain that safe levels of discharge are not being exceeded, 7.1.3 Geoseismology . General. - Faults, vibrationsg, and tsunamis are the major earthquake- induced phenomena to be considered in the siting and the design of nuclear facilities (including fuel reprocessing plants). All of these are important - for some sites along the West Coast of the Unites States; on the other hand, vibratory effects are generally the sole concern in the eastern part of the country. -9 The rapid growth of the nuclear power economy has focused consider- able attention on the unknown and/or imperfectly understood aspects of earthquakes as related to reactor siting and design.lo In general, this lack of knowledge has made it imperative that conservative estimates and evaluations of the critical geoseismological and engineering design param- eters be made for nuclear facilities in all parts of the country. In many areas, 1t appears that earthquake-induced phenomena can be adequately considered through currently acceptable engineering practices; however, in some highly seismically active regions, the high degree of geoseismo- logical conservatism requires that unique and presently unproved designs be considered. With regard to the needed improvements in predicting faulting, shaking, and tsunami effects at potential nuclear facility sites, it is emphasized that, since there is no quantitative way to predict earthquakes, empirical and somewhat indirect approaches to the problem must be used, One of the principal means for studying earthquake phenomena is, of course, through the observation of earthquake events. Since large earthquakes occur rather infrequently and there are currently few, if any, positive corre- lations between the occurrence (as to place and time) of earthquakes and measurable changes in the physical and/or chemical properties of rocks that comprise the earth's crust, major improvements in defining and deter- mining the geoseismological factors pertinent to reactor siting and design of nuclear facilities will require intense and concerted efforts in the geological, seismological, and engineering disciplines. Structural designs for accommodating moderate amounts of differential ground displacement and for ensuring plant survival for most of the con- ceivably strong ground motions appear to be attainable; however, demon- strable proof of these designs is needed to ensure the functioning of all components or systems that are directly or indirectly related to the containment of radioactivity. Faulting., - The exact mechanisms for generating earthquakes are not known, but it is generally agreed that faulting is the cause of most of the large shallow earthquakes in California and in other tectonically similar areas of the world. In the continental United States, historic 7-10 faulting has been largely confined to the area west of the Rocky Mountains (see Fig. 7.1), and there is good reason to believe that faulting in the foreseeable future will continue to be restricted primarily to this area. Through observations of historic surface breaks, we can make a general evaluation of the length of the main fault trace, its location, and the amount of displacement that may be expected when an earthquake of a spe- cific magnitude occurs along a major fault. However, it is much more difficult to determine the nature and extent of the many secondary, minor, or subsidiary faults that commonly occur adjacent to, or near, the main fault traces. At the present time, relatively little is known about the characteristics of faulting that occurs at areas located away from the main fault traces. Thus, secondary or minor faulting is one of the prin- cipal problems in the siting and design of nuclear facilities in seismic- ally active areas, Another important and controversial problem in siting concerns the degree of activity of faults., Many faults in the western part of the United States can be clearly labeled as active, while others have been determined to be unquestionably inactive. However, the large number of faults that lie between these two extremes are probably of greatest con- cern, At present, it is difficult to determine precisely enough the date of the most recent motion along such faults., In addition, it is difficult, in many cases, to state what elapsed period of time after the last movement along a fault would provide assurance that the fault is inactive and that no further movement would occur in the foreseeable future, Since the majority of faults have not moved in historic times, we must rely on geologic relationships and seismological evidence to provide data regard- ing the tectonic activity of many faults. Shaking, - Small earthquakes, which occur over most of the earth's crust, may cause localized shaking. However, during shocks of great magnitude, extremely large areas, covering hundreds or thousands of square miles, may be subjected to severe shaking. The strong earth motions assoclated with larger shocks should be recognized and taken into account in the siting and design of fuel reprocessing plants. However, at the present time, there is some uncertainty in defining the characteristics of these expected motions. ORNL-DWG 67-10281R2 Fig. 7.1. Locations of Known Surface Faulting Accompanying Farthquakes in Historic in the United States. Times TT-. 7-12 Strong-motion accelerometers are currently being used to record severe shaking in the United States. Although ground accelerations have been recorded for a relatively large number of small shocks, only a few large earthquakes have been recorded on strong-motion instruments, The charac- teristics of these ground accelerations have been thoroughly analyzed, and this information, along with other seismic and geologic data, is commonly used to estimate the amount and the nature of the expected ground motions at other sites. Because of the general lack of instrumental records of strong ground motions, and because there 1s no suitable theoretical basis for predicting these motions, we do not have an acceptable quantitative method for precisely determining accelerations, veloclties, and displace- ments of the ground motions for a given earthquake at most sites. Thus, estimates of the important design parameters of ground motion must be conservative. Earthquake-induced ground motions in soils are usually larger than those in hard rock; however, it is not possible to state precisely the magnitude of the difference. Most of the studies of the intensity of shaking in various types of rock and soil have been conducted with small shocks, and there is no satisfactory technical method for extrapolation to larger shocks. Recently, determinations of the dynamic properties of soil sections and theoretical considerations have yielded important rela- tionships concerning some soil conditions and ground motion characteristies; however, before reliable predictions can be made for the majority of soil conditions, extensive laboratory and field investigations must be carried out. In addition to their amplification and/or attenuation properties, solls may also fail or be displaced, via consolidation, differential com- paction, sliding, and liquefaction, as a result of earthquake-induced ground oscillations, Tsunamis, - Tsunami and tsunami-generated oscillations are potentially dangerous to nuclear fuel reprocessing installations at coastal sites, since they may cause damage to the plant and water intake structures by means of runups and/or withdrawals. Most tsunamis are thought to be gen- erated by vertical displacement of the subsea bottom. About 60% of all recorded tsunamis originate in the Pacific Ocean, where large earthquakes 7-13 occur along deep, bordering trenches. Runup heights in Japan, Kamchatka, Chile, Peru, Alaska, and the Hawaiian Islands have often exceeded 30 ft; however, during historic times, the West Coast of the United States has apparently not been subjected to runup heights greater than 16 ft. The relatively low runup along the West Coast of the United States is attrib- uted to the coastal shelf. In contrast to the experience on the West Coast, damaging tsunamis have not been recorded along the Gulf and Eastern Coasts of the United States. Chances that a tsunami will be generated by a local disturbance off the West Coast are thought to be limited, since apparently no locally generated damaging tsunamis have occurred there in the past. At many sites, the prediction of maximum runup height is difficult because relatively obscure coastal and submarine features tend to amplify waves., Since some sites have consistently high waves and others have consistently low waves, the best guideline for prediction along the West Coast appears to be previous experience with tsunamis, regardless of the direction of approach. With the exception of the Alaskan tsunami that occurred in March 1961 and caused runups as high as 16 ft at Crescent City, California, the recorded runups along the West Coast have not exceeded the tidal range of about 6 ft, 7.2 Geographic Factors The primary consideration in acquiring a site for a fuel reprocessing plant is to provide sufficient distance between the plant and private lands to ensure that the general public will not be harmed by either normal oper- ations or by credible accidents. Second, the site should be located at a place where the aggregate cost of raw materials, transportation of materials to the plant, manufacturing, and transportation of finished products to the market will be at a minimum.11 In present plants, the basic raw materials are water, nitric acid, solvent, and aggregate for concrete, Either a railroad spur or a waterway with barging facilities is a practical neces- sity since some spent-fuel shipping casks weigh 50 to 100 tons. Paved highways are required for trucking smaller casks, raw materials, finished 7-1L products, and waste. Manufacturing costs are dependent on an adequate supply of skilled labor and the prevailing wage scales in the vicinity. Conveniently located housing and community facilities are desirable. Long commuting distances and poor facilities (as well as an undesirable climate) tend to result in a large labor turnover. The plant must have adequate acreage for possible future expansion, adequate soil or rock foundations to support heavy concrete structures, and reliable electric power; the latter should preferably be available from two independent sources. JIdeally, the plant should be located relatively near nuclear power reactors and sites for disposal of high- and low-level waste. The following sections briefly describe considerations that affect the site size, the surrounding population density, the land usage in the vicinity, and the relation of the plant to other nuclear facilities. Regions in the United States having certain desirable and undesirable features will be delineated. 7.2.1 Site Size The site boundary is determined most accurately and restrictively by the requirement that the direct exposure of the surrounding public to radio- active gaseous or ligquid effluents must be maintained at allowable levels. These considerations will be discussed in detail in Sect. 8. Penetrating radiation that escapes through the shielding used in the plant is not normally a consideration., For example, the penetrating radiation from 0 an unshielded nuclear excursion of lO2 fissions would cause whole-body exposures no greater than 25 rem at distances of only about 350 m.12 Studies at I—Ia_nford]j’uL indicate that controlled areas extending 0.5 to 1 mile from the plant are desirable for the control of '"nuisance con- tamination" resulting from a temporary loss of control of relatively small quantities of radioactive materials. Such minor releases might result from outside decontamination operations on large pleces of process equip- ment or shipping casks. This is not an absolute limitation; it is possible (i.e., at increased cost) to house those facilities that would potentially disperse low-level contaminants. It was found that the routine release of noxious nonradiocactive chemicals to the atmosphere (most significantly 7-15 NOZ) would dictate a site boundary about 1 mile from the stack., This is also not an absolute limitation, since such gases may be removed from stack effluents to practically any extent regquired using present technol- ogy. The discharge of low-level radioactive effluents is determined pri- marily by the relative flow rate of groundwater and surface water as a function of distance from the plant, 7.2.2 Population Density of the Surrounding Area Federal regulations (1LOCFR100) specify that there shall be a zone of low population surrounding a reactor plant. The primary concern is to prevent population groups from receiving somatically or genetically significant doses of radiation. The costs of indemnification are also of concern; claims resulting from overexposure to radiation resulting from accidents would probably be directly proportional to the number of persons involved. QGuthrie and Nichols have estimated that monetary losses of $50,000, $10,000, and $2000 would result from exposures of greater than 100, 10 to 100, and 1 to 10 times the allowable annual industrial radia- tion dose respectively.8 It was estimated that severe contamination, resulting in long-term evacuation and total loss of property value, would cause an average monetary loss of $10,000 per person, Minor contamina- tion, which would necessitate short-term evacuation, washing nonporous surfaces, and replacing or recovering porous surfaces such as sidewalks, pavements, roofs, etc., was estimated to result in monetary losses of $1500 per person, Minor contamination, which would require roofs, streets, and buildings in urban areas to be hosed was estimated to result in a monetary loss of about 5 mills per square meter of projected surface, 7.2.3 Land Usage Special considerations are required when fuel reprocessing plants are located in areas where there is a mechanism for reconcentration of the radioactive effluents and a pathway for ingestion by the general public., Because certain radionuclides (e.g., 9OSr, and 13705, see ref. 8) are known to concentrate in crops, the restrictions on low-level liquid waste effluents that are subsegquently to be used for irrigation may be 7-16 more severe than if the water were used only for drinking., Deposition of radiciodine from gaseous wastes on grass, followed by the cow-milk pathway to the thyroid of small children, may result in maximum permissible air concentrations lower by a factor of 700 than those for inhalation. Special congsiderations are also required when fuel reprocessing plants are located near other plants whose products are very sensitive to radiation (e.g., the photographic industry), 7.2.L Relation of the Plant to Other Nuclear Facilities The fuel reprocessing plant should be designed and located to take into account adjacent nuclear facilities, including reactor plants, other reprocessing plants, and waste disposal sites. Effluents from the plant must not mask nuclear instrumentation at adjacent sites. Accidents in the plant should not cause unduly hasty and unsafe evacuations of adjacent sites. In addition, the effluents from each plant must be restricted in such a manner that thelr combined effect does not endanger the health and safety of the surrounding public. In practice, the effect of these re- strictions has been minimal at production plants and national laboratories; the cost of engineered features is generally offset by the decreased cost of logistics. 7.2.5 Regional Distribution of Potential Sites in the United States We have gathered information that may be of value in selecting potential sites, based on surrounding population density, distance to a population center, and seismicity. The attached packet of overlay maps includes: (1) a base map showing population densities, by county, of the United States minus Alaska and Hawaii; (2) an overlay of major towns for all communities having a population of 20,000 or more; (3) an overlay showing all presently used railroad lines; and (L) an overlay indicating the major seismic areas of the country, 15 The colored base map shows population density ” according to the following color code: : e B ¢ ] B B R R B R S B B BB B B B B B B 7-25 <10 persons/mile2 purple 10-30 pérsons/mile2 blue 30-100 persons/mile® green 100-300 persons/mile2 yellow 300-1000 persons/mile2 orange >1000 persons/mile2 red The first overlay (red) indicates the location of all single commu- nities having populations of 20,000 or more (according to the 1960 U.S. census), plus a number of towns that have grown to this size on the basis of more recent census estimates, Federal power districts are also indi- cated, The 250 U.S. metropolitan areas with populations ranging from 52,000 to about 10,000,000 were plotted first. Then all towns with populations as low as 20,000 persons were plotted. The sizes of the circles indicate, in general, the sizes of the cities, and they were plotted with a radius based on the distance to the edge of concentrated housing plus 10 miles. On this basis, the circles have the following diameters: 20,000-50,000 20-mile diameter 50,000-100, 000 25-mile diameter 100,000-250, 000 30-mile diameter 250,000-1,000,000 35-mile diameter >1,000, 000 LOo-mile diameter, or 10 miles beyond the edge of the metropolitan area It is interesting to note that a large number of the communities in the 20,000-50,000 population range were found to be suburbs of the larger metropolitan districts. In order to complete this survey, it is recommended that all towns with populations as low as 10,000 be plotied in order to discover twin cities and tri-cities having combined populations of 20,000 or more. These data are now in hand, In addition, it must be noted that, for the nonmetropolitan-area towns, actual city limit population values were used and that the actual built-up areas may include 20 to 100% more people. 7-26 The second overlay (blue) shows the location of all major rail lines that are presently in commercial use,16 but does not include a number of presently unused branch lines. These branch lines are in varying degrees of disrepair, and their capacity for handling heavy fuel casks would have to be determined in the event that interest in their use should be indi- cated, The third overlay (green) shows the seismic risk zones from the seismic risk map of Algermissen.lT The four zones may be expected to approximately represent expected damage and intensity. Zone O contains the areas in which earthquake damage is not expected to occur, and where Modified Mercalli (M.M.) intensities in excess of IV have not been observed, Zone 1 is composed of areas of expected minor damage, where M.M, intensities in excess of VI have not been observed., Zone 2 contains areas where moderate damage may be expected. Zone 3 contains areas where major destructive earthquakes have occurred in the recorded past. There is no clear distinction between zones 2 and 3 on the basis of expected intensity, other than that catastrophic earthquakes have occurred in zone 3, 7.3 References 1. U.S. Department of Commerce, Weather Bureau, Meteorology and Atomic Energy, AECU-3066 (July 1955). 2. W. M, Culkowski, Estimates of Accumulated Exposures and Environmental Buildup of Radioactivity, TID-7593 (1959), pp. 89-99. 3. F. A, Gifford, The Problem of Forecasting Dispersion in the Lower Atmosphere, USAEC pamphlet (1961). li. W. M. Culkowski, Deposition and Washout Computations Based on the Generalized Gaussian Plume Model, CR0O-599 (1963). 5. Manual of Surface Observations (WBAN), Circular N, 7th ed., Weather Bureau, Air Weather Service, and Naval Weather Service (November 1961), p. 100. 6. I, A, Singer, Steadiness of the Wind, BNL-11282 (1967). 7. A. C. Chamberlain and A. E, J. Eggleton, "Washout of Tritiated Water Vapour by Rain," Int. J. Air Wat, Poll. 8, 135-49 (196L). I e K 10, 11. 12, 13. 1. 15. 16, 17. =27 C. E. Guthrie and J. P. Nichols, Theoretical Possibilities and Conse- quences of Major Accidents in U233 and Pu239 Fuel Fabrication and Radioisotope Processing Plants, ORNL-3LL1 (April 196L). John E. Galley (ed.), Subsurface Disposal in Geologic Basins — A Study of Reservoir Strata, Memoir 10, The American Association of Petroleum Geologists, Tulsa, Okla., 1968, T. F. Lomenick and C. G. Bell, Earthquakes and Reactor Plant Design, ORNL-NSIC 28 (in press). F. C. Vilbrandt, Chemical Engineering Plant Design, p. 366, McGraw- Hill, New York, 19,2. H, T. Williams et al., Safety Analysis of Enriched Uranium Processing, NY0-2980 (Mar, 18, 1960). R. J. Sloat, Progress Report — Study of Land Requirements, IS0-668 (Jan. 1, 1967). R. J. Sloat, Study of Land Requirements, IS0-842 (May 8, 1957). L. H. Long (ed.), '"U.S. Population by States and Counties; Land Area,™" pp. 301-20 in The World Almanac, Newspaper Enterprise Association, Inc., 1967. Major American Railroads, Rand-McNally and Company. S. T. Algermissen, "Seismic Risk Studies in the United States," to be published in the Proceedings of the Fourth World Conference on Earthquake Engineering. 8-1 8. HEALTH AND SAFETY ASPECTS OF PLANT .SITING The principal criterion for judging the adequacy of a site for a fuel reprocessing plant is the provision that no undue risk exists with regard to public health and safety in the surrounding areas. Present and fore- seeable technology requires that such plants routinely discharge small quantities of radioactive materials to the atmosphere; for this reason, and also because of the large inventory of physiologically hazardous materials, there is always a small, but finite, probability of a more massive discharge. The magnitude of the routine discharge and the prob- ability of a more massive discharge are determined by the inventory of radioactive materials and by the design features of the plant, Present licensing procedures for fuel reprocessing plants apply existing federal regulations for radiation protection (lOCFR?O),1 licens- ing of production and utilization facilities (lOCFRSO),2 and siting of nuclear reactors (lOCFRlOO),3 wherever applicable, to the plant under study. The safety of a proposed facility is determined by evaluating, as a unit, the proposed plant and the site., The design features of the plant, together with the geological, hydrological, seismological, and meteorological characteristics of the site, are analyzed to determine whether the proposed design is adequate to maintain the barrier between radiocactivity and the surrounding population under adverse environmental conditions such as earthquakes, tornados, and floods. The consequences of releasing radioactive effluents during normal operations as well as during "upper limit accident" conditions, are evaluated using environ- mental characteristics of the site. The calculated concentrations of normal plant effluents are compared with the values published in 10CFR20; the engineered features for prevention and mitigation of the consequences of accidents are compared with the guidelines of 10CFR50; and the calcu- lated doses received by a member of the general public from postulated accidental releases are compared with the guidelines specified in 10CFR100, If, by employing conservative assumptions, it can be demonstrated that engineered safety features and releases under all credible conditions are within the guidelines, then the plant and the site are considered acceptable 8-2 The following sections of this chapter present estimates of the effect of health and safety considerations on the siting of spent-fuel processing plants. These include the consequences of an expanding nuclear economy on the worldwide distribution of long-lived volatile radionuclides, local envirommental effects of the routine release of radionuclides, and the effects of credible accidents. Section 8,1 presents estimates of the worldwide distribution of 85Kr and 3H in an expanding nuclear economy, assuming that these nuclides are released quantitatively to the atmosphere and the hydrosphere. These estimates, together with those of following sections, lead to the conclu- sion that worldwide pollution hazards will be avoided and local operating personnel will be protected by the necessary expedient of providing engi- neered safety features and site boundary distances that ensure appropriately low radiation exposures of members of the public at the site boundary. Section 8.2 presents estimates of the effect of routine releases of radiocactive materials from spent-fuel processing plants. The consequences of, and site boundary distances dictated by, routine releases from fuel processing plants were estimated assuming (1) ORNL meteorological condi- tions, (2) the complete release of noble gases and tritium, (3) iodine decontamination factors of 2000 (present technology) and lO7 in plants for processing highly irradiated fuels that have decayed 150 and 30 days, respectively, and (4) a particulate-release-rate model that agrees satis- factorily with existing data. For reference purposes, the acceptable concentrations at the site boundary were selected as one-third of the air concentrations listed in 10CFR20, Appendix B, Table II, Column 1, with the exception that the 1311 concentrations were further reduced by a factor of 700 to account for the grass-cow-milk pathway to the thyroids of small children. The downwind consequences resulting from the routine release of radio- nuclides from a plant processing light-water reactor (LWR) fuel (postirra- diation decay period of 150 days) or a plant processing fast breeder reactor (FBR) fuel (decay time of 30 days) are estimated to be controlled by the release of noble gases and iodine, It is concluded that equipment for removing 50 to 99% of the noble gases is necessary in plants of 8-3 capacity more than a few tons per day; more efficient iodine removal than that demonstrated in present technology is required for LWR plants of capacity greater than about 6 to 10 tons/day, whereas DF's for iodine as high as 108 may be required for FBR plants, Section 8.3 presents estimates of the effect of releases of radio-~ active effluents in "upper 1imit accidents.! The consequences of upper limit accidents were estimated assuming that the acceptable annual dose commitments resulting from exposure to the cloud or inhalation at the site boundary are those recommended by the National Committee on Radiation Pro- tection for annual occupational exposure. Although the assumed acceptable dose commitments have been employed only for reference purposes, they may be plausible on the basis that the ratio of benefit to probability of exposure is believed to be greater for an individual of the general popu- lation living near the site boundary than for a worker in the plant. The meteorological and dose commitment analysis was based on the assumptions of flat downwind terrain and exposure to the radiocactive cloud. The consequences of downwind ground contamination and additional exposures by such phenomena as reentrainment were not considered as mecha- nisms that would limit plant siting. Excessive levels of ground contami- nation would cause inconveniences, require expensive decontamination procedures, and result in property loss; however, they would probably not present an unavoidable threat to the health and safety of the public, It is concluded that the confinement and ventilation systems in spent-fuel processing plants remove particulates of nonvolatiles dispersed under accidental conditions to such an extent that the upper limit acci- dents are controlled by the release of such volatile and semivolatile materials as the noble gases, iodine, ruthenium, cesium, and tellurium. Credible upper limit accidents in well-desighed facilities for interim storage of wastes, either in liquid or solid form, are estimated to be inconsequential with respect to those from processing operations in the plant. 8-L 8.1 Buildup of PKr and °H in an Expanding Nuclear Power Industry As the free world's nuclear power production increases, the buildup of BSKr in the atmosphere and 3H in the hydrosphere may become important. Therefore, estimates of dose equivalents to the year 2000 from a uniform worldwide distribution of these radionuclides have been made, Estimates of the amual production of 85Kr and 3y are based on the AEC's projected civilian nuclear power economy in the United States and in the free world.h’5 In Fig. 8.1, which shows the growth of the nuclear power industry, foreign capacity in the year 2000 is assumed to be equal to the estimates of capacity in the United States at that time. Thermal power generation was estimated by assuming load factors of 0.8 to 1980 and 0,7 at the year 2000, and a thermal efficiency of 0,31. Thus, in the year 2000, the free world's nuclear capacity for continuous operation is estimated to be 1 million electrical megawatts and 3.3 million thermal megawatts. The rates of production and accumulation of 85Kr and 3H are shown in Fig. 8.2. Production rates were based on an assumed core irradiation of 20,000 Mwd/metric ton and a specific power of 25 Mw/metric ton. The accumulated quantities of 85Kr and 3H were obtained by allowing each radionuclide produced in the immediately preceding 5-year period to decay for 2.5 years and adding this value to the previously accumulated quantity (corrected for decay for 5 years). Accordingly, in the year 2000, 85Kr production will be 520 megacuries/year, and 3000 megacuries will have been accumulated., Tritium production will be 15 megacuries/year, and 96 mega- curies will have been accumulated. 8.1.1 85Kr Distribution and Dose Equivalent The concentration of 85Kr in the atmosphere was estimated by assuming complete mixing of the 85Kr and the air throughout the first 8 miles of the atmosphere. Within this zone, 85Kr was assumed to be distributed according to the density mass of air. Above 8 miles, the tropopause would inhibit rapid mixing into the stratosphere.6 Rainout was consid- ered negligible, since calculations indicated that the atmosphere ORNL-DWG 66-10262 THERMAL POWER ELECTRICAL POWER NUCLEAR POWER (Mw) 1960 1970 1980 1990 2000 CALENDAR YEAR ENDING Fig. 8.1. Estimated Growth of Civilian Nuclear Power in the Free World. o ORNL-DWG 66-10261 10 5 85 ) Kr ACCUMULATED (curies) 2 10° 5 2 85 Kr PRODUCTION RATE (curies /year) W o8 3 Q 2 5 5 O <] hs 3H ACCUMULATED S 2 (curies) w L o S 107 O 5 3H PRODUCTION RATE (curies/year) 2 10° 5 2 10° 1960 1970 1980 1990 2000 2010 CALENDAR YEAR ENDING Fig. 8.2. Estimated Production of O°Kr and H from the Nuclear Power Industry of the Free World. 8-7 contained more than 95% of the stable krypton as compared with the 7 oceans, Figure 8.3 shows the estimated whole-body exposure from BSKr as a function of elevation, A maximum dose rate of 1.8 millirems/year in the first one-fourth mile of the atmosphere can be compared to an average background radiation of 100 millirads/year (to skin) near sea level and to permissible whole-body exposures of average population groups of 170 millirems/year, and of members of the public of 500 millirems/year, as recommended by ICRP and FRC.B’9 8.1.2 Tritium Distribution and Dose Equivalents Practically all of the tritium in irradiated fuel elements may be released to the emviromment during spent fuel processing., This release is assumed to occur as HTO, either as tritiated water or as tritiated water vapor., The volumes of circulating waters in the world, listed in Table 8.1, were used to calculate the concentration of tritium in the environment, It was assumed that: (1) tritium was mixed in oceans and seas to a depth of J,0 m, (2) all the water in stream channels and in the first 10 km of the atmosphere was circulating, (3) only the portion of the groundwater located in the root zone was available for mixing, and (L) complete isotopic dilution occurred in these waters. As shown in Fig. 8.1, the estimated dose equivalents to body tissue due to inhalation of air and absorption throufih skin, and to ingestion of surface water for the year 2000. Nonuniform distribution of 3H in rainwater and surface water has been indicated by Libby in his claim that 50% of the tritium containing 34 are 7.2 x 107 and 1.4 x 1072 millirems/year, respectively, released from the detonation of thermonuclear devices in 1958 had fallen between 30° and 50° north latitude.lo releases of 3H from fuel reprocessing plants, then approximately 10% of the earth's surface will receive one-half of the total 3H. Thus, the If this occurs in the case of dose equivalents in this temperate zone may be five times the calculated average, ORNL-DWG 66-2712 101 100 DOSE RATE (mrem/yr) 3 3 no 10-4 O 1 2 3 4 S 6 7 8 ELEVATION {(miles) Fig. 8.3. Dose Rate from 85K:r' in the Atmosphere According to Elevation. 8-9 Table 8.1. Volumes of Circulating Water in the World & Volume of Water (m North Latitude Total (30°-50°) : | 16 15 Oceans and seas, in surface IO m 1.y x 10 1.43 x 10 Stream channels, average 1.17 x 1072 2.51 x 10%2 Atmospheric moisture, average 1.29 x lOlJ'L 1.72 x 1013 Subsurface water in the root zone 2.50 x lOlLL 5.38 x 1013 . : 16 15 Total circulating water 1.48 x 10 1.50 x 10 8-10 ORNL-DWG 66-8842 DRINKING WATER S H o INHALATION OF AIR DOSE RATE (mrem/ yr) N S o 2 -6 10 1965 1970 1975 1980 1985 1990 1995 2000 YEAR Fig. 8.4. Dose Rate Received by Body Tissue from 3H That Is Inhaled and Ingested in Drinking Water. 8-11 8.2 Routine Release of Radionuclides to the Atmosphere Present technology requires that fuel reprocessing plants continuously discharge off-gas and ventilation air to the atmosphere. Nonradiocactive gases are generated in some process operations; for example, air is sup- plied deliberately to some process vessels for such purposes as pneumatic liquid level determination, mixing of solutions by sparging, and maintain- ing nonflammable concentrations of gases and vapors. Since absolutely leak-tight containment barriers are impractical, a flow of ventilation air from normal working areas to enclosures (glove boxes, cells, canyons, ete.) containing radioactive materials in process equipment 1is required to main- tain a contamination gradient. By a variety of mechanisms, radiocactive gases, vapors, and aerosols of liguid and solid particles tend to become entrained in these off-gas and ventilation streams. The absolute removal of all radiocactive materials from these streams prior to discharge to the atmosphere is impractical, The policy for the routine discharge of radioactive effluents to the enviromment is to maintain the rate of release of radicactive materials at the lowest practical level consistent with current technology by care- ful control and c amtinuous monitoring. In any event, the consequences of the release must be within the limits established by federal regulations (LOCFR20), which have the intent of providing that negligible risk to the health and safety of the public will result. This policy is achieved by (1) striving to maintain process vessel enclosures free of mobile radio- active materials in order to minimize the possibility that the ventilation air will become contaminated, (2) maintaining the flow rate of the off-gas that contains (or comes in contact with) mobile radiocactive materials at the minimum practical level, (3) employing devices such as scrubbers and filters to remove as much of the radioactive material from the effluent as is practical, and (L) discharging the effluent through stacks to provide effective atmospheric dispersal. 8.2.1 Sources of Routine Releases The rate of routine release of radionuclides to the atmosphere from fuel reprocessing plants as a function of capacity (Table 8.2) was esti- 8-12 Table 8.2. Estimated Routine Release Rates for Radionuclides as a Function of Reprocessing Plant Capacity Release Rate per Unit of Throughput LWR Fuel FBR Fuel Reprocessing Reprocessing Plant?® PlantP Noble Gas 85y 1.0 1.0 33¢e 0.1 Tritium 1.0 1.0 Halogens 0.001 107" Particulates® 1.2 x 1078 8.5 x 10710 IWR fuel irradiated to a burnup of 33,000 Mwd/metric ton, at a specific power of 30 Mw/metric ton, and allowed to decay for 150 days. Off-gas rate = 1000 cfm per metric_ton per day. Filter effluent = 0,0012 mg of solution per m3. Solution concentration = 0.3 kg of fuel per liter, bLMFBR (mixed core and blankets) irradiated to a burnup of 33,000 Mwd/metric ton, at a specific power of 58.2 Mw/metric ton, and allowed to decay for 30 days. Off-gas rate = 70 cfm per metric ton per day. Filter effluent = 0.0012 mg of solution per m3, Solution concentration = 0,3 kg of fuel per liter, CParticulate release rates are assumed to scale approximately as the 0.6 power of the plant throughput rate. The rates given are estimated for a plant with a capacity of 260 metric tons per year, 8-13 mated, based on current technology for LWR fuel reprocessing plants and foreseeable technological developments for plants that will process FBR fuels, The corresponding release rates, in curies, may be obtained as the product of the fractional release (Table 8.2), the fuel processing rate (in metric tons/day), and the concentration of the isotopes in a metric ton of fuel (Table 8.3). These values permit a preliminary esti- mation of site sizes that would result from the effect of routine releases. Section 8.l will present an analysis of tradeoffs that can be made in site size through the use of additional engineered safety features. Noble Gases. - A total of approximately 0.001 ft° (STP) of the noble gases He, Kr, and Xe is generated in each megawatt-day of reactor opera- tion. The radioisotopes of physiological hazard significance that remain after 30 or more days of postirradiation decay are 85Kr and 133Xe. Un- vented fuel contains approximately 0.3 curie of 85Kr for each megawatt-darr of burnup. Unvented fuel contains about 1300 curies of 133Xe per mega- watt of thermal power after 30 days of decay and negligible quantities after 60 days of decay. In preparing Table 8.2, it was assumed that these gases will continue to be released quantitatively from LWR fuel reprocessing plants as the fuel is chopped and/or dissolved, It was assumed that, in plants for reprocessing FBR fuels after 30 days of decay, the gas would be held up (in a charcoal bed) for a period of 18 days to effect an order-of-magnitude reduction in the 133Xe activity. Several processes (employing charcoal adsorption, liquid nitrogen, Amsco, or fluorocarbon scrubbing, or perm- selective membranes), within moderate extensions of current technology, may be employed to remove 90 to 99% of both xenon and krypton if required because of particular site limitations or a strict adherence to a policy of maintaining '"lowest practicable" release rates. Release rates, partic- ularly for 85Kr, would be lower for reactor fuels that use the vented fuel concept. Tritium, - Approximately 0,025 curie of -H is formed for each megawatt-day of reactor exposure., The common and most stable compound, HTO, is practically unrecoverable by present technology after it has been mixed with water, Present plants discharge tritium essentially quanti- 8-1L Table 8.3. Radionuclide Content of ILWR Fuel Decayed lgO Days and Mixed Core-Blanket LMFBR Fuel Decayed 30 Days Concentration (curies/metric ton) Concentration (curies/metric ton) Nuciide In LWR Fuel In IMFBR Fuel Nuclide In LWR Fuel 1In IMFBR Fuel 3y 692 932 1317 2,17 139,000 85kr 11,200 10,200 1321 - 11300 895y 96,000 637,000 133%¢ - 7,100 sy 76,600 43,400 13k 213,000 29,000 Py 76,600 43,500 136 20.8 28,800 Ly 159, 000 921,000 13705 106,000 109,000 P 276,000 2,100, 000 05, 430 523, 000 P 518,000 2,660,000 140r, 195 601, 000 Mo , 1610 e, 56, 700 1,480,000 990 - 1730 Lk, 770,000 1,280,000 991¢ 14.2 1.9 h3p,, 6L 641y, 000 103, 89,100 1,760,000 W Ty 51.0 185,000 1065, 110,000 1,290,000 WTpy 99,1400 353,000 103mgy, 89,100 1,760,000 1L9py, - 61.5 iy, - 12,600 Bl 1150 1690 115me 4 L. 3 269 1525, 11.5 10.5 2L, 86. 3 76.7 1558, 6370 79,100 1255, 20.0 6720 160y, 300 9460 125, 8130 19,600 239 17.4 7220 125, 3280 6860 238p, 2810 11,200 12Ty 6180 61,100 239 330 3530 LeTpg 6110 61,800 2o, L78 1260 L29mne 6690 181,000 2blp, 115, 000 600, 000 12974 11290 116,000 b1y 200 1570 1320 - 1170 2heqn 18,000 65,500 1291 0.038 0.053 2llon 21,90 1240 %These data are taken from Tables 3.9, 3.15, 3.33, and 3.39. 8-15 tatively to the enviromment in off-gas and low-level liquid waste.ll Complete release of tritium to the atmosphere, the planned means of dis- posal at the MFRP plant,12 is assumed in Table 8.2. Advanced technology, employing either vented fuel elements or a high-temperature oxidation process after the fuel has been chopped, may reduce the rate of release of tritium from fuel processing plants by factors of 10 to 100, Halogens. - Of the fission-product halogens, only the isotopes 131 and 129I are physiologically significant after 30 days or more of post- he 1311 contents of reactor fuels are approximately irradiation decay. T 0.07 and 2400 curies per megawatt of thermal power after decay times of 150 and 30 days respectively. The 1291 content is about lO_6 per megawatt- day of fuel exposure. In current technology, iodine reports, almost completely, to off-gas systems as 12, HI, or iodine-organic compounds that are generated in such process operations as chopping, dissolving, and evaporation. Current off-gas trains use caustic scrubbers, which remove approximately 90% of the iodine, and silver nitrate towers, which remove about 99% of the remaining iodine. Through 1962, such devices were used to maintain an average 131I release rate to the atmosphere of approximately 0.3 curie/day at NRTS, HAPO, SRP, and ORNL.D’ It is assumed that plants for reprocessing fuels that have decayed at least 150 days will routinely release 0.1% of the iodine. However, plants for reprocessing fuels after a decay period of 30 days will require develop- ment of techniques for maintaining the fractional 131 of 10-7. I release in the range Particulates. - The common chemical forms of the fission products other than the noble gases, tritium, and halogens have sufficiently low vapor pressures that the predominant mechanism of release to the off-gas systems is by entrairment of particulates. While several semivolatile fission products (Tc, Se, Ru, Cs, and Te)} are known to concentrate in off-gases from certain process oper'atj'.ons,]'L‘L the general experience at ORNL in fuel reprocessing operations has been that particulates in off-gas streams have essentially the same relative content of fission products as 8-16 the fuel being processed. The explanation is that most of the aerosol in the ventilation streams consists of liquid particles that have become entrained in off-gases that have contacted radicactive solutions. The liquid particles probably have the same fission product content as the original solution since the off-gas streams generally have high relative humidities. The particles that dry after being deposited on ventilation ducts and filters largely tend to remain fixed and to contribute little +o the routine release of nomvolatile fission products. (However, they may be the source of a serious accidental release if there is a means for sudden and massive reentrainment.) At ORNL it has been found that the off-gases from aqueous fuel reproc- essing operations contain particles of aqueous solutions at a concentration of approximately 10 mg/m3 (i.e., the concentration of water particles in fog) and that there are equal weight fractions of particles in the size ranges less than O.L y, O.4 to 1.3, 3 to 5, and greater than § u'15 Also, it is known that the weight distribution of particles less than about 5y in size is relatively constant even if there is gross entrain- ment of larger particles. Typical deep-bed sand or High Efficiency Particulate Air (HEPA) filters used in processing plants would quantita- tively remove 100% of the particles greater than about 3 , in size and about 99.98% of the particles less than 3 .4, which have the size distri- bution indicated above, From these data, it is estimated that the concentration of aerosol in the filter effluent is of the order of 0,0012 mg/mB. Assuming that the radioactive solutions in the plant contain 300 g of fuel per liter (typical of the dissolver and accountability tanks, which contribute significantly to the off-gas) and have a specific gravity of about 1.2, the estimated concentration of fuel in the filter effluent is 0.3 x lO_l2 metric ton of fuel per cubic meter of air, The estimated fractional release of fuel to the atmosphere from a l-metric ton/day plant for processing >150-day-decayed LWR fuel, using current technology, is 1.2 x 10_8, assuming a combined dissolver and vessel off-gas flow rate of 1000 ¢fm. By comparison, the dissolver and the vessel off-gas flow rates are 400 and 620 cfm, respectively, at the NF'S plantl6 and approximately 500 and 1000 cfm at the Hanford Purex plant. 8-17 It is estimated that the flow rate of the dissolver-vessel off-gas at the MFRP plant will be 250 efm, +° The estimated fractional release from a l-metric ton/day plant corre- sponds to0 dally release rates of 0,037 curie of mixed fission products, 0.0006 curie of 2OSr, 0.007 curie of P5Zr->Nb, 0.00l curie of T%°Ru 0.0005 curie of lthe, and 0,00003 curie of Pu. By comparison, the 2 average dally release of nonvolatile fission products from the three Hanford processing plant stacks includes 0,01l curie of 95Zr-%Nb, 0.007 curie of 103Ru, 0.006 curie of *0%Ru, 0.001 curie of “lGe, and 0.00003 curie of total alpha emitters (presumed to be Pu).17 It is estimated that the daily release of particulates from the MFRP plant stack will consist of less than 0,006 curie of mixed fission products and less than 0.002 curie of alpha activity from plu.toniu‘m.l2 The estimated daily release of particulates from the 5-metric ton/day BNFP plant consists of less than 0,17 curie of mixed fission products and less than 0,000l curie of alpha activity from plutonium; this corresponds to a fractional release of about 1 x 10-8.18 It is estimated that technological developments will permit the dissolver and the vessel off-gas flow rates to be reduced to 20 and 50 cfm in l-metric ton/day plants that would process 30-day-decayed FBR fuel, If such is the case, the routine release of particulate activity should be lower than from current plants, in spite of the higher specific activity of FBR fuels. Tt is assumed that the routine release of radiocactive particulates to the enviromment will increase in direct proportion to the vessel off-gas flow rate in plants having larger throughput rates. The fuel inventory of individual process vessels will not increase in direct proportion to the production rate because of the necessity for multiple equipment lines to permit continuity of operation and the use of progressively more con- tinuous equipment. The routine release to the off-gas system is roughly proportional to the area of the interface between the radioactive solid or solution and the gaé. Radioactive aerosols are entrained in off-gas streams primarily by sparging (usually at a fixed rate of approximately 1 scfm/ftz), but also by diffusion and recoil from surfaces, As a first 8-18 approximation, continuous equipment will have a greater surface-to-volume ratio, which will offset the effect of larger process vessels. 8.2.2 Local Environmental Consequences of Releasi@gfiBSKr and °H Many pathways have been postulated by which radionuclides may be transmitted through the environment and thereby contribute to the total dose received by man.19 A generalized model that relates the principal parameters involved in estimating the external dose is as follows:2o t 2 ext - D; 3 [t toy v(t)] = Q 33 P35k (%) C; s [y(t)] dt, (1) where ext . . . s Dijk [t15 by, v(ty)] = total external dose to radionuclide 1 in pathway j at location k for an individual of age Y(tl) at the beginning of exposure, Qij = quantity of radionuclide i released that is entering or available to pathway Jj, Pijk(t) = concentration of radionuclide i in pathway j at location k during time t per unit of radionuclide initially available, and n Cij[Y(t)] dose rate to the reference organ of an individual of age y per unit concentration of radionuclide. The total external dose due to radionuclide i in pathway j at location k, accumulated from time t, to t, by an individual of age Y(tl) at the beginning of exposure, is the integral of the product of the level of contamination (the quantity Qij and the concentration Pijk) and the dose rate term, Cij' The later term includes all necessary factors that account for the habits and characteristics of the individual. With minor changes, the same expression can be used to estimate internal dose., For internal dose, the Cij term denotes the dose commitment in the (t2 - 1) days following a one-day exposure of the individual, 8-19 According to the International Commission on Radiological Protection, the entire human body is the critical organ for exposure to 85Kr.8 The principal mode of exposure is submersion in contaminated air. Body tissue is the critical organ in the case of exposure to tritium as tritiated water or tritiated water vapor. However, the external dose resulting from sub- mersion in air containing HTO vapor is limited to areas where the skin has minimal thickness, because of the limited penetration range of tritium's beta particle. Prior studies at Hanford and Oak Ridge have demonstrated that not all modes of exposure, or pathways contributing to the same mode, are of equal 21,22 The modes of exposure considered in this analysis will importance., include ingestion, inhalation (and accompanying skin absorption), irradia- tion from a contaminated surface, submersion in contaminated water, and submersion in contaminated air. These egtimates of dose consider only the dose to "standard! man. Procedures for Estimating Permissible Release. - Acceptable release 85 ing plant located at the Oak Ridge National Laboratory. This selection rates for ~“Kr and 3H were investigated for a hypothetical fuel reprocess- was made since information was already available on some of the environ- mental factors that influence the dispersion and possible reconcentration of fission products that may be released. Average annual downwind air concentrations are calculated by a modi- fied Gaussian plume formula as follows: S R 2,032F(93).Q 2 Xox) = ) L S s C ‘:"_h_z : (2) -t ¥ u, . i=1 2 1 Zj(SX)Z where X(9x) = average annual concentration along a 22.5° arc at distance x in direction 3 (curies/m3), F(es)i = fraction of time that the wind is in direction g, for stability S and wind speed group i, Q = initial emission rate (curies/sec), c}(Sx)Z = vertical dispersion coefficient at distance x for stability S (m), 8-20 E(eS)i = average wind speed in direction g, for stability S and speed group i (m/sec), h = stack height (m), R = index denoting wind-speed groups, S = index denoting stability parameter, This expression is obtained by integrating the Gaussian plume formula over the crosswind direction and distributing the results uniformly along the entire arc. Since the average wind-speed vector and its frequency of occurrence are used, calculations yield average annual air concentrations, Applications of this technique have been demonstrated previously by Culkowski. 2> Equation (2) is modified to include washout and fallout by multiplying by the appropriate correction factors., Corrections for washout and fallout are based on the work of Chamberlain and Slade respectively.zh’25 These corrections are as follows: _ _ X Ywashout ~ S*P [ T(63), } ’ (3) where ) is the washout coefficient (sec_l); and 12 U 9 h° Qfallout = exp {- (2/n) L—:-‘(—g'g-):f ;@-fi; exp{- m] dxj, (L) where Vg is the deposition velocity (m/sec). Equation (L) can be evalu- ated numerically, based on curves of o, values given by Hilsmeier and Gifford.26 Figure 8.5 shows the calculated air concentrations at the ground surface for a l-part/sec release from a 100-m stack located at ORNL. The most recent meteorological data reported by Hilsmeier are used in these calculations.27 Concentrations shown in Fig. 8.5 can be compared with others that include fallout, washout, and changes in stack height; by this process, average annual doses can be estimated for a variety of conditions. 8-21 ORNL-DWG 66-1024 36°15 84°37'30" 84°30' 84°22'30" 84°15' 84°07'30" 84°00' 1 T 1 [ KENTUCKY 7 I J l ‘ 1078 TENNESSEE % cnggfm éf‘TKYE \NORRIS DAM /MISSISSIPPI ‘I ALABAMA ‘ GEORGIA IN 5x 10_9\_J OLIVER SPRINGS , ¢ -8 5x10 MILES B g gxig® ‘ KINGSTON MELTON HILL DAM a—TENNESSEE RIVER \__~ -8 3x10°8 LENOIR CITYN -0 - _ / 22 Fig. 8.5. Average Annual Air Concentrations at Ground Surface in Parts per Cubic Meter. Source height, h = 100 m; source strength, Q = 1 part/sec. 36°00' 36°07'30" 35°52'30" 8-22 Washout. - Washout coefficients for soluble gases have been calcu- lated by Chamberlain, using the assumption that the rate of absorption is 2,28 controlled by the rate of gas diffusion to the raindrop. Since the solubility of krypton in water is small (1.85 x 10710 T g of krypton per gram of water at equilibrium),' it was assumed that the solubility limit controls the amount of krypton absorbed. The solubility of 85Kr in rain- water, even when released at 1 curie/sec, would be limited by the stable krypton in the atmosphere (about L x 1073 g/m3 near sea level).29 It was further assumed that krypton is washed out of the atmosphere, beginning at an average height of 1 mile. This assumption is based on the height of rain-bearing cumulus clouds and on the extent of vertical development of radioactive clouds released as a point source. The average intensity of rainfall is about }} mm/hr in the Oak Ridge area,BO and, at equilibrium, 2 X ].O-]'LL g of krypton per second could be absorbed in a column of the atmosphere 1 mile high and 1 em’ in area. About 5 X ].O-)‘L g of stable krypton per square centimeter is contained in the atmosphere to a height of 1 mile. Based on these considerations, the average washout coefficient was calculated to be: _2x lO‘lLL g of Kr sec™’ em™? = = =) x 107 sect . 5 x 107" g of Kr cm A The washout coefficient of tritiated water vapor (HTO) has been estimated from Chamberlain's calculations for 802 deposition in rain- 2l cient of the vapor in air. Therefore, the following expression was used to calculate Aypo for a Ji-mm/hr rainfall: water, It was considered t0 be proportional to the diffusion coeffi- D _ HTO _ -1 2 802 where AHPO = washout coefficient of HTO vapor (sec-l), washout coefficient of S50, (2 x lO"LL sec_l),zh o ! oo = diffusion coefficient of HTO vapor in air (0.23 cmz/sec),31 o Il 30 diffusion coefficient of SO2 in air (0.115 c1'r12/sec).2)‘L 8-23 Loss of HTO from a raindrop to the atmosphere was assumed to be negligible. This assumption is valid if the distance the raindrop falls below the con- taminated cloud is small as compared with the relaxation length.% A washout coefficient of L x 10“)‘L se with that indicated by Chamberlain and Eggleton, ¢~ for HTO vapor is consistent 32 Similar values can also be deduced from published data on the comcentration of tritium in the atmosphere and in rainwater., For example, the maximum concentration of tritium, in tritium units (TU), was reported to be lO6 in hydrogen,33 35 L Assuming the average water content of air to be 8.6 g/m3 (at 50% relative 10° in water vapor,33 2 x 107 in methane,Bh and 1.l x 10° in rainwater. humidity and 20°C) and using the values of TU listed above, the concentra- 11 tion of tritium in the atmosphere is estimated to be 2.9 x 10~ curie/mB. The tritium content in a column of the atmosphere 1 mile high and 1 m2 in 8 curie. The rate of tritium removal from a l—m2 area -12 area is .7 x 10~ by a h-mm/hr rainfall would be 5.1 x 10 washout coefficient is calculated to be: curie/sec. Therefore, the 12 curie/sec _ 5.0 x 10~ -1 h.7 x 10'8 curie A = 1.1 x lO-LL nec Since the annual frequency of a lil-mm/hr rainfall in Oak Ridge is only 0.037, the average annual ground-level air concentrations are not reduced significantly at these washout coefficients. Fallout. - If the sorption of a radionuclide by the ground surface is irreversible, the flux of the radionuclide to the surface does not depend on the amount already deposited.36 Chamberlain describes the rate of deposition for such a system in terms of a deposition velocity. The following equation is used to estimate the deposition velocity of gases . 28 or very small particless — (5) 1n (ku%ZlD-l) vg(Zl) = *Relaxation length is the distance in which the isotopic composition of the raindrop decreases by l/e. 8-24 where V_ = deposition velocity (cm/sec), k = von Karman's constant (0.L), u* = friction velocity (cm/sec), Zl = reference height above ground surface at which the concentration of the radionuclide is measured (cm), D = molecular diffusivity (cmg/sec). By assuming u% = LO cm/sec (appropriate to the Oak Ridge area),37 Z, = 100 ¢m, and D = 0,15 cm2/sec (diffusion coefficient of krypton in nitro- gen),38 the deposition velocity of krypton is 1.7 em/sec., For tritiated water vapor, with D = 0,23 cm2/sec, the deposition velocity is 1.8 cm/sec.Bo The retention of krypton by the soil is assumed to be limited by the adsorption capacity of the soil for krypton. The retention of krypton by soil can be estimated, assuming that the amount of adsorbed krypton is proportional to the surface area of the soil. From measurements of krypton adsorption on charcoal (2 x 10"6 g of krypton per gram of charcoal at 25°C and ]_O—3 mm Hg partial pressure)39 and the ratio of soil area to charcoal surface area (O.OS),LLO the adsorption of krypton by soil is estimated to be 107 g per gram of soil (or 1.2 x 107" g/cm3 for a soil density of 1.2 g/cmB). The rate at which krypton is deposited on the soll is esti- mated as the product of the deposition velocity (Vg = 0,017 m/sec) and the krypton concentration in the atmosphere (X = L x 1073 g/mB), or 6.8 x lO_5 g n™° sec™t, At this rate, the soil will probably become saturated with krypton and may not act as a perfect sink for the addition of 85Kr. The amount of 85Kr adsorbed on the soil at equilibrium is assumed to be directly proportional to the ratio of radioactive and stable krypton in the atmosphere, For a 85Kr release rate of 1 curie/sec, the soil load (at equilibrium) at the point of maximum ground-level air concentration would be the product of 1.2 x 107! g/em> (soil) and 4 x 1077 g/m° divided by the product of L4 x 1073 g/m3 (air) and 397 curies per gram of 85Kr, or L.8 x 10-11 curie/cmB. Since the adsorption of 85Kr by the soil may not be an irreversible process, the net flux of 85%r to the soil (g m™° sec'l) may change as the 8-25 soil approaches saturation. The deposition velocity calculated from Eq. (L) can be used to estimate only the initial flux of 85Kr to the soil (and cloud depletion by fallout). The flux to the soil would be expected to diminish with time until steady-state conditions are attained. The mechanisms by which HTO vapor may be retained by the soil would probably include adsorption, condensation, and exchange with soil moisture. Evaporation, evapotranspiration, and soil drainage would act to redistribute the deposited material., Water vapor (HQO) in the atmosphere would also be acted upon by these mechanisms and would compete with HTO for retention by the soil, In the absence of isotopic fractionation, the ratio at equi- librium of the deposition rate of HTO vapor to H20 vapor would be directly proportional to the ratio of their respective concentrations in air, A deposition velocity of 0.018 m/sec can be used to estimate the flux when the soil acts as a perfect sink, Assuming an average water vapor content in the atmosphere of 8.6 g/m3, the flux of water vapor to the soil due to fallout would be 0.15 g n™° sect (4.7 x 10° g m™° year'l). The average rate of rainfall in Oak Ridge is 1.1 g n™¢ sec™t., For a frequency of rainfall of 0,037, the quantity of rainwater deposited each year is 1.3 x lO6 g/mg. These rates imply that, if the soil acts as a perfect sink for water vapor fallout, the soil would receive an amount of water equivalent to a continuous rainfall of about 0. mm/hr. Obviously, this does not occur; thus the soil would not act as a perfect sink for either HTO or E,0 vapor, and the flux of HITO vapor to the soil would be expected to vary with time. Only a free water surface, such as the Clinch River, can be assumed to act as a perfect sink for HTO vapor that is released from a stack. Further studies are necessary to evaluate the flux of both 3H and 85Kr to the soil during transient and steady-state conditions. As a first approximation, the following conservative assumptions are made: (1) the contaminated cloud is not depleted of 85Kr and HTO by fallout; (2) the quantity of 85Kr retained by the soil or by the Clinch River is proportional to the ratio of radioactive and stable krypton in the atmosphere; (3) the quantity of HTO retained by the soil is propor- tional to the ratio of HTO vapor and HEO vapor in the atmosphere; and (L) the Clinch River is a perfect siik for HTO vapor. 8-26 Krypton and H2O vapor may be adsorbed on particles in the atmosphere and, therefore, be deposited on the ground with these particles. The quantity of krypton associated with particles is estimated by assuming that the air contains 1.L x 10_)'L g of particles per cubic meter (average of city atmosphere)hl and, as an upper limit, that these particles can adsorb as much krypton as charcoal (2 x 10_6 g of krypton per gram). Adsorption of krypton on particles is estimated to be 3 x 10_10 g per cubic meter of air, which is negligible as compared with the krypton in the atmosphere (4 x 1073 g/m3). Assuming that charcoal particles can retain two layers of water vapor, the adsorption of water vapor by the particles is estimated to be 7 x lO"5 g per cubic meter of air., This value is negligible as compared with that of water vapor in the atmosphere (8.6 g/m>). Dose Estimation Models. - Methods described and parameters given in ICRP Publication 28 are used to convert concentrations (X in curies/mB) to estimates of dose equivalents to "standard! man from submersion in a contaminated cloud, from ingestion, and from inhalation. In particular, Egs. (12), (13), and (20) in ref. 8 are used, and equilibrium conditions are assumed where appropriate. These dose equations are summarized in Table 8.L. Submersion dose rates in contaminated water were calculated by assum- ing that the body is in the center of a sphere and receives equal quantities of radiation from all directions.22 Other assumptions included: (1) the radius of the contaminated fluid is large as compared with the range of beta particles and to the half thickness of the fluid for gamma rays, (2) an effective energy that is equal to the average energy of the beta particle is absorbed, and (3) penetration distance for the beta particle in the body is short, thus limiting beta radiation to skin and subsurface tissue, The following expressions were derived to calculate dose equiva- lents at the surface of a body submerged in contaminated fluid: For 85KI': R = 0.26 X rems/hr For “H: R =1.1x 1072 X, rems/hr, i Table 8.l4. Equations to Calculate Dose Equivalents (rems per week) to Standard Man® LO-hr Week Exposure 168-hr Week Exposure Exposure Critical ModeP Organ 3H Aggkr 3H 85Kr Inhalation and skin I I absorption Total body 1.2 x 10 Xa 3.6 x 10 Xa Inhalation and skin L A absorption Body tissue 2.0 x 10 Xa 5.8 x 10 Xa Ingestion Total body 0.67 X L9 X Ingestion Body tissue 1.1 Xfi 3.2 Xw o | N Submersion in air Total body 9.2 x 10° X, h.0 x 10)'L X ~ Submersion in air Skin 3.9 x 102 Xa 1.7 x 103 Xa External exposure, 2.5 ft above con- taminated ground 1 5 surface Total body 2.4 x 10 X, 1.0 x 10" X a . . . . Dose rate, in rems/week, when the concentration in air, X expressed in units of curies/m3, as Or the concentration in water, X,, is Exposure mode and critical organ for inhalation and skin absorption, ingestion, and submersion in air are based on information contained in ref. 8. 8-28 where Xw is the concentration of 85Kr or 3H in the fluid in microcuries per gram of fluid. Hine and Brownell describe the derivation of equations that relate to the calculation of dose rates in air from beta emitters associated with an infinite plane of negligible th:'_ckness."L2 Equations (10), (11), (20), and (21) in ref. L2 are selected for calculation in cases where the energy- dependent parameters are those adapted for dose estimates in soft tissue. Equation 9-30 from work by Morgan and Turner is used to calculate the dose due to gamma emitters when the source is of infinite planar extent and infinite thj.ckness.LL3 External dose equations listed in Table 8.L for soil contaminated with 85Kr are then derived from the expected soil load (L,.8 x 107H curie/cmB) at the maximum air concentration (1.6 x 10-6 uc/cmB). The range, in aluminum, of the average-energy beta particle from BSKr is used to estimate the thickness of contaminated soil contributing to the beta radiation dose and, thus, the amount present per unit area. The beta radiation dose rate is calculated by assuming that this amount of 85Kr is spread uniformly over the surface without taking self-absorption within the soil layer into consideration, Fstimated Dose Equivalents. -~ For the purposes of this analysis, we have chosen 85Kr and 3H release rates of 0.55 and 0.03)4 curie/sec respec- tively. These release rates correspond to a reprocessing plant with a capacity of about 6 metric tons/day (a fuel exposure of 33,000 Mwd/metric ton and a specific power of 30 Mw/metric ton). All of the 85Kr is assumed to be released to the atmosphere. It is assumed that 0,0085 curie of 34 per second is released to the atmosphere as HTO vapor and 0.0255 curie of 3y per second is discharged to the Clinch River at mile 20.5 (below the Oak Ridge municipal water intake and above the water intake for the Oak Ridge Gaseous Diffusion Plant) as liquid waste and is diluted with L900 £1° of river water per second. Other schemes of 3H release, such as the dis- tillation of 3H-bearing liquids and release to the stack as water vapor, are possible, but would require an appropriate adjustment in the dose estimates that follow. 8-29 Surface water in the area can be contaminated directly by fallout and washout of 85Kr and 3H, as well as by the direct release of HTO in liquid waste. Clinch River water is assumed to equilibrate with 85Kr at the maximum specific activity expected in the atmosphere up to the solu- bility limit of krypton in water. Soils that equilibrate with 85Kr or 3H from the overlying atmosphere are assumed to retain these materials. The contribution, by washout, is based on the deposition rates that are calcu- lated in the northeast sector; and these are the maximum rates. Assumptions made for the addition of 85Kr by fallout and 3y by washout would then give conservative estimates of concentrations in Clinch River water. Figure 8.6 shows the average annual dose equivalents in millirems per year, to the total body for submersion in air containing 85Kr. Exposures are assumed to be continuous (168 hr/week and 50 weeks/year). These average dose rates were calculated from the ground-level air concentrations (Fig. 8.5) that result from a l-curie/sec release rate and a negligible cloud depletion by washout and fallout. Figure 8,7 shows the estimated dose rates for continuous exposure, in millirems per year, at a distance 2.5 ft above a ground surface contaminated with 85Kr. Tonizing radiation associ- ated with tritium on the ground surface would be shielded effectively by 2.5 ft of air, Table 8.5 contains the estimated annual dose equivalents, to the standard man working at the Oak Ridge Gaseous Diffusion Plant (CRGDP) or residing in Oak Ridge, due to the release of 0,55 curie of 85kr and 0.034 curie of 3H per second in the emwironment. Periods of occupancy are L0 hr/week and 50 weeks/year for the ORGDP employee, and 168 hr/week and 50 weeks/year for the Oak Ridge resident. A '"less than” sign preceding certain values reflects a conservative estimate. The critical modes of exposure are submersion in air for 85Kr, and inhalation and absorption through the skin for 3H. The estimated total-body exposure, due to re- leases from a 6-ton/day plant, is about 90 millirems/year for the standard man residing in Oak Ridge. Interpretation of Results. - The Federal Radiation Council (FRC), in consideration of a linear relationship between biological effect and dose, background radiation, benefits and risks to be derived from radiation use, 8-30 ORNL-DWG 64-8697R2A 84°37'30" 84°30' 84°22'30" 84°45' 84°07'30" 84°00" T T T 1 j _l.fl KENTUCKY 7 = 5 A LAKE § - TENNESSEE cnsgsmfl CITY ‘o\?:: NORRIS DAM [MISSISSIPPI ) ALABAMA \ GEORGIA IN 1055__/} = " - 4~ o Q QO M ot 6& OLIVER o SPRINGS e i 18 o o0 "M N 2 345 'R R MILES n 5 Tw KINGSTON 100 MELTON HILL DAM o N TENNESSEE RIVER N /\ 20 LENOIR CITYN " . i L 1 Fig. 8.6. Average Annual Ground-Level Dose Equivalents, in milli- rems/year, to Total Body for Submersion in Air Containing 85Kr. Source height, h = 100 m; source strength, Q = 1 curie/sec. 8-31 ORNL=-DWG 64-6697R3 B4°37°30" 84°30" 84°22'30" 84°15' 84°07'30" 84°00" T gl l hmzmucmr / — \ ob f,/\&__',w % TENNESSEE % LAKE 0'0 i ,NFVORRIS DAM " AN CiTY’ /MiSSiSSlPPG ] ALABAMT\ GEQRGIA ‘N | , ' - P 0.027 ‘ (,@él = 3 ’ HINES » - . CLINTON X S (F‘é 0‘:?:.‘ g < o % v ¢ of qfi' ¢ sc OLIVER \ 054 SPRINGS 7 ©76 S : ] - \J \ 'é‘,;,o 0 ’0.09 00 w ,{’J”UAK A ——— 1 % ‘ ..v’&v L N [’_értp v 2 :JCB o f / : " \ Y-12/ N / HARR:MAN_/J , a\?‘? < - 0 12345 - o A.’--.—-\w\ ™ O X i}§§\¥\w MILES 3 | . S i st KINGSTON © 2 L7 /’/ -2 1 LT \ TENNESSEE RIVER 1 : ! ‘:{ f } f" v /_\ 0-06& oo P \L/ LENOIR CITYN Lo / / m : Fig. 8.7. Average Annual Dose Equivalents, in millirems/year, at a Distance 2.5 ft Above Ground Surface Contaminated with 5Kr 100 m; source strength, Q = 1 curie/sec. height, h = Source 8-32 Table 8.5. Estimated Annual Dose Equivalents, in millirems, Received by the Standard Man due to a 6-metric ton-per-day Reprocessing Plant Located at ORNL Dose Rate (millirems/year) Mode of Reference Exposure™ Organ® Employee of ORGDP Oak Ridge Resident Krypton-85 Submersion in Total body 13 88 air Submersion in Total body <0.006 <0,006 water Contaminated Total body 0.03 0.2 ground (2.5 ft above surface) Tritium Inhalation and Body tissue 0.43 1.9 skin absorp- tion Ingestion of Body tissue 10.0 <0.08 water Submersion in Skin 0.009 0.06 air Submersion in Skin 0.22 <0.001 water aExposure mode and reference organ for submersion in air, inhalation and skin absorption, and ingestion of water is based on information contained in ref. 8. 8-33 and other factors, established, as its basic recommendation, that the annual radiation exposure to the whole bodies of individuals in the general population (exclusive of natural background or medical exposures) should not exceed 0.5 rem.9 In the event of widespread radioactive con- tamination, and because of uncertainties in the relationship between average and maximum exposure, the FRC suggests the use of the arbitrary assumption that the majority of individuals do not vary from the average by a factor greater than 3. Thus, the use of 0,17 rem for the annual whole-body exposure of average population groups is recommended. When the size of the population group under consideration is sufficiently large, consideration must also be given to the contribution of the genet- ically significant population dose. According to the FRC, "The use of 0,17 rem per capita per year, as described in paragraph 5.4 as a technique for assuring that the basic Guide for individual whole body dose i1s not exceeded, is likely in the immediate future to assure that the gonadal exposure Guide is not exceeded.” These guides are essentially in agreement with current recommendations of the ICRP and NCRP. Each agency also encourages that every reasonable effort be made to keep exposures as far below the offered guidance as practicable, In current reports, the ICRP and NCRP list the total body as the critical organ and submersion in a semispherical infinite cloud of radio- 85y 8,hl active gas as the critical mode of exposure for However, the basic recommendations in effect at the time these reports were published considered the whole body and the blood-forming organs as a unit, and, as mentioned above, even the genetic dose was partially related to whole- body dose. Because of the rather short range of the beta radiation from 85Kr, only a small fraction of the total mass of the blood-forming organs or the testes would be exposed to a significant part of the beta dose to skin; however, this might be as much as 1 g of red marrow (e.g., in the skull). The mass of 1 g was previously used as a basis for dose assess- ment.lLS In later publications of the ICRP, the principle of averaging the dose over organs and tissues is stated without qualification., This L6 penetrate well below the skin layer, as shown subsequently, a significant principle would permit a higher dose, Since the beta radiation does 8-3L volume of body tissue would be irradiated at 50% or greater of the surface skin dose. If this tissue is to be limited to 1.5 rems per year, an increase by about a factor of 3 or slightly more might be warranted. Krypton-85 decays principally by emitting a 0.5l4-Mev photon 0.7% of the time and a beta particle of 0.695 Mev maximum energy 99.3% of the L7 body submerged in a semispherical infinite cloud containing 85Kr is com- time. Calculations indicate that the total dose at the surface of a posed of about 99% beta and 1% gamma, The ranges in tissue of the beta rays of maximum and average energy are estimated to be 2.6 mm and 0.55 mm respectively. A considerable fraction of the beta particle energy will be deposited, on the average, in the epidermal (range in thickness, 0.023 to 0,070 mm) and dermal (average thickness, 0.70 mm) layers of the skin of the total bod;)r.)'L8 Thus, there is reason to reevaluate the total body 85 as the critical organ from submersion exposure to “Kr as a function of depth-dose relationships. For the complete release of 85Kr and 3H from a fuel reprocessing plant sited at ORNL, 85Kr would be of greater dose potential to man than 3H. Of the modes of exposure considered, submersion in contaminated air would deliver the largest dose, that is, about 90 millirems per year for a 6-ton/day plant. As explained above, current guidance for total-body exposure to 85Kr limits the maximum permissible dose of individuals in the general population to 500 millirems per year (and of average popula- tion groups to 170 millirems per year). The potential dose resulting from the release of 3Y in liquid waste is small because credit can be taken for dilution in the Clinch River in which flow is substantial (1,919 ft3/sec) and the river is not used as a source of municipal water. Dose estimates by the ingestion of water (10 millirems annually) at ORGDP would increase in direct proportion to a reduction in flow rate and increase by a factor of 3 if the water were used as a municipal water supply. Disposal of 3H in water vapor released to the stack may be one way to reduce the potential exposure from ingestion of water. i 8-35 Economic benefits would be expected to accrue from large processing plants, but remote siting may not be a practical method for restricting population exposures in the future. This is the justification, therefore, to continue research and development studies, now in progress, to reduce the amounts of 85Kr and 3H released and to understand more completely the fate of these radionuclides after discharge to the envirorment. 8.2.3 Local Envirommental Consequences from All Routine Releases Although the routine releases of 85Kr and 3H were emphasized in the preceding section, the absolute removal of all other radioactive materials from gases and vapors prior to discharge to the atmosphere is impractical. Of the remaining radionuclides, 1311 is known to be important because of reconcentration that occurs in the grass-cow-milk pathway to the thyroids of small children., Less experimental information is available on the behavior of 1291 in the erviromment, but the assumption will be made that the grass- cow-milk pathway is the dominant mode of exposure from this radioisotope. The controlling pathways for exposure from particulates of mixed fission products and actinides are, alsc, not well understood. However, it is known that, under some circumstances, such effects as reconcentration in fish or crops and resuspension may be important. In this analysis, it will be assumed that the major exposures from the atmospheric release of particulates will result from direct inhalation of the contaminated air,. Based on the results of the preceding section, it will be assumed that the maximum acceptable average annual concentrations of 85Kr and 3H in air at the boundary of a fuel reprocessing plant site are 1 x 1077 and 7T x lO"8 curies/m" respectively. These are the values recommended by 10CFR20, Appendix B, Column II, and correspond to annual whole-body exposures of 170 millirems. Experimental evidence has suggested that the average annual concentration of 1311 4n air, as provided by 1OCFR20, should be reduced by a factor of about 700 to account for deposition followed by the grass-cow-milk pathway.h9 It is assumed that this same reconcentration factor of 700 should be applied to 1291, but that, in addition, another factor of 10 is required to account for the relatively longer effective half-1ife of 1291 on grass. GConsequently, the assumed 8-36 1317 ana 1291 at the maximum acceptable average annual concentrations of site boundary are 1 x 1071°/700, or 1.l x 1073 2 x 10'11/700/10, or 3 x 10'15 curie/mB, respectively. The assumed acceptable average annual air concentrations of particulates containing curie/mB, and mixtures of radionuclides are weighted average values that were derived using one-third of the 10CFR20 concentrations for specific nuclides and relative radionuclide concentrations from Table 8.3. These assumed values are 1 x 1070 3 x 10710 L x 10713 curie/m3 for the mixed actinides from either type of fuel. curie/m3 for mixed fission products from the LWR fuel, curie/m> for the mixed fission products from FBR fuel, and Maximum site boundary distances dictated by the routine release of radionuclides to the atmosphere were estimated by assuming average annual concentration parameters that prevail in the direction northeast of ORNL (Fig. 8.5). Figure 8.8 compares this concentration parameter for the northeast direction at ORNL with corresponding parameters that have been estimated for the Hanford,so, NRTS,Sl and Savannah River Sites.h9 The dashed curve labeled "I" shows the concentration parameter for iodine at ORNL that would result if the iodine were depleted from the plume with a deposition velocity of 0,04 m/sec.52 The ORNL, Hanford, and NRTS data presented in Fig. 8.8 are based on meteorological calculations averaged over annual-weather conditions, but they are known to be reasonable based on long-term environmental monitoring studies. The Savannah River data are derived from results of air sampling studies for 131I made at the site boundary over a periocd of one year. The Savannah River data reflect the depletion of iodine in the plume. Table 8.6 presents estimates of the site boundary distances and resultant average annual concentrations of the various species of radio- nuclides that would be dictated by routine releases from conceptual IWR and FBR plants sited at ORNL. These estimates assume that the plume is not depleted by deposition, fallout, and washout. Table 8.6 also gives estimates of the average annual concentrations of radionuclides at the site boundaries of the NFS,© MFRP,12 and BNFPL® plants. These latter results were taken from the Safety Analysis Reports for the three plants; thus the assumptions made in the calculations are not necessarily the same 8-37 ORNL DOWG 69-13160 10~6 HANFORD ° Kr - ~ O < n 1077 £ . Eé o - 2 N\ OrnL J B \ Kr & \ & " I 2 \ = \ — é 10-8 \\ \ > N \ b \ > - SRP \ o - I \\ U o \ i \ - \ jo-9L—L 11 11 Lo b el l L1111 | io 100 DOWNWIND DISTANCE (km) Fig. 8.8. Effect of Downwind Distance on the Average Annual Down- wind Ground Concentration per Unit Emission Rate from a 100-m-tall Stack. Table 8.6. Fraction of Maximum Permissible Average Annual Air Concentrations Resulting from the Routine Release of HRadionuclides at the Site Boundaries of Existing, Proposed, and Conceptual Private Industrial Fuel Processing Plants (260 days of operation per year) s s Average Fuel Characteristics Distance Annual Fraction of 1/3 x(LOCFRZ0) Concentrations at Site Boundary®’P Plant Specific Decay to Site Aeolian Capacity Burnup Power Period Boundary Dilutign 8 1 129. 131 Fission Product Actinide Plant (metric tons/day) (Mwd/ton) (Mat/ton) (days) {¥am) (sec/m”) SKr-133%e 3y 1431 Solids Solids NFS 1 20,000 32 150 1.5 2.2 x 1077 0.23 0.002 047 0.0007° - (3,300,000) (18,000) (3.1) (~1) - MFRP 1 L3,800 30 160 0.6-3 1.1 x 1077 0.12? 0,005 0.23 <0,0005 <0,11 (3,300,000) (100, 000) (3.1) (<2.2) (<0.63) BNFP c.8 35,000 L0 160 2 C.7 x 10'8 0.2h 7 0,02 0.27 0.003 0.017 (1.4 x 107 (600, 000) (21) (60) (3.5) LWR 1 33,000 30 150 <0.6 6.3 x 1077 0.58 5 0.054 Q.15 0.003 0,021 (2.9 x 10%) (180,000) (0.56) (13) (0.43) LWR & 33,000 30 150 0.5-6 1.8 x 1077 1.0 7 0.093 0,25 0,002 0.018 (1.7 x 10") (1,100,000} (3.4) (41) (1.3) LWR 36 33,000 30 150 529 3.0 x 1078 Lo 4 0.093 0.25 0.001 0,009 (1.0 x 10°) (6,500,000) (20) (120) (3.8) FBR 1 33,000 28 30 <0.6 6.3 x 1077 0.9 0.073 0.52 0.0003 0,008 (4.6 x 107) (240,000) (3.6) (h.5) (0.16} FBR 6 33,000 58 30 1.5-10 1.1 x 107/ 1.0 0,079 0,56 0,0001 0,003 (2.8 x 107 {1.450,000) (22) (9.0) (0.31) FBR 36 33,000 58 30 7-L2 1.9 x 1070 1.0 4 0.079 0,56 0.0001 0.003 (L.7 x 10%) (8, 700,000) (130) (sk) (1.9 aThe reference values selected are one-third of the concentrations found in 10CFR20, Appendix B, Table II, Column 1, They are 1 x lO-?, 7 x 10'8, 1 x lO-lo, 3x lO_lO, and L x lO-13 for 8SKr —133Xe, 3H, mixed LWR fission products, mixed FBR fission products, and mixed actinides respectively. The 10CFR2Q value for 1311 was reduced by a factor of 700, resulting in a reference concentration of 1.k x 10-13. The 10CFR?0 value for 13T of 1. x 10-13. The 10CFR20 value for 1?91 was reduced by a factor of 7000, resulting in a reference concentration of 3 x 10-15. I was reduced by a facter of 700, resulting in a reference concentration b . \ . Release rates, in curies/year, are given in parentheses. i alf - il - i | [ i il o il ) Al » sk Y N g ik i i ] o " il L] o i It i g¢-g 8-39 as those employed for the present analysis of conceptual plants. The com- parisons are of value in that they reflect the range of results that can be obtained through the use of various assumptions and computational techniques, as well as point out differences that may exist in meteorclog- ical conditions from site to site, The large site boundary distances that are estimated for plants of high capacity provide incentive for removal of a larger fraction of the noble gases and iodine than was assumed in Sect. 8.2.1. This will be considered further in Sect. 8.l after estimates are presented of the site boundary distances that are dictated by upper limit accidents, 8.3 Accidental Releases of Radioactive Materials Fuel processing plants utilize three barriers for the confinement of radioactive materials. Accidents may cause the primary barrier to fail and, in turn, radioactive gas, liquid, or aerosol (usually under pressure) to be discharged to the second barrier. The first confinement barrier consists of the process vessels, the associated interconnecting piping, and the highly efficilent vessel off-gas train. The second barrier is the thick concrete cell wall, which is designed to provide radiation shielding and to limit the effect of the maximum explosion in a process vessel within the cell to minor leakage of air or gas to the third barrier, The latter barrier, an industrial building, surrounds all penetrations in the cell walls. Under normal conditions, outside air is drawn into the building through (1) a roughing filter, (2) a check valve and another roughing filter to the cells, and (3) a ventilation duct (where it mixes with the effluent from the off-gas train) and HEPA or deep-bed filter to blowers, which exhaust to a stack. Normally a portion of the ventilation air from the building does not pass through the cells but flows directly, through a suitable restriction, to the upstream side of the filters. In an acci- dent situation, in which one or more cells may become pressurized, this latter flow tends to maintain the building at a negative pressure with respect to the environment. Glove-box facilities have three barriers - the box, the laboratory, and the building — which have comparable confine- ment potential to the vessel, cell, and building. Mobile materials in storage canals are confined by a container, the water, and a building. 8-L0 Potentially, liquid waste management facilities also have three barriers of confinement — the tank, a vault, and a building, In present practice, however, massive failure of the tank (such as by a hydrogen-air explosion in the vapor space), resulting in significant pressurization of the vault, is not considered credible because of the assumed reliability of preventive measures; therefore, the third barrier (a backup floor pan and a building) may not be considered necessary. By making the more pessi- mistic assumption that a hydrogen-air explosion in a waste tarnk is credible, it is assumed in this study that either the waste tank or the vault (which is possibly vented through a large pipe to other vaults or to cells of the processing plant) is desighed to contain the explosion (a maximum of ~100 psig in the vapor space of the tank), resulting in only minor leakage that is confined to a building and routed through a filtered ventilation system. The following basic assumptions were made for the purpose of assessing the effects of credible accidents in fuel reprocessing plants: (1) The secondary containment barrier (cell, vault, water in the storage pool, and ventilation-filter system) and the building can, and will be, designed to maintain their confinement potential following exposure to any credible internal forces. (2) Process and confinement systems can, and will be, designed in such a manner that exposure to credible external events or forces (loss of power, earthauake, tornado, flood, hurricane, impaction by moving vehicles, etc., but not including acts of war) will not impair the ability to shut down the plant safely and maintain safe shutdown conditions. The following sections will describe more detailed assumptions that have been made with respect to the properties of fuel reprocessing plants and waste management facilities, estimates of the fractiomal release of radioactive materials resulting from accidents, a model for the assessment of downwind consequeunces of a release, and implications of the estimated dose rates as a function of distance downwind, 8-L1 8.3.1 Assumed Properties of Fuel Reprocessing Plants Properties of fuel reprocessing plants as a function of capacity (see Table 8.7) have been assumed for the purpose of estimating the fractional release of radiocactive materials in the event of an accident, With a few exceptions, the contaimment and confinement features that were selected represent either present or only moderate extensions of current technology. Future large-capacity plants will, undoubtedly, have many properties different from those selected; however, it is assumed that the important derived numbers (i.e., the quantities of radioactive materials released in accidents) will remain unchanged or decrease with advancing technology., The assumed properties are for central plants processing spent fuels from light water (IWR) or fast breeder (FBR) reactors employing unit oper- ations of chop-leach, solvent extraction, and ion exchange. A schematic drawing of the type of plant that is assumed is shown in Fig., 8.9. It is assumed that spent fuels are stored prior to processing in water-filled canals, High-level wastes are assumed to be either pot-calcined immediately and stored in water-filled canals for two years prior to shipment or stored for two years in an acid solution and then calcined prior to shipment. Low-level wastes are assumed to be discharged predominantly to the atmos- phere. Intermediate-level wastes (spent solvent, resins, etc.) are assumed to be fixed in asphalt, polyethylene, or concrete; and hulls are assumed to be stored in vaults in relatively small containers. Process Equipment. - It is assumed that the concentrations of fuel (U + Pu) in aqueous solutions in the head ends of the IWR and FBR plants are 0,3 and 0.1 metric ton/m3 respectively. The volume of fuel solution in a single vessel was kept rzlatively small, 3 to 30 m3, by assuming that the relative processing rate will have increased by a factor of 3 (because of more continuous equipment) by the time that 18-metric ton/day IWR plants or 9-metric ton/day FBR plants are built, and that the 36-metric ton/day plants for IWR and FBR fuels consist, respectively, of two 18-metric ton/day and four 9-metric ton/day independent modules. Multiple tarnks of these assumed sizes, in separate compartments to prevent interaction in the event of an accident, would be employed if additional capacity is needed for head-end equipment (dissolver, and accountability and solvent extrac- 8-L2 Table 8.7. Assumed Properties of Reprocessing Plants and Waste Storage Facilities Fuel Processing Rate (metric tons/day)? LWR Fuel FBR Fuel 1 6 36 1 6 36 Processing plant Total dissolver solution, m-/day 3,33 20 120 10 60 360 No, of independent lines 1 1 2 1 1 L Relative processing rate/line 1 1 3 1 3 3 Max. head-end vessel capacity, m> 3.33 20 20 10 20 30 Total cell capacity/line, m 2333 11,000 14,000 7000 1,000 21,000 No, of cells/line 7 1 1k 7 1 1, Cell size, m 333 1000 1000 1000 1000 1500 Cell ventilation rate, m>/min 66.7 200 200 200 200 300 Total ventilation rate/line, n/min 700 L4200 1200 2100 L4200 6300 Ventilation train® F,M M F,M F,AM FAM FAM Total off-gas filow rate 28 85 255 2.0 4.0 2l Off-gas train® s,I,F S,I,F S,T,F s,I,F S,I,F S,IF Interim® liquid waste (acid) storage facility Tank volume (80% filled), m 812 3785 3785 990 3785 3785 No., tanks required for 2-year 2 3 10 2 3 13 accumulation Off-gas flow rate/tank, m-/min 6.1 28 28 7.4 28 28 Off-gas train® C,F C,F c,F c,F C,F c,F Vault ventilation rate, m>/min 6.1 56 22l 7.4 56 280 Ventilation train® ¢,F,M ‘C,F,M C,F,M c,F,M C,F,M C,F,M Interim® waste solids storage canal Length for 1k.6-m width, m 5.8 35 210 7.1 L2 250 Ventilation rate, m>/min 170 1000 6100 210 1200 7300 Ventilation train® C,F C,F ¢,F C,F C,F C,F A 1,0-metric ton/day plant processes 260 metric tons of uranium + plutonium per year, Ps = caustic scrubber; 90% removal of iodine. = silver tower; 99% removal of iodine. = activated charcoal filter; 99% removal of iodine, metal mesh or silica gel; 99.9% removal of Te, Cs, and Ru, = high-efficiency iodine removal units: iodine DF of 10. = steam condenser; discharges air at 100°F and 100% relative humidity. = either reliably-protected HEPA or deep bed filter. Normal effluent = Q,00L2 mg/m3. Accident effluent = 0,02 mg/m>. Mo years, HoOoQ k= P 3 i ORNL DWG &8-10577RI 14 CELLS HIGH LEVEL LIQUID WASTE 30x30x40 FT. 2 YR INTERIM STORAGE — = — 3-1,000,000 GAL TANKS ‘e o el ; ., - LT SAND FILTER - s CRasa WEwl oS [AerER BETiid 8.9. Spent-Fuel Processing Plant with a Capacity of 6 Metric Tons/Day. - 330 — 320 — 310 — 100 — 90 — B0 —70 —60 — 50 — 40 — 30 =20 — 40 £1-8 8-Ll tion feed tanks) or plutonium storage (tanks of the assumed maximum size packed with borosilicate-glass raschig rings containing solution at a 3. plutonium concentration of 0,25 metric ton/m Process Cells. - Process cells are assumed to have reinforced (1 to 2% steel) concrete outer walls that are approximately 5 ft thick, rein- forced concrete partition walls between cells that are approximately 2 ft thick, and volumes 50 to 100 times greater than the maximum vessel. Such cells, roughly 25 to 35 ft cubes that have secured roof plugs, could withstand a sustained pressure of 30 to 50 psig or the detonation of 25 to 50 1b of TNT at their geometric centers without rupture. It is assumed that 7 cells are used in small plants; however, more compartments (i.e., 1l cells per process line) are used in plants having capacities greater than 6 metric tons/day. Vessel Off-Gas System. - Present technology, with a trend toward relatively lower off-gas flow rates per unit of plant capacity, is assumed for LWR plants, It is also assumed that FBR plants will be designed to minimize the vessel off-gas flow rate to approximately 70 cfm in a l-metric ton/day plant and that this flow will vary directly with plant capacity but inversely with relative processing rate. The off-gas is assumed to pass through a train (wet scrubber, solid halogen absorber, and filter) to effect partial removal of iodine, semivolatile fission products, and particulates and to discharge to the ventilation system on the upstream side of the ventilation filter. It is assumed that the wet scrubber serves to retain about 93% of the iodine in a relatively nondispersible form and that the off-gas train for FBR plants 7 will include devices which will provide a cumulative DF of 10' for iodine. Ventilation System. - The ventilation air exhaust is assumed to consist of the air flow from the cells (at 0.2 air change per min) plus an additional 50% that flows directly from the building (third containment barrier). This stream is filtered, passed through metal mesh or silica gel for 99.5% removal of ruthenium vapors, and finally exhausted to the atmosphere through a 100-m-tall stack., In addition, the FBR plant is assumed to be equipped with activated charcoal filters for 99% removal of lodine. The filter system is assumed to be composed of either a sand 8-L45 filter or roughing and HEPA filters with equivalent reliability and integrity. Independent process lines are assumed to have independent ventilation systems. Facility for Interim Storage of Liquid Wastes. - The interim liquid waste storage facility is assumed to provide for two-year storage of acid waste (at a concentration of 0,0l gal per Mwd of burnup) consistent with a maximum tank size (80% filled) of 1,000,000 gal and at least 30% spare tankage. The off-gas stream — 1000 c¢fm for a 1,000,000-gal tank — is assumed to pass, first, through a condenser (which would condense and recycle the distillate to the tank in the event of loss of coolant), then through a filter, and finally be discharged to the ventilation system for the vault. The latter ventilation system collects the small purge flow from each tank vault (plus the caubined off-gas from all tanks) and discharges it through a backup condenser, filter, and ruthenium removal device to a 100-m stack. The tarks and/or the vault are assumed to be designed to withstand a hydrogen-air explosion (an internal pressure of ~100 psi) without rupture, possibly by venting to other tanks or vaults. The tanks, vaults, and ventilation system are assumed to be designed to withstand the effects of the maximum earthquake, Canal for Interim Storage of Waste Solids. - The canal for interim storage of waste solids (i.e., calcined waste) is assumed to provide for a two-year accumulation of 6-in,-diam by 10-ft-long pots, each containing fission products from 1,100 Mwd of burnup at an average solids concen- tration of 1.0 x lO_h ftB/de. The pots are assumed to be covered with at least 20 ft of water, The ventilation system for the canal and build- ing provides 12 air changes per hour to minimize fog formation. The ventilation system is assumed to be exhausted through a dehumidifier and HEPA filters at the roof of the building. 8.3.2 Analytical Models and Mechanisms of Accidental Release Mechanisms that tend to negate the primary confinement barrier (proc- ess vessels, associated piping, and the efficient, low-flow off-gas system) have the potential of releasing radiocactive materials to the atmosphere through the ventilation system, The following sections will describe 8-L6 models for predicting the fractional release, discuss dispersive mecha- nisms, and present estimates of the fractional release to the atmosphere from upper limit accidents, The designs of models for the release of radiocactive materials depend on whether the material is released to the ventilation system as a gas (or vapor) or as an aerosol. Gas or Vapor. - Certain of the fission products (the noble gases, halogens, and semivolatiles) may escape from the primary containment barrier in gaseous form. The release to the environment from such sources is relatively easy to predict; it is the fractional release from the vessel mitigated by the removal efficiency of the devices in the 85y and l33Xe, may be released essentially quantitatively from process vessels. ventilation train, The noble-gas fission products, dominated by Devices for partial removal of noble-gas fission products are not used in present commercial reprocessing plants, but several types of devices have been proposed for this application.sl’53 The halogens, dominated by 1311 5 HL, or organic iodides. Since these compounds have high vapor pressures at and 1291, may be volatilized from process operations as I room temperature, they are not appreciably removed by filtration, Usually, activated charcoal filters may be relied upon to remove 99% of the iodine from a ventilation stream, especially if most of the iodine is in the form of 12 or HI (the typical forms released from most process operations). Certain other fission products, notably (in approximate order of importance) Ru, Cs, Te, Tc, and Se, may be classed as semivolatiles since gases or vapors of these elements may result from certain abnormal process operations. The oxides of Se and Tc are completely volatilized at temp- eratures in the vicinity of 200°C, while the normal oxides of Ru, Cs, and Te require temperatures generally greater than 750°C.5h Under highly oxidizing conditions in acid solutions, ruthenium may form the tetroxide, which has a boiling temperature of approximately 80°C. A slight excess of KMnOh in an acid uranyl nitrate solution at 80°C will result in the volatilization of 70 to 804 of the contained ruthenium in 5 to 10 min.55 For this reason, highly oxidizing conditions are avoided in present fuel 8-L7 reprocessing plants. Evaporation and complete boildown of a nitric acid solution of fission products will result in the volatilization of 10 to 20% of the ruthenium.sé’57 Once airborne, the vapor tends to rapidly deposit on metal surfaces and decompose to the relatively nonvolatile dioxide. For this reason, a "bucket of Brillo" (i.e., a tank packed with stainless steel mesh) has been found to be effective for removing ruthenium from off-gas and ventilation streams at the Savannah River Plant., ©Silica gel absorbers, operating at about 70°C, were found to remove 99.6% of the ruthenium from waste calciner off-gas at Idaho.58 Radioactive Aerosols, - The aerosol that would be dispersed in cell air by an accident would consist of a dispersion of a radiocactive solu- tion, solid particles, or smoke. The physical properties of aerosols are such that they effectively restrict the escape of radioactive particles to the environment. This is seen commonly in practice since, through the use of appropriate deentrainment mechanisms, the condensate from the L 10 10_6 of the activity of the solution. Gravitational settling serves evaporation of a radiocactive solution may be made to contain only 10~ to 1limit the maximum aerosol concentration; we have been able to demon- strate this through an approximate correlation of the solution concentra- tion in air or vapor arising from cooling towers, evaporators, and air- 59 sparged vessels. This correlation is shown in Fig. 8,10. In order to properly describe the release of aerosols from a cell, we must be able to ascribe removal efficiencies to filters and to cracks in cell walls. For superficial velocities less than approximately 0.15 ft/sec, it has been found that an aerosol formed by vigorous mixing of a solution with air is metastable and has a concentration of the order of 10 mg/m3. This metastable concentration is approximately equivalent to fog, which has a concentration of approximately 10 mg/m3 and a particle size of approximately 10 y. For orientational purposes, a 1l-in.,/hr rain with a mass mean particle size of 3000 u has a concentration of 1000 mg/mB. At ORNL,59 the particle size distribution of the metastable aerosol in a ventilation stream downstream from the source has consistently been found to have the particle size distribution shown in Fig. 8.11. Another piece of relevant information reported by Garneréo is that the weight distribu- 8-L8 ORNL-LR-DWG 47936 R2 10* / /. ® - / ~. 3 ® 5?10 - Z O — < ° - z L) QO Z o OV w 102 |- g > v ® = e POINTS CALCULATED USING COOLING TOWER DATA & A THOREX SPARGING DATA 5 ) = " FOG = » g s 10 l | 1 l 1 0 2 4 6 8 10 12 MINIMUM VERTICAL VELOCITY IN LINE (ft/sec) Fig., 8.10. Effect of Minimum Superficial Velocity in an Off-Gas Line on the Concentration of Liquid Solution Particles Resulting from Vigorous Mixing of a Solution with Air, Density of solution, 1 g/cc. ot Wi 8-L9 ORNL-LR-DWG 48665 R2 0 B A i o _ /S 3 r L N ° N w [ Q_ — D:.._ §_ - ® Ol v v v L | 20 o0 90 PERCENT LESS THAN STATED SIZE Fig. 8.11. The Particle Size Distribution of a Stable Aerosol Which Has Encountered Several Changes of Direction in a Pipeline, 8-50 tion of particles smaller than 10 to 20 y will be fairly constant, even if there is gross entrainment of larger droplets. The knowledge that this distribution is fairly constant and constitutes approximately 10 mg/m3 may be used to estimate the approximate concentration of particles smaller than a given size, even in an air stream which is very concen- trated with liquid droplets. Practically, it is possible to assign efficiencies to an absolute filter and calculate the effluent concen- tration. In evaluating the concentration of aerosols in air which leaks from a cell, it is considered that the aerosol must follow many small tortuous paths in its escape through 5 ft of concrete. The evaporator deentrain- nent studies by Walsh and Schlea®l indicate that a single right-angle impingement will reduce a liquid aerosol concentration to 10 mg/m3 or less. Fine heavy-element dust would be reduced to the order of 1 mg/mB, and the concentration of smoke in leaked air would probably be no more than approximately 100 mg/m3. These numbers are primarily of use in estimating the radiation dose to plant operating personnel, Essentially all of the material that escapes from the cells through cracks during a period of temporary pressurization would be routed through the filtered cell ventilation system., Junge62 and Friedlander63 have observed that the particle size distribution of airborne aerosols is remarkably constant or '"self- preserving." Small particles tend to agglomerate rapldly by Brownian motion, while large particles are removed by impingement or sedimentation, Friedlander suggests that a quasi-stationary state exists such that the rate at which matter enters a differential size is equal to the rate at which matter is lost by sedimentation. It has also been determined that the stable concentration of small particles (less than about 3 w in diameter), because of agglomeration, is consistently less than a few grams per cubic meter after the aerosol has been permitted to '"age' for 6l-66 a few seconds or minutes. Friedlander has proposed the following formulation for the differential concentration of particles in a metastable aerosol as a function of size: 8-51 dn = k’r dr (6) where n = number of particles per unit volume, k’ = a constant, r = radius of the particle, a = a constant with a value of ~-1 to -1.5, By converting to a mass concentration and integrating from r = 0 to “ r = r, the concentration of particles with less than a given diameter is: - C(*? | (7) where . C( 0.3 (11) These expressions for aerosol concentration are compared with experi- mental data for a wide variety of heavily concentrated and turbulent aerosols (smoke, flyash, DOP, etc. in air, and water droplets in air and steam) in Fig. 8.12. Expressions (10) and (11), for concentrations of monodispersed aerosols with a half-life of 10 min, provide a practical upper bound for the concentrations of solid particles in air. A better description of aerosols containing liquid particles, is provided by expression (7) when a is approximately equal to -2, Based on the maximum concentration of particulates as a function of particle size (see Fig. 8.12) and assuming that the efficiency of deep- bed sand or HEPA filters is 100% for particles 0.3 , in diameter, the predicted concentration of particles in the effluent from absolute filters is 0,02 mg/m3. Cheever determined experimentally that the maximum concen- tration of plutonium particles in the effluent from a 30-in.-deep sand filter, occurring at the optimum superficial velocity for a penetration of 4.8 ft/min, was 0,02 mg/m3 (ref, 68). This experiment was performed under conditions that are very unlikely to occur in accident situations; the filter influent concentration was 100 mg/m3, and the count-mean particle size was only 0,07 y because the aerosol had aged for only a few seconds, Cheever also found that an HEPA filter removed an additional 99% of the particles in the effluent from the sand filter. From these data and known characteristics of filter systems, it is assumed that filter effluent concentrations of 0,02 mg/m3 or less are attainable in practice, regardless of the mass concentration of the influent. The release of radioactive material through the cell ventilation system by a mechanism that generates aerosols is estimated to be as follows: 8-53 ORNL DWG 68-5834R2 CONCENTRATION OF PARTICLES WITH DIAMETER LESS THAN D, C(mg/m3) |05h T I lll'llll ] T lllllll T 1 Illllll I 1 llllll‘ T T II||II: [ ® MISCELLANEOUS AEROSOLS OF 3 " SOLIDS IN AIR ° ] | © ENTRAINED HpO DROPLETS IN _ STEAM. SUPERFICIAL VELOCITY o - 2-4 ft/sec . ° A ORNL VENTILATION STREAM €=2900 D3 0" I~ — I03 _; - 102 — [ ¢=9700 D4 D<0.3 ) 10 &~ = I B — OI 1 1 1 4 IIII {1 1 lllllll 1 3 ) lLl-lll 1 ] lllllll i 1 i L Lttt |02 |10 | 10 102 103 PARTICLE DIAMETER, D(u) Fig. 8.12. Mass Concentration of Solid and Liquid Particles in Aerosols. 8-5h DP R =0C= [th + VC "-l:||<:"-=_1 D ] , (12) f C where R = the quantity of a component (in one metric ton of fuel) that is released to the atmosphere, C = mass concentration of particles in the filter effluent = 2 x 1070 kg/m”, Dp = concentration of the component in the particles, weight fraction, Df = concentration of the component in the fuel, weight fraction, F, = total flow rate in ventilation system, m3/min, FC = flow rate from the cell in which the aerosol has been generated, m3/min, t = duration of the source term for aerosol generation, min, V = volume of the air in the cell (evaluated at one atmosphere of 3 pressure) following the dispersion of aerosol, m~”, When the aerosol is formed essentially instantaneously, as in an explosion, the rate of release to the atmosphere will decrease exponenti- ally with a mean life of F_/V_ (which is assumed to be 5 min). Dispersive Mechanisms. - Mechanisms for the dispersion of gases and aerosols in cells include chemical explosions, fires, nuclear excursions, and leakage. Some properties of explosions relative to the containment potential of cells are shown in Table 8.8. All of the quantities of the limiting explosive materials are very large as compared with their credible inventories in a process vessel. The allowable quantities are even larger if the cells are vented to another confinement zone of large volume (i.e., the cell-canyon concept used at Hanford and planned for MFRP). It is assumed to be incredible that the cell would first fill with hydrogen or solvent vapor and then explode. The flow rate of cell ventilation air is sufficient to dilute any radiolytic H2-O2. The most serious fires in a fuel processing plant would be those involving plutonium, that is, solvent or ion exchange resin loaded with 8-55 Table 8.8. Estimated Properties of Explosions That Could Occur at the Center of a 10-m> Cell® with 5-ft-thick Reinforced (1 to 2% Steel) Cell Walls Without Rupture Total Energy Pressure Release at Cell Wall Source of Explosion (Btu) (psig) 30 1b TNT 5}, 000 <100 500 m’ of 4O vol % H, in air 1,500, 000 50 1120 m> of 5 vol % propane in air 1,500,000 50 150 1b of "Red Oil™" ~1,500,000 50 300 1b of sodium in water ~1,500,000 50 Nuclear burst of 3 x 10°°0 fissions® 9,100,000 ~0.7 a - - - Inside dimensions. bMaximum burst of lO16 fissions/liter in a tank containing 30,000 liters of solution at a temperature of 85°F, 8-56 plutonium. Purex-type solvent will burn at the rate of about 1 in., of depth per hour and generate approximately 20,000 Btu/hr per ££° of burning surface. Experience in gloved enclosures has shown that fires covering the entire area of the floor of the enclosure tend to self-extinguish in a matter of minutes because of depletion of the oxygen. This has been true even in well-ventilated enclosures because the pressure increases to several inches of water and reverses the flow through the intake. It has been observed that ion exchange resin loaded with plutonium nitrate can ignite spontaneously at about 120°C and burn (in the absence of air), liberating about 540 Btu/lb. Experience has shown that the initial burst resulting from a super prompt-critical nuclear excursion in a solution is limited to a maximum of 1019 fissions per m3 of solution.69 At this fission density, the void coefficient caused by the generation of radiolytic gas (~1.L m> of gas, at STP, per m° of solution) is sufficient to override the effect of high reactivity addition rates. Assuming that the temperature of the solution is 85°F (the yield would be lower if the temperature were higher), this burst would increase the temperature to bolling. If the solution is not rendered permanently subcritical by the initial or succeeding bursts or by ejection of solution, it may possibly boil to dryness. The dried solids, if not subcritical because of low density and lack of moderation, probably would be dispersed by one last burst, Assuming that all of the solution in an equilateral cylinder with a volume VT is involved in a nuclear excursion, the upper limit yield of the initial burst (and probably the most powerful burst, resulting in the generation of a radiolytic gas void fraction, at STP, of 1.4) is 1019 v 77 fissions. When boiling begins, the steam void coefficient (2.3 x 10 m>/fission) (ref. 70) would limit individual bursts to approximately 6.1 x 100 1.08 v/3, and the bubble rise rate is about 12 m/min, the period betseen 1/3 . mi fissions/mB. Since the height of the solution in the tank is bursts is approximately 0,09 V n. The total time required for boil- down of the solution (assuming 2.1 x 106 Btu/ton) is about 52 Vl/3 min, The sudden generation of radiolytic gas or steam in the solution would cause an inertial force to be exerted against the walls of the [T 8-57 tank, An overestimate of the maximum amount of work that can be done in deforming the vessel, taking no credit for free expansion into the vapor space of the tank, may be calculated71 assuming that the liquid and gas expand reversibly against the plastic flow pressure of the vessel. Assum- ing a gas void fraction (at STP) of 1.L per burst, no more than about 0.5% of the energy released in the burst could do pressure-volume work agalnst a resisting pressure of 200 psig. An unrestrained cylindrical tank of characteristics assumed in this study could, theoretically, withstand repeated bursts without rupture. [The rupture strain of 30LL stainless steel is 0.65 (ref. T1)]. 8.3.3 Method of Analysis of the Downwind Consequences of a Unit Release of Radioactive Material The method that has been selected for investigation of the environ- mental consequences of an accidental release of radicactive material from a fuel reprocessing plant consists, first, of the examination of a "unit" release of activity and, second, the application of the resulting data to actual releases which could be expected from the various credible accidents. Two different mixtures of isotopes have been considered. These mixtures (listed in Table 8.3) simulate the fission product and actinide contents of typical spent LWR fuel and IMFBR core and blanket fuel mixtures which will actually be encountered. It is assumed that the LWR fuel has decayed for 150 days prior to processing and that the LMFBR material has decayed for 30 days. For these mixtures, a "unit! release is defined to be the release of all materials that are associated with 1 kg of fuel; therefore, the funda- mental calculations have been performed on this basis., Initially, no provision is made for differences in the chemical or physical behavior of the various isotopes, and it is assumed that all of the components in a unit release escape. However, as will be shown, it is possible to treat differences in behavior and thus account for variations in the release fraction due to filtration, chemical reactions, and other processes that affect some of the components but not others. 8-58 We have investigated both the external gamma dose and the external beta dose that result from direct exposure to the radiation flux origi- nating in the plume and from the internal radiation dose received as a result of the inhalation of radiocactive material by a receptor submerged in the plume. The calculation of both types of doses depends on a know- ledge of the concentration of radiocactive material in the plume as a function of time and space. The concentrations have been computed by T2 using the "Gaussian Plume" formula'® and by utilizing the source term and ground reflection correction described by Binford, Barish, and Kam.73 The source term is derived using the assumption that a unit quantity of radio- active material is released into the processing building, where it is instantly and uniformly mixed with the air in the building, It is further assumed that a constant fraction of the building volume is being discharged from the stack per unit time., These assumptions lead to the following expression for the concentration at the space point (x,y,2z), relative to an origin of Cartesian coordinates at the stack orifice, and at time < after the release has occurred: A, alxfe 1) 2, 2 aqe e oy (13) X(Xayaza’t) = [ -22/23§ x e 2nuayaz + e s + = x/u _(2h + Z)z/Qgg] where X(x,y,2,t) = concentration, (curies/m>), q = initial release, (curies), u = wind speed in the x-direction, (m/min), X,¥,% = space coordinates (m), oy(x), gé(x) = horizontal and vertical dispersion param.eters,72 respectively, (m), a = exhaust rate, (min"l), A = decay constant, (min-l), h = effective stack height, (m), + = time since release, (min). 8-59 Decay will be neglected for the mixtures under consideration so that \ is set equal to zero. Moreover, the value of the concentration at the plume center line (y = 0) at ground level (z = -h) is of great interest, Under these conditions, aqea(x/u - ) —h2/20§ oy 3, =0, T < x/u . This expression is proportional to the inhalation dose rate at ground level at the plume center line and very nearly proportional to the external beta dose. The time integral from t = 0 to vt » » is then proportional to the total dose. This integral, 0 -h2/29'§ e X(x,0,-h,r) dy = e — (15) nuy_J, is independent of a, the exhaust rate. The dispersion parameters oy(x) and Oi(x) are monotonic increasing functions of the downwind distance, x; however, they also vary with atmos- pheric stability. For a given value of x, the dispersion parameters decrease with increasing stability. It is, therefore, necessary to specify the degree of atmospheric stability in order to select the appropriate set of values for the parameters. For the purpose of investigating the external doses, two sets of atmospheric conditions have been utilized: (1) "Most Representative Conditions," where the wind speed has been chosen to be 100 m/min (3.73 mph) and slightly unstable (C) con- ditions are assumed to prevail, (2) "Inversion Conditions," where the wind speed is 50 m/min (1.86 mph), moderately stable (F) conditions prevail, and an inversion "l1id" exists just above the stack orifice. To account for the latter, the vertical dispersion parameter is modified by being held constant once it reaches the value gi(x) = n/2.15 (see ref. T2). 8-60 Many other combinations of wind speed and stability conditions are possible; however, it is believed that these two are reasonably typical, cover most of the likely situations that may arise, and permit valid inter- polation to other cases which lie in between, The inhalation doses have been computed on a somewhat more comprehen- sive basis. As suggested above, many different combinations of wind speed and stabllity conditions are possible. Thus the inhalation doses have been computed for each of six different stability conditions, the results have been plotted on a single graph, and the envelope of the curves thus obtained have been utilized to estimate the inhalation dose to be expected at each ground level point downwind on the plume center line. In all cases, it is assumed that the effective stack height, h, is 100 m. (The effect of stack height on ground-level concentration will be discussed in detail in a later section.) Finally, it should be pointed out that all of the doses computed below assume exposure of the receptor during the entire course of the accident, External Beta and Gamma Doses. - These doses stem from direct exposure of the receptor to the radiation flux in the plume. Because of their short range, only the beta particles that originate in the vicinity of the receptor contribute to the dose. Hence, the dose rate may be assumed to be propor- tlonal to the concentration of beta emitters at the location of the recep- tor.73 The gamma dose, on the other hand, requires a space integration over the entire volume of the cloud in order to sum the photon flux incident on the receptor. 13 which was originally developed in order The computer program PLUME, to calculate internal iodine and external iodine and noble-gas doses following a reactor accident was used to perform these calculations. Input for the beta dose calculation is the average energy per disintegration, the equivalent number of curies, and a numerical constant to convert Mev/m3 into dose units. Input for the gamma calculation consists of the gamma- emitting inventory, divided into nine energy groups, and the appropriate cross sections and buildup factor parameters. The results are displayed in Figs. 8.13 and 8,1. Note that, in these cases, there is no physical T 8-61 ORNL DWG. 68-5836R! 1072 . . : "INVERSION" — Y- < = DOSE ~ / Sao 7 >~ / ~o 3 ~ 10°F ~ J \\ / "MOST N ! REPRESENTATIVE" Y-DOSE / - / 107t i / ] / / / / ' ' fae—— "MOST " ,' sl | REPRESENTATIVE" | 5 | B-DOSE i = ! le—"INVERSION" 2 ! / B- DOSE o ' ! W oét i O Qo o7t 10-' :. -9 o L : ! 0~ | 10 102 DISTANCE DOWNWIND (km) Fig. 8.13. External Radiation Dose due to All Fission Products Released in Reprocessing 1 kg of LWR Fuel. DOSE (rem/kg) 8-62 ORNL DWG 68-5835-RI lo-| 1 I IIIIII[ | 1 IIIIIII I 1 T 1T F 1T 11 "INVERSION" y -DOSE — 7 s e 7 > = / - — \\ / o / Y . / “MOST » ~ /| REPRESENTATIVE S -DOSE |0-3 — ’I 7 / / / / ! / I / l, / -4 107~ | "MOST / | REPRESENTATIVE" ! . j«——— B-DOSE "INVERSION" | [T——~#-DosE | I 10-%}— | ! 10"6 - |0-7 — IO"B 1 1 ||||||| I 1 Illlll[ ] 1 1 L1111 107! | 10 DISTANCE DOWNWIND (km) Fig. 8.1L4. External Radiation Dose due to All Fission Products Released in Reprocessing 1 kg of LMFBR Mixed Core and Blankets. 8-63 separation of the components and that all of the isotopes present are assumed to behave similarly. Inhalation Dose Calculations., - These radiation doses result from inhalation of the contaminated air in the plume and from subsequent depo- sition of radioactive material in the various organs of the body. The rate of intake of radiocactive material is proportional to the breathing rate and to the concentration of the radiocactive material at the location of the receptor. The total intake is simply the time integral of the product of these two quantities. For the purpose of these calculations, it will be assumed that the receptor is located at ground level on the center line of the plume and that the exposure lasts for the duration of the accident so that the intake rate is integrated over infinite time. If the breathing rate is assumed to be constant, the total intake is: 2 2 -h /202 I-= !fl—eu—— curies (16) nuay 3, where B is the breathing rate in m3/min and the other symbols are as defined previously. For this study, B has been taken to be 2.08 x lO_2 m>/min. The quantity I/q is the total amount of activity inhaled per curie of originally released material. It should be noted that the spatial variation of the inhalation dose is independent of the amount of released material, the wind speed, and the breathing rate. Thus, for a given wind speed, breathing rate, and quantity of material released, the expression -h2/272 g = — (17) Ty Ty, is the same function of x, regardless of the amount of material that is released. This function, normalized to unity at its maximum value, has been plotted in Fig. 8.15 for each of the six stability conditions, An envelope enclosing the six curves has been drawn; this envelope permits estimations of the dose at each point downwind by using a knowledge of the dose at any given point, ORNL DWG 68-5842R2 RELATIVE DOSE o -2 1 Lt al ] l Lttt . L1 L L1111 10 0~ Fig. 8.165. 10° 10! 02 DISTANCE DOWNWIND { km) Generalized Curve for Estimating Inhalation Doses as a Function of Distance Downwind (h = 100 m). The fraction of released material that is inhaled by a person 400 m downwind is estimated to be 4.3 x 10 7. 9 19-8 8-65 The dose at 400 m under A (extremely unstable) conditions is the maximum and has been chosen for reference. At a wind speed of 100 m/min, the total intake following the release described above is 4.3 x 1073 ue per curie released. This factor has been utilized as input data for a computer program INREM,W'L which, given the quantity of radioactive material inhaled, computes the dose to the most important organs as a function of time after inhalation. The program takes into consideration uptake by the various organs, effective half-life, and the age of the receptor, which, for this study, was chosen to be 20 years. The INREM Computer Code. - The rate of intake of radiocactivity is the primary radioactivity input for calculating the cumulative dose equivalents by the INREM Code. These estimates of dose are compiled for the various body organs from inhalation or ingestion of radiocactivity programmed as continuous or intermittent intakes as a function of age. The parameters in the dose equations change as a function of time as the person ages during the time of intake or during the period of interest (which may be longer than the period of intake). This code, as currently dimensioned, has the capacity to handle 110 radionuclides and 1l body organs. The model, programmed for all organs except the gastrointestinal (GI) tract, is written as follows: Dinltystysty) = 51 f Lilt = t),8)85,(0 - %) U NCEESE t t 1 s -t b x exp[—f RCY; dy]ds} dt (18) b=t where Din(tl’tZ’tb) = cumulative dose equivalent (rems) received during the time interval tl to t2 from the ith radionuclide in the nth organ, resulting from intake during this time interval by an individual born at tb’ 8-66 t, = time (days) of initial intake relative to time of release (t = 0 at time of release), t, = time (days) at end of period of interest relative to time of release, t, = time (days) of birth relative to time of release, c.*- il time (days) after release, s = time after intake relative to time of release, Ii(t) = intake (uc/day) of ith radionuclide at t, mn(t) mass (g) of the nth organ at t, fin(t) = fractional absorption (dimensionless) of the ith radio- nuclide in the nth organ at t, ain(t) = effective absorbed energy (Mev) of the ith radionuclide in the nth organ at t, and Xin(t) = effective elimination constant (dgy_l) of the ith radio- nuclide in the nth organ at t. The variables tl’ t2, tb, whereas the variables I(t), mn(t), fin(t), gin(t), and xin(t) are functions t, and s are measured relative to release, of the age of the individual. The code uses Eq. (18) for ingestion of con- taminated food and water, or inhalation of contaminated air, and calculates the cumulative doses to all organs except the GIL tract. When the age-dependent cumulative dose equivalents to the GI tract are to be calculated, the (MPC)a or (MPC)w is used in the following way: t D, (tq,% ) = o 2 syz (BT by “""ny"(mcjiyz A LI - t)st] m/my (6 - 5)) 65, (8 - )/eyg 1 x £5,(b - 8)) /8 dt (19) where 8-67 Diyz(tl’tZ’tb) = cunulative dose, equivalent (rems) to a critical segment of the GI tract, received during the time interval t; to t, from inhalation (y = 1) or inges- tion (y = 2) of the soluble (z = 1) or insoluble (z = 2) form of the ith radionuclide for an intake during this time interval by a person born at t, I’ = intake (cc/day) of air (y = 1) or water (y = 2), 3 ff’ maximum permissible concentration (uc/cc) of the ith radionuclide in air (y = 1) or water (y = 2), where the ith radionuclide is soluble (z = 1) or insoluble (z = 2), tl = time (days) of initial intake relative to time of release, t, = time (days) at end of period of interest relative to time of release, tb = time of birth relative to time of release, H e ~~ ct S i intake (uc/day) of the ith radionuclide at t, s = standard man index, £ = age index, m, = mass (g) of the critical segment of the GI tract for the gth age group, E5p = effective absorbed energy (Mev) of the ith radio- nuclide in the critical segment of the GI tract in the gth age group, fig = fractional intake of the ith radionuclide reaching the critical segment of the GI tract in the gth age group. Calculations were made with the INREM code to determine the dose commitment for the first year following inhalation (which, in this model, is the highest annual dose commitment) and also the dose commitment for a period of 50 years following the intake. The complete output data from 8-68 INREM have been reproduced in Tables 8.9 through 8.12. The results at 1,00 m were then utilized to obtain the doses at each point downwind by means of the generalized curve of Fig. 3.15. In order to allow for differences in chemical and physical behavior of the various isotopes, the isotopes may be divided roughly into cate- gories, depending upon their volatilities, as follows: 1. Volatile fission products: noble gases, halogens, tritium 2. Semivolatile fission products: Ru, Te, Cs, Tc, Se 3. Nonvolatile fission products: all other fission products li. Nonvolatile actinides: plutonium and transplutonic elements The dose commitment to the various organs, as well as to the whole body, by these categories is given in Table 8.13. Deposition. - In all of the foregoing calculations, it has been tacitly assumed that there is no depletion of the plume by deposition, fallout, or rainout. Consequently, the results thus far obtained are conservative in that some depletion of the plume due to these mechanisms will occur. On the other hand, the deposition of relatively large quanti- ties of an extremely toxic substance, such as 9OSr or plutonium, on the ground in and around a highly populated area may give rise to a serious 69 hazard, Criticality Accidents. - One possible cause of a serious accident in a fuel reprocessing plant is inadvertent criticality that results in a nuclear excursion, Aside from damaging mechanical effects, such an accident would augment the inventory of fission products to an extent depending on the number of fissions taking place during the excursion. In order to assess the additional radiation doses that would result from such an incident, a "unit" nuclear excursion of 3.7 x 1018 fissions has been investigated. The iodine isotopes and the noble gases and their daughters were considered to be of primary importance., The internal dose due to iodine and the external dose due to both iodine and noble gases plus their daughters, have been calculated for both the "most representa- tive" and "inversion' conditions, using the PLUME computer program.73 ND. O~ D WA~ Table 8.9. NUCLTDE LABEL H-13 303 SR=-8% a9 SR-90A 901 SR=-908 902 SR=90QC 903 ¥=90 43 ¥Y-91 48 IR=-9% 65 NB-~G95 67 TC-99 79 RU-1Q03 a8 RU=-106 97 RH=103Mm a9 AG-111 114 CO-115M 125 SB~124 159 SN-125 1561 SB~125% 162 TE-125M 163 TE-12T7M 169 TE-127 17¢ TE-129M 176 TE-129 177 1-131 187 CS~134 327 CS-13é& 207 €S-137 210 BA~ 140 221 LA-140 222 CE-141 227 CE-144 233 PR~143 237 ND=-147 246 PM- 4T 247 sM=-181 255 FU~-152 328 EU-1585 262 TB-' 60 272 NP-239 330 PU-238 280 pPU-239 281 PU-240 282 pUY=241 283 AM-241 1 CM=242 2 CM~244 3 TOTAL TOTAL 800Y 3. TOQOE-O7 4.70C3E-02 4.5C6E-04 6.587€-03 3.6TTE-02 2.035E-04 5.984E-03 1.857E-02 3. 854E-C3 2.974E~09 2.924E-04 4.9728-03 2-9085—07 5. 269€-0n7 1.053E-0¢ 3.821E~-Q6 1.759E"07 1, 267E~04 9, 96TE-06 &. 3%3E-CS 9. 420E-07 1. 248BE-04 3.63ME-0Q7 2.425E-"8 4.985E~02 4,970F-07 1.423E-02 1.320€-05% 1. T60E-06 3.978E~04 1.5863E-01 3.340E~06 2.523E-07 1.413E-03 1.125€-05 1.718€E-CH 1. 971E-04 1.834E-"5 1.310€E-08 4.482E-02 4.913F-03 T.116E-03 7.239E-02 3.186E-03 1.737E-01 4.104E-02 6.552€-01 BONE NO DATA 1.,679E-01 6. T58E-C3 9,880E-02 5.516E-01 7-647E-03 2+235E-01 TalblE-02 2.885E-0Q2 T+435E~-09 5.908E-04 3.563€-02 S.T740E-07 2.3TE-0Q¢ NO DATA 1.007€-05 2.977E~0Q¢& S5.558E~0% 5.0853€E-053 3.23CE~-04 4.325E-08 5.197E-04 9, 281E-07 NO BATA 2.971E-Q2 24282E-07 2. 299E~-02 1.636E-04 1.001E-05 4,757E-03 2.547E 00 6. T729E-05 3.2C7E-06 2.843E-02 1. 390E-04 4. TQ2E-06 1. 374E-Q3 1. 4€3E~04 1.947E~-07 1.7€2E 00 2.003F-01 2.901E-01 3.8525E 00 3.962E~-02 2+566E 00 6.179E-01 1.283E 01 MUSCLE NO DATA NG DATA NC DATA NO DATA NG DATA NO DATA NC DATA NO DATA NG DATA NO DATA NC DATA NO DATA NQ DATA NO DATA NO DATA ND DATA NO DATA NO DATA NO DATA NC DATA NO DATA NO DATA NO DATA NO DATA 1,702€E-02 5.018E-07 2+261E~02 1.084E~07 NO DATA NO DATA NO DATA NO DATA NO DATA NC DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NQ DATA NO DATA NO DATA NO DATA NO DATA S.962E-02 THYRCID NOC DATA NO DATA NO CATA NO DATA NC DATA NO CATA NOC DATA NG DATA NO DATA NGO DATA NO DATA NO DATA NC DATA NG DATA NO DATA 24332E-08 6.823E-08 5.680E-07 1.663E-05 9.120E-05 3.,210€-06 1.917E-04 T« T80E-07 1.361E~05 NG OATA NO DATA NO CATA NO OATA NC CATA NO DATA NO DATA NO DATA NC DATA NO CaATA NO DATA NG DATA NC DATA NC CATA NO DATA NO DATA NO CATA NO DATA NC DATA NO CATA NG DAYTA ND CATYA 3.178E~-04 LIVER NO DATA NCQ DATA NO DATA NO DATA NO DATA NO DATA NOC DATA 3.034E-02 1.660E-02 1.102E-08 NO DATA NO DATA 4.TB9E-07 1.017E-06 3.295E-05 14675E-09 8.113E-08 61 TOE-06 3.,360E-05 1.669E-04 2.095E~06 3.265E-04 5.8461E-07 NO DATA 8,623E-02 T«614E~07 3.341€-02 2+366E-0T b.T14E-06 3,258E-03 1.324E 00 2.689E-05 3.893E~CH 4.654E-03 1,10 2E-04 4.505E-06 b.3465E-04 NO DATA 2.127E-08 2. 799E-01 3.068E-02 f.b444E-02 1.965E-01 4.482E-02 2.668E 00 6.525E-01 5.417€ 00 w3 Intake perlod of l-day duration; begins at age 20 K IDNEYS NO DATA NO DATA NO CATA NG DATA NO DATA NO DATA NO CATA 3.840€-02 1.803E-02 14388E-07 2,448E-03 6.872E~-02 1,903E-06 2-916E"06 2.6665'05 NO DATA NO DATA NO CATA 2.856E-04 l.419E=-03 1,781E-05 2.60BE~-03 4.641E-06 NO DATA 3s051E-02 4. TOLE-O7 l«281E-02 T+661E-08 NO DATA 1.591E-03 7.292E-01 1.559€-05 1.825E~06 6.596E-03 5.249E~05 5.073E-06 T+915E-04 6.042E‘05 6+639E-Q8 2.087E~01 2.,288E-02 3.3‘“E—°2 3.665€E-01 2.4234E-02 T.765E~-01 1.915e-01 2.535E 00 Inhalation dose commitment (in rems) integrated over 1 year SPLEEN NO DATA NO DATA NO DATA KO DATA NO DATA NO DATA NO DATA 24304E-02 1.456E-02 NO DATA NO DATA NO DATA 1.332€-0¢ NO DATA NO DATA NO DATA NG DATA NO DATA T.235E-05 3.594E-04 4.,512E-06 6.607E-04 1.176£-06 NO DATA 6.427E-02 5.481E-07 2.784E-02 9.614E-08 NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO OATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA ND DATA 1.308E-01 TESTES NO DATA NO DATA NC DATA NQ DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NGO DATA NO DATA NG DATA NDO DATA NQO OATA NQ DATA NO DATA &+.171E-05 3.780E-04 4+898E-06 6.345E-04 1.126E-06 NQ DATA NO DATA NO DATA ND DATA NO DATA NG DATA NO DATA ND DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NQO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA 1.08QE-03 Internal Dose at L4100 m Downwind Following the Release of 1 kg of LWR Fuel OVARIES NGO NO NO NO NO NO NO KO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO kO NO NO NO NO NO NO NO NO NG NG NO NO NO NQ NO 0.0 DATA DATA DATA DATA OATA DATA DATA CATA DATA DATA DATA DATA DATA DATA DATA DATA OATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA CATA DATA DATA DATA DATA DATA DATA DATA DATA DATA CATA DATA DATA DATA DATA DATA 69-8 =z 0 I DO BNPRL WA NUCLIDE H-3 SR-89 SR=-G0 A SR-908 SR-90C Y-90 Y-91 TR=-95 NB-95 TC-99 RU-1N3 RU-176 RH-103M AG-111 CD-11%M S8-124 SN=-125 sB-125 TE-1254 TE-127M TE-127 TE-129M TE-329 1-131 CS5-134 CS~-136 £5-137 BA~-140 LA-14D CE-141 CE-l44 PR=-1413 ND-147 PM-147 SM-151 EU-1%2 EuU-155 TB-160 NP-239 PU-238 Py-239 PY-240 PU-241 AM-24" CM-242 CM=244 TOTAL LABEL 3013 38 501 902 9013 43 48 6% 67 79 88 c7 89 114 125 159 161 162 163 169 179 176 177 187 327 207 210 221 222 221 238 237 246 247 258 323 262 272 310 280 281 282 283 1 2 3 Table 8.9 (Continued) LUNGS SOLUBLE INSOLUBLE NO DATA 2.751e-05 NO DATA T.311E-02 NO DATA 3.423E-01 ND DATA 3.433€E-01 NO DATA 3.423E-01 NO DATA 6.968E-03 NO DATA 1.428E~Q1 NO DATA 2.314E-~01 NGO DATA 1.423E~31 G.36BF~-10 5.485E~06 NO DATA 2.,BE6E-02 NO DATA 1.898E 00 NO DATA T« 267E-08 NO DATA 3.964E-05 NO DATA 3.33%E-05 T.519€-06 1.241E-04 NO DATA 6.590€-06 3. 946E-04 7,.601E-03 NO DATA 6.969E-04 NO DATA 4.2711€-03 NO DATA 2.21%9E-0% ND DATA 5.612E-03 NO DATA 6.225E-06 NO DATA 1.915€-07 G.179E-D3 4.522E-01 6.215€-C8 3.329E-06 3, ¢13€E-073 1.772E-01 1.322€~07 2.T15E-04 NO DATA 3,517E~05 NO DATA 1,005E~02 NO DATA 3.144E 00 NO DATA 1.065E~-02 NO DATA 6.161E-06 NO DATA 2.576E-Q2 NO DATA 1.980€E-04 NG DATA 1.532E-05 NO DATA 2.180E-03 NO DATA 2.519€E-04 NO DATA 2.481€E-07 NG DATA 1.135€ 0C NO DATA 1.243E~-01 NO DATA 1.801£-01 NO DATA 4.234E-02 NO DATA 4, 682E-02 NO DATA 2.520E QO NO DATA 6.061E-01 1.319€-02 1.2C4E 01 G.l. SOULUBLE 6.2REE-CB 9. £90E-023 2.3220€-G2 2.220€-03 2.320€-03 1.740E-02 2.40TE~-Q2 2.507€-02 2.353E-02 1.843€-07 4,047E-02 1.242€-01 2.701E-05 3.443E-05 6. TO0TE-Q6 1.,56BE-0% 4.542€-06 2,893E-04 Z-ngE'04 2.B0TE-C4 9,2326E-05 8-695E-04 1.548E=~05 G+ BSTE-0S 1.53%€E-03 S.448E-CH 4.B15€E~04 &, 810E-05 B8.994E~05 2.5T75€-03 2.332€-02 6.305E~05 4. £33E-06 1.806E-02 1. 30¢4E~05 Se224E-07 le44T€-C4 2.T25€E-05 5'269E-07 4.25%E-04 4,5G6E-C5 T«237E-05 3. 4B2E-04 3.028E-05 2.725E-03 3.77CE-0Q4 2.505E~01 TRACT INSOLUBLE 3, 143€E-06 1. T44E-D2 3. 8656E-03 3, 8¢ 6E~-03 3. B64E-03 2+320E-0G2 2. 889E-02 2.507€~02 2.353£-02 4.300€-07 8. 094E-03 1.862€~01 4.052E-05 4,304E-35 1.006E-05 1. 96CE~05 6. 056E-06 3.693E~04 1.485E~04 6.23BE~-C4 1.84TE~D4 1.522€~03 3.837€~05 1,971E~-07 2.T64E-02 1.8%0E-06 1.204E-02 9. 766E-05 1l.124€E-04 2.575€-03 3. 4976-01 7.005E-05 4.633E-06 2.257€E~03 1,492E-2% 1. 045E-06 1,447E-04 3.40TE-05 Te 9C3E~J7 5.996E-05 8. 685E-05 5.224E-04 3.,634E-05 3.4OTE-03 4.524E-04 T.269€E-01 0L-8 Table 8.10. NO. DO~ Wi NUCL IDE H-3 SR-89 SR-90A SR-cDB SR-90C Y-90 ¥-91 ZR-95 NB-95 TC-99 RU-1013 RU-106 RH-103M AG-111 CO-115H 58-12¢4 SN-125 $SB-125 TE-125M TE-127M YE-127 TE-129M TE-129 I-121 CS-134 CS-13¢ CS-137 3A-140 LA-140 CE-*4l CE~144 PR-163 ND-147 PM-147 SM-151 EU-152 EU-155 T8-160 NP-239 PU-238 PU-239 PU-240 PU-241 AM-241 CM-242 CM-244 TOTAL LABEL 303 38 901 902 9n3 43 48 -3 67 79 8e 97 k] 114 125 159 161 162 163 1£9 179 176 177 187 327 2" 7 210 221 222 227 238 237 246 247 25% 328 262 272 330 280 281 282 283 1 2 3 TOTAL BOOY 3. 7C0E-07 4.T734E-02 4.5CHE-04 6.408E-03 5.753E-01 2.035E-04 6.361E-03 1.877€E-02 9.859E=93 2- 97“E'09 2.924E-04 4.972E-03 2.908E-07 5.269E-07 1.056E~06 3.821E-06 1.759E-07 1.268E-0¢ A, 96TE-06 4.353E-05 9.420E-0T7 1l 2‘089-04 3.621E-07 2.425E-08 5.090£-02 4. 9T0E-Q7 1.463E-02 1.323%E-05 1. T60E-CE 3.979€-04 Z2.130E-01 3.340E-06 2 523e-07 2.926E-03 3.4T7T1E-05 4.727€E-06 3-565E-04 1.875E-05 1.310nE~08 1.7176 00 2.236F-01 3.234E-01 1.24€€ QO 1.154F-21 2.158E-01 T.89S¢E-01 5.540F 00 BONE NO DATA 1.690E-01 6. T758E-03 9.912E-02 8. 63CE QO T.647F-03 2.2684E-01 Te244E-02 2.887€-02 7.435E-09 5.908E-04 2,563E-02 S.T40E-07 2,371E-06 ND CATA 1.0CRE-0OS 2.977E~-08 6.0864E-04 5.,053E-05 3,230E-04 4.325E-06 5.,197E-04 9,281E-07 NO DATA 3,383€E-02 2.382€-07 2.738E~02 1.636E-04 1.8C1€-05 4,758E-03 3,940E 00 6. T2SE-05 3.2076-06 7.953€-02 B.644E-04% 2.360E-°5 3.140€-03 1. 500E"04 1.947€-07 6.823E 01 9.212€ 00 1.333€E 01 6.048E 01 1,761E 00 3.250E 00 1.331E Q1 1.B821E 02 MUSCLE NC NO NG NG NC NQO NC NO NG NO AC NQ NO NO NO NG NO NO NO NO NO NG NG NC DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA THYRCQID NG DATA NC DATA NC CATA NO DATA NO CATA &0 CATA NC CATA NO DATA NO CATA NO CATA NO CATA NG DATA NG CATA NG DATA NG DATA 2¢332E-08 6.823E-08 5.680E-07 1.663E-05 9.120E-05 3.210E~06 1.917E-04 1. 780€E-07 1.361E-05 8.770E~0Q2 §.018€-07 2.692E-02 1.084E-07 NO DATA NC DATA NO DATA NO DATA NO DATA NO DATA NO DATA NG DATA NOC DATA NO DATA NO DATA NO DATA NC DATA NG DATA NC DATA NO DATA NO DATA NO DATA 1.146£-01 NO NO NC NO KO NO NG NO NC NO NG NO NC NO NC NO NO NO NO NO NO NC 2,178E-04 DATA DATA DATA DATA DATA CATA DATA DATA DATA DATA CATA DATA DATA CATA CATA CATA CATA DATA CATA CATA DATA CATA LIVER NO DATA NO DATA NO DATA NO DATA NO DATA NGO DATA NO DATA 3.060E-02 1-661E-02 1.102€-08 NO DATA ND DATA 4,T89E-07 1.017€E-06 3.297E-05 1.675E~09 8.113E-08 6.176E-06 3.360E-05 1.6649E-G4 2.095E-06 30265E-04 5.861E~07 NO DATA 9.022E-02 T.614E-07 3.549E-02 2.366E-07 b, T14E-06 3.258E-03 1.609E 00 2.689E-05 3.893E-06 9. £40E-03 1.48‘0E-0" 5.181€-06 6.995E£-04 NO DATA 2.127¢-C3 9.T64E 00 1.259E 00 1.823E 00 3,137€E 00 6.10TE~-Q1 3.312E 00 S.T43E 00 2.744E 01 Intake period of l-day duration; begins at age 20 K1ONEYS NO CATA NO CATA NQ CATA NG DATA NO CATA NC DATA NG CATA 3.894E-02 1.804E-02 1.388E~-07 2e446E-03 6.BT2E-02 1.902E-06 2.916E-06 2e669E-05 NO CaTA NO DATA NOQ CATA 2+856E-04 1.419€-03 1.781€-05% 2.608E-03 4.641E-08 NO DATA 34056E-02 4. TC4E-O7 1.285E-02 T.661E~08 NO DATA 1.592€-03 9.939E-01 1.559€E~05 1.825€E-06 1.366E-02 1.620E-04 2.522E-05 1.806E~03 6.176E-05 6.639E-08 7.280E 00 9.5C1E~0C1 1.376E 00 £4969E O B.650E-01 9.817€E~01 3.684E 00 20229E 01 Inhalation dose commitment (in rems) integrated over 50 years SPLEEN NO DATA NO DATA NG DATA NO DATA NO DATA NG DATA NO DATA 24336E~-02 1.456E-02 NO DATA NG DATA NO DATA 1,332E-0¢ NO DATA NO DATA NO DATA NO CATA NO OATA T.235€-05 3.59%E=04% 4.512€E~06 6.6C7E-04 1.176E-08 NO DATA 6.813FE-02 5.4Bl1E-07 3.C06E-02 9.614E~-08 NO DATA NO DATA NO DATA NO DATA NO DATA NC DATA NO DATA NO DATA NG DATA NO DATA ND DATA NO DATA NO DATA NO DATA NDO DATA NO DATA NO DATA NO DATA 1.372E-01 TESTES NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NU DATA NO DATA NO DATA NO DATA NC DATA NO DATA NO DATA NO DATA NO DATA 6.171E-05 3.780E-04 4.89BE-06 6.345E~04 1.126E-06 NG DATA NG NG NO NO NO NO NO NG NO NO NO NO NO NO NO NO NO NO ND NOD NO NO 1.CBGE-03 DATA OATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA Internal Dose at LOO m Downwind Following the Release of 1 kg of LWR Fuel OVARIES NO NO ND NO NO NO NO NO NO NO NO NO NO NO NO NOD NO NO NO NO NO NG NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO NO 0.0 DATA DATA DATA DATA DATA DATA DATA DATA DATA OATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA CATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA DATA T.-8 2 a J3O0@M~4OWV S U NUCL I DE H-3 SR-89 SR=G0A SR-90B SR-90C Y-90 Y-91 IR=-95 NB-95 ¥C-99 RU-103 RU-106 RH-103M AG-111 CO-115M SB-124 SN-125 sB-125 TE-125M TE-127M TE-127 TE-129M TE-"29 1-131 CS-114 £s-136 Cs-137 BA-140 LA-14) CE-141 Ce-144 PR=143 ND=-147 PM-~147 SM-151 FU-152 FUu-18%5 T8-160 NP=230 PU-238 PU=-239 PU-240 P1=241 AM-241 CM=-242 CM-244 TOTAL LABEL 303 38 901 902 903 43 48 65 67 79 88 97 8% 114 12% 159 161 162 163 169 170 176 177 187 327 207 210 221 222 227 238 237 246 247 255 328 262 272 323 280 281 282 283 1 ? 3 LUNGS SOLUBLE INSOLUBLE NO DATA 3.,154E-05 NO DATA T.317€-02 NO DATA 3. 896E-01 NO DATA 3. 896E-01 NO OATA 3.8G6E-01 NO DATA 6.G68E-0Q3 NO DATA 1.430E-01 NO DATA 2.319E-01 NO DATA 1.423€-01 9.368E-10 6.245€E-06 NO DATaA 2.866E-02 NO DATA 2.021€ 00 NO DATA T.286TE=-D% NO DATA 3.964E-05 NO OATA 3.335E-05 T.529€-06 1,244E-04 NC DATA 6+ 590E-06 4.199E-D4 B.364E-03 NO DATA 6.979E-04 NG ODATA 4.318E-03 NO DATA 2+21%E-05 NO DATA 5.,612€E-03 NG DATA &6.225E-06 NO DATA 1.,915€E-07 1.045€6-02 4.9TE-01 6.215E-08 3.329E-08 4.303E-03 2,011E~01 1.322E-07 2.715E-06 NO DATA 3,517€E~-05 NO DATA 1.0C05£~02 NO DATA 3.313E Q¢ NO DATA 1.065€-04 NG DATA 6.161E-06 NO DATA 2.839E-02 NO DATA 2.2%3€-04 NO DATA 1,.731€E-05 NO DATA 2.373E-03 NO DATA 2.529E-04 NO DATA 2.481E-07 NO DATA 2.255E 0O NO DATA 2.489E-01 NO DATA 3,605E-01 NO DATA 8.0¢L1€E-C2 NG DATA 5.330E-02 NO DATA 2.587E 00 NO DATA 6.B65E-01 1.518E-02 1.416E 01 1 1 G.l. SOLUBLE 6.2BEE-C8 G9.€E9CE-Q2 2.320E-03 2.320€E-02 2.320E-03 1. 740€E-C2 2."0 TE-OZ 2.5CTE-02 2.353E-C2 1.843E-07 4,047E-02 1.242E-01 2.701€E-05 3. 443€E-05 b.T707E-06 1.56B8E~05 4q542€~06 3e693E-~C4 2.9T1E~04 2.807€E-C4 9,236E~C5 B.ES55E~04 1.949E~-05 9.857E~-06 1.935€-03 9.448E-08 4.B15E~-04 6.510E-05 8.994E-05 2.575€E-013 2.322E-013 6.,3CCE-05 4.623E-06 1.804E-C3 1.304£-05 S.224E-07 ls44TE-D4% 2.T25E-05 5. 265E-07 4,255E-04 4.99LE-05 T.237E=-05 3.,482E-0C4 3.028E-05 2.725€-03 3, 7TT0E-C4 2.505E-01 Table 8.10 (Continued) TRACT INSOLUBLE 3.143E-06 lo 7‘0‘05‘02 3. 866E-013 3.866E-03 3. B6EE-Q3 2.320€E-02 2.507E-02 2.353€-02 4+ 300E-07 8. 094E-03 1l.862€E~01 4,052E-05% 4 304E-05 1. 006E-05 1.96CE~-D5 6. 056E-06 3. 693E-04 1.485E-04 6o 238E-—04 1. 847E-04 1.522E-03 3.897€-05 1.971E-07 2. TE4E=-02 1.890E-06 1. 204€E-02 9. T66E-05 1. 124E-04 2.575E-013 3.49TE-01 T.005€E-05 4.633E-06 2.257E-03 1.492E-05 1-045E‘06 1.44TE-O4 3.407E-05 T+ 202E-C7 5.105E-04 S« 994E-Q5 B. 685E~-05 5.224E-04 3, 634E-05 3.407€E-03 4,524E=04 T.26%€-01 2L-8 Table 8.11. Internal Dose at 400 m Downwind Following the Release of 1 kg of LMFBR Fuel Tnhalation dose commitment (in rems) integrated over 1 year NO. NUCLIDE LABEL TOTAL BODY BONE MUSCLE THYRCID LIVER KIDNEYS SPLEEN TESTES OVARIES 1 H-3 303 4.98B4E-07 NO DATA NO DATA NO DATA NO DATA ND DATA NO DATA NO DATA NO DATA 2 SR-B9 38 3.120E-02 1.114E 00 NO DATR NO CATA NO DATA NO DATA NO DATA NO DATA NO DATA 3 SR-30A 901 2.553E-04 3.829E-023 NO DATA NG OATA NG DATA NO DATA ND DATA NO DATA NO DATA 4 SR-50B 902 3. 7326032 5.,598E-02 NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA 5 SR=90C 903 2.083E-02 3.125¢-01 NO DATA NQ DATA NC DATA NO DATA ND DATA NO DATA NO DATA & Y-90 43 1.156E=04 44343E-03 NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA 7T Y-91 48 3.466E-02 1,295¢€ Q0 NO DATA NQ DATA NQ DATA NO OATA NO DATA ND DATA NO DATA 8 IR-35 &5 l.413E-01 5.433E-01 NC DATA NO DATA 2.308E=-01 2.922E-01 1.753E-01 NO DATA NO DATA 9 NB-§5 67 5.060E-02 1.481E-01 NO DATA NO DATA 8.523E-02 9.260E-02 To4T4E-02 NO DATA ND DATA 0 MO-99 17 5.096E~0D6 NO DATA NO DATA NO DATA 2,699E-05 6.493E-05 NO DATA NG DATA NO DATA LT TC-99M 78 1.110E-C8 5.550E-11 NO DATA NO CDATA 5.999€E-10 8.418E-09 NO DATA NO DATA NO DATA 12 TC-939 19 3.100E-09 7. T49E-0D9 NO DATA NO DATA 1.14BE~08 la446E-07 NO DATA NO DATA NO DATA 13 RU-103 88 5.77T7E-03 1,167E-02 NO DATA NO DATA NO DATA 4.831E-02 NO DATA NQ DATA ND DATA 14 RU-106 37 1.564E-02 1.121e~01 NO DATA NO DATA NO DATA 2.162€E-01 NO DATA NO DATA ND DATA 15 RH=103M 8% 5.737E-06 1.132E-05 NO DATA NQ CATA 9.449E-06 3.755€E-05 2.629E~-05 NO DATA NO DATA 16 AG-111 114 1.752E-05 T.883E-05 KD DATA NO DATA 3.382E-05 9.695E-05 NO DATA NO DATA NO DATA 17 CD-115M 125 6.393E~06 NO DATA NO DATA NO DATA 2.001E-04 1.8619E~04 NO DATA NO DATA NG DATA 1B S5B-124 159 3, 396E~N4 8.950E-06 NO DATA 2.073E-08 1.489E-09 No CDATA NO DATA NO DATA NO DATA 19 SN-125 161 5.,885E-05 9. 959E-04 NO DATA 2.282E-05 2.714E-05 NO DATA NO DATA NO DATA NG DATA 20 58-125 162 3.147E-04 1.416€E-03 NO DATA 1.411E-06 1.533¢-05 ND DATA NO DATA NO DATA NO DATA 21 TE-125M 163 2.158E-05 1.094E-04 NO DATA 3.600E-05 T.275E-05 6.184E-04 1.566E~-04 1.336E-04 NO DATA 22 Te-127M 169 4,304E-04 3,194E-Q3 NO DATA 9.017€-04 1.65CE~03 1.403E-02 3.553E-03 3.737€E-03 NO DATA 23 TE-127 170 9. 544E-06 4.382F-05 NO DATA 3,252E-05 2.123E-05 L.804E-04 4.5TLE~-05 4.962E-05 NO DATA 24 TE-129M 176 3.371E-03 1.404E-02 NG DATA 5.179E-03 8.821E~03 T.046E-02 1.785E-02 1. 714E-02 NO DATA 25 TE-129 177 9, 817E-06 2.510E-05 NO DAYA 2.104E-05 1.585€E-05 1.255E-04 3.179€E-05 3.045E-05 NO DATA 26 TE-132 191 3.488E-05 5.680E-05 NO DATA 44.389E-05 4.882E~-05 3.,621E-04 .1 THE~DS T+5T2E-05 NO DATA 27 1-131 187 1.553E-03 NO DATA NO DATA 8.716E-01 NO DATA NO DATA NO DATA NO DATA NO DATA 28 [-132 192 2.370E-06 NO DATA NO DATA 9, 725E~04 NO DATA NG OATA NO DATA NO DATA NO DATA 29 CS~-134 327 6.788E~-03 4,045€-03 1.049E-02 ND DATA 1.174E-02 44154E-03 8.751E~03 NO DATA NGO DATA 30 (S-13s 207 6. BA2E-04 3.299E-04 6.949E=-04 NO DATA 1.054€E-03 6.513E-04 T.590E-04 NO DATA NO DATA 31 CS-137 210 1.464E-02 2.3&64E-02 2325E~02 NO DATA 3.435E-02 1.318E-02 2.863E-02 NO DATA NO DATA 32 BA-140 221 1. 606E-02 1.99¢e-01 1.318E-04 NO DATA 2.BT7BE-04 9.318E-05 1.169€E-04 NG DATA NO DATA 33 LA-140 222 2.137E-03 1.215€E-02 NO DATA NO DATA 8.151E-03 NO DATA NO DATA NO DATA NO DATA 34 CE-14! 227 8.42CE-03 1.007e~01 NG DATA NO DATA 6.895E-02 3.368E-02 NO DATA NO DATA NQ DAYA 35 CE-144 238 2.597E~-01 4.234E 00 NO DATA NO DATA 2.201€ @O 1.212€ 00 NO DATA NO DATA NO DATA 36 PR~143 237 3.099E-03 6.244E-02 NO DATA NO DATA 2+496E-02 l.446E~02 NO DATA NG DATA NO OATA 37 ND-147 246 9.15CE-04 1.163E-02 NO DATA ND DATA 1.412¢-02 6.,619E-03 NO DATA NO DATA NO DATA 38 PM-147 267 5.019E~03 1.011E-01 NO DATA NO DATA 1. 653E-02 24342E-02 NO DATA NG DATA NO DATA 39 PM-149 251 8.139E-08 1.031E-06 NGO DATA NO DATA 2.185E-07 2.954E~0T7 RO DATA NG DATA NO DATA 40 SM-151 255 4.587E-05 5.667E~04 NO DATA NO DATA 4+.495E-04 2.141E~-04% NO DATA NG DATA NO DATA 41 EuU-152 328 1.569E-06 4.293E-06 NO DATA KC DaATA 4.114E-06 4. 632E-06 ND DATA NO DATA NO DATA 42 EU-155 262 2.456E-03 1.713E-02 ND DATA NO DATA T.933E-03 9.866E~03 NO DATA NG DATA NO DATA 43 TB-140 272 5.TB4E-N4 4.615E-03 NO DATA NO DATA NQ DATA 1.905E-03 NO DATA NO DATA NO DATA 44 NP-239 330 5. 435E-06 8.081E-05 NO DATA NO DATA 8.828E-06 2. 755E-05 NO DATA NO DATA NO DATA 45 py-238 280 1.787E-01 7.023E 00 NO DATA NG DATA 1.116E 00 8.319E-01 NO DATA NO DATA NO DATA 46 PpU-239 281 5.255E=-02 2.142E 00 NC DATA NO DATA 3.282E-01 2.448E-01 NO DATA NO DATA NO DATA 4T PU-240 282 6.342E-02 2.585E 00 ND DATA NO DATA 3.961€-01 2.954E-01 NO DATA NO DATA NO DATA 48 PUY-24) 283 3, T77E-01 1.835E 01 NO DATA NO DATA 1.025E 00 1.912e 00 NO DATA NO DATA NO DATA 49 AM=-241 1 2.501E-02 3.110E-01 NO DATA NO DATA 3.518E-01 1.753E~01 NO DATA NO DATA ND DATA 50 (CM=-242 2 T.452E-01 1.121E 01 NO DATA NO DATA 1.165E 01 3.391€ 00 NO DATA NO DATA NO DATA 51 CH-244 3 2.044E-02 3.077e-01 NO DATA NO DATA 3.250E-01 9.539E~02 NO DATA NO DATA NO DATA. TOTAL 2.094E 00 5.036E 01 3.456E-02 8,788E-01 1.791E 01 9.002E 00 3.101E-01 2+1117E-02 0.0 Intake period of l-day duration; begins at age 20 £L-8 Zz o . Vo4O NH BN~ NUCLIDE H-3 SR-89 SR~90A SR-908 SR-90C Y-90 Y-91 IR-55 NB~95 MO-99 TC-99M TC-99 RU-103 RU-106 RH-103M AG-111 CO-115M $8-124 SN-125 sB-125 TE-125M TE-127M TE-127 TE-129M TE-129 TE-132 1-131 1-132 Cs-134 CS-136 €5-137 BA-140 LA-140 CE-141 CE-144 PR=-143 ND-147 PM-147 PM-149 SM-151 EU-152 EU-155 TB-160 NP-239 PU-238 PU~239 PU-240 PU-241 AM-241 CM-262 CM=-244 TOTAL LABEL 303 38 901 902 903 43 246 247 251 255 328 262 272 330 280 281 282 283 LUNGS SOLUBLE INSCLUBLE NO DATA 3.758E~05 NO DATA 4.851E-01 NO DATA 1.945E-01 NO DATA 1.945E~01 NO DATA 1.945E~01 NO DATA 3.957E-02 NO DATA 8.271E~01 NO DATA 1.T60E 00 ND DATA T.306E-01 NO DATA 9. 236E-05 3.060E~10 8.497€E~08 9. 764E-10 5, TI16E-06 NO DATA 5.661E-01 ND DATA 5.972E 00 NG DATA 1.434E-04 NO DATA 1.318E-03 NO DATA 2.,025E-04 6. 683E~08 1.103E~-04 NO DATA 2.204E-03 9, 804E~04 1.888E-02 NO DATA 1.509E-03 NO DATA 4,222E-02 NO DATA 2+ 248E-04 NO DATA 1.,516E-01 NO DATA 1.683E-04 NO DATA 5.5T4E~04 NO DATA 1.227€-02 NO DATA 1.625€-08% 1,250E~03 6.157T€E-02 8.606E-05 4,610E-03 3,715€E-03 1.822€-01 1.608E-04 3.302E-01 NO DATA 4.271E-02 NO DATA 2.128E-01 NO DATA 5.226E 00 NO DATA 9.880E-02 NO DATA 2.235E-02 NO DATA 9.148BE-02 NO DATA 2,279E-06 NO DATA 8,0715E~-04 NO DATA 1.398E-05 NO DATA 2.7T18E-~02 NO DATA T.945E-03 NO DATA 1.029E-04 NO DATA 4.523E 00 NG DATA 1.330€E 00 NO DATA. 1.605€ @G0 NO DATA 2+ 209E~01 NO DATA 3,675E-01 NO DATA 1.100E 01 NO DATA 3.018€E-01 6.199E-03 3. 682F 01 Table 8.11 (Continued) 6.1« TRACT SOLUSBLE INSOLUBLE B.46TE-08 4.233E-06 6.430E-02 1.157E-01 1.314€-03 2.190E-03 1.314E-03 2.190E-03 1. 314E-03 2.190E-03 9. 8TYE-03 1.3176-02 1.394E-01 1.673E-01 1.908E-01 1.90BE-01 1.2086-01 1.208E-01 3.289E-05 2.349E-04 1.572E=-06 3. 143E-06 1.921E-07 4.482E-07 7.994E-02 1.599€-01 3,906E-01 5.859E-01 5,329E-04 7,.994E-04 1.145E-03 1.431€-03 4.073E=05 6.109E-Q5 1.394E-05 1.742E-05 1.519€-03 2.026€~03 9., 1756-04 9, 1756-04 6.432E-04 3.216E-04 2.TT5E-03 6.16TE=03 9,357E-04 1.8T1E-03 2.349E-02 4.111E-02 5.269E=04 1.054E-03 5. 412E-04 G.4TO0E-04 6.314E-04 1.263E-02 4,340E-05 1.302E-04 2.634E-04 32.T64E-03 1.308E-04 2,616E-03 4.951F-04 1.238E-02 T.919E-02 1.188E~01 1.0926-01 1.365E-01 5.45)1F~02 5.451E-02 3,876E~03 5.B814E-01 5,850E~02 &.500E-02 1.681€-02 1.681E-02 6.414E-03 8,017E-03 5.587E-06 6.984E-06 54326E-05 6.08TE-0S 4,765E-07 9,53%E-07 1.803€~03 1.803E-03 8.594E-04 1.074E~03 2.1866-04 3,279E-04 1.69¢E-03 2.035E-03 S5.345E=04 6.414E-04 6.450E-04 7. T4 CE~D4% 1.817€6-03 2.725E-03 2.37176-04 2,852E-0Q4 1.190E-02 1,488E-02 1.383€ 00 2.455E 00 ML-8 Table 8.12. NO. — SO ®NN P WA NUCLIRDE H-3 SR=-89 SR-90A SR-S0B SR=30C Y-90 Y-91 IR=-95 NB-95 MO-99 TC-G9M TC-99 RU-103 RU-106 RH=-103M AG-111 CO~115M $8-124 SN-125 SB-125 TE-125M TE-12TM TE-127 TE=-129M TE-129 TE-132 -131 I-132 CS-134 €S—-136 C5-137 BA-140 LA-140 CE-141 CE-144 PR=143 ND-147 PM-147 PM-14G SM~151 EU-152 EU-155% TR-16C NP-239 PU-238 PL-239 PU=240 PU=-241 AM=-261 CM-242 CM-244 TOTAL Internal Dose at 400 m Downwind Following the Release of 1 kg of LMFBR Fuel o Inhalation dose commitment (in rems) integrated over 50 years Intake period of l-day duration; begins at age 20 LABEL 33 38 201 902 903 43 48 6% 67 77 78 79 a8 97 89 114 12% 159 161l 162 163 1&9 170 176 177 191 187 192 327 207 20 330 280 281 232 283 TOTAL BODRY &, 9B4E-OT 3. 141FE-02 2.553E-04 3. 744€-03 3.26CE-01 1.156E-0% 3.511E-02 1.428E-01 5.063E-02 5.096E-06 1.11GE-08 3.100€-09 5. TTTF-03 1.564E-02 5.737E-04% 1.7528-05 6. 398E-06 3,396E-06 5., 885E-0Q5 3. 150E-04 2.158F=05 4.304E-D4 9. 544E-06 3,371E-03 9.817€-06 3.488€~-05 1.553€~03 2.370E-06 6.932%€-03 6,882E-04 1.504E-02 1. 4608E-02 2.137E-03% Bs422E-03 3.541E-01] 3.099E-03 9.150FE-04 1.039e~02 8.139E-0D8 1.415F~04 4,316E-06 4. 444E-01 5.912E-04 5. 435E-06 6.842E 00 2.392€E 0C 2.883E 00 6.502E 00 9. 058E-0} 9.422E-01 3.931E-01 2.190FE 01 BONE NO DATA 1.122€E 00 3.829E-03 5.616E~02 4,889E GO 4e343E-03 1.312¢ Q0 5.512g-01 l.482E-01 NO DATA 5.550E=-11 7.T49E~-09 1.167€-02 1.121E=-01 1.132E-05 T.883E~-05 NG DATA B.961E-0Q6 S.98CE-04 1.507E-03 1.094E-04 3.194E-03 4.282€-05 1.404€-02 2.,510€E-05 5.680E-05 NG DATA NG DATA 4. 606E-03 3,299E~-04 2.815E‘02 1.990E-01 1.215E-02 ]oOOTE‘OI €.549E €0 b6.244E~02 1.163E-02 2.824F~01 1.031E-06 3,525€-03 2. 154E-05 3.91‘?E'°2 4. T30E-03 8.081E-05 2.719E 02 9. 854E 01} 1.188E 02 3.166FE 02 1.383E 01 L.419E 01 6.630E 00 8.560FE 02 MUSCLE NO DATA NO DATA NO DATA NC DATA NQ DATA ND CATA NC DATA NO DATA NC DATA NO DATA NC DATA NO DATA NC DATA NO DATA NCQ DATA NC DATA NC DATA NGO DATA NO DATA NO DATA NO DATA NO DATA ND DATA NO DATA ND DATA NO DATA ND DATA NO DATA 1.194E=-02 £.949E-Cé 2:768E‘02 1.318E-04 NO DATA NO DATA NG DATA NO DATA NG DATA NO DATA ND DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA 4.,045E-02 THYRCID NC DATA NO DATA NC OATA NO DATA NG DATA NO DATA NC CATA NG CATA NO DATA NO DATA NC DATA NO CATA ND DATA NO DATA NO CATA NO CATA NG CATA 2.073E-08 2.282E‘°5 l.411E~06 3.400E-05 9.017E-04% 3.252E~05 5.,179€E-03 2.104E-05 4.389E~05 8,716€E-01 9,725E-04 NO DATA NO CATA KO DATA NO DATA NG DATA NG DATA NO CATA NO DATA NO EATA NO DATA NO DATA NC DATA NO DATA NG DATA NO DATA NC CATA NC DATA NO DATA NO DATA NO DATA NO DATA NG [ATA NO DATA 8.788E-01 LIVER NO DATA NO DATA NO DATA NO DATA ND DATA NO DATA NDO DATA 2.328E-01 80527{‘02 2+699E-05 5.999E-10 1-149E-08 NO DATA NO DATA 9.449E-06 3.382€E-05 2-0025*04 1.489E-09 2+714E-05 1.534E-05 7.275E-05 1.650E-03 2.123E-05 8.821E-03 1.585E-05 4.882E-05 NO DATA NO DATA 1.228E-02 1.054E-Q3 3.649E-02 2.BT8E-04 8.151E-03 6.896E-02 2.675€ 00 2+456E-02 1.412€~02 3.423E~02 2.185E-07 6.050E-04 4»T30E-06 B.T19E-03 ND DATA 8.82BE~Q6 3.,892E 01 1.346E 061 1.625E 01 1.637€ 01 4,7T94E 00 1.446E 01 2,B60E 00 1.103E€ 02 KIDNEYS ND DATA NO DATA RO DATA NO DATA NO DATA NG DATA NO DATA 2¢963E-01 9.265E-02 6.493E~05 8,41BE~09 l.446E-07 4,831E-02 2.162E-01 3,.755€E~05 9.695E~05 1.621E-04 NO CATA NO DATA NO DATA 6.184E-04 1.403E-02 1l.804E-04 T.046E~02 1.255€E-04 3,621E-04 NO DATA NO DATA 4#.161E-03 6.513E-04 1.321E-02 9.318BE-05 NG DATA 3.369E-02 1.652E 0Q l.446E-02 6.619E-03 4.,850€E-02 24954E-07 6.605E-04 2.303E-05 2.251E-02 1.948E-03 2. 758e-05 2.902E 01 l.016E 01 1.226E 01 3.114E 01 6.790E 00 4,287 00 1.835€ 00 9.803E 01 SPLEEN NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA 1.778E-01 7.4 T8E-02 NO DATA NO DATA NO DATA NO DATA NO DATA 2.629E-05 NO DATA NC DATA NO DATA NO DATA NO DATA 1.566E-04 3.554E-03 4.5T1E-05 1.785€-02 3.179€-05 9.1 T4E-05 NO DATA NO DATA 9.276E-03 7.590E~04 3.091E-02 1.169E-04 NO DATA NO DATA NO DATA NG DATA NO DATA NO DATA NO DATA NOG DATA NO DATA NO DATA NO DATA NO DATA ND DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA 3.154E-01 TESTES NO DATA NO DATA NO DATA NGO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO QOATA NO DATA NG DATA NO DATA NC DATA 1.336E-04 3.737£-03 4+962E-05 1 .TL14E-Q2 32045E-05 T.872E-05 NO DATA NC DATA ND DATA HO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NGO DATA NO DATA NO DATA NO DATA NO DATA RO DATA NO DATA NO DATA NO DATA 2.117E~-02 OYARIES NO DATA NO DATA NO DATA NO DATA NO DATA NG DATA NO DATA NO DATA NO OATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA ND DATA NO DATA NO DATA NO DATA NO DATA NQO DATA NO DATA NO DATA ND DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA NO DATA SL-8 £ =] -* O AR W e NUCLIDE H-3 SR~-89 SR-90A SR-909 SR-90C Y-90 Y-91 IR-G5 NB-9% MO-99 TC~99M TC-99 RU-103 RU-106 RH-103M AG-111 CD-115M Sp-124 SN-125 $B-125 TE=-125M TE-127M TE-127 TE-129M TE-129 TE-132 1-131 1-132 cs-134 €5-136 Cs-137 BA-140 t A-140 CE-141 CE-144 PR-143 ND-147 PM-147 pPM=149 $M-151 EU-152 EU-155 TA=-160 NP=-239 PU-238 PU-239 PU=-240 PU-241 AM-241 CM-242 CM-244 TOTAL LABEL 303 38 901 902 903 43 48 65 67 77 78 79 88 97 89 114 125 159 161 162 163 169 170 174 177 191 187 192 327 207 210 221 222 227 238 237 2646 247 251 255 32s 262 272 330 280 281 282 283 1 2 3 LUNGS SOLUBLE INSCLUBLE NO DATA 4, 247€-05 NO DATA 4,855E-01 NO DATA 2.207E-01 NO DATA 2.207€-01 NO DATA 2.207€-01 NO DATA 3,957£-03 NO DATA 8.284E-01 NO DATA 1. 764E 00 NO DATA 7.307£-01 NO DATA 9.236€-05 3.060E-10 B.457E-08 9, 764E-10 6.509£-06 NO DATA 5.,662E-01 NO DATA 6.359€ 0Q NO DATA 1.434€-04 NO DATA 1.318€-03 NO DATA 2.025€-04 6.691E-06 1.105E-04 NO DATA 2,204E-03 1.0436-03 2,078E-02 NO DATA 1a511£-03 NO DATA 4.269E-02 NO DATA 2.24BE-04 NO DATA 1.516E-01 NO DATA 1, 683E-04 NO DATA 5,574E-04 NO DATA 1.227E-02 NO DATA 1.625E-05 1.4236-03 6.T67E-02 8.606E-05 4,610E-03 4,424E-03 2,068E-01 1.608E-04 3.302€-01 NO DATA 4.271E-02 NO DATA 2.128E-01 NO DATA 5,507E 00 NO DATA 9,880E-02 NO DATA 2.2356-02 NO DATA 1.008E-01 NO DATA 2.279€-06 NO DATA 9.187E-04 NG OATA 1.581E-05 NO DATA 2.953E-02 NO DATA 7.974€-03 NO DATA 1.029E-04 NO DATA 8.986€ 00 NO DATA 2.662E 00 NO DATA 3,213E 00 NO DATA 4,206E-01 NO DATA 4,184E-01 NO DATA 1.130€ 01 ND DATA 3.419E-01 7.145E-03 4.,561E 01 Table 8.12 (Continued) GaTe SCLUBLE 8.44TE-08 6.430E-02 1.314€E-03 1l.314E~-02 1,314E~03 9. 87SE-03 1.394£-0C1 1.908E-01 1.208€-01 3.289E~-05 1.572E°06 1.921€-07 Te994E-02 3.606E-01 5. 329€-04 1.145E-03 4.073E~05 1.394E-05 1.519E-03 9,175E-04 6.432E-04 2. 775€E-03 9,35TE~-04 2.349E-02 54269E~04 5.4126-04 60 314E-C4 4.340E-05 2. 6234E-04 1.308E-04 4,951E-04 7.919E-02 1.092E-01 5.451€-02 3,87€E-03 5.850E-02 1. 681E-02 6.414E-03 5.587€E-06 5.32€E-05 4,T69E-07 1,803E~C3 8.594E-C4 20186E‘°4 1.69¢€E-03 5.345E-04 6. 450E-04 1.817E-03 2.377E-04 1.19QE-02 1.877E~-C4% 1.383E 00 TRACT INSOLUBLE 4. 233E-06 1.157E-01 2.150E-03 2.190€-03 2.190E-03 1.317E-02 1.673E-01 1.908£-01 1.208€-01 2.349E-04 3.143E-06 4.482E-07 1.599E-01 5.859E-01 Te 994 E-04 1.431E-03 6. 109E-05 1.742E-05 2,026E-03 9. 175€E-04 3. 21 6E-04 6,16 7E-03 1. 871E"03 4.,111E-02 1.054E-03 9., 4T0E-04 1,263E-02 1. 30 2E-Q4 3. T64E=03 2.616E-03 1.238E-02 1.188E-01 1.365E-01 5.451E-02 5. 814E-01 6,500E-02 1-681E'°2 8,017E-03 6.984E-06 6,08 TE~05 9,53%E~07 1.803E-03 1,07 4E-03 3,279E-04 2,035€E-03 &6, 414E-04 T. 740E=~04% 2.725E-03 2., 852E-04 1.488E-02 2,253E-04 2.455E 00 9L-8 Table 8.13. 8-77 Summary of Maximum Inhalation Dose Commitments® at 4,00 m Downwind Following the Release of 1 kg of LWR or Mixed LMFBR Fuel from a 100-m Stack Whole Body Bone Lungs Liver Thyroid IWR Fuel - First-Year Dose Commitment Volatile fission products 0.00000039 - 0.000018 - 0.0000136 Semivolatile fission 0.0695 0.0898 2.56 0,120 0.000303 products Nonvolatile fission 0,242 3.74 .83 1.38 - products Plutonium 0.129 5.78 1.48 0.552 - Transplutonic elements 0,215 3.22 3.17 3.37 - Total 0.655 12.8 12.0 5.42 0.000318 LWR Fuel - Lifetime Dose Cormitment Volatile fission products 0.00000039 - 0, 0000317 - 0, 0000136 Semivolatile fission 0.0709 0,0983 2.76 0.126 0,00030L products Nonvolatile fission 0.839 13.7 5.13 1.6h - products Plutonium 3,51 151. 2.9 16.0 - Transplutonic elements 1,12 18.3 3.33 9.67 - Total 5.54 183. 1.2 27.4 0.000318 IMFBR Core-Blanket Fuel - First-Year Dose Commitment Volatile fission products 0.00156 - 0.0123 - 0.872 Semivolatile fission 0,019, 0.169 6.98 0,0578 0,00621 products Nonvolatile fission 0.610 8.29 10.4 2.68 - products Plutonium 0.672 30.1 7.68 2.87 - Transplutonic elements 0.791 11.8 11.7 12.3 - Total 2.09 50.1 36.8 17.9 0.879 IMFBR Core-Blanket Fuel - Lifetime Dose Commitment Volatile fission products 0,00156 - 0.0123 - 0.873 Semivolatile fission 0.0L80 0.174 7.40 0.0604 0,00621 products Nonvolatile fission 1.01 15,5 10.9 3.1 - products Plutonium 18.6 806. 15.3 85.0 - Transplubonic elements 2.2h 34.6 12.0 22.1 - Total 21.9 856. 45.6 110. 0.879 a In rems. 8-78 - The input data for this calculation are presented in Table 8.1, and the results are given in Fig. 8.16. In some instances, iodine may be retained on a charcoal filter; in - such cases the thyroid dose shown in Fig. 8.16 would be reduced by a factor equal to one minus the filter efficiency. For very efficient filters, virtually all of the iodine would be retained; the whole-body dose would then be due only to the noble gases produced during the excur- sion and to those that result from the decay of the iodines trapped on the filter. The external gamma-ray dose delivered due to noble gases alone is shown in Fig, 8.17. " Validity of These Calculations. - The foregoing methods for estimat- ing the downwind radiation doses following a nuclear accident have been developed using the "Gaussian Plume” model.72’75 Implicit in this deriva- tion are the assumptions that the degree of atmospheric stability, the " wind speed, and the wind direction remain unchanged during the entire , course of the incident, Although the results have, in most cases, been extrapolated to a distance of 100 km from the stack, it is extremely doubtful whether this . model is valid for distances of more than 20 or 30 km. At a speed of 100 m/min, it would require 17 hr for the plume to extend for a distance of 100 km. However, it is almost certain that variations of the weather conditions, both with time and distance, would occur, Moreover, the model also assumes flat, featureless terrain and does not take into account the various topographical features such as hills, valleys, and lakes. No provision is made for the presence of buildings and other structures, which may affect the behavior of the effluent either because of proximity to the emitting source or because of modification of the behavior of the plume in the vicinity of the receptor. Items such as these must be handled on an individval basis, and, at present, there seems . to be no obvious way of generalizing the results of these effects. In all cases, it has been assumed that the release takes place at an - elevation of 100 m. This assumption produces somewhat lower ground-level concentrations than would a similar release that occurs at ground level. i 8-79 Table 8.1k, Source Terms for Criticality Accident (based on 3.7 x 1018 fissions) Isotope Yield x(sec_l) q (curies) 131; 0.029 9.96 x 107/ 2.9 1321 0.0LL 8.02 x 10> 352.9 1337 0.06% 9.25 x 100 60.1 134y 0.076 2,20 x 1071 1,672.0 135¢ 0.059 2.89 x 10™° 170.5 83my., 0.00L8 1.0L x 1074 L8.5 85my.. 0.015 L1 x 1075 66.2 8%y 0.027 1.18 x 107% 399.6 88y, 0.037 6.95 x 1072 2572 . 89k 0.046 3.63 x 1072 16,698.0 133my, 0.0016 3.49 x 1070 0.56 133%e 0.065 1.52 x 1070 9.9 135my 0,018 7.0 x 107 1,332.0 135¢e 0. 062 2,11 x 10™° 130.8 138y, 0.055 6.79 x 107k 3,734.5 88Rb Same ag 88Kr 257.°2 1T %%y Same as O%Kr 16,698.0 18805 Same as 13805 3,734.5 133Xe Same as 1331 60.1 TII 135mye 308 of 1351 51,2 1357 708 of 1357 119.3 8-80 ORNL DWG. 68-5845R2 "INVERSION" - "MOST REPRESENTATIVE" WHOLE BODY? »"~~~_ 7/ \\ THYROID o' ,’ \\‘ J B e /"INVERSION" AP ~£_ THYROID ,l\ \\ s N ~ 1 / 2 I ] 10 } ! ! REPRESENTATIVE" ) S = ,‘ WHOLE BODY 7 I s i / Sl,ofl— | ! - ! ] © ! ! o I ] " i ~ ; : g (] . M ) ® o wr £ e Ll DD o 07| o 08} |o'7 - L L 0 10 | 10 10 DISTANCE DOWNWIND { km) Fig. 8.16. Radiation Dose due to Volatile Fission Products Produced During a Nuclear Excursion (Based on 3.7 x 10 18 Fissions). vl 8-81 ORNL DWG 68-5844R1 to-! T T T T T T T T TT7 1072 J 10-3 - g “INVERSION" $ CONDITIONS -4 o - o 0 "MOST REPRESENTATIVE" CONDITIONS 1078 - n 10~© L . 0-7 vl | Ll L Ll il 1o~ | 10 102 DISTANCE DOWNWIND (km} Fig. 8.17. Whole-Body External Gamma-Ray Dose due to Noble Gases Produced During a Nuclear Excursion (Based on 3.7 x 10 18 Fissions). (Note: 1In this case it is assumed that all of the iodine isotopes are retained on filters.) 8-82 However, except for extremely stable meteorological conditions, the difference is quite small once the peak ground concentration produced by the elevated release has been passed. For example, under C-conditions, the ground concentration from a release taking place at an elevation of 100 m reaches 75% of that from a similar ground-level release at a distance of 1.2 km from the point of release, The wind speeds used in these calculations have been chosen to be 100 m/min and 50 m/min because this range of wind speed is reasonably characteristic of many locations. However, as can be seen from Eq. (16), the dose is inversely proportional to the wind speed. Finally, it should be pointed out that all of the doses calculated are those which are delivered at ground level on the plume center line. To obtain off-center-line ground-level doses, it is necessary to multiply the results by the quantity fiy2/2g2 e I where y is the distance (in meters) normal to the plume center line at which the dose is required, and oy (in meters) is the horizontal disper- sion parameter., For convenience, values of 7y and o, for the various stability conditions are shown in Figs. 8.18 and 8.19. Despite the obvious shortcomings of the procedures outlined, it is believed that they will, at least, produce order-of-magnitude results. These procedures will permit the development of sufficient "feel! for the magnitude of the various credible accidents so that the problem of siting can be approached in a quantitative manner, 8.3.4 Downwind Consequences of Upper Limit Accidents Upper limit accidents were determined using the assumed properties of fuel reprocessing plants (Sect. 8.3.1) and models and mechanisms described in Sect. 8.3.2 such that the release of noble gases, "fresh" fission products, iodine, semivolatile fission products, nonvolatile fission products, and plutonium is maximized, The computed fractional releases from the most significant accidents are summarized in Table 8.15. map: ol 8-83 ORNL—-DWG 63 —547R1 A~ EXTREMELY UNSTABLE B -~ MODERATELY UNSTABLE G — SLIGHTLY UNSTABLE D~ NEUTRAL E - SLIGHTLY STABLE F - MODERATELY STABLE Oy HORIZONTAL DISPERSION GOEFFICIENT (m) 2 5 10° 2 5 10? 2 5 10 x, DISTANCE FROM SOURCE (m) Fig. 8.18. Horizontal Dispersion Parameter as a Function of Distance Downwind. Fig. 8.19, Downwind, 3010° 0y , VERTIGAL DISPERSION COEFFICIENT (meters) ' 8-8L ORNL-DWG 63 -518 !O. A — EXTREMELY UNSTABLE B8 - MODERATELY UNSTABLE C — SLIGHTLY UNSTABLE D — NEUTRAL E — SLIGHTLY STABLE F — MODERATELY STABLE 10° 2 5 103 2 5 104 2 5 10° x. DISTANGE FROM SCQURGE (meters) Vertical Dispersion Parameter as a Function of Distance o 8-85 Table 8.15. Accidental Releases from Fuel Reprocessing Plants as a Function of Capacity Release (kg of Fuel® Unless Otherwise Indicated) from Plants of Capacity (metric tons/day) of: IWR Fuel Reprocessing Plant FBR Fuel Reprocessing Plant Accident 1 6 36 1 6 36 Nuclear Excursion in Head End Duration, min 78 140 140 110 140 160 No. of fissions 2.7x10°° 1.6 x 10 1.6 x 102 8.0 x 10°° 1.6 x 107 2.4 x 10% Noble gas, % 100 100 100 100 100 100 Todine, % 30 30 30 1.0 1.0 1.0 Volatile fission products 300 2000 2000 10 20 30 Semivolatile fission 1 6 6 1 2 3 products Nonvclatile fission products 0.00043 0.0046 0.0046 0.00082 0.0020 0.003% Transplutonic elements 0. 00043 0.0046 0,00L6 0.00082 0,0020 0.0035 Pu (head end) 0.00043 0.00L46 0.0046 0.00082 0.0020 0.0035 Pu (Pu storage tank)b 0.035 0.37 0.37 0.018 0. 04k 0.075 Noble-Gas Release 85%r 4 133¥e, curies 70,000 420,000 2,500,000 350, 000 2,100, 000 13,000,000 Halogen Release 131y 780 4700 1k, 000 7.8 L7 70 1297 36,000 220,000 660,000 360 2200 3300 Semivolatile Release Semivolatile fission 1 6 6 1 2 3 products Release of Nonvolatiles Semivolatile fission 0.00075 0.00L45 0.0045 0.0018 0.0037 0.0055 products Nonvolatile fission products 0.00075 0.0045 0.0045 0.0018 0. 0037 0.0055 Transplutonic elements 0.00075 0.,0045 0.0045 0.0018 0,0037 0.0055 Plutonium Release Plutonium 0,045 0.27 0,27 0.016 0,032 0,048 #The release of a component of the fuel is the product of these numbers and the concentration of that component in a kilogram of average LWR or IMFBR fuel, bThe nuclear excursion in the Pu storage tank 1s estimated to have the same yield and duration, but would release only "fresh" fission products and plutonium. 8-86 Site boundaries dictated by the upper limit accidents were estimated assuming that the maximum acceptable annual dose commitments resulting from exposure to the cloud or inhalation at the site boundary are those recommended by the NCRP for annual occupational exposure. These emergency dose commitments are compared with those of lOCFRlOO,3 an Isochem land requirements study,76 and an ORNL study involving peacetime applications of nuclear explosiveseo in Table 8.16. The assumed acceptable dose commit- ments have been employed only for reference purposes, but are believed to be reasonable in view of the very low probability of occurrence of the assumed upper limit accidents. The maximum site boundaries (Table 8.17) for all LWR plants and the l-metric ton/day FBR plant are determined by the whole-body dose resulting from the release of volatile "fresh" fission products from a nuclear excur- sion (30% and 1% release of iodines from LWR and FBR plants, respectively, plus 100% release of the noble gases). Site boundaries for the larger FBR plants are determined by the thyroid dose resulting from a silver tower explosion, which is assumed to release 0.1% of the equilibrium inventory of iodine. In Table 8.18, the total dose commitments resulting from various upper limit accidents at the accident-dictated site boundaries of these conceptual plants are compared with estimated dose commitments at the site boundaries of the NFS, MFRP, and BNFP plants, Noble Gases. - In plants that will partially remove the noble gases from off-gas streams, the upper limit accident involving these gases is considered to involve the complete release of the contents of a storage vessel that contains a 7-day accumulation of krypton and xenon. A release 85 a maximum (at 1400 m) downwind whole-body dose of 5 rems. This quantity of approximately 6,400,000 curies of ~“Kr plus l33Xe is required to cause represents the total accumulation of these gases over 890, 148, and 25 days in LWR plants with capacities of 1, 6, and 36 metric tons/day, respec- tively, and the total accumulation over 680, 115, and 3 days in FBR plants with capacities of 1, 6, and 36 metric tons/day respectively. The release of the 7-day accumlation of O2Kr and 13%Xe in a 36-metric ton/day FBR plant would result in a whole-body dose of greater than 5 rems within distances of about 2.3 km. 8-87 Table 8,16. Compariscn of Assumed Maximum Dose Commitments for Individuals in the General Population as a Result of Upper Limit Accidents with Those Given in 10CFR100, an Isochem Land Requirement Study, and a Study for Excavation of a Sea-Level Canal with Nuclear Explosives This Study 10CFR100 Isochem Study Nuclear Excavation Studyd Maximum Approximate Maximum Maximum Maximum Maximum Annual Dose Total Dose Total Dose Total Dose Annual Dose Total Dose Commitmeng Gommitment b Commi tment Commi tment Commitment Commitment (rems/year”) (rems/50 years ) (rems®) (rems/lifetime) (rems/year) (rems/70 years) Whole body 5 50 25 25 3 10 Red bone marrow 5 3 10 Head and trunk 5 (Gonads s 1o Lens of eyes o 8 15 Skin 30 300 15 30 Thyroid 30 30 300 300 15 30 Bone 30 500 300 15 30 Hands, forearms, 75 38 75 feet, and ankles Other single organs 15 90 (liver) 150 8 15 18 (lung) %These data are maximum permissible annual doses for occupational exposure as recommended by NCRP, bThese data represent the approximate 50-year dose commitment resulting from a single intake of mixed spent reactor fuel such that the maximum annual (first-year) dose commitments do not exceed those given in the first column. ®10CFR100 provides reference values of total whole-body and thyroid dose (incurred during passage of the radiocactive cloud) for use in the evaluation of reactor sites with respect to potentizl reactor accidents of exceedingly low proba- bility of occurrence and low risk of public exposure to radiation. dThese data are proposed maximum acceptable dose commitments for use in planning for the construction of a sea-level canal with nuclear explosives, They are considered applicable to special radiation protection problems in which an assessment of risk vs benefit would dictate greater annual dose commitments than those recommended by the ICRP, FRC, and IAEA. 8-88 Table 8.17. Site Boundaries (Distance from the 100-m Stack) Determined by the Maximum Upper Limit Accidents in a Spent-Fuel Processing Plant? Distance to Site Boundary (km) for Reprocessing Plants of Capacity (metric tons/day) of: IWR Fuel FBR Fuel Accident 1 6 36 1 6 36 Nuclear excursion 0.hh 2.0 2.0 1.2 2.0 2.8 Helease of': Noble gases (1.1)b (6.6)b (39)b (5-5)b (33)b 2.2 Halogens (0.2)° (2.0)P (2.9)° (23)P 1.0 2,3 Semivolatiles (l?)b 0.4 0.Ll (h7)b (O.93)b 1.0 Nonvolatiles (o.ol;)]D (0.23)b (0.23)b (0.27)° (0.55)° (0.81)P Plutonium (0.9)° (5.2)° (5.2)° (1.6)° (3.2)° (L..8)° %These boundaries are selected such that the maximum annual (first-year) dose commitment to the critical organ will not exceed that recommended by the NCRP for annual occupational exposure. bThe maximum acceptable dose commitment is not exceeded at any distance downwind, The numbers in parentheses are the maximum percentages of the maximum acceptable dose commitment, which occur 40O m downwind of the stack, Bt Table 8.18., Estimated Lifetime Dose Commitments to Critical Organs Resulting from Upper gimit Accidents at NFS, MFRP, BNFP, and Conceptual Plants for Processing LWR and FBR Fuels®? Birkn Conceptual LWR Plants of Capacity: Conceptual FBR Plants of Capacity: 1 Metric 6 Metric 36 Metric 1 Metric 6 Metric 346 Metric Type of Release NF3 MFRP BNFP Ton/Day Tons/Day Tons/Day Ton/Day Tons/Day Tons /day "Fresh" fission products Total number of fissions 10%° 1020 1018 2.7 x 10°° 1.6 x 10°% 1.6 x 10°% 8.0 x 10°° 1.6 x 101 2.} x 10°% Thyroid dose commitment, rems ~2 26 - 9.4 30 30 0.65 1.0 1.3 Whole-body dose commitment, rems 0.09 0.002° 5.0 T.0 T.0 5.0 5.0 5.0 Noble gases (BSKr and 133Xe) Release, curies - - - 70,000 420,000 2,500,000 350,000 2,100,000 13,000,000 Whole-body dose commitment, rems - - - 0,054 0.18 1.0 0.18 0.88 L.h Halogens (1311 and 12%1) Release, curies 1.7 1.2 1.1 3.1 28 55 1100 6500 9700 Thyroid dose commitment, rems - 0.017 - 0.05 0.2 0.5 L.6 22 27 Semivolatile fission products Release, curies - - 1900 760 4500 hs00 3600 7300 11,000 106Ru, curies - - 1500 410 2500 2500 1300 2600 3900 Lung dose commitment, rems - - ~0.0007¢ 2.7 8.9 8.9 5.0 7.9 13 Nonvolatile fission products and transplutonics Release, curies 1,1 5 120 3.3 20 20 37 4 111 hhce, curies - - 23 0.58 3.5 3.5 2.3 L.7 7.1 zh2Cm, curies 1.7 0,011 0,068 0.068 Lung dose commitment, rems - - <0.0007° 0.008 0.03 0.03 0.0L 0.06 0.07 Bone dose commitment, rems (~0.02) 0.075 - 0,024 (0.005) 0,077 (0,0L7) 0.077 (0.017) 0.060 (0.024) 0,10 {0.0Lk) o0.12 (0.05) Plutonium Release, alpha curies 0.65 <3 0.11 0.16 0.98 Q.98 9.30 0.61 0,91 Bone dose commitment, rems 13 <0,0007 6.7 (0.26) 22 (0.8) 22 (0.8) 8.6 (0.3) 14 (0.35) 18 (0.7) Distance to site boundary, km 1.5 0.6 2 0.4 2.0 2.0 1.2 2.0 2.8 68-8 #The underlined numbers are those that fix the radial distance to the site boundary. bThe numbers in parentheses are the first-year dose commitment for those cases in which the first-year dose commitment is not equal to the lifetime dose commitment. “The Allied Chemical Corporation reports the external exposure dose from beta and gamma radiation. 8-90 Fresh Fission Products. - Fresh fission products would be generated in a nuclear excursion., A nuclear excursion in a head-end vessel of maxi- mum capacity, resulting in complete boildown of the solution, is assumed. After boildown and dehydration, the reaction would terminate in the assumed vessels because of the low effective density of the fissile mater-~ ial (~3 g per cm3 of uranium plus plutonium in calcined solids is assumed). The thermal power of the nonvolatile fission products (the fresh fission product heat is significant for the first 1 to 2 hr following the excur- sion) would then calcine the solids; these solids would probably subse- quently melt through the vessel, flow onto the cell floor, and resolidify. It is assumed that the initial rupture breaks the off-gas line and that all of the steam generated in the boildown phase (containing all of the noble gases, 30% of the iodine, 20% of the semivolatile fission pro- ducts, and particulates of solution have the average concentration of nonvolatile fission products and plutonium) is discharged to the cell atmosphere and exhausted through the ventilation system. It is assumed that 99.5% of the semivolatile fission products are removed from the hot (air and saturated steam at ~100°C) ventilation stream by passage through metal mesh or silica gel absorbers. The ventilation systems of FBR plants are assumed to incorporate activated charcoal filters for removal of 9% of the iodine. The particulate release is calculated using the model presented in Sect. 8.3.2. The doses delivered by a nuclear excursion are dominated by the whole-body dose that results from exposure to the radiocactive cloud of fresh fission products (Fig. 8.17). Release of Iodine Inventory. - It is assumed that a fire or explosion in a solid halogen absorber would completely release the contained equi- librium concentration of 131I and a two-year accumulation of 1291. It is assumed that approximately 93% of the iodine collected by pretreatment in a wet scrubber is not dispersible. FBR plants are assumed to utilize charcoal filters that remove 99% of the remaining iodine, The thyroid dose which results from the explosion of a silver reactor is obtained by properly prorating the 1317 and 1297 doses at 0.4 km, as found in Tables 8.9 and 8.11, and applying the generalized dose curve (Fig. 8.15). e e 8-91 Release of Semivolatiles. ~ It is assumed that a total of 0.1% of the semivolatiles in the largest vessel is released by a mechanism other than a nuclear excursion (i.e., a tank boildown or an inadvertent addition of oxidants to a process vessel). The upper limit accident in a waste tank for interim (2-year) storage of mixed fission products would release a smaller amount of ruthenium by comparison, In evaluating the waste tank accident, it is assumed that coolant is lost from the tank and that the tank leaks, disdharging steam to the vault ventilation system and its condenser. The distillate, con- taining about 20% of the semivolatiles, is assumed to be returned to the tank, but an aerosol composed of particulates containing 204 of the con- centration of semivolatiles in the waste is discharged through the filters. The release from this source is insignificant (semivolatiles content, lOS rems at 0.l km downwind) have little meaning other than to show why such maxi- mum theoretical accidents must be rendered incredible through the use of appropriate engineered safety features. Another related type of accident in an acid waste tank, also consid- ered only of a theoretical nature since it depends on a very improbable combination of circumstances, involves simultaneous failure of the coolant for the coils of the tank and the off-gas condenser. In the event of such an accident, the contents of the tank would boll down on essentially the same time scale as that discussed previously. Because of the low heat capacity of the air and typical ventilation ducts, a mixture of air and saturated steam at approximately 100°C could pass through the off-gas and ventilation filters and be exhausted to the stack. Certain of the semi- volatile fission products (in particular, ruthenium tetroxide, which has a boiling point of ~80°C) may be carried by this stream. It is assumed (as in the case of the nuclear incident discussed previously) that 20% of the ruthenium is volatilized during the boildown phase and that, of this, 99.5% is removed by deposition on metal or on the filter. The remainder of the semivolatile fission products might be evolved in the calcination phase, but the off-gas line is assumed to cool following cessation of steam flow, permitting essentially complete (by comparison) removal of the semivolatile fission products by deposition and filtration. Mitigation of Accldents. - The authors stress that such accidents as the one denoted as '"maximum theoretical" may be converted to the tolerable category, in terms of consequences, by proper forethought and design. For 8-97 example, the effects of the postulated hydrogen-air explosion can be mitigated by one of the following (and possibly by others, limited only by the ingenuity of the designers): (1) Increase the reliability of preventive measures for control of the purge air flow and the hydrogen concentration, (2) Enclose the waste tarks within a building that is ventilated through a condenser and filter. (3) Design the tank and/or the vault to withstand a pressure of about 100 psig without rupture. () Decouple the tank from the vault. Use a pressure suppression and/or pressure relief system in the tank, Vent the vault to a containment system with large capacitance or to a pocl of water for steam suppression, (5) Use titanium tanks and self-boiling wastes to ensure effective purging of the hydrogen by steam. 8.3.6 Consequences of the Leakage of High-Level Wastes to the Ground Radiocactive waste solutions that are released by tank failure might be routed through the geologic formation lying between the tank site and the nearest surface drainageways. Since analyses must be made using specific site conditions, a hypothetical tank site at Oak Ridge was chosen for didactic purposes. This site was considered to be located in Conasauga shale on a promontory, with intermittent surface streams passing to the east, south, and west of the tank site. The shale formation is quite impermeable, and the movement of water is restricted so that it flows only along bedding planes. Samples of the Conasauga shale were obtained below the highly weathered zone in a direct path toward the surface streams., These samples were acid- ified for the removal of calcite, and the exchange capacities were determined by the calcium titration method of Jackson.77 A mean value of 11 + 1 meg/10C 3 at 85°C showed a hydrogen ion consumption of 260 meq/100 g, which would be sufficient to neutralize g was obtained., Overnight refluxing in 7 M HNO 8-98 the entire contents of an acid waste tark within a distance of 30 ft from the tank. In the case of acid waste, it was assumed that neutralization of the acid by calcite in the formation would result in a calcium salt system. In this system, strontium was assumed to compete with calcium without selectivity of either ion, alggough strontium might be slightly more selectively sorbed than calcium, For the sorption of strontium from neutralized wastes, and for the sorption of cesium and ruthenium, information on the sorption properties of Conasauga shale were obtained from previous laboratory studies.78-82 The quality of the groundwater was assumed to be similar to that of Clinch River water, which has a total cation (calcium and magnesium) concentration of about 0,002 meq/ml.83 Seepage rates were assumed to be characteristic of the area surrounding Waste Pit 2, where the average seepage rate from 1953 to 1958 was 3900 gal/day through an average side- wall area of 9000 ft2 (ref. 8lL). This corresponds to a mean superficial velocity of 0.06L ft/day. A mean groundwater velocity of 0.67 ft/day was used, which implies approximately 10% efficiency of contact between the shale and solution, If the initial seepage rate were maintained, the daily seepage rate from the acid waste tank (filled to a height of 35 ft with 106 gal of waste) would be 2275 gal. The seepage from the neutral- ized waste tank (filled to a height of 36 ft with 1.25 x 106 gal of waste) would be 2340 gal. Dispersion properties of solution in the formation (Fig. 8.21) were estimated from the results of a chloride tracer test conducted at the 8L, These data indicate an effective plate height of 16.5, according to the notation of G-lueckaui‘.85 site. Calculation of Radionuclide Movement. - In addition to the assumptions outlined above, it was further assumed that the waste would move longitudi- nally through a zone 75 ft wide, with a height equal to the original liquid level in the waste tank, to surface water at a distance of 200 ft., No allowance was made for lateral dispersion, but the spread of the solute was assumed to occur according to Glueckauf's model for the elution of a band of solute through a linear ion exchange column, The porosity of the e 8-99 ORNL-DWG 64-916 97 7 T 95 = ./ B O ’ E 90 — L | x | . N = d Z 80 Tt ¢ S / 5 ® © / g 60 T 7 T g / N / & 40 7 - 52 . ]! N o 1 - 9 20 /7 ) 5 i/ J £ 10 % t = 140 hr T ; = | T . Q 5 %// t' = 87 hr 1 | | 3 ® T / N = 4.3 - 8 2 - d = 200 ft - & | / | S / HETP = 46.5 ft [ 5 | | - S 0.5 / F ] | " 0.2 10 20 50 100 200 500 1000 TIME (hr) Fig. 8.21. Dispersion Properties of Chloride in Conasauga Shale at a Four-Acre Tank Site. 8-100 shale effectively contacted by solution was assumed to be 25%, with a grain density of 2.6L g/ml. If a leak were to develop in a waste tank, the amount of solution lost to the formation would be limited by the ability of the formation to accept the solution. During percolation of the waste solution, the groundwater concentration in the zone of migration would be increased, returning to normal when the waste solution was again displaced by the local ground- water, Movement and dispersion of the specific radionuclides were esti- mated by using Glueckauf's model in order to describe the dispersion of the unsorbed anions and correcting for retention of the radionuclides by the formation, as discussed by Inoue and Kaufm.an.86 However, due to the variable concentration of electrolyte in the groundwater, the retention factor was not constant with time. In addition, radiocactive decay was considered, The results of calculations for the movement of 9OSr from an acid 9OSr activity at the tank are shown in Fig. 8.22. The initial peak in surface drainageway occurs at about 1 year and is due to the relatively slight sorption of strontium by the shale in the presence of high concen- trations of electrolyte. With time, these high concentrations of salt are diluted and replaced by fresh groundwater, and a second concentration peak occurs after about 150 years. The relative magnitude of these two peaks depends on the total quantity of electrolyte released to the formation, If, after a leak occurs, the waste solution is pumped from the ground, the initial rapid movement will not be observed due to the removal of the excess electrolyte, Furthermore, in the case of 9OSr in an acid waste system, an appreciable fraction of the total radioactivity could be removed (Table 8.20). For neutralized waste, the precipitation of strontium, in addition to the increased probability for ion exchange, prevents 9OSr from attaining any significant concentration at the surface drainageways. The high affinity of the Conasauga shale for cesium deters movement of 13705 50 that radicactive decay occurs before significant concentrations would be observed in either acid or neutralized waste systems. The relatively rapid ORNL DWG 68-5839R1 10°2} 10~3 0%+ 10-S1 10-S 10~ 1078 10-9} io"ll lo7t2|- 'o‘l3 e FRACTION OF INITIAL ACTIVITY IN THE TANK AT THE TIME LEAK STARTS | ( LB LR 1 ¢ 1llr]TI 1 I Illllll T 1 rVrryrty | YEAR LEAKAGE 100 DAYS J | MONTH I T0T~8 1 1 lJllllI 1 1 l_LJllll C.l 1.0 0 TIME AFTER LEAK STARTS (yeors) Fig. 8.22., Strontium-90 Activity in the Groundwater at a Point 200 ft from a Leaking Tank of Acid Waste. 8-102 Table 8.20, Recovery of Radionuclides from the Soil After a Leak Has Developed in a Waste Tank Percentage Recoverable ILsotope Acid Waste Neutralized Waste P05y 88