'OAK RIDGE 'NATIONAL LABORATORY MARTIN MARIETTA | * OPERATED BY ~ MARTIN MARIETTA ENERGY SYSTEMS, INC. - FOR THE UNITED STATES ' DEPARTMENT OF ENERGY ORNL/TM-9756 o e N‘! TN oy MR LT P s Y s It x @ L COVER Extended Storage-in-Place of MSRE Fuel Salt and Flush Salt Karl J. Notz DIETRIRUTION OF THIS DOSUW ST IS UNLIKIT S Eooad el ) i o Lo I 3 TR E““i‘i OV te O Printed in the United States of America. Available from National Technical Information Service U.S. Department of Commerce 5285 Port Royal Road, Springfield, Virginia 22161 NTIS price codes—Printed Copy: A07; Microfiche AO1 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the U nited States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents thatits use would notinfringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereot. e ORNL/TM--9756 DE86 001720 Nuclear and Chemical Waste Programs EXTENDED STORAGE-IN-PLACE OF MSRE FUEL SALT AND FLUSH SALT Karl J. Notz Chemical Technology Division With contributions from: R. C. Ashline, Chemical Technology bivision D. W. Byerly, University of Tennessee L. R. Dole, Chemical Technology Division D. Macdonald, Martin Marietta Energy Systems, Inc., Engineering T. E. Myrick, Operations Division F. Jo. Peretz, Martin Marietta Energy Systems, Inc., Engineering L. P. Pugh, Operations Division A. C. Williamson, Martin Marietta Energy Systems, Inc., Engineering Date Published: September 1985 NOTICE This document contains information of a preliminary nature. It is subject to revision or correction and therefore does not represent a final report, Work Sponsored by U.S. Department of Energy Surplus Facilities Management Program OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37831 operated by MARTIN MARIETTA ENERGY SYSTEMS, INC. for the U.S. DEPARTMENT OF ENERGY under Contract No. DE-AC05-840R21400 & DISTRIZUTION OF THIS DOCIITIENMT IS UMLIRITED AB STRACT * * * * e . -* L] ® o * e 1. EXECUTIVE SUMMARY < o « « & l.1 HISTORY o o o o o o o &« 1.2 PROJECTIONS « ¢ ¢ o & & 1.3 CONCLUSIONS &« o « o o INTRODUCTION « o ¢ o o o o TORY e - e o . . . o * e DESCRIPTION o« o o ¢ o & PRIOR STUDIES « o ¢ o OPTIONS « o o o o o o & PROJECTIONS 4« o ¢ o o ¢ o 4,1 RADIOACTIVITY o ¢ o o 4.2 RADIOLYSIS o o s o o & 4,2.1 Pressure Rise . 4.2.2 Corrosion Rates 4.,2.3 Use of Getters . HIS 3.1 3.2 SURVEILLANCE AND MAINTENANCE 3.3 3.4 4.3 INTEGRITY OF THE FACILITY 4.3.1 Cell Penetrations 4,3.2 Water Control . » iii - * 4.3.3 Secondary Containment 4 GEOLOGY AND HYDROLOGY . o> ENTOMBMENT + ¢« o o « o 6 UTILIZATION o o o o o 7 FINAL DISPOSAL « o o » 4e7.1 WIPP v o o o o o 4,7.2 Commercial Spent 4.8 IN-CELL ALTERATIONS . 4,9 CONTINGENCY PLANNING . CONCLUSIONS o ¢ o o = o + @ 5.1 EVALUATION OF OPTIONS . 5.2 SELECTION OF PLAN « « « CONTENTS Fuel Repository 4.7.3 Greater Confinement Disposal . . Appendix Appendix Appendix Appendix Appendix Appendix A. B. C. D. E. F. BIBLIOGRAPHY AND REFERENCES o« ¢« ¢ ¢ ¢ ¢ o o o« & SEMIQUANTITATIVE EVALUATION OF SIX MSRE OPTIONS WIPP WASTE ACCEPTANCE CRITERIA .« & & 4 o ¢ o o ANALYSIS OF MSRE CELL PENETRATIONS . « «. . . . BUILDING 7503 DRAINAGE SYSTEMS « o ¢ o ¢ « & & GEOLOGIC INVESTIGATIONS RELATIVE TO MOLTEN SALT REACTOR DECOMMISSIONING e o o o« ¢ o ¢ o o o lo 17 18 27 27 45 46 50 52 54 55 60 62 62 63 64 65 66 67 6/ 68 71 71 73 75 79 85 93 111 125 EXTENDED STORAGE-IN-PLACE OF MSRE FUEL SALT AND FLUSH SALT Karl J. Notz ABSTRACT The solidified fuel salt and flush salt from the Molten Salt Reactor Experiment (MSRE) have been stored at the QOak Ridge National Laboratory (ORNL) since the reactor was shut down in 1969. The fluoride salt eutectic, containing 37 kg of uranium plus plutonium and fission products, is safely con- tained in three heavy-walled Hastelloy tanks, which are located inside a reinforced concrete cell. Removal of these salts to a remote location is not feasible until an appropriate repository has been identified, built, and placed in operation. Since this may take many years, extended storage—-in-place was criti- cally evaluated. The evaluation, which involved a preliminary assessment of several options for enhancing the integrity of in-place storage, including containment improvement, the addi- tion of chemical getters and neutron poisons, and entombment in concrete, showed that this approach was a rational and safe solution to the problem for the short term. Entombment is essentially nonreversible, but the other options are open- ended; they do not 1limit the future selection of a final dispo- sal option. Specific actions and improvements that would enhance safe containment during extended storage and would also be of future benefit, regardless of which disposal option is finally selected, were identified. 1. EXECUTIVE SUMMARY 1.1 HISTORY The Molten Salt Reactor Experiment (MSRE) was concluded in 1969 after several years of well-planned and highly successful work. This homogeneous reactor concept was based on the thoriunrU-233 fuel cycle and used a molten fluoride eutectic as the operating medium. This work is thoroughly documented. At shutdown, the fuel salt containing most of the uranium and fission products was divided and drained into two separate tanks, thus ensuring criticality safety. The flush salt, containing 1 to 2% of the uranium and fission products, was drained into a third tank. The salts were allowed to cool and freeze, thereby precluding any leakage and decreasing further the already-low corrosion rate. The drain tanks are made of heavy—-walled Hastelloy N, a special alloy created for the program, which has superior strength at high temperatures and outstanding corrosion resistance toward the eutectic fluoride system used for the MSRE. These tanks are contained within a hermetically sealed, stainless steel—lined, reinforced-concrete hot cell, located below grade except for a double set of roof plugs. A surveillance and monitoring program, which includes daily and monthly measurements, has been in force since shutdown. There is also an annual reheat (but not hot enough to remelt) to recombine any fluorine that might have formed from radiolysis by (a,n) reactions on the fluoride salt. There have been no adverse incidents or releases of radioactivity since the reactor was shut down 16 years ago. Several prior studies have been made of decontamination and decommissioning (D&D) of the facility, based on removal and reprocessing of the salts and on the assumption that a site or repository would be available to accept this material. In fact, there is no such site or repository at present. Nor is it feasible to reprocess the salts without major construction of such capability, whflch would be very costly and would introduce a finite probability of radiation exposure or release. The present study focused on extended storage and any enhancements to the storage mode that would benefit the storage period and also be beneficial in view of eventual final disposal. Time frames of 1 to 20 years (short term), 20 to 100 years (near term), 100 to 1000 years (midterm), and more than 1000 years (long term) were considered. 1.2 PROJECTIONS The most important of these is the radioactivity projection, which was carried out to 1 million years by using the ORIGEN2 code. (Prior projections were truncated at 5 and 20 years.) This projection showed two major aspects: decay of fission product (FP) activity, as antici- pated; and decay/ingrowth/decay of actinide activity, which is somewhat unusual and derives from the slow ingrowth of the U-233 decay chain. The FP activity has declined by a factor of 50 since discharge, will decline by another factor of 8 at 100 years after discharge, and will essentially disappear at 1000 years. The fission products are the major source of beta and gamma activity. The actinide activity will initially decay by a factor of 4 at 1000 years but will then grow back in to about its origi- nal level at 40,000 years before commencing final decay. The actinide behavior is the net result of U-232 (half-life of 72 years) and plutonium decay, along with U-233 (half-life of 158,000 years) ingrowth and then decay. Each actinide decay spawns an additional five or six alpha decays (along with some beta—gamma activity). The alpha activity is important for several reasons: it is a long—term source of neutrons from (a,n) reactions on Be-9, F-19, and Li-7, which are major components of the fluoride eutectic, and it contributes about 50 W of decay energy for a very long time. The neutrons must be considered in shielding calculations but do not pose an undue problem. The decay energy is, indirectly, a problem, not because of heat but because of radiolysis. Radiolysis of fluoride yields fluorine plus free metal. At slightly elevated temperatures (>150°F) recombination is rapid enough to preclude a buildup of fluorine; however, at lower temperatures, free fluorine will eventually form (after an incubation period of >5 years) and will then continue to be produced., Unless the radiolysis problem is brought under control, disposal of the salts in the fluoride form will present a signi- ficant problem. One possible way to limit the formation of free fluorine is by the addition of a getter - an active metal that reacts readily with fluorine. The physical integrity of the drain tanks and cell is of obvious importance. This factor was considered in terms of penetrations and control of water. No deficiencies that would jeopardize extended storage are evident at this time in either area, but some additional work in these areas would be beneficial. 1.3 CONCLUSIONS At this time, extended storage of the solidified fuel salts is the most prudent and rational course. Actions that can be easily taken out- of-cell to enhance storage should be implemented. An eventual decision to remove the fuel salts to a final disposal site will be tempered by site availability in 20 years or so. A future decision concerning whether to reprocess or not will be controlled by the ability to limit radiolysis, which is an open question at this time but must be resolved in the interim. 2. INTRODUCTION The MSRE was a graphite-moderated, homogeneous—fueled reactor built to investigate the practicality of the molten—salt reactor concept for application to central power stations. It was operated from June 1965 to December 1969 at a nominal full-power level of 8.0 MW. The circulating fuel solution was a eutectic mixture of lithium and beryllium fluorides containing uranium fluoride as the fuel and zirconium fluoride as a che- mical stabilizer. The initial fuel charge was highly enriched 235U, which was later replaced with a charge of 233y, Processing capabilities were included as part of the facility for on—line fuel additions, removal of impurities, and uranium recovery. A total of 105,737 MWh was accumu- lated in the two phases of operation. Following reactor shutdown, the fuel salt was drained into two critiéally safe storage tanks and isolated in a sealed hot cell, along with a third tank containing the flush salt. When the reactor was first shut down, the assumption was made that the facility and the fuel and flush salts would probably be utilized again at some later date. Therefore, the shutdown procedure was essen- tially a mothballing operation followed by surveillance and maintenance (S&M) procedures. These procedures were designed to ensure safe tem- porary storage and to maintain the operational capability of the facility. Some years later, it became evident that the facility would not be restarted, at least not as a molten—salt reactor. A decommissioning study was done in which two options were considered: (1) removal of the fuel and flush salts, followed by complete dismantling of the hot cells; and (2) removal of the fuel and flush salts, followed by entombment of the structural compoments within the reactor and fuel drain cells. The second option is far less costly than the first, but both are quite expensive because of the removal of the radioactive salts, which must be processed and repackaged. Obviously, both options require a repository which will accept the processed and repackaged salts. No such repository currently exists; nor is there any assurance that one will be available in the reasonably near future, even though there are several possibilities. Therefore, our task is to determine what steps should be taken to extend the storage period safely and what measures, if any, should be adopted to enhance the present storage conditions. In dealing with these questions, there are several time frames that represent various operational limits. These can be defined and described as 1. 3. 4. follows: Short term: 1 to 20 years. This is the period during which in=-cell operations can still be carried out with confidence. Final disposal options will be more clearly defined at the end of this period. During this time span, the fission product activities will decay to about two—-thirds of their present level. Near term: 20 to 100 years. During this period, institutional control can be maintained; therefore, it is the logical time to transfer the fuel (and flush) salts to the final repository. The fission product activities will decay to about 15% of their present level during this time span. Midterm: 100 to 1000 years. This is a reasonable lifetime for a man-made concrete structure. During this time, the fission product activities will decay to essentially zero, while the U-232 and plu- tonium activities will decay to about 3% of their present level; however, the U-233 activity will grow in to about 180% of its pre- sent level. Long term: 1000 to 1l million years. Geology will be the controlling factor for this period. During this time, only the U-233 decay chain will have any significance. Its activity will peak after 40,000 years at about 900% of the present level and will then decay permanently. 3. HISTORY 3.1 DESCRIPTION The primary reactor and drain system components are contained within two interconnected cells; the coolant and fuel processing systems are located separately within adjoining cells (Figs. 1 and 2). The reactor and drain tank cells are sealed pressure vessels that serve as secondary containment for the fuel. The reactor cell is a 24-ft-diam steel tank, while the drain tank cell is a stainless steel—lined reinforced concrete rectangular tank. Each cell has removable roof beams and shield blocks with a stainless steel membrane seal that must be cut open for access. The coolant system and fuel process systems are located within shielded cells that are kept at a slightly negative pressure and are swept by a containment ventilation system. Access is gained via removable top shield plugs. The associated equipment is housed in a steel—concrete— transite structure that has containment features. Both the containment cells and the high-bay area are maintained under negative pressure, with an active ventilation system consisting of centrifugal fans and roughing and HEPA filters that exhaust through a 100-ft steel discharge stack. The reactor heat dissipation system included a salt-to-air radiator exhausting through a steel stack and a drain tank for storage of the coolant salt, where this material now resides. This stored coolant salt is essentially nonradioactive. Ancillary facilities at the site include an office building, a diesel generator house, a utility building, a blower house, a cooling—water tower, and a vapor condensing system. These facilities have been described in detail in other reports.“’5’15’18 The three cells of most concern to this study (reactor, drain, and pro- cessing) are described in terms of penetrations and "containment enve- lopes” in Sect. 4.3 ffjor components of the material of construction, « A fuel-salt drain tank is shown in Fig. 3; the properties and the Hastelloy N (also c4lled INOR-8), are listed in Table 1,19 The presence o!&the solidified, stored fuel and flush salts is the most significant aspect of the MSRE. More than 4600 kg of fuel salt and 4300 kg of flush salt, containing about 37 kg of uranium (primarily U-233) ORNL OWG 63-4347 ¥ 2 3 4 5 6 T 8 - 9 E FUEL PUMP ' A INTENANGE OFFICE LUBE OIL 5YS BATTERY MAINTE : ROOM SHOP CHEMICAL SERVICE COOLANT PUMP L ARBORATORY TUNNEL UBE OIL SYSTEM D . TO FILTERS AND STACK gy Ty CELL VENTILATION 2 t AND BLOCK 1 VALVE C — fl_ o rrr——— R, . SPECIAL i JlTrans room EQUIP ROOM i fruustlfi 10 VAPOR STORA MAINTENANCE FQ STORAGE COND SYSTEM PRATICE CELL £OGLANT PUMP B COOLANT SALT IUID WASTE DRAIN TANK CELL GULATOR CELL A - FUEL SALT FUEL FLUSH FUEL DRAIN TANK NO 2 FUEL SALT ADDIT TANK ToR FUEL DRAIN THERMAL TFANK NO. ¥ SLOWERS SHIELD BLOWER *ELEC SERVICE AREA BELOW HOUSE ; RADIATOR HBLOWERS Fig. 1. Basement level floor plan of Building 7503. 01 ORNL DWG. 64-597 /—0 1, l ll_,l - 7 /] _ , E / STark 30-TON GRANE 3 AND 10-TON CRANES e === Pl waintenance ; | } 1 CONTROL | C01 aNT 1t ROOM | i {'saLT pump ! = — - — —J L B Lo F / REACTOR CELL be====—= _1| | S!‘%E“’ - : - : ocks ] b . FUEL SALT - ; : | A ADOITION . ANNLLUS L g / ' SHIELD BLOCKS STATION e _ N g | o 341 swiElp - o -1 ratecks ExSE:TrEd‘\,W \ I' HANG : Sy r—— LIQUID WASTE CELL ; DECONTAMINATION |- ] CELL T .- _ .. \RAD!ATOR - FUEL - i STORAGE .. nY. Ea | A TANK = 1 || RADIATOR b i . BYPASS DUCT FUELJ i i SN - PROCESSING b ¥ . CELL AL - ot CODLANT SALT Tl DRATN TANK = FUEL “REACTOR VESSEL FLUSH THERAMAL SHIELD FUEL TANK DRAIN _INE R e Lruee brain NO TANK NO 2 MSRE PLANT LAYOUT, ELEVATION Fig. 2. Elevations view of Building 7503. T 12 OQRNL-LR-DWG 61719 INSPECTION, SAMPLER, AND LEVEL PROBE ACGESS STEAM QUTLET STEAM DOME GONDENSATE RETURN WATER DOWNCOMER INLETS CORRUGATED FLEXIBLE HOSE STEAM RISER BAYONET SUPPORT PLATE BAYONET SUPPORT PLATE HANGER CABLE STRIP WOUND FLEX!BLE HOSE WATER DOWNCCMER ‘e-untlu [} GAS PRESSURIZATION AND VENT LINES T INSTRUMENT THIMBLE FUEL SaLl 57YSTEM FILL AND DRAIN LINE SUPPORT RING FUEL SALT DRAIN TANK BAYONET HEAT EXCHANGER THIMBLES (32) TANK FiLL LINE [} o o PN L & @ THIMBLE POSITIONING RINGS FUEL SALT SYSTEM FILL AND DRAIN LINE — TANK FiLL LINE MSR Primary Drain and Fill Tank Fig. 3. Fuel-salt drain tank. Table 1. 13 Composition and properties of INOR-8 (also known as Hastelloy N) Chemical compositon, 7 Ni Mo Cr Fe, max C Ti + Al, max S, max Mn, max Mn, max Si, max Cu, max B, max W, max P, max Co, max Physical properties: Density, 1b/in.3 Melting point, ° Thermal conductivity, Btu/(hefte«°F) at 1300°F Modulus of elasticity at ~1300°F, psi Specific heat, Btu/lb+°F at 1300°F Mean coefficient of thermal expansion, 70 to 1300°F range, in./in.*°F Mechanical properties: Maximum allowable stress,? psi 1000°F 1100°F 1200°F 1300°F 17,000 13,000 6,000 3,500 0.317 2470-2555 12.7 24,8 x 10 0.138 8.0 x 10°6. 8ASME Boiler and Pressure Vessel Code, Case 1315. 14 and 743 g of plutonium (primarily Pu-239) are present in the drain tanks. Calculated fission product activities (mainly beta—-gamma) of these salts, decayed to 1985, total about 32,000 Ci. The alpha activity from tran- suranic isotopes and their daughters amounts to about 2000 Ci. These isotopes are divided roughly 99%Z in the fuel salt and 1% in the flush salt. The total alpha activity of the fuel salt is very high, about 400,000 nCi/g, while that of the flush salt is about 6000 nCi/g. The total decay heat at present is about 200 W, with three—-fourths coming from the beta-gamma component and the remainder from the alpha emissions. The compositiogs of the fuel, flush,and coolant salts are given in Table 2, The latter salt, which served as a secondary coolant, is not radioac-— tive and is stored in the coolant cell. It is of interest primarily because it contains 338 kg of high-purity Li-7. As expected, the radiation hazards associated with the stored fuel are significant. Gamma and neutron dose rates within the reactor and storage cells are in the 103 rad/h range. Some of this radiation results from (a,n) reactions with the eutectic salt base. The high radiation field also causes the generation of some free fluorine, which slowly accumulates. Since recombination (with the metal simultaneously set free in the radiolysis) is accelerated by increased temperature, the salt is reheated periodically; however, it is not melted. The stored salts are in a stable, noncorrosive state, as dry frozen solids. In addition to the stored fuel, the principal areas of concern at the MSRE are the reactor components and process equipment remaining in the below-grade containment cells. These components are internally con- taminated and, in some cases, highly neutron activated. Exposure rates of up to 2200 R/h have been measured in the reactor vessel, attributed primarily to Co—-60. The inventory of residual radioactive materials in the reactor and fuel processing cells is estimated to be several thousand curies, with the majority of that activity being associated with fission and activation products. The remaining cells, process piping, and asso-— ciated operating areas are known to be slightly contaminated. The readily accessible areas of the reactor building (including the reactor bay and office areas) are generally uncontaminated and are being used for laboratory and office space, as well as for storage of various materials. 15 Table 2. Stored MSRE salts Fuel salt Flush salt?® Coolant salt? Total mass, kg 4650 4290 2010 Volume, ft3 at room 66.4 69.9 42.5 temperature Density, g/cm3 2.47 2.17 2.17 Composition, mol % LiF 6445 66D 66D BeF 30.3 34b 34b ZrFu 5.0 - - UF1+ 0.13 — —— Uranium content, kg U-232 c c U-235 0.85 0.09 - U-236 0.04 0.00 - U—238 20 01 0.].9 - Total 36.46 0.49 Plutonium content, g Pu-239 657 13 - Pu-240 69 2 —— Other Pu 2 0 - Total 728 15 —_ Lithium composition, 7% Li-6 - 0.009d 0.009 Li-7 - 99,9914 99.99] 4Trace—element analyses of 39 batches used for both salts gave 16 ppm Cr, 39 ppm Ni, and 121 ppm Fe. Twelve other analyses of the flush salt gave 38, 22, and 118 ppm, respectively. (Note: Could the Cr and Ni have been interchanged?) In another series of 22 batches, the corresponding values were 19, 25, and 166 ppm. bRreported values. Analytical data for batches 116—161 gave 63 and 37%, calculated from reported values of 12.95 wt % Li, 9.75 wt % Be, and 77.1 wt 4 F. For batches 101-130, the calculated composition was 64.3 and 35.7%. CPresent at 220 ppm, U-basis. dFor batches 116~142. The values are 0.0065/99.9935 for batches 143-161. 16 The MSRE facility appears to be structurally sound and capable of retaining the current radionuclide inventory. No significant spread of contamination or personnel exposure has occurred since facility shutdown. 3.2 SURVEILLANCE AND MAINTENANCE A comprehensive maintenance and surveillance program is provided to ensure adequate containment of the residual radioactivity at the MSRE. 11 Routine inspections of the containment systems and building services, radiological surveillance of operating areas and ventilation exhaust, stored~salt monitoring (temperature and pressure), and periodic testing of safety systems are performed as part of this program. In addition, the fuel and flush salt are reheated in order to allow recombination of fluorine,and the containment cells are subjected to a leak test, both on an annual basis. Facility maintenance includes general repairs, exhaust duct filter changes, and instrumentation and controls main- tenance. Consolidation of the surveillance instrumentation and periodic heater and controls tests are planned as major improvements to the current program. The salt storage cells have been under a planned program of regular surveillance and maintenance since the reactor was shut down in late 1969. This program includes daily observations of certain parameters by the Waste Control Operations Center and monthly observations by the Reactor Operations Group. The daily observations include measurements of internal cell temperature, building—air radioactivity, pressure differen- tials of the cell ventilation system, and stack off-gas radioactivity levels in terms of alpha, beta-gamma, and radioiodine. These obser- vations are recorded on log sheets (the logs have been kept since 1969). The radioactivity data are now also computerized and coordinated through the central Waste Control Operations Center. A primary input point to the Center is located in the control room in 7503. The monthly checklist involves a complete walk-through inspection of the entire facility plus recordings of in-cell temperatures and sump levels at seven locations. Any necessary maintenance items are noted and added to the list of work to be scheduled. (These logs have also been kept since 1969.) 17 On an annual basis, the three fuel and flush salt tanks are reheated to recombine any elemental fluorine that may have formed from radiolysis. The acceptable temperature range for the reheating is >300°F (to ensure that the diffusion rate is fast enough) but <500°F (well below the melting point of the salts). The reactor and drain tank cells are also leak-tested annually, and a check is made of about 40 equipment items. In addition, special checks are made of the ven- tilation system, two of the sump pumps, and the DOP efficiency of the ventilation filter. During the course of this study, it became apparent that data on building groundwater were needed. Therefore, two additional items were added to the S&M procedures: a periodic check of the radioactivity of the building sump discharge, and a periodic check of the operating fre- quency of the building sump pump. The radioactivity measurements, which were initiated in June, have consistently shown no activity above background. The sump pump monitoring was started on August 19, and daily readings show an operating cycle of about 1 h/d. Rainfall during the period was greater than normal. The pump capacity is about 35 gal/min. These measurements will be continued for at least 1 year in order to establish a data base that covers omne complete annual weather cycle. 3.3 PRIOR STUDIES Decommissioning of the MSRE presents some unique problems because of the presence of the fuel and flush salts. Plans for site decommissioning will first have to address the issue of disposition of the fuel. 1In the early studies, it was generally assumed that the stored fuel would be removed from the MSRE cells, with or without reprocessing (fluorination) to strip out the uranium and some of the fission products, and then sent to a final repository or to retrievable storage; however, these operations are complex and potentially hazardous and, therefore, expensive.13’15’1‘5,18 Thus far, no need for the recoverable U-233 or for the cell space has been established. Therefore, there has been no incentive from that direction to proceed with decommissioning. Further, there is no known site that will accept the fluoride fuel salt for disposal in its present condition, either now or in the near future. Consequently, it appears 18 that the solidified salts must continue to be stored in their present mode and location, regardless of future possibilities, needs, or deci- sions. Fortunately, this is possible because the salts, their contain- ment, and the facility are in excellent condition. However, it is essential to address two issues not directly considered in prior studies: l. Is it rational to continue storing these salts in their present situation and, if so, what guidelines should be followed to enhance this storage in terms of safety and stability for an extended period of time? 2. What other options exist for the future, and what steps should be taken in the near term to facilitate future decision—-making? This study gives an affirmative answer to the first question and identifies specific areas where improvements should be made. It also pinpoints needs that should be addressed in the short term to provide necessary information for future selection among available options. Finally, a preliminary work plan is outlined to achieve the above objectives. 3.4 OPTIONS Prior studies of MSRE decommissioning have focused on the final disposition of both the salts and the facility. Selection among options involved many possible decision points, including the following: A. Should the salt be chemically processed to separate the uranium (and some fission products) from the bulk of the fluoride salts? B. Should the fluoride salts be chemically converted to another form? C. Should the salts (with or without processing and/or conversion) be packaged and sent to an off-site repository or an on-site storage/disposal area, or should they be disposed of in the MSRE facility itself? D. Should the MSRE facility be totally or partially dismantled and portions of it entombed in concrete? At present, there is no firm basis for selecting among these options. Options A or B, if exercised, would require sophisticated chem- ical processing of highly radioactive materials; new flowsheets and 19 processing equipment would also be needed (and the equipment would ulti- mately have to be disposed of). The total cost of processing might be as high as $10 to $20 million. No choice can be exercised on option (C) at this time because there is no off-site or on—-site location that can accept this material at present, and we do not have an adequate basis at present to recommend permanent disposal in the MSRE facility itself. Option D could cost as much as $10 million and requires a prior decision on options A, B, and C. In short, we need to defer a final decision. Continued storage of the fuel and flush salts is forced upon us by external conditions. The issues of concern, then, revolve around three questions that need to be considered: l. Is is reasonable and safe to plan continued storage-in-place for up to 20 years or so? 2. Before deliberately embarking on such extended storage, what, if anything, should be done to enhance the storage environment? 3. During this storage period, or prior to it, what should be done to facilitate a future decision of a final nature? In order to gain a sense of perspective, it is instructive to define a range of options and systematically evaluate them against various criteria. The six options that have been treated this way as part of the evaluation are listed and defined in Table 3. (Option O, which is essentially to continue the present practice, is not acceptable for an extended period of time.) Options 1, 2, and 3 focus on the final decision and bracket a wide range of possibilities; this provides a broad, long~term perspective. Options 4, 5, and 6 focus on three variations of enhanced storage-in-place in order to bring short-term and near—-term needs into view. The steps used for assessing each option are as follows: l. List the process operations involved and compare them qualitati- vely. 2. Evaluate them semiquantitatively on the bases of (a) process readi- ness, (b) economics, (c) short—term hazard, (d) long-term hazard, and (e) conservation (i.e., beneficial utilization of the buildings and contained materials). 20 Table 3. Options for the Decontamination and Decommissioning of the MSRE Option Action to Be Taken 0 Continue as is, with no decision. Requires continued surveillance and maintenance, introduces maximum uncer- tainty, and eventually still needs a decision between the "real” options outlined below. I Complete dismantlement of hot cells following removal of fuel and flush salts. 2 Entombment of reactor and drain tank cells following remo- val of fuel and flush salts. 3 Entombment of reactor and drain tank cells, with fuel and flush salts in-place. 4 Enhanced near-term storage of solidified salts (which still leaves all other options available). 5 Enhanced storage in-place, including remelt and addition of getters (which still leaves optioms 1, 2, 3, and 6 available). 6 Enhanced storage in-place, including remelt, addition of getters, and repackaging (which still leaves options 1, 2, and 3 available). 21 3. Identify the limiting factor(s). Of necessity, these evaluations are subjective and only semiquantitative at this stage. The first evaluation (Table 4) lists the major process steps that would be required for the various options and indicates which steps are required in each case. The steps were defined in such a way that each represents a more—~or—less equivalent effort or cost. Evaluation then simply requires counting the number of steps for the particular option in question. Clearly, from this point of view, options 1 and 2 are pro- hibitively complex (and costly) and would not be selected unless there were an overriding reason for doing so. The second evaluation involves "scoring™ each option for every applicable process step, as identified in Table 4. These results are tabulated in Appendix B. Scoring, which was done on five different bases, was set up so that a low score is more favorable than a high score. The available scoring ranges were selected to balance the rela- tive importance of each of the five factors, on the assumption that near—term hazard, long—term hazard, and conservation were each about twice as important as either process readiness or economics. Scores for each criterion, as well as totals, are summarized in Table 5. The totals were converted to a more normal scale of O to 100, with a high score more favorable, by taking reciprocals of each total and then nor- malizing by dividing by the largest value (the 0.0161) and multiplying by 100. On this basis, Options 3, 4, and 5 are distinctly'superior to Options 1, 2, and 6. The third evaluation (Table 6) identifies limiting factors that would disqualify an option from further consideration at present. While this evaluation is the least quantitative, it is, nonetheless, highly significant because it seeks to identify controlling factors. The limiting factors are those which, if not within an acceptable range, would eliminate that option no matter how favorable the other aspects might be. Thus, in terms of the five criteria already established, Options 1 and 2 should not be considered further at this time. When two additional criteria were introduced, based on anticipated public reac- tion and on reversibility regarding future choice of options; Option 3 was also removed from consideration. 22 Table 4. Process steps required for MSRE options Option Process step 1 2 3 4 Melt salts (and add getters) Remove salts Build salt process facility Process or convert salts/U Package salts (and U) Dismantle/dispose of facility Transport packaged products PO XK R KR oKW | I Store/isolate products Decontaminate cells/equipment Dispose of liquid LIW | I | Remove equipment Dispose of solid LIW Dismantle cell structure Dispose of solid LLW/rubble PP PSP R K b e K R ) e e Restore area Seal pipes/penetrations Internal entombment i KoK e Ko s t { External structures Stabilize drain—tank cell - - - X Continue surveillance - - - X Number of more-or—-less equal process steps involved 15 11 3 3 Table 5. Summary of semiquantitative evaluation Option Factor 1 2 3 4 5 6 A. Process readiness 47 41 15 15 19 23 B. Economics 54 47 15 15 20 28 C. Near-term hazard 66 47 14 16 21 31 D. Long—~term hazard 4 6 12 17 12 10 E. Conservation 14 6 6 6 6 6 Total 185 147 62 69 /8 98 Reciprocal 0.0054 0.0068 0.0161 0.0145 0.0128 0.0102 Normalized 34 42 100 90 80 63 tc 24 Table 6. Limiting factors for MSRE options Factor 1 2 3 4 5 A, Process readiness Decision now? XXX XXX Decision in 5-20 years 777 7277 B. Economics XXX XXX C. Near—term hazard XAX XXX D. Long-term hazard Minimal for all options E. Conservation XXX F. Institutional perception of anticipated public reaction Decision now XXX Decision in 5-20 yearsP 777 G. Reversibility/future XXX XXX XXX choice of options AThere is no identified location at this time where the removed salts could be taken. bpublic reaction may become more rational with time, or institutional leaders may regain public confidence. 25 The results of all three evaluations are consolidated in Table 7. In summary, in terms of these three evaluations, and at this time: l. Options 1 and 2 are rejected on the basis of each evaluation. 2. Option 3 is rejected, based on limiting factors but recognizing that public reaction may change in future years, and that lack of reversibility may not be detrimental when judged against more definitive criteria. 3. Option 6 deserves further consideration based on the semiquan- titative evaluation. This leaves Options 4 and 5, and possibly Option 6, available for short-term selection. These options are addressed in Sects. 4 and 5. Some topics relevant to the other options are also addressed; however, much of the information presented is of broad and general applicability, and is germane to all options. 26 Table 7. Options and evaluations for decontamination and decommissioning sScored Option Description ab B¢ cd 0 Continue as is, with no decision. Requires con- tinued surveillance and maintenance, introduces maximum uncertainty, and eventually still needs a decision between the "real” options outlined below. 1 Removal of fuel and flush salts; complete dis- 15 185 5 mant iement of hot cells. 2 Removal of fuel and flush salts; entombment of 11 147 4 reactor and drain tank cells. 3 Entombment of drain tank cell, with fuel and 3 b2 2 f lush salts in-place. 4 Enhanced near-term storage of solidified salts 3 6Y U (which still leaves all other options available). 5 Enhanced storage in-place, including remelt and 4 78 v addition of getters (which still leaves Options 1, 2, 3, and 6 available). 6 Enhanced storage in-place, including remelt, 6 98 0 addition of getters, and repackaging (which still leaves Options 1, 2, and 3 available). 81n all cases, a low score is superior. bNumber of process steps involved. CSemiquantitative rating of process steps. dNumber of Limiting factors. 27 4. PROJECTIONS 4.1 RADIOACTIVITY The original decay calculations, made using ORIGEN and MSRE-specific S Subsequently, these data have input data, were truncated at 5 years. been extrapolated to the year 2000.1¢ However, any decisions regarding permanent emplacement require decay data out to a million years, as well as a detailed knowledge of radiological properties in the 10— to 10,000~ year time span. Ideally, the entire calculation (generation and deple- tion) should be repeated using ORIGENZ2, the updated and improved version of ORIGEN. However, ORIGEN2 does not have a molten salt reactor model, and construction of such a model would be prohibitively expensive. Therefore, the approach was to take the old ORIGEN output data at discharge and then make the decay éalculations with ORIGEN2. When these two sets of data at 5 years after discharge were cross—checked, only inconsequential or readily explainable differences were noted. This gives us confidence in using the ORIGEN-generated results as our starting point. The truncated ORIGEN data were presented in four groupings: fission products in both grams and curies, and actinides in both grams and curies. At least one activation product (Zr-93, from the fluoride salt mixture) was included in the old ORIGEN runs. The two sets of lists (grams and curies) were generally mutually exclusive; that is, isotopes listed on the grams output (because they were present in a quantity greater than 0.1% of the total mass) were generally not included on the curies output (where they appeared if contributing greater than 0.17% of the total radioactivity). By combining these two sets of lists, we believe that the input data for our ORIGEN2 calculations were acceptably complete. The summary results are given in Tables 8 through 11 (fission pro- ducts and actinides; grams and curies) for these times: RE SR SR SR Y Y ZR ZR ZR ZR ZR ZR ZR NB NB MO MO yO 87 88 89 90 89 91 90 91 92 93 Sy 95 96 93 95 95 97 98 MO100 TC aUu g° 99 RU101 RU102 RU103 RU104 RU106 RH103 PD105 PD106 PD107 AG107 SB125 TE125 TE127M TE128 TE1294 TE130 1127 1129 XEB129 Cs137 BA137 BA138 Lat139 CE140 CE1W CE142 CE144 PR141 ND143 ND144 ND145 Table 8. grams of fission products and daughters DISCHARGE 8.480E+00 5.640E+00 S.573E+0Q0 9.893E+01 6.660E+01 T.458E+00 4,440E+00 9.730EB+01 1. 100E+Q2 1.190E+02 1.1B0E+02 9.306E+00 1. 140 E+02 0.0 4,397E+00 2.040F+01 3.250E+01 3.160E+01 2.850B+01 2.980E+01 0.0 1.950E+01 1.560E+Q1 2.292E+00 7.880E+00 2.238E+00 5.07T0E+01 1.820E+01 4.610E+00 5.680E+00 0.0 6.176 E-01 0.0 4.175E-01 9.46 QF+00 8.859E-01 2.0908+01 0.0 1.490E+01 1.287E+02 5.560E+00 T.440E+01 1.700E+02 1.890E+02 1. 439E+01 1.730E+02 3.979E+(1 1.700E+02 1.670=2+402 1. 130E+02 1.070E+02 SUMMARY ORIGEN2 output for MSRE salts: TABLE: 28 CONCENTRATIONS, GRAMS ENTIRE CORE (MSRE WITH U-233 FUEL) 1.01IR 8.480E+00 S.640E+00 3.705E-02 9.660E+01 7. 214E+01 9.846E-02 6.768E+00 1.047E+02 1. 100E+02 1. 190E+02 1. 180B+02 1.779E-N 1. 140E+02 3.979E-06 2. 194E-01 3.772E+01 3.250E+01 3.160E+01 2.850E+01 2.980E+01 9.697E-05 1.950E+0% 1.560E+01 3.659E~-03 7.880E+00 1.125E+00 5. 299E+01 1.820E+01 5.723E+00 5.640E+00 6.018E~07 4.809E-01 1. 404E-01 4.092E-02 9.460E+00C 4.T31E-04 2.090E+01 3.889E-01 1.579E+01 6.972E-07 1.258E+02 8.5C0E+00 T.440E+01 1.700E+02 1.890E+02 5.972E-03 1.73CE+02 1.633E+0 1.844E+02 1.670E+02 1.365E+02 1.070E+02 10.01R 8.480E+0Q0 S.64 0E+00 9.390E-22 7.798E+01 7.217E+01 1.201E-18 2.580E+01 1.048E+02 1.100E+02 1.190E+Q2 1.180E+02 6.080E-17 1.140E+02 1.379E-04 T-416E~17 3.812E+01 3.250E+ 01 3.160E+01 2.850E+01 2.980E+01 9.687TE=-04 1 .950E+01 1.560E+01 2.357E-28 7.880E+00 2.309E-03 5.299E+01 1.820E+01 6.846E+00 5.640E+00 6.018E-06 5.057E-02 5.767E~01 3.4188-11 9.460E+0Q0 Y.67T3E-33 2.090E+01 4.299E~-01 1.579E+01 6.972E-06 1.021E+02 3.211E+01 T.440E+01 1.700E+02 1.890E+02 2.183E-33 1.730E+Q2 5.395E-03 1.8B44E+02 1.670E+02 1.528E+02 1.070E+02 100.0%R 8.480E+00 5.64 QE+0Q 0.0 S.154E+00 T.217E+01 0.0 9.424E+01 1.04 8E+02 1.100E+02 1.190E+02 1.180E+02 ¢.0 1.140E+02 4.393E-03 0.0 3.812E+01 3.250E+01 3.160E+01 2.850E+01 2-S7S5E+G? 9.696E-03 1.950E+01 1.56 0E+G1 0.0 7.880E+Q0 3.060E-3C 5.299E+01 1.820E+01 6.848E+00 5.64 0E+00 6.018E-05 8.368E-12 6.280E-01 0.0 9.450E+00 0.0 2.050E+01 4.299E-01 1.579E+01 6.972E-05 1.277E+01 1.215E+02 7.4 0E+01 1.700E+02 1.890E+02 0.0 1.730E+02 8.319E-38 1.844E+02 1.670E+02 1.528E+02 1.070E+02 1.0KY 8.480E+00 S.640E+00 0.0 4.552E-09 T.217E+401 0.0 1.034E+02 1.048E+02 1.100E+02 1. 189E+02 1« 180E+Q2 0.0 1.140E+02 5« 290E~=02 0.0 3.812E+01 3.250E+01 3.160E+01 2.850E+01 2-370E+0i 9.681E-02 1.350E+01 1.560E+01 0.0 7.88CE+0QQ 0.0 S5.295E+01 1.820E+01 6.848E+00 5.639E+00 6.018E-04 0.0 6.28B0E-01 0.0 9.460E+00 0.0 2.090E+01 4,299E-01 1.579E+01 6.972E-04 1.1888~-08 1.343E+02 7.440E+01 1.700E+02 1.890E+02 ¢.0 1.730E+02 0.0 1.844E+02 1.670E+02 1.528E+02 1.070E+02 10.0KY 8.480E+00 5.640E+00 0.0 0.0 7.21TE+01 0.0 1.034E+02 1.048E+02 1.100E+02 1. 185E+02 1.180E+02 0.0 1.140E+02 5.369E-01 0.0 3.812E+01 3.250E+01 3.160E+01 2.850E+01 2.8858+01 9.541E-01 1.950E+01 1.56Q0E+01 0.0 7.880E+00 G.0 5.299E+01 1.820E+C1 6.84BE+00 5.634E+00 6.0158-03 0.0 6.280E-01 0.0 9.460E+00 0.0 2.090E+01 4.299E-01 1.578E+01 6.371E-03 0.0 1.343E+02 T.440E+ 01 1.700E+02 1.890E+02 0.0 1.730E+02 0.0 1.844E+02 1.670E+02 1.5288+02 1.070E+02 100.0KY 8.480E+00 S.640E+00 0.0 0.0 7.217E+01 0.0 1.034E+02 1.04BE+02 1.100E+Q2 1.137E+02 1.180F+02 0.0 1.180E+02 5.270E+00 0.0 3.812E+01 3.250E+0Q1 3.160E+01 2.850E+N Z2e152E+01 8.27TE+Q0 1.950E+01% 1.560E+01 0.0 7.880E+00 0.0 5.299E+01 1.820E+01 6.848E+00 5.580E+00 5.986E-02 0.0 6.280E-01 0.0 9.460E+00 0.0 2.090E+0 4.299E-Q1 1.572E+01 6.9578=02 0.0 1.343E+02 7.440E+01 1.700E+02 1.890E+Q2 0.0 1.730E+02 0.0 1.844E+02 1.670E+02 1.528E+02 1.070E+02 1.0MY 802+00 UOE+CO 17E+0Q1 4 6 0 0 2 0 1.034E+02 1.048E+02 1.100E+02 7.564E+01 1.180E+0Q2 0.0 1.140E+02 4.335E+01 0.0 3.812E+01 3.250E+Q1 3.160E+01 2.850E+01 1.151E+00 2.865E+01 1.950E+01 1.560E+01 0.0 7.880E+00 0.0 5.299E+01 1.820E+0Q1 6.8482+00 5.069E+00 5.708E-01 0.0 6.280E~-01 0.0 9.460E+00 0.0 2.090E+01 4.295E-01 . 1.511E+01 6.821E-01 0.0 1.343E+02 7.440E+01 1.700E+02 1.850E+02 0.0 1.730E+02 0.0 1.844E+02 1.670E+02 1.528E+02 1.070E+02 ND146 ND14g ND15G PM147 Puidgy SM147 smMtus SM150 SM151 SM152 EU 151 EU152 EU153 EU154 EU155 GD154 GD155 SUMTOT TOTAL DISCHARGE 8.410E+01 4,630E+01 1.900E+01 4,.011E+01 4.912B-02 2.390E+01 0.0 2.710E+01 5.585E+00 1.460E+01 0.0 3.1108-02 5,450E+00 1.303E-01 7.629E=01 0.0 0.0 2.709E+03 2.7T09E+03 SUMMARY Table 8. TABLE: 29 (continued) CONCENTRATIONS, GRAMS ENTIRE CORE {MSRE WITH U-233 FUEL) 1.01IR 8.410E+01 4.630E+01 1.900E+01 3.080E+01 1.069E-04 3.321E+01 4,9018-02 2.T10E+01 5.542E+00 1. 460E+01 4.285E~02 2.955E-02 5.450E+00 1. 202E-01 6.634E-01 1.00698-02 9.951E-02 2.709E+03 2.709E+03 10.0YR 8.410E+01 4.630E+01 1.900E+01 2.856E+00 1.166E-28 6.115E+01 4$.912E-02 2.710E+ 01 5.171E+00 1. 46 1E+0Q1 4.140E-01 1.868E-02 5.450E+00 5.820E~02 1.886E-01 7.210E-02 5.743E-01 2.709E+03 2.709E+03 100.0YR 8.410E+01 4.630E+01 1.300E+01 T.345E-10 0.0 6.401E+01 4.912E-02 2.7T10E+01 2.585E+00 1. 46 2E+01 3.000E+00 1.90 3E-04 5.450E+00 4.118E-05 6.453E-07 1.303E-01 7.629E-01 2.709E+03 2.709E+03 1.0KY 8.410E+01 4,.630E+01 t-900E+01 0.0 0.0 6.401E+01 4,912e-02 2.710E+01 2.524E-03 1. 462E+01 5.582E+Q0 2.286E~-24 S.450E+00 1.294E-36 0.0 1.303E-01 7.629E-01 2.709E+03 2.709E+03 10.0RY 8.410E+O 4.630E+01 1.9CO0E+O1 0.0 0.0 6.401E+01 4.9128-02 2.710E+01 1.582g~-33 1.462E+01 5.585B+00 0.0 5.450BE+00 0.0 0.0 1.303E-01 7.6298-01 2.709E+03 2.TC9E+ Q3 100.0KY 8.410E+01 4.,630FE+01 1.9C0E+Q1 2.0 0.0 6.401E+01 4.912E-02 2.T10E+01 0.0 1.462E+01 5.585E+00 0.0 5.450E+00 0.0 0.0 1.303E-01 7.629E~01 2.709E+03 2.709E+03 1.0H4Y 8.4 10E+01 4.530E+01 1.9G0E+01 .0 01E+0Q1 12E-02 10E+01 62F+01 85E+00 0 0 6 4 2 o 1 5 0 S5.450E+00Q 0 0 QOLFCUVMEONWYEFEC 1.303E-01 7.629E-01 2.709E+03 2.709E+03 SR Sk Y Y ZR ZR NB NE NE TC 89 90 90 91 93 95 934 95 95¥ 9s RU103 RU106 RH103X RH106 PD107 SN123M SB125 TE125H4 TE127 TE127M TE129 TE129HM 1129 CS5134 €5137 BA137H CETul CE142 CE1u4 PR144 PR1GYN PMI4T PM148H SM151 E0152 EU154 EU1SS SUMTOT TOTAL Table 9. DISCHARGE 1.620E +05 1.350E+04 1.360E+04 1.830E+05 2.991E-01 2.000E+05 0.0 1.720E+05 4.149E+03 5.054E-01 7.400E+04 7-.491E+03 7.401E+04 8.949E+03 2.902E-03 1.390E+0Q2 6.380E+02 1.870E+02 3.2792+04 3.940E+03 9.790E+04 2.670E+04 2.631E-03 5.519E+00 1.120E+04 1.050E+04 4.101E+05 4.153E~-06 1.270E+05 1.280E+405 0.0 3.720E+04 1.050E+03 1.470E+02 5.380E+00 3.519E+01% 3.550E+(2 1.800E+06 1. 800E+Q6 SUMMARY 30 ORIGEN2 output for MSRE salts: curies of fission products and daughters TABLE: RADIQCACTIVITY, CURIES ENTIRE CORE (MSRE WITH U-233 FUEL} 1. 01IR 1.077E+03 1.318E+04 1. 319E+04 2-416E+03 2-991E-01 3.824E+03 1. 412E-02 8.584E+03 2.837E+01 5.054E-01 1.181E+02 3.766E+03 1.065E+02 3.766E+03 2.902E-03 0.0 4.967E+02 1.216E+02 3.782E+02 3.862E+02 9.282E+00 1.426E+01 2.789E-03 3.944E+00 1.094E+04 1.035E+04 1. 702E+02 4.153E-06 5.211E+04 5.211E+04 6.254E+02 2.856E+04 2.286E+00 1. 459E+02 5. 113E+00 3.246E+01 3.087E+02 2.068E+405 2.068E+05 10.01R 2.729E-17 1.064E+0U4 1.064E+004 2.9U8E-14 2.991E-01 1.307E-12 1.135E~-01 2.901E-12 9.694E-15 5.054E-01 7.611E-24 7.730E+00 6.861E-24 7.730E+00 2.902E-C3 0.0 5.224E+01 1.275E+01 3.160E=-07 J.226E-07 3.281E-29 5.0418-29 2.789E=-03 1.9148-01 8.889E+(03 8.409E+03 6.220E-29 4.1538-06 1.722E+01 1.722E+G1 2.066E-01 2-649E+03 2.492E-24 1.361E+02 3.232E+00 1.572E+01 8.773E+071 4.159E+04 4.159E+04 100.0¥R 0.0 1.209E+03 1.269E+03 ¢.0 2.99 1E-01 0.0 2.824E=-01 52E-01 OO OO0 20E-26 2U4E-26 02E-03 OO0 a0 OOVLDLHOD 48 E-09 09E-09 789E~03 1.389E-14 1. 111E+03 1.051E+03 0.0 4.153E-06 2.655E-34 2.6558-34 3.186E-36 1.247E-07 0.0 6.804E+01 3.292E-02 1. 11 2E-02 3.002E-04 4.730E+03 4.730E+03 1.0KY 0.0 6.212E-07 6.214E-07 0.0 2.990E-01 89E-03 34E-06 83E-07 43E-02 55E=-22 95E-34 = OEF PR CUUa4O]OO0ONOoOUOOOOORUOOOOOOOPO 1.159E+00 10.0KY 1.076E+00 100.0KY 0 0 0 0 859E-01 16E-01 50E-M T1E-03 T6E-03 0. 0. 0. OO 2. 0. 2. 0. 0. 3. OQ 0. 0. 0. 2. 0. 00 0. G. 0. 0. 0. 2. 0. 0. 0. 0. 4.153E~-06 0. 0. 0. 0. Q. 0. 0. 0. 0. 9. 0 2 0 0 6 ¢ 0 0 0 8 0 0 0 0 0 0 0 7 0 0 0 0 1 0 0 0 0 0 0 0 0 0 2 B1E-01 9.281E-01 1.0MY 01E-01 06E-01 52E-02 08E-03 6BE-03 53E-06 WOOODOOOODO0ODOFOOOONOUCOOOONCOODOOa OO0 aOmOco0O 8 & & 2 4 r B R 84BN T F MY AR NN A s s A Y Y R WOOOCOCOO0OO0OOOMO00CO0OOUOOOOO 000 QUOoOOEEQQUoOeO 56E-01 3.956E-01 HE 4 PB206 PB207 PB208 BIZ209 EA226 TH229 TH230 TH232 0232 0233 0234 U235 U236 7238 NP237 PU239 PU240 pPu2u41 AM241 SUMTOT TOTAL Table 10. DISCHARGE OO0 O0OOOCO - - - - - - - - - - 0 0 0 0 0 0 0 0 0 7 846E+00 3.232E+04 2.911E+03 9.860E+02 6.800E+01 2.370E+03 0.0 6.223E+02 7.501E+01 8.T42E+Q0 2.793E-01 3.937E+04 3.937E+04 SUMMARY 31 ORIGEN2 output for MSRE salts: grams of actinides and daughters TABLE: CONCENTRATIONS, GRAMS ENTIRE CORE (MSRE WITH U-233 FUEL) 1.0YR 1. 0u45E-02 9.789E-15 8.4 14E-14 6.277E-02 5. 396E-06 3.587E-08 1.389E~-01 8.111E-03 1. 979E-086 7. T7T1E+00 3.232E+04 2.9%1E+03 9.860E+02 6.801E+01 2.370E+03 7.761E-04 6.223E+02 7.500E+01 8.331E+00 6.894E-01 3.937E+04 3.937E+04 10.0YR 1.051E-01 3.171E-10 8.777E-11 6.435E-01 5.926E-04 3.5828~06 1.388E+00 8.111E-02 1.980E-05 T.1268+00 3.232E+04 2.911E+03 Q.862E+02 6.808E+01 2.370E+03 3.273E-02 6.221E+02 TW493E+ 01 5.402E+00 3.586E+00 3.937E+04 3.937E+04 100.0YR 8 .354E-01 2.170E-06 5.077E-08 4.437E+00 5.9 1E-02 3.535E-04 1.382E+01 8.107E~-01 1.990E-04 2.996E+00 3.231E+04 2.910E+03 9.878E+02 6.878BE+01 2.370E+03 1.065E+00 6.205E+02 TLU22B+00 7.096 E-02 7.868E+00 3.937E+04 3.937E+04 1.0KY 4.4587E+00 4.088E-03 8.727E-06 7.207E+00 5. 7948400 3.108E-02 1.323E+402 8.064E+00 2.089E-03 5.161E-04 3.218E+04 2.903E+03 1.003E+03 7.542E+01 2.370E+03 7.028E+00 6.046E+02 6.T46E+0 1 1.085E-20 1.875E+00 3.337E+04 3.937E+04 10.0KY T.195E+01 2.036E+00 8.893E-04 T.207TE+00 4.4C5E+02 1.217E+00 8.765E+02 7.647E+01 2.802E-02 1.207E-41 3.094E+04 2.830E+03 1.139E+03 1.162E+02 2.3708+03 8.847E+00 4,.666E+02 2.598E+01 0.0 1.015E~06 3.937B+04 3.937E+04 100.0KY 1.142E+03 2.208E+02 6.807E-02 7.207E+00 9.361E+03 9.393E+00 9.958R+02 4,560E+02 3.917E-01 0.0 2.08B7E+04 2.193E+03 1.56 3E+03 1.414E+Q2 2.370E+03 8.593E+00 3.492E+01 1.866E~03 0.0 0.0 3.937E+04 3.938E+Qu 1.0MY 3.6 19E+03 2.352E+03 1.303E+00 7.207E+00 2.861E+04 1.576E+00 1.947E+01 7.7T12EB+01 4.045E+00 0.0 4.081E+02 1.711E+02 1.596E+03 1.376E+02 2.370E+03 6.4 20E+00 1.525E-10 0.0 0.0 0.0 3.938E+04 3.938E+C4 TL207 TL208 TL209 PB209 PB210 PB211 PB212 PB214 BI210 BI211 BI212 BI213 BI214 P0O210 PO212 Po213 PO214 P0O215 PG216 Pc218 AT217 RN219 RN220 RN222 FR221 RA223 RA224 RA225 RA226 AC225 aca27 TH227 TH228 TH229 TH230 TH231 TH234 PA231 Pa233 PA234H 0232 0233 U234 0235 u23e U238 NP237 PE238 PU239 PU240 PO241 Am24n cnzu2 SUMTOT TOTAL Table 11. curies of actinides and daughters DISCHARGE S1E+01 302402 30E+02 CUVOoOCOODUNOOOCOFO BOE+01 30E+02 30E+Q2 30E+02 30E+02 a2 @ 4 & % ¥ B & & 5+ & 3 B % M B s & & B s ¥ s e % s s e OO0 U0 R0O0 a0V a00OLOO0OWNO OO0 COCUNOOLOONOCODOCUooCOoOUNOoOocOoO~NOoOo 1.680E+02 3.130E+02 1.820E+01 2.132E-03 4.401E-03 7.971E-04 0.0 1.000E+00 3.8702+01 1.710E+01 9.0102+02 9.590E~ Q1 2.450E+01 2.553E+03 2.553E+03 SUMMARY ORIGEN2 output for MSRE salts: TABLE: ENTIRE CORE 1.0YR 7.084E~10 5.672E+01 6.385E~04 2.956E-Q2 3.591E-10 7. 104E-10 1.579E+02 3.547E-08 3.591E-10 7. 104E-10 1.579E+02 2.956E-02 3.547E-08 1. 164E-10 1.011E+02 2. 892E~-02 3.546E~-08 7.104E-10 1.579E+402 3.548E-08 2.956E-02 7.104E-10 1.579E+02 3.548E-08 2.956E-02 7. 104E-1¢C 1.579E+02 2.956E- 02 3.548E~-08 2.956E-02 7-104E-10 7.006E-190 1.573E+02 2.956E-02 1. 638E-04 2. 132E-03 7.97T1E~04 4.511E-08 5.473E-07 7.371E-04 1.664E+02 3.130E+02 1.820E+01 2. 132E-03 4.402E~03 7.971E-04 S5.473E-07 1.090E+00 3.870E+01 1. 710E+01 8.586E+02 2.367E+00 5. 193E+00 2.525E+03 2.525E+03 10.0YR 6.46TE-08 S.618E+01 6.382E-03 2.955E-01 3.395E~07 6.486E-08 1.564E+02 3.542E-06 3.396E-07 6.486E-08 1.564E+02 2.955E-01 3.542E-06 3.396E~07 1.002E+02 2.891E-01 3.542E-06 6.486F=08 1.564E+02 3.543E-06 2.955E-01 6.486F~08 1.564E+02 3.543E-06 2.955E-01 6.486F-08 1.564E+02 2..955E-01 3.543E-06 2.955E-01 6.U82E~08 6.396E-08 1.562E+02 2.955E~-01 1.638E-03 2.133E-03 T7.971E-04 4.5148-07 2.3088-05 7.971E-04 1.526E+02 3.130F+Q2 1.820E+01 2.1338-03 4.406E-03 7.971E-04 2.308E-05 1.040E+00 3.869E+01 1.708E+01 5S.567TE+02 1.231E+01 4.511E-06 2.206E+03 2.206E+03 32 RADIOACTIVITY, CURIES (MSRE WITH U-233 FUEL) 100.0%R 3.144E~-06 2 .36 8E+01 6 .354E-02 2.942E+00 1.94 1E-04 3.153E-06 6 .590E+01 3.495E-0Q4 1.94 1E-04 3.153E-06 6.590E+01 2.942E+00 3 .495E-04 1.941E-04 4,222E+01 2.878E+00 3.495E-04 3.153E-06 6.590E+01 3.496E-04 2.94 2E+00 3.153E-06 6.530E+01 3.496FE-04 2.94 2B+00 3.153E~06 6 .590E+01 2.942E+00 3.496E-04 2.94 2B+ 00 3.153E-06 3.110E-06 6.590E+0Q1 2.9 2E+00 1.637E-02 2.136E~-03 T.97T1E-04 4.511E-06 7.508E-04 7.97T1E-04 6.415E+01 3.129E+02 1.819E+01 2.136E-03 4.452E-03 7.971E-04 7.508E-04 5.108E-01 3.859E+01 1.692E+ 01 7-313E+00 2.701E+01 0.0 9.704E+02 9.704E+02 1.0KY 4.492E-05 4,.086E-03 6.07T9E-01 2.815E+01 3.073E-02 4,504E-05 1.137E-02 3.074E-02 3.073E~02 4.504E-05 1.137E=-02 2.815E+01 3.074E~-02 3.073E=-02 7.286E-03 2-754E+01 3.073E-02 4.504RE-05 1.1378-02 3.074E-02 2.8 15E+01 4.504E-05 1.137E-02 3.074E-02 2.815E+01 4,5048-05 1.137E-02 2.8B15E+01 3.074E-02 2.8 15E+01 4.504E-05 4.442E=-05 1.137E~-02 2.815E+01 1.628E~01 2.170E-03 7.971E-04 4,503E-05 4 .956E-03 7.971E~Q4 1.105E=02 3.116E+02 1.815E+01 2.170E-03 4.881E-03 7.971E=-04 4.956E-03 4.173E-04 3.760E+01 1.538E+01 1.118%-18 6.439E+00 0.0 6.1U9E+02 6.149E+02 10.0KY 4,396E-04 1.105E-09 4.029E+00Q 1.865E+02 1.204E+0C 4.4 Q9E-04 3.074E-09 1.204E+00 1.204E+00 4.4 09E-04 3.074E-09 1.865E+02 1.204E+00 1.2C4E+00 1.970E-09 1.825E+02 t.2C4E+0Q0 4.409E-04 3.074E-09 1.204E+G0 1.865E+02 4.409E~04 3.074E-09 1.204E+00 1.865E+02 4.409B-0U 3.074E-09 1.865E+02 1.204E+00 1.865E+02 4.409E-04 4.348E-04 3.074E-09 1.865E+02 1.544E+ 00 2.463E-03 7.971E-04 4.407E-04 6.239E-03 7.9712~-04 2.584E-40 2.996E+02 1.769E+01 2.463E-03 7.520E-03 7.971E-04 6.239E-03 5.541E-35 2.901E+01 5.922E+00 0.0 3.487=~06 0.0 1.857E+03 1.857E+03 100.0KY 2.795E-03 1.544E-08 4.577E+00 2.119E+02 9.287E+00 2.803E-03 4% .298E-08 9.289E+00 9.287E+0Q0 2.803E-03 4.298E-08 2.119E+02 9.289E+00 9.287E+00 2.754E-08 2.073E+02 9.287E+00 2.803E-03 4.298E-08 9.291E+00 2.119E+02 2.803E-03 4.298E-08 9.291E+00 2.119E+02 2.803E-03 4.298E-08 2.1T19E+02 9.291E+00 2.119E+02 2.803E~-03 2.764E-03 4.298E-08 2.119E+02 9.208E+00 3.3818-03 7.971E-04 2.803E-03 6.060E~03 7.971E-04 0.0 2.021E+02 1.371E+01 3.381E-03 9.1498-03 7.971E-04 6.060E-03 0.0 2.171E+00 4.25UE-Q4 0.0 0.0 ¢g.0 2.006E+03 2.006E+03 1.0MY 3.442E-03 1.595E-07 8.9518-02 $.144E+00 1.558E+00 3.452E-03 4,%938E-07 1.558E+00 1.558E+00 3.452E-03 4,.438E-07 4.144E+00 1.558E+00 1.558E+00 2.844E-07 4.055E+0Q0 1.558E+00 3.452E-03 4.438E-07 1.559E+0Q 4.144E+CO 3.4528-03 4.438E-07 1.559E+00 4 . 144E+Q0 3.4522~-03 4.438E-07 4.184E+00 1.559E+00 4.144E+00 3.452E~03 3.404E-03 4.438E-07 4,144E+Q0 1.557E+00 3.452B-03 7.970E-04 3.452E-03 4.,527E-03 T.970E-04 0.0 3.9528+400 1.069E+00 3.452E-03 8.909E-03 7.970E-04 4,527E-03 0.0 S7E-11 8E+01 5.382E+01 33 Discharge (these are the input data, as taken from ORIGEN) 1 year 10 years 100 years 1000 years 10,000 years 100,000 years 1 million years Data were also calculated for those times used in the original ORIGEN rums: 384 d (1 year and 19 d, corresponding to January 1, 1971; used in the old ORIGEN calculations) 749 d (2 years and 19 d) 1115 d (3 years and 19 d) 1480 d (4 years and 19 d) 1845 d¢ (5 years and 19 d) Agreement between the two outputs was checked at the 1845-d output and found to be excellent, with minor exceptions. A detailed listing is given in Tables 12 through 15. In summary, the exceptions are as follows: Fission products (in grams): Of a total of 45 isotopes, 31 agreed exactly and the other 14 differed by no more than 47%. The total FP mass was 2710 g in both cases. Fission products (in curies): Of a total of 29 isotopes, 7 agreed exactly, 6 differed by minor amounts (1 to 4%), 3 differed signifi- cantly (17% less, 50% less, and 2507 more), and 13 differed by large factors (20 to 800). All of the large differences were in minor isotopes that contributed less than 0.l1% of the total activity, which agreed very well (5.70 x 10% vs 5.71 x 10% Ci). The causes of these discrepancies are twofold: revised half-lives, and a numeri- cal error in ORIGEN for the very low activities., Actinides (in grams): Excellent agreement for all nine isotopes; two agreed exactly, five were within 1%, one minor constituent (Pu-241) was within 14%, and the major contributor (U-233) was within 2%Z. The difference for U-233 was traced to a revision in the 34 Table 12. Reconciliation of ORIGEN/ORIGEN2 Results I: Actinides and Daughters, in Curiles Curies after 5 yr Isotope Half-life ORIGEN ORIGEN2 Comments In secular equilibrium with U-232¢ In secular equilibrium with U-232 In secular equilibrium with U-232 secular equilibrium with U-232°€ T1-208 3.05 min 5.82 El Pb=-212 10.6 hr 1.62 E2 Bi-212 60.6 min 1.62 E2 Po-212 0.3 usec 1.03 E2 1.04 E2 R —t o Po-216 0.15 sec 1.62 E2 a In secular equilibrium with U-232 Rn-220 55.6 sec 1.62 E2 a In secular equilibrium with U-232 Ra-224 3.64 d 1.62 E2 a In secular equilibrium with U-232 Th-228 1.91 yr 1.62 E2 1l.61 E2 In secular equilibrium with U-232 U=-232 72 yr 1.60 E2 a 7-233 158 E3 yr 3.13 E2 a U~234 245 E3 yr 1.82 E1 a Pu-238 88 yr 1.09 EO 1.08 EQ Pu-239 24.1 E3 yr 3.87 El a Pu-2450 6540 yr 1.71 El a Pu-241 l4.4 yr 6.88 E2 7.07 E2 ORIGEN used half-life of 13.0 yr Am-241 432 yr 6.97 EO 7.40 E0 Daughter of Pu-241; ORIGEN half-life was 470 yr Cm—242 163 d 3.81 E-2 0,97 E-2 Assumed ORIGIN included a precursord Subtotal 2.37 E3 2.40 E3 Difference due mainly to Pu-24] Total 2.38 E3P 2.40 E3 Difference due mainly to Pu-241 4Same wvalue as ORIGEN. bThis total includes some minor contributors not included in the subtotal. CBranching decay product of Bi-212. dAm-242 m (152 yr); 0.04 Ci at discharge would explain the difference. 35 Table 13. Reconciliation of ORIGEN/ORIGEN2 Results I1: Actinides and Daughters, in Grams Grams after 5 yr Isotope Half-l1ife ORIGEN ORIGEN2 Comments U-232 72.0 yr 7.49 EQ 7.47 EQ U-233 158 E3 yr 3.30 E4 3.23 E4 ORIGEN used half~life of 162 E3 yr U-234 245 E3 yr 2.94 E3 2.91 E3 ORIGEN used half-life of 247 E3 yr U-235 704 E6 yr 9.86 E2 a ORIGEN used half-life of 711 £6 yr U~236 23.4 E6 yr- 6.80 E| 6.81 El ORIGEN used half-life of 23.9 E6 yr U-238 4,47 E9 yr 2.37 E3 a ORIGEN used half-life of 4.51 E9 yr Pu-239 24.1 E3 yr 6.31 E2 6.22 E2 ORIGEN used half-life of 24.4 E3 yr Pu-240 6540 yr 7.74 E1 7.50 E1 ORIGEN used half-life of 6758 yr Pu-241 14.4 yr 6.03 E0 6.86 EO ORIGEN used half-life of 13.0 yr Subtotal 4,01 E4 3.94 E4 Difference due mainly to U-233 Total 4.01 E4 3.94 E4 Difference due mainly to U~233 2game value as ORIGEN, I1I: 36 Table 14. Curies after 5 yr Reconciliation of ORIGEN/ORIGEN2 Results Fission Products, in Curies Isotope Half-1ife ORIGEN ORIGEN2 Comments Sr-89 50.5 day 8.67 E-4 .016 E-4 Numerical error in ORIGIND Sr-90 29.1 yr 1.19 E4 a Y-90 64 hr 1.19 E4 a Y-91 58.5 day 8.43 E-=3 .059 E-3 Numerical error in ORIGIND Zr-95 64.0 day 7.31 E=2 .042 E-2 Numerical error in ORIGIND Nb-95 m 87 hr 7.76 E-4 .03l E-4 Numerical error in ORIGINC Nb-95 35 day 8.56 E-=2 ,096 E-2 Numerical error in ORIGIND Ru-103 39 day 1.79 E~7 .005 E-7 Numerical error in ORIGIND Rh=103 m 56 min 8.94 E-8 .049 E-8 Numerical error in ORIGINC Ru—-106 368 day 2.30 E2 2.32 E2 ORIGIN used half-life of 366 days Rh-106 30 sec 2.30 E2 2.32 E2 In secular equilibrium with Ru-106 Sn=123 m 40 min 7.99 E=2 0 ORIGIN used half-life of 129 daysd Sb-125 2.77 yr 1.83 E2 1.80 E2 ORIGIN used half-life of 2.70 yr® Te-125m 58 day 8,70 El 4.4 El Daughter of Sb-125; revised branching ratiof Te-127 m 109 day 1.09 EO .032 EQ Numerical error in ORIGINP Te-127 9.35 hr 5.39 E-1 .31 E-1 Numerical error in ORIGINC Te-129 m 33.6 day 6.40 E-10 .008 E-10 Numerical error in ORIGIND Te-129 $9.6 min 2.05 E-10 .005 E-10 Numerical error in ORIGINC Cs~134 2.06 yr 1.00 EO a Cs-137 30.0 ¥r 9.95 E3 a Ba-137 m 2.55 min 9.30 E3 9.43 E3 Ce~141 32.5 day 1.52 E-9 003 E-9 Numerical error in ORIGIND Ce~144 284 day 1.41 E3 a Pr-144 17.3 min 1.41 E3 a 37 Table 1l4. (continued) Reconciliation of OQRIGEN/ORIGEN2 Results II1I: Fission Products, in Curies (Concluded) Curies after 5 yr Isotope Half-life ORIGEN ORIGEN2 Comments Pm—-147 2.62 yr 1.02 E4 .98 E4 Pm-148 m 41,3 day .61 E~8 .004 E-8 Numerical error in ORIGEND Sm-151 90 yr 1.42 E2 a Eu~152 13.6 yr 4,02 EO 4,16 EO ORIGEN used half-life of 12 yr Eu-154 8.6 yr 2.83 El 2.34 E ORIGEN used half-life of 16 yr Eu-155 4.96 yr 5.14 E1 17.52 El ORIGEN used half-life of 1.8 yr Gd-162 8.6 min 8,27 E-2 0 ORIGEN used half-life of 1,00 yr8& Tb~162 m 7.6 min 8.27 E~2 0 In transient equilibrium with Gd-162 Subtotal 5.71 E4 5.70 E4 Total 5.71 E4 5.70 E& 4same value as ORIGEN. bA11 of these had half-lives of 32 to 109 days. ORIGEN gave the correct values at | yr, but then changed to a longer half-life (about 50% longer). Later versions of ORIGEN do not have this error, “Daughter of a precursor that suffers from the error described in ff "b". doRIGEN reversed the half-lives of $n~-123 and Sn~123 m. €0RIGEN used a longer half-life during the first year. fORIGEN used 47%; later value is 23%. 8This is an erronecus value used by ORIGEN, later versions of ORIGEN used 10.4 min, before going to 8.6 min. 38 Table 15. Reconciliation of ORIGEN/CRIGENZ Results IA'H Grams after 5 yr Fission Products, in Grams Isotope Half-life ORIGEN ORIGEN2 Comments Rb-87 4,7 E10 yr 8.48 EO a ORIGIN used half-life of 5.0 E10 yrP Sr-88 stable 5.64 EQ a Y-89 stable 7.24 E1 7.22 E1 Sr-90 29.1 yr 8.44 E1 8,77 El ORIGIN used half-life of 28.1 yr Zr-90 stable 1.57 El a Zr-91 stable 1.05 E2 a Zr—-92 stable 1.10 E2 a Zr-93 1.5 E6 yr 1.19 E2 a Zr-94 stable 1.18 E2 a Mo-95 stable 3.97 El1 3.81 El Zr-96 stable l.14 E2 a Mo-97 stable 3.25 El a Mo-98 stable 3.16 El a Te-99 2.1 ES yr 2.99 E1 2.98 El Mo-100 stable 2.85 El a Ru-101 stable 1.95 El a Ru-102 stable 1.56 El a Rh~103 stable 5.30 E1 a Ru~-104 stable 7.88 EO a Pd~-105 stable 1.82 El a Pd-106 stable 6.77 EO 6.78 EO Pd-107 6.5 E6 yr 5.64 EO a ORIGIN used half-life of 7.0 E6 yrP Te-128 stable 9.47 EO 9.46 EO I-129 1.57 E7 yr 1.58 El a ORIGIN used half-life of 1.70 E7 yrb Te-130 stable 2.09 El a Cs-137 30.0 yr 1.14 E2 1.15 E2 Ba-137 stable 1.97 El a 39 i Table 15. (continued) Reconciliation of ORIGEN/ORIGEN2 Results IV: Fission Products, in Grams (Concluded) Grams after 5 yr Isotope Half-life ORIGEN ORIGEN2 Comments Ba-138 stable 7.44 El a La-139 stable 1.70 E2 a Ce-140 stable 1.90 E2 1.89 E2 Pr-141 stable 1.84 E2 a Ce~-142 stable 1.73 E2 a - . . . ; c gg_%zz ggzbée i.Z% gfl i.gg EEI Assume ORIGEN included a precursor Nd-144 2.1 E15 yx 1.52 E2 a ORIGIN used half-life of «P Nd-145 stable 1.07 E2 a Nd~146 stable 8.41 El a Pm—-147 2.62 yr 1.10 El 1.06 El Assume ORIGEN included a precursord Sm=-147 1.06 E11 yr 5.46 El 5.35 El ORIGIN used half-life of «P>€ Nd-148 stable 4,63 E1 a Nd-150 stable 1.90 El a Sm—150 stable 2.71 E1 a - Sm~-151 90 yr 5.21 EO 5.37 ED ORIGIN used half-life of 87 yr Sm=-152 stable 1.46 El a Eu=153 stable 5.49 EO 5.45 EO Subtotal 2.71 E3 a Total 2.76 E3f 2,71 E3 a5ame value as ORIGEN. bror very long-lived isotopes the difference in half-lives has no effect over a 5-yr period. Cpr-143 (13.6 days); 6 grams at discharge would explain the difference. dNd-147 (11 day); 1.6 grams at discharge would explain the difference. €ORIGIN had more precursor, Pm—-1l47. fThig total includes some minor contributors not included in the subtotal. 40 half-life from 162,000 to 158,000 years. This accounts for the U-233 and also for the small difference in the total actinide mass of 4.01 x 10" g vs 3.94 x 10% g, The difference for Pu-241, a minor constituent, also results from a revised half-life of l4.4 years vs the prior value of 13 years. ® Actinides (in curies): Excellent agreement; of a total of 17 isoto- pes, 11 agreed exactly, 5 were within 6%, and an explanation was found for Cm~242, a minor constituent where ORIGEN2 was 75% low. A minor precursor for Cm—-242, Am—242m (152—years half-life), would not have been shown in the summary ORIGEN printout because it was below the 0.1%Z cutoff; however, after 5 years, it would contribute the major part of the total Cm—242, which has a half-life of only 163 d. The old ORIGEN calculations were quite thorough, including allowance for activation of the eutectic salt and rather precise modeling of the U-235 and U-233 fuelings and operating cycles. Allowance was made for the continuous gas sparging, which resulted in partial stripping of tri- tium and rare gases; also, the fluorination processing after the U-235 operation, which removed not only U, but also H, He, Se, Br, Kr, Nb, Mo, Tc, Ru, Te, I, Xe, and Np, was taken into consideration. Both Tc-99 and I-129, of potential concern for permanent emplacement, were in the grams printout. We were able to identify only two factors that were not included: activation of corrosion products, and activation by neutrons from (a,n) reactions on the Be-9 and F-19 in the eutectic salt mixture. Both of these are very minor contributors. A helpful view of the ORIGEN2 results is given in Figs. 4 and 5. Figure 4 shows the total FP activity and the major contributors out to 1 million years. Figure 5 shows the total actinide activity and the major contributors, also to 1 million years. Both drawings were made to the same scale. The former is essentially all beta-gamma contributors, while the latter is mostly alpha emitters. The actinide decay chain does include T1-208 (from the U-232 chain), which is noteworthy because of a very energetic (2.6 MeV) gamma accompanying the beta decay to 298ph. 1In each case, a relatively few isotopes contribute most of the activity over a rather distinctive time span. Two observations are particularly noteworthy: 41 ORNL DWG 85-438 106 105 104_ 103 — 1()2 | TOTAL RADIOACTIVITY (Ci) 101 | 100 | i — I 1 l | | TOTAL FISSION PRODUCTS AND DAUGHTERS —-— Ce-144 /Pr-144 AND Pm-147 ---------- Sr-90/Y-90 AND Cs-137/Ba-137m \ ————— Zr-93/Nb-93m AND Tc-99 (The Zr-93 is an activation product.) — ——— - = w— e W S SW — | | | | 1 10-1 10° Fig. 4. 10 102 10° 104 10° 106 TIME AFTER DISCHARGE (years) Fission product activity of MSRE fuel and flush salts. 42 ORNL DWG 85-439 106 T 1 I I I 105 - — TOTAL ACTINIDES AND DAUGHTERS —-— U-232 CHAIN ........... U_233 CHA'N ~ 104~ / eeee- Pu-239, Pu-240, AND - o Pu-241/Am-241 >. = > - o CaFo 1.55 0.43 Zr Zr + 2F5 » ZrFy 6.4 0.12 Ti CTi + 2F5 > TiFy 4.5 0.09 Al Al + 1.5F, » AlF3 2.7 | 0. 24 Mg Mg + Fp » MgFp 1.74 0.23 54 Other factors to consider are the manner and the degree of contact between the solid salt and the getter metal. Since the F, has found its way to where pressure can be measured, simply adding metal flakes, chips, or sponge to the void space above the salt may be adequate. If finely divided metal were also added to the bulk salt (while in a molten con- dition and then allowed to freeze), recombination within the solid salt would be enhanced. Metal with a density nearly identical to that of the salt is required in order to achieve good dispersion throughout the salt. This can be achieved by alloying Be, Mg, or Ca with Al, Zr, or Ti in suitable proportions. Before the addition of getters is undertaken, additional data are needed in two areas: the storage tanks themselves, and the performance of candidate getter metals. Concerning the former, a means for monitor- ing for Fj release should be included. This could be done either via measurements of pressure rise in the sealed storage tanks or via periodic analyses for F; in swept samples. Meanwhile, tests should be conducted with various metals to determine the actual efficacy of each in reacting with F, at ambient temperature and in various configurations. 4.3 INTEGRITY OF THE FACILITY The apparent integrity of the hot cell and the contained fuel salt and flush salt is very high. In many ways, the facility already has the characteristics of an engineered repository. The uranium is in a solid form and is being held in a configuration that is safe against criticali- ty. The containers are made of heavy-walled Hastelloy N (INOR-8), which is highly resistant to corrosion by fluoride. They are stored in a heavily shielded hot cell made of reinforced concrete with 3-ft-thick walls (largely underground) and a double~layer 5-ft-thick roof. The cell, lined with a welded skin of 1/8-in. stainless steel, is gas tight and leak-tested annually; it is monitored for temperature and pressure on a regular basis. However, the facility was not designed for permanent or extended emplacement, and three factors need attention — cell penetra- tions, control of groundwater, and secondary containment — before this can be seriously considered. These factors are addressed in Sects. l].. 30 ].""'40 3.3. 55 4.3.1 Cell Penetrations The hot cells were built for complex operations involving material transfers and extensive use of electrical heating. All functions were fully instrumented with redundant power supplies, control circuits, tem- perature and pressure readouts, etc. Therefore, the three main operating cells (reactor cell, drain tank cell, and fuel processing cell) have numerous penetrations for electrical signals plus a large number of material transfer lines through the outer cell walls. There are also two large cell-to-cell openings: a 36-in. opening between the reactor and drain tank cells, and a l4-in. opening between the drain tank and repro- cessing cells (the latter is sealed). These openings carry insulated, heated lines for the transfer of molten salt, plus other lines that almost fill these openings. Tables 2] and 22 list the penetrations for the reactor cell and drain tank cell, respectively. A more comprehensive tabulation, including the processing cell, is given as part of Appendix D. In summary, these penetrations can be grouped and described as follows: Reactor Cell Numbers I—XXIV: Penetrations are from 4 to 36 in. in diam (many 8 and 24 in.); most are fitted with multiple lines, for materials, electrical, ther-— mocouples, and off-gas. Only one is a "spare;" four carry dual water lines. Drain Tank Cell Numbers 1—30: Single penetrations are from 3/4 to 14 in. in diam; they are used to move material [steam, water, helium, and molten salt; nine were used by the Chemical Technology Division (Chem Tech) for reprocessing] and are located as follows: - South wall: 18 lines, mostly l-in.; - West wall: 2-3/4-in. lines; —_ North wall: the nine used by Chem Tech and the l4-in. salt transfer opening; eight of the nine are 1-1/2-in., and one is 4-in. i1l Iv y 1! Vil Yill Ix I Xl ill Xill {1V Y XVl X¥I1 WL X1X XX ixl iE1 FX1 X Huitiple &0 Hultiple 44 fultiple 44 Hultiple &0 Muitiple &% Gpen Open Huitiple 8 Blanked in Cell Multipie Hultiple i Single Furnace Hultiple Hultiple Hultiple Heltiple Hultipie Furnace Hultipie Muitiple fipen Duct i Hultiple 38 Open Pl G = P B P Y PR e e Ll O Table 21. Reactor Leak DBetectors Electrical Electrical Thermocouples Instrumentation Samplier ffgas Sampler Fuel Pump Aux: Fiping Neutron lastrusent Tube Fuel Puap Liquid Level Fuel Puep Aux. Piping Coaponent Coolant Air Coolant Salt to HX Water Lines Spare Water Lines Water Lines Water Lines Coolant Salt to Radiator Ottgas Otfgas Cell Exhaust Duct Thermocouple Brain Tank Cell intercon Sump? Cell Location Ft-in., Degrees 836 634 834 834 836 847 B47 836-9 834-5 B44-4 B36-9 829-10 840-10 839-9 839-% B39-9 §39-9 839-9 837 839-9 839-% 824-10 836 .B25-2 Bottos Center 15 30 43 b9 75 1o 13 125 143 153 160 143 1790 183 200 203 210 220 220 22 230 243 323 330 MSRE reactor cell penetration list e ————— - ————— Sub-unit sizes TH3/4% 2B%1/2" 2583 /B* IPS 60 € 3/B*1PS Iel/2%, §4374%, I91VIFS Arcess frea Size Reference Drawings Generals EGGD-40704,41487,41489,41450 South ES5A 24" DKKD-40974 EBED-418463 EBBB-41B&4 DJJD-35494 DIID-40495 60 % 1/4" tubing South E5A 24" DKKD-40976 EBBD-4i863 EBBD-41B64 EMNI-56230 ENMI-5624h 6 % 1*1FS, 38 ¥ 3/4"1PS South ESA 24" DKKD-40976 EBED-41863 EBBD-41B64 EMMI-56230 EMMI-54246 & * 1“1PS, 38 % 3/4°IPS South ESA 24" DKKD-40976 EBED-41863 EBBD-41864 DHHB-35567 South ESA 28" DKKD-40976 EBRD-41863 EBBD-41B44 DHHB-35567 High Bay 4 DKKD-40%73 DKKD-30974 {2 #1/2%) High Bay 4" DKKD-40973 DKKD-40974 DBBC-4133% Servite Tunnel (B" DKED-40717 EKKD-40735 High Bay 36" DKKD-40716 EKKD-40715 EHHA-4179% Special Eq. Ra. Special Eq. Re. Special Eq. Ae. Coolant Celi Coalant Cell Coolant Cel} Coolant Cell Coclant Cell Coolant Cell Coolant Cell Coolant Cell Coolant Cell {07 Tunnel West Tunnel Drain Tank Cell 24" b 6! 30" 28" 3&* DKKD-40973 DKKD-40%75 EGBD-55411 EJID-535428 DKKD-44718 EKKD-40737 EGBD-35411 DKKD~-40714 EGGD-55411 EKKD-40711 EGBI-3349B DKKB-40746 DKKD-40741 DKKD-40740 DKKD-40741 DKKB-40740 DKKD-40741 DEKD-40740 DEKD-20741 IKKR-40740 DKKG-40741 DKKD-464712 EBBI-35498 DEKB-40740 DKKD-40741 DKKD-40740 DRKD-40741 DKKD-4471G £EKKD-40749 DKKD-40976 EBED-41863 EBBD-4:B4&4 DHHB-55367 EEKE-40713 ERBI-55493 10 % lon Chasber Buides 6 ¢ 1/4* tubing 2+ 1"1PS, 11 ¥ 1/2*1P5 Jet" 2 ¥ 1-18 38 ¢ 3/8"1PS 9¢ Table 22, MSRE drain tank cell penetration list 1L, Type Subsets lsage Lell Locatian Accesc frea Size Reterence Drawings Sub-unit sizes 1 Single | Steam fros Domes South Wail South ESA 3" in 4° DDKD-40945 ERGD-55425 2 Single 1 Stear froam Domes Couth Mall Scuth ESA 3" in &° DDKD-30945 EBGD-554%5 3 Single | Water to Steam Domes South Wall South ESA I DOrD-40948 EGGD-35425 4 Bingle 1 Water to Steam Doaes South Hall South ESA 1" DDKD-40948 ERGD-5542 3 Single | Water ‘ South Wall South ESA 1" [DED-4094E DKXB-412533 & Single 1 Hater South Kall South ESA 1" BOED-40948 7 Single 1 Spare South Wall Bouth E54 Iy DiKD-40948 B Singie ! Spare South Hall South ESA I DBED-40948 g Single { Heliua South Wail South ES8 1/2% in 1° DI¥D-40548 10 Sirgle ! Heliua South Wall South ESA 1/2* in 1" DDKD-40948 it Single 1 Heliua South Wail South €54 172" in 1" DDET-4G948 {2 Single { Heliue South Wall South ESA 1/2% in 1" DDER-A094E i3 Single I Helium South Wall South ESA 142" in 1* DDKB-40948 14 Single 1 Heliua South Wail South ES4 1/2" in 1° DDKD-40948 15 Single 1 Heliua South Wall South ESA /2" in 1" DDKD-40748 14 Single ! Helium Gouth Wall South ESA §/2" in 1" DDKD-40948 17 Single I Sump Discharge South Wall Waste Cell 314 DOKD-40948 DKKB-41280 DKKB-412681 18 Single 1 Sump Discharge South Wall Waste Cell 34" DDKD-40948 DEKB-412B0 DKKR-41281 19 Single 1 Air to Sump West Wall West of Bldg. 3/4" DDKD-40948 DEKB-31280 s Single 1 Air to Sump West Wall West of Bldg., 3/47 DDKD-40948 DKKE-412B0 24 Single {Chea. Tech.) North ¥all Fuel Proc, Cell 1-1/2° DKED-40949 22 Singie {Chen. Tech.) North #all Fuel Proc. Cell i-1/2" DriD-40949 23 Single (Chem. Tech.) Narth Wall Fuel Froc, fell 1-1/2° DEKD- 40949 24 Singie (Ches. Tech.!} North #all Fuel Proc. Cetl 1-1/2° DKED-40%49 5 Single {Chem. Tech.} North Wall Fuel Proc. Cell t-1/2° DEXD-40349 z Single {Chea. Yech.) North Wall Fuel Proc, fell &7 DKED-40249 27 Single {Chem. Tech,) North Kali Fuel Proc., Cell 1-1/2° DKKD-40749 Z Singie {(Chee. Tech.) North Wall Fuel Proc. tell 1-1/2° DKED-40949 29 Single {Chem. Tech.) North Mail Fuel Froc. fell 1-1/2° DKKD-40949 in Furnace Salt Transfer North Wall Fuel Froc. Leil 14" DKED-4094% Ai-1 to 34 BHank 29 Instrueentation East Wall Narth ESA 3ig" DKKD-40947 [HHB-35347 B-1 to 34 Bank 36 Instrumentation East Kall North ESA 374" DKED-40947 DHHB-55347 £-1 tp 34 Bank 36 Theraocouples East Wall Korth ESA 34" DKkD-40947 DHHE-33567 D-1 to 34 Bank 16 Thereocouples East Wall North ESA 3/4" DKKD-401947 E-1 to 36 Bank 14 Thermocouples East Wall North ESA iy DExD-40947 f-37 to 50 Bank 24 Electrical East Wall Rorth ESA 316" in 3/4" DEKD-40947 EMMI-31636 §-37 to &0 Bank 24 Electrical East Wall North ESA /16" in /4" DEKD-80947 EMNI-S1636 £-37 tp &0 Bank 24 Elertrical East Mall North E5A 346" in 3/4" DKKD-40947 EMMI-51456 p-37 to b0 Bank 24 Electrical East Wall North ESA 316" in 374" DEED-40947 EMMI-T1656 LS o 60 Bapk n & Bank — T T o e Hultiple [~ Hultiple Hultiple Single Multiple Hultiple Aultiple 3R} T o I X T O r-J 4 Electrical 4 Electrical Spare Spare Cover BGas Spare Spare Spare Component Coolant Air Component Conlant Air ! Component Coolant Air 4 BP Cell 12 Leak Detector 12 Leak Detector - d d (2 East Hall Bact Wail East Hall Eact Wall East Wall East #all Eact Kall East Wall East Wall East Wail East KWall East Wall East Mall East MWail Table 22. North EGA North ESA North ESA Merth ESA North ESA North E54 Horth E54 Morth ESA Korth ES4 North ESA North ES3 North ESA South ESA South ESA 34167 in 3747 DYKD-40947 EMMI-D1ASH 316" in 374" DEED-40947 EHMI-51656 6"‘ b" b‘! in o o :, b R 1-1/2" in 4" (continued) DEXD-40947 FRED-40947 DEKD-40947 EGGE-41584 3% 1/27IFS DXED-40747 BEKD-409487 DEED-40747 DEKD-40747 EGBE-31684 DJJA-41877 DIJA-41BBO 3% /4IRS [XkD-40947 EGGE-41884 DIJA-41579 DJJA-41BRO 3% 3/401P8 DKKD-40347 EGGE-41894 DJIA-41879 LJIAR-41880 DYED-40547 EGGE-1684 DJJA-41879 BJJA-41880 4 1/2°1PS DKKD-40947 ERED-41863 EBBD-41BAT DJID-354%4 DJJD-55495 12 * 1/4" tubing DKKD-40947 EBER-41BAT EEBD-41863 DJID-33494 DIJD-55495 12 # 1/4" tubing 8§ 59 Banks A~l! to -60, B-1 to -60, C~1 to -36, D-1 to =36, and E-1 to -36: All are 3/4-in. penetrations of the east wall and are used for for instruments, thermocouples, and electrical supplies. Numbers GR: All are 6—-in. penetrations of the east wall, most with multiple lines for cover gas, cooling air, and leak detectors; five are "spares.” Reprocessing Cell Numbers 31-54: Specific penetrations have not been identified, but these functions are provided: - material transfer: at least 4 (HF, H,, F,, off-gas); > - H-series line heaters: 28; ‘ - equipment heaters: 57; - thermocouples: 94; — limit annunciators: 15, These penetrations are well-engineered and have proven to be reliable through many years of operation and monitored storage. They, along with the reactor and drain tank cells, were pressure—~tested to 45 psi (gage) prior to placing the MSRE into hot operation. All external penetrations are welded into place and were offset or shielded to mini- mize radiation leakage. The numerous electrical penetrations were sealed at installation. The transfer lines are, or can be, sealed on both sides of the cell wall with Gray-loc closures or by welding. There is no cause for concern for the short term, but these penetrations should be reviewed from engineering and corrosion aspects before extended storage is adopted as a planned policy. Some of the transfer lines probably should be capped, rather than just being valved off. Many of the electrical penetrations are of no further use and could also be sealed. A more stringent analysis is appropriate for extended storage. The large cell- to-cell opening also deserves a stringent analysis for long-term storage if that is to be pursued. Again, for the short-term, the two—cell system appears to be more than adequate, but extended storage might benefit from specific modifications. 60 An engineering study of the penetrations, closure options, and possible containment envelopes was started late in FY 1985 but could not be completed with the available funds; the results obtained thus far have been reported and are included as Appendix D. It is anticipated that this work will be completed in FY 1986, or as soon as funding is pro- vided. The important features of this study to date are: 1. A sequence of three possible containment envelopes was defined, for future consideration: the drain tank cell; the drain tank plus reactor cells; and the drain tank, reactor, and reprocessing cells. Eight types of cell penetrations were identified and catalogued. Three types of closure options were identified for these eight types of penetrations: (1) capping/welding of unneeded lines that emerge from a multiple penetration; (2) total capping of a multiple penetration where none of the individual lines are needed anymore and are cut off behind the new seal-plate; and (3) use of a threaded, pressure-tight plug on electrical lines, as was done on the original design with spare lines. Individual penetrations were catalogued in terms of containment envelope and closure options. The identification of possible future uses for each of the penetrations was started but not completed. Finally, it was suggested that a visual inspection be made of all accessible penetrations to determine the present condition of each. This will be done as soon as possible because of the obvious impli- cations if any penetrations were to be found in poor conditiom. 4.3.2 Water Control 4. If water should enter the cells, it could cause major problems such increased corrosion rate of the Hastelloy tanks; slow reaction with the fluoride salts or rapid reaction with any fluorine (from radiolysis) to yield HF; radiolytic production of Hj and 05, which is an explosive mixture that could cause pressure buildup or could increase oxidative corro- sion; and movement of radioactive material along transport paths provided by water. 61 Water entry has never occurred over the past 20 years. Each cell does have a sump where any water would be collected (if it did enter) and could then be pumped out or jetted out. The cells are leak—-tested annually and are known to be leak tight; therefore, water entry is theoretically not possible, even if the level of the groundwater were to rise above that of the cell floor. Even so, groundwater is a long—term concern since the water table is not very far down at the MSRE site. However, the building was constructed with French drains below the footers and has a building sump with a minimum elevation of 812 ft. The lowest cell level is the drain tank cell, at 814 ft for the cell floor, which is above the building sump level. Although the building sump pump is known to operate occasionally, the frequency of operation and the volume of water pumped have not been recorded or measured previously. As a part of this study, a pump monitor was recently installed and has been in operation since August 19, 1985. Readings were taken daily, except on weekends. During the first three weeks, the west pump operated an average of 1.06 h/d, with a daily high of 1.5 h and a daily low of 0.6 h. The east pump, a backup system, did not come on during this period. It is planned to continue gathering these data as part of our ongoing S&M. Because of the importance of water control for extended storage, an engineering overview was prepared. This report is attached in its entirety as Appendix E. The information contained within it was obtained from engineering drawings and reports; highlights from the report include: l. A description of the radioactive liquid waste system. 2. A description of the nonradiocactive drainage, which includes ground- water collected by the building French drain system and water entering floor drains outside the cells, which is then routed to the building sump cited earlier. 3. A description of the storm water system, which collects water from roof drains and discharges to a separate catch basin. 4., A description of the sanitary wastewater system, which is discharged to a septic tank and drain field system. 62 4.,3.3 Secondary Containment In context of this discussion, secondary containment refers to the external building structure surrounding the hot cells (i.e., the outer shell, the building ventilation system, the high-bay area and overhead crane, and the control rooms). This is the containment that is relied on during open—cell maintenance or in the event of a release from one of the cells. Secondary containment will be required and must be fully functional in case it should become necessary to open one of the cells for modifications, additions, removal of extraneous equipment, or treat- ment of the salts. Secondary containment is not required during standby mode, which is the usual condition of the facility. The present status of the secondary containment system is functional. The high-bay area can be closed and ventilated through the cell ventilation system, while maintaining a negative preséure inside the containment zone. The overhead crane is still operational. The remote maintenance control room has deteriorated somewhat, and the zinc bromide viewing windows have been drained (but are still usable). The main control room contains most of the instrumentation being used for monitoring and surveillance. The useful lifetime of the secondary containment system is definitely limited in the absence of a planned program to maintain this capability. Any operations that may need to be done prior to committing to extended storage should be carried out in the reasonably near future. In doing so, it would be possible to still utilize experienced personnel who have a firsthand knowledge of the facility from prior experience when the MSRE was an active project. 4.4 GEOLOGY AND HYDROLOGY A geologic and hydrologic overview of the MSRE site was prepared by a consultant in this field. A copy of his report is included as Appendix F. The more significant of his findings can be summarized as follows: 1. The general geologic conditions at the MSRE site are favorable for continued storage. The area has a low seismic risk and has been relatively stable for millions of years. Other waste storage facili- ties (solid LLW, TRU waste, hydrofracture) have been sited in Melton Valley because of its suitability for such applications. 63 2, Water is a significant concern because the water table is not far below the surface and because water is a potential problem for two reasons: it can hasten corrosion, and it can provide a pathway or mechanism for movement of any released radioactivity. In planning for the future of MSRE and evaluating various options, the time scale must be considered since each time frame has its own characteristics. The various time frames are defined in Sect. 1. For purposes of this evaluation, two key characteristics are: the types of physical control that can be applied, and the nature of the radiocactive decay which is occurring. Seismic factors are of minor concern to us through the near term (up to 100 years) since the reinforced concrete cells and heavy-walled Hastelloy tanks will provide ample strength during the early years. Water intrusion, however, must be cdnsidered even for the short term. Data are needed on water behavior in the immediate environment. As listed in the geologist's analysis, data are needed on the following: (1) characterization of the earthen materials at or near the site; (2) the site design itself, including all engineered improvements; and (3) site hydrology. 4.5 ENTOMBMENT Entombment in cementitious material has been surveyed by L. R. Dole of this Laboratory, whose complete report was given as Appendix E in the evaluation study that preceded this report.29 His survey identifies several aspects of this concept that could be detrimental, thus indi- cating a need for caution since entombment is nonreversible for all prac-— tical purposes. It also points out the potentially harmful effects of water, which suggest restraint at present and also support the need for more data in this area (as noted in the previous section). While entombment is very attractive and might, in fact, become the option of choice at a future time, it is premature at present because we are not in a position to make a permanent commitmént. This will continue to be the case until we have definitive data concerning the radiolysis of fluoride. The negative aspects of entombment include the preclusion of direct surveillance, the possible net weakening of the overall structure from shrinkage or expansion mismatch, and the introduction of water from 64 the grout mixture itself. In addition, the solidified grout could act as a wick to transport moisture to metal surfaces and thereby increase corrosion. It also fills up void space that might better be used as a sink for intruded water or as a reservoir for a desiccant (e.g., unslaked lime or Portland cement). On the other hand, a properly engineered grout could add strength and provide a diffusion barrier against the movement of water either in or out of the cell. The projected lifetime of a grout can be measured in millennia in a dry climate. A more-detailed survey is required to iden- tify any areas where specific data need to be gathered, or generated, in order to have the necessary technical basis for a future decision regarding entombment. Radiolysis of the water contained in grout is also a possible drawback. The resulting oxygen and hydrogén could lead to excessive pressure or an increased corrosion rate; however, inhibitors or catalytic recombiners could be used to mitigate this effect. The use of other entombment materials, such as lead or sulfur, has been suggested, but these materials have not been examined in detail. 4.6 UTILIZATION The MSRE building facilities are being used extensively, par-— ticularly the office areas and the storage space in the high-bay and receiving areas. Approximately 50 people from the Health and Safety Research Division are housed in the office complex, and this number is expected to grow. The history of the Laboratory clearly shows an ever increasing need for office space. Therefore, the office facilities should be retained in any event. The need for storage space for low-level radioactive samples and materials has also continued to grow. At present, the building provides space for soil samples (about 100 55-gal drums) and Cs-137 sources (about 260 in 200-1b lead pigs), and miscellaneous equipment (e.g., about 20 remote manipulators). These storage facilities definitely need to be retained. Even a conservative estimate suggests an additional 20-year lifetime for the offices and the storage space. This is commensurate with the anticipated short-term extended storage of the fuel and flush salts. 65 Looking beyond utilization already extant, the hot cells themselves are a valuable asset whose future utility should be preserved. Because of activation, the reactor, drain, and processing cells are likely can- didates for entombment unless a suitable major project comes along. Even with entombment of these three cells, however, they could first be filled with waste removed from other cells. The remaining cells could easily be used for storage purposes or, with modification, for a suitable project. 4.7 FINAL DISPOSAL The long—~term alpha activity of the U-233 decay chain may require eventual removal of the fuel and flush salts, even if the radiolysis problem can be controlled with getters. In the latter case, it would be possible — and highly desirable — to leave the salts in their present Hastelloy N tanks since these tanks should serve as excellent primary containers. There are three possible options for permanent disposal of the tanks and contents: 1. the WIPP (Waste Isolation Pilot Plant), which is now under construc- tion in New Mexico; 2, a spent fuel repository, which is planned for a western location; or 3. on-site greater confinement disposal (intermediate—depth burial), which has been considered for 0Oak Ridge. These three options are discussed in the following sections. In the event that the radiolysis problem cannot be controlled with getters, thus making reprocessing necessary, the separated uranium and TRU waste would still be candidates for the same three final options. However, in this case, with the fluoride removed and the residue repackaged, acceptance criteria should be easier to meet. It should be recognized that, ini- tially, all disposal sites will be conservative in specifying their acceptance criteria but that, with the passage of time and accumulation of experience, the potential for some relaxation in these criteria is very real, especially for a one~time interment that could be given addi- tional overpacks. 66 4.7.1 WIPP Since the WIPP is actually under construction and has been planned for many years, the waste acceptance criteria (WAC) for WIPP have been the subject of extensive study, review and comment, and revisions. Even so, these WAC are not yet final and, as already pointed out, tend to be conservative. Appendix C summarizes the WAC for remotely handled (RH) waste, the Certification Compliance Requirements (CCR), and some general comments. The WAC and CCR are not always in agreement. Those require- ments that concern the MSRE fuel salt are briefly discussed below. Size limit: Diameter of tanks too large (50 vs 26 in.) but length satisfactory (86 vs 121 in.) Weight: Each tank (~1300 kg) plus salt (~2300 kg) is just within the limit (3636 kg); addition of getter would make them overweight. Surface dose: Surface dose probably greater than the 100-rem/h limit; also, neutron dose much higher than 75 mrem/h, as limited by the CCR (although the WAC requires only that neutron doses >75 mrem/h be reported). Thermal power: Well below the limit of 300 W/package; drain tanks are about 100 W each. Criticality: About 10 times the limit of 1.9 g/L; actually, about 17 g/L. Pu-239 equivalent: Probably about the limit of 1000 plutonium- equivalent Ci/package. 4,7.2 Commercial Spent Fuel Repository This offers some attractive features., Certainly, the MSRE fuel and flush salts qualify in a generic sense. The MSRE D&D is funded through the commercial branch of the SFMP since the MSRE was intended for com— mercial application. The fuel salt is a spent fuel, and the flush salt contains some spent fuel. Most of the WIPP restriction would not be applicable to a spent fuel repository, which would be designed to handle higher doses, thermal power, criticality, and Pu=-239 equivalent. One possible problem is the tank diameter. A repository might not be designed for a package diameter of over 4 ft, even on an exception basis. 67 4.,7.3 Greater Confinement Disposal Every major DOE site (Oak Ridge included) has potential problems with oversized, overweight, or otherwise out-of-specification TRU waste, for which intermediate~depth disposal is being considered. The MSRE fuel salt 1s an obvious candidate for such disposal since the large tank diameter would not be a problem in this situation. Shipment would also be minimized if an Oak Ridge site is developed in the future. 4,8 IN-CELL ALTERATIONS A number of in-cell alterations might be performed in the short term (within 10 to 20 years) to enhance the condition of the fuel salts during extended storage and also to help set the stage for eventual removal. As a step toward the eventual removal of the salts, it would be highly desirable to place the cell and the salts into a suitable condition during the short term so that the removal operation can be carried out as simply and easily as possible. The following possibilities are presented and then evaluated in terms of both enhanced storage and eventual removal: l. removal of steam domes (and addition of neutron poison); 2. addition of fluorine getter, without remelting; 3. remelting of salt, with addition of fluorine getter; 4. repackaging of salt into smaller containers, after remelting with addition of fluorine getter; 5. disconnection of lines to and from the drain tanks; and 6. installation of additional instrumentation. It is assumed that the fuel salts will eventually have to be removed to a more permanent location because of the long—-term activity from the U-233 chain. It is also assumed that it will be feasible to do so with the salts packaged in their present containers. With these assumptions, it is clear that the steam domes must be removed eventually; therefore, this should be done in the short term while experienced personnel are still available. Once the steam domes with attached bayonet tubes are removed, the thimbles (32 per tank) can be used to insert neutron poisons. These can be in the form of Cd, B, or Ge metal rods. The earlier discussion on radiolysis is a strong argument for the addition of a fluorine getter. In fact, if an effective getter is not 68 added, disposal of the fuel salt as the fluoride will probably not be acceptable. The simplest way to include getter is to add it to the void volume in each tank. This might require cell entry; of course, the pre- ferred method would be to supply getter through the addition tube used previously for adding salts. 1If cold laboratory tests show that the getter must be dispersed within the solid salt to be effective, then remelting would be required. While this operation is possible, it might cause some further corrosion to the tanks. Dispersal of added getter could be accomplished via sparging with dry helium through the dip tubes already provided for this purpose. In the event that the salt needs to be melted in order to add getter, this would be the logical time to repackage into smaller con- tainers (if this step is needed). However, the incentive to repackage is small since the controlling factor is radiolysis and, if radiolysis can be controlled by the addition of getter, this can be done just as well in the present drain tanks. The lines for the vessel should not be disconnected until there is assurance that the salt need not be repackaged or reprocessed. Such assurance cannot be provided until actual data are in hand to verify the addition of getter as a viable method for limiting fluorine formation. This may require the installation of additiomnal instrumentation. 4.9 CONTINGENCY PLANNING The MSRE has had an exceptionally clean history following shutdown of the reactor in December 1969. Because of this long period of trouble- free surveillance and maintenance, formal contingency planning and/or emergency planning has received virtually no attention. However, since it is now becoming clear that extended storage in the present location will be required, such planning is appropriate in the immediate future. Obviously, no major contingencies are expected; nor are they likely to occur., On the other hand, it is prudent to carry out a structured analy- sis in order to systematically investigate the possibilities, to identify any weaknesses that could or should be remedied in advance, and to be prepared in case the unexpected happens. A preliminary list of possible events includes, but is not necessarily limited to, the following items: 5.1 71 5. CONCLUSIONS EVALUATION OF OPTIONS Evaluation of the options, data, and constraints discussed in the body and appendixes of this report led to the following conclusions: 1. 2. 3. Extended storage is not only feasible but preferable at this time because Ae b. the fuel salt is presently in a safe place; there is no available repository or disposal site to take it if it were removed, either with or without reprocessing, at this time; the full range of future disposal options or ldcations is not yet known ; any enhancement actions taken at this time would probably benefit whichever option is finally exercised; and extended storage would provide time to collect additional data that would be of direct help in selecting a final option at a future date. Entombment should be deferred because: de b. such action is not necessary at this time; it is essentially nonreversible and would be premature at this time; and additional technical data are needed before such a decision could be supported. Some enhancements in direct support of extended storage should be evaluated in the near future. For example: Ae cell penetrations should be reviewed by Energy Systems Engineering, inspected by the Operations Division and, where con- sidered necessary, sealed permanently; cell internals should be reviewed to identify interfering struc— tures, if any, and plans made for their removal; remelting of the fuel salt, with possible addition of a fluorine getter (and possibly a neutron poison) should be evaluated for implementation since this would eliminate both the need for annual reheating and long—-term concerns about radiolysis-produced fluorine; 4. d. 72 repackaging (if needed, it should be done in conjunction with remelting) should also be evaluated; enhancements seem prudent since the alpha activity (and the resulting neutrons from a,n reactions) will decline only slightly in future years, even though the gamma activity will drop signifi- cantly; these enhancements would not interfere with the exercise of options at a later date and would, in the interim, improve on the already sound and secure condition of the facility; and these enhancements, while not necessary for continued short-term storage, should be made prior to initiating extended storage while the assured capability to do so still exists, in terms of both people with firsthand knowledge and equipment that is still func- tional. The following pertinent data (or studies) should be obtained (or made) over the next few years: Ae b. Coe observations of groundwater levels around the site; possible improvements of a civil engineering nature, to lower the groundwater level; Hastelloy N corrosion behavior in the presence of moisture plus alkali, fluoride, or both; suitability of candidate materials for use as a fluorine getter, in—cell desiccant, tailored grout, and neutron poison; and contingency plans developed, along with a review and update of S&M procedures. Beneficial utilization of the facility should be encouraged and aided for the following reasons: A b. to help fill directly our own facility requirements; office space at the site is fully utilized today, and this need is expected to grow; the high-bay area is being used to store bulky, mildly radioactive samples,and this need is expected to continue; the three more active hot cells (reactor, drain tank, and repro- cessing) could be used for storage of other radiocactive wastes; the remaining cells could be used for hot operations or radio- active waste storage; and 69 (1) ventilation failure; (2) extended power failure; (3) water ingress from leakage; (4) electric heater failure; (5) thermocouple failure; (6) overheating during annual reheat; (7) loss of instrument air; (8) a catastrophic event such as a tornado, earthquake, plane crash, or flood; (9) sabotage or terrorist action; (1Q0) possible radiation exposure via plume or ingestion; and (l1) possible radiation exposure to site opera-— tors. In addition to the above incidents, a contingency analysis should also cover three broad categories: (1) information sources, (2) iden-— tification of equipment, and (3) physical condition of components. In the event of a contingency, needed information must be easily and T quickly available. A concise "contingency plan,” with readily available backup information in various specific areas, is probably appropriate. In addition, easy and positive identification of components, controls, etc., may be essential. The recently completed color coding scheme is a significant step in this direction. The coding employed is pink for routine maintenance and surveillance, whether daily, monthly, or annual, and yellow for possible future use in remelting, transferring, etc. (If these colors seem unusual, they were chosen to match the markup of drawings used to do this work.) In the event of a serious contingency, it might be of considerable benefit if the frozen fuel were in a condition whereby it could either be sealed off and left in place for a long time or, alternatively, transported elsewhere. These factors would, of course, be enhanced if the fuel had previously been remelted and getters added, or remelted and repackaged, or if interfering structures (e.g., steam domes) had been removed. ' At this time, work is scheduled for the next fiscal year on a con- tingency plan, along with a review and update of S&M procedures. f. 73 the Li-7 content of the cooling salt is nonradiocactive and may have enough value to justify recovery. 5.2 SELECTION OF PLAN Based on the above considerations, an outlined plan and approxi- mate schedule are proposed: 1. 2. 3. 4e d. b. d. 1986 — complete (or expand on) work in progress: penetrations — complete the engineering evaluation and provide recommendations; groundwater — continue data collection on radioactivity and sump pump operation; complete an on-site survey and inspection of drainage system and components; radiolysis — conduct a peer review of the overall problem and potential solution; prepare a technical plan for pressure or fluorine "sniffer” measurements on the two drain tanks; make a decision if reheating is to be discontinued (in order to obtain direct data on radiolysis); and conduct preliminary laboratory- scale tests with potential fluorine getters; review and update the S&M procedures and prepare a contingency plan. 1987 — upgrade facility and prepare for new work: ae b. Coe seal or improve cell penetrations, as appropriate; install fluorine radiolysis test equipment; and prepare a technical plan for removal of steam dome and other internals in order to evaluate the work involved. 1988-95 — collect data: ade b. fluorine radiolysis in the drain taunks; and laboratory—scale tests on fluorine getters., 1996 — Make the following decisions: a. b. Ce d. €. to remelt and add getter, or not; to remove steam dome and internals, or not; to reprocess, or not; to repackage, or not; and selection of final disposal site, based on information available at that time. 75 Appendix A. BIBLIOGRAPHY AND REFERENCES 76 Appendix A. BIBLIOGRAPHY AND REFERENCES The MSRE Project is extensively and thoroughly documented by reports, drawings, photographs, and miscellaneous documents. A com— puterized listing of these has been prepared by Park Owen and Nancy Knox, of the ORNL Remedial Action Program Information Center. There are approximately 1450 entries in this file, categorized as follows: Type of entry No. of entries Progress report ~140 Technical report CF memo ~370 TM report ~240 ORNL report V70 Miscellaneous 14 Journal article 32 Drawing 19 Photograph ~500 Patent 4 Conference paper §50 Correspondence 11 Total ~1450 Copies of most of these documents are in the MSRE file, which is stored in the basement of Building 7503. The Information Center also has copies of many of the reports. These documents are a merger of files previously maintained by individuals who were closely associated with the project. (Unfortunately, several extensive files have been lost over the years.) Most of the photographs were supplied by Luther Pugh. There is also an excellent collection of construction drawings main- tained by Martin Marietta Energy Systems, Inc., Engineering on file in Building 1000. A relatively small number of reports have provided all the background information used for this study. A listing, in chronolo- gical order, is provided on the following pages. The chronological num- bers are used as reference call-outs in the main body of the report. 1. 2. 3. 7 8. 9. 10. 11. 12, 13. 14. 15. 77 Ruth Slusher, H. F. McDuffie, and W. L. Marshall, Some Chemical Aspects of Molten-Salt Reactor Safety: (1) Dissolution of Coolant and Fuel Mixtures in H50, (2) A Portion of the System LiF-BeF,-H,0 at 25, 60 and Near 100°C, ORNL-TM-458, December 14, 1962. P. N. Haubenreich, Inherent Neutron Sources in Clean MSRE Fuel Salt, ORNL-TM-611, August 27, 1963. J. R. Engel and B. E. Prince, Criticality Factors in MSRE Fuel Storage and Drain Tanks, ORNL-TM-759, September 14, 1964. R. C. Robertson, MSRE Design and Operations Report — Part I — Description of Reactor Design, ORNL-TM-728, January 1965. R. B. Lindauver, MSRE Design and Operations Report - Part VII - Fuel Handling and Reprocessing Plant, ORNL-TM-907R, December 28, 1967. R. B. Lindauer, Processing of the MSRE Flush and Fuel Salts, ORNL-TM-2578, August 1969. R. C. Steffy, Jr., Inherent Neutron Source in MSRE with Clean U-233 Fuel, ORNL-TM-2685, August 10, 1969. We R. Grimes, "Molten—-Salt Reactor Chemistry”, Nuclear Applications & Technology 8, 137, February 1970. M. J. Bell, Calculated Radioactivity of MSRE Fuel Salt, ORNL-TM-2970, May 1970. P. N. Haubenreich, Fluorine Production and Recombination in Frozen MSR Salts after Reactor Operation, ORNL-TM-3144, September 30, 1970. R. H. Guymon, MSRE Procedures for the Period Between Examination and Ultimate Disposal, ORNL-TM-3253, February 10, 19/1. Roy E. Thoma, Chemical Aspects of MSRE Operations, ORNL-4658, December 1971. P. N. Haubenreich and R. B. Lindauer, Consideration of Possible Methods of Disposal of MSRE Salts, ORNL/CF-72-1~1, January 28, 1972. E. L. Compere et al., Fission Product Behavior in the Molten Salt Reactor Experiment, ORNL-4865, October 1975. C. D. Cagle and L. P. Pugh, Decommissioning Study for the ORNL Molten-Salt Reactor Experiment (MSRE), ORNL/CF-77/391, August 25, 1977. 16. 17. 18. 19. 20, 78 EBASCO Services, Inc., Technical Report - Feasibility Study: Disposal of MSRE Fuel and Flush Salts, Letter Report (unnumber), October 1980. D. R. Simpson, Preliminary Radiological Characterization of the Molten Salt Reactor Experiment (MSRE), ORNL/CF-84/92, September 20, 1984. F. J. Peretz, Preliminary Decommissioning Study Reports — Volume 5: Molten Salt Reactor Experiment, X-0E-231, Vol. 5, September 1984. T. E. Myrick, The ORNL Surplus Facilities Management Program Long- Range Plan, ORNL/TM-8957, September 1984. K. J. Notz, Feasibility Evaluation of Extended Storage-—-in-Place of MSRE Fuel Salt and Flush Salt, ORNL/CF-85/71, March 1985. 79 Appendix B. SEMIQUANTITATIVE EVALUATION OF SIX MSRE OPTIONS Table No. Evaluation Basis B.1 Process Readiness B.2 Economics B.3 Short-Term Hazard B.4 Long~Term Hazard B.5 Conservation 80 Table B.l. Semiquantitative evaluation of MSRE options: A. Process readiness?® Option Process step 1 2 3 4 5 6 Melt salts (and add getter) Remove salts Build salt process facility? Process or convert salts/U2 Package salts (and U) Dismantle/dispose of facility Transport packaged products b Loow s N W | i i N Store/isolate products Decontaminate cells/equipment Dispose of liquid LLW Remove equipment Dispose of solid LIW Dismantle cell structure Dispose of solid LLW/rubble R W U = W= W W NN N W Restore area Seal pipes/penetrations Internal entombment i v e W v o Wb | i I External structures Stabilize drain-tank cell - - - 5 5 Continue surveillance - - - 5 5 5 Total 47 41 15 15 19 23 aRange: 0 to 5. (A low score is superior.) The rating considers the cost of developing the process, as well as the uncertainty of success- ful development. The time delay involved is not a part of the rating since ample time is presumably available. bNo process flowsheet has been defined at this time. CNo facility which will accept the packaged products has been iden- tified at this time. 81 Table B.2. Semiquantitative evaluation of MSRE options: B. Economics@ Process step Option 3 Melt salts (and add getters) Remove salts Build salt process facility Process or convert salts/U Package salts (and U) Dismantle/dispose of facility Transport packaged products Store/isolate products Decontaminate cells/equipment Dispose of liquid LLW Remove equipment Dispose of solid LIW Dismantle cell structure Dispose of solid LLW/rubble Restore area Seal pipes/penetrations Internal entombment External structures Stabilize drain-tank cell Continue surveillance Total 54 Ww =~ U = P =~ U1 NN W L o W 47 15 15 20 238 4Range: 0 to 5. (A low score is superior.) relative cost of each process. The rating considers the 82 Table B.3. Semiquantitative evaluation of MSRE options: C. Short-Term hazard? Option Process step 1 2 3 4 5 6 Melt salts (and add getters) 3 3 - - 5 w | i ( Remove salts v A% | 1 | i Build salt process facility ot o Process or convert salts/U 10 - - - - Package salts (and U) Dismantle/dispose of facility Transport packaged products e NN | ) ! i Store/isolate products Decontaminate cells/equipment Dispose of liquid LIW Remove equipment Dispose of solid LLIW Dismantle cell structure Dispose of solid LLW/rubble — = N = 0~ 0 W N~ U | | 1 ! 1 Restore area Seal pipes/penetrations Internal entombment | — BN w & Wun 1 | 1 External structures Stabilize drain-tank cell - - - 8 8 8 Continue surveillance - - - Total 66 47 14 16 21 31 3Range: O to 10. (A low score is superior). The rating considers the expected operator exposure, public exposure, probability of an adverse release, and possibility of sabotage on-site, during transport, or in the final location. Short term means less than 100 years, nominally 5 to 20 years. 83 Table B.4. Semiquantitative evaluation of MSRE options: D. Long-Term hazard?® Option Process step 1 2 3 4b 5 6 Melt salts (and add getters) - - - -~ - - Remove salts - - - - - - Build salt process facility - ~ - - - - Process or convert salts/U - - - - - - Package salts (and U) - - - - - - Dismantle/dispose of facility - - - - - - Transport packaged products - - - - - - Store/isolate products 1 1 - - - - Decontaminate cells/equipment - - - - - - Dispose of liquid LLW 1 - - - - - Remove equipment - - - - - - Dispose of solid LIW 1 - - - - - Dismantle cell structure - - - - - - Dispose of solid LLW/rubble 1 - - - - - Restore area - - - - - - Seal pipes/penetrations - 2 5 5 5 5 Internal entombment - 2 5 - - - External structures - 1 2 1 1 1 Stabilize drain~tank cell - - - 10 5 3 Continue surveillance - - - 1 1 1 Total 4 6 12 17 12 10 aRange: 0 to 10. (A low score is superior.) The rating considers the expected public exposure, probability of an adverse release, and possibility of inadvertent intrusion. Long term means greater than 100 years, nominally 1000 years or longer. bnot really a viable long-term option; would eventually have to be supplanted by one of the other options. 84 Table B.5. Semiquantitative evaluation of MSRE options: E - Conservation® Option Process step 1 2 3 4 5 Melt salts (and add getters) - - - - - Remove salts - - - - - Build salt process facility - - - - - Process or convert salts/U - - - - - Package salts (and U) - - - - - Dismantle/dispose of facility - - - - - Transport packaged products - - - - - Store/isolate products 1 1 - - - Decontaminate cells/equipment - - - ~ - Dispose of liquid LIW 1 - - - - Remove equipment - - - - - Dispose of solid LIW 1 - - - - Dismantle cell structure - ~ - - - Dispose of solid LIW/rubble 1 - - - - Restore area (land) - 5 5 5 5 Seal pipes/penetrations - - - - - Internal entombment - - - - - External structures (buildings) 10 - 1 1 1 Stabilize drain-tank cell - - - - - Continue surveillance - - - - - Total 14 6 6 6 6 aRange 0 to 10. (A low score is superior.) The rating considers near—term utilization of the buildings and long-term utilization of the land. 85 Appendix C. WIPP WASTE ACCEPTANCE CRITERIA C.l. WIPP WASTE ACCEPTANCE CRITERIA FOR REMOTELY HANDLED (RH) TRU WASTE C+.2. WIPP CERTIFICATION COMPLIANCE REQUIREMENTS C.3. WIPP: GENERAL COMMENTS 86 C.l. WIPP WASTE ACCEPTANCE CRITERIA FOR RH TRU WASTE Source: WIPP-DOE-069, Rev. 2, Draft C, dated August 1984. Combustibility: Any combustible TRU waste shall be packaged in a noncombustible container. Immobilization: Required if more than 1 wt % is in form of particles <10 microns, or if more than 15 wt % is <200 microns in diameter, Par- ticulate desiccants (e.g., Portland cement) are exempted. Sludges: Shall be packaged such that internal corrosion of the container does not occur. Liquid Waste: Free liquids are not allowed; minor amounts in cans or bottles are acceptable, Explosives and Compressed Gases: TRU waste shall contain no explosives or compressed gases, as defined by 49 CFR 173, subparts C and G. Pyrophoric Material: No more than 1 wt % may be pyrophoric forms of radionuclide metals, and these shall be generally dispersed in the waste. Pyrophoric materials other than radionuclides shall be ren- dered safe by mixing with stable materials (e.g., glass or concrete). Radioactive Mixed Waste: No hazardous waste is permitted unless it is also co-contaminated with TRU waste. Reactive material shall be identified. Corrosive materials must be neutralized, rendered non-— corrosive, or packaged in a manner to ensure container adequacy through the design lifetime. Container: May be the RH container itself, or it may be the container inside an overpack. It must meet design conditions for "Type A" 87 packaging, 49 CFR 173.412(b), specified by DOT. Container (and labeling) shall be certified for a design life of at least 20 years from date of certification. Package Size and Handling: Maximum size 26" 0.D. (0.66 m) and 121" long (3.1 m) including the pintle. Must have an axial pintle, of a design acceptable to WIPP, and no other lifting devices. Waste Package Weight: Shall not exceed 8000 lbs (3636 kg). Surface Dose Rate: Not greater than 100 Rem/hr at any point. Neutron contributions greater than 75 mRem/hr shall be reported in the data package. Surface Contamination: Smearable contamination no greater than 50 x 10712 Curies per 100 cn? for alpha and 450 x 10712 per 100 cm? for beta- gamma. Thermal Power: Not greater than 300 watts per package. Nuclear Criticality: Shall not exceed 1.9 g/liter of fissionable isotopes (averaged over 5 liters with a maximum 50% void space). If such reasonable distribution cannot be assured, then the canister is limited to 240 g total (in Pu—-239 fissile gram equivalents). The canister may be loaded with DOT 17C drums which will provide inter- nal partitioning and increase (sic) the limits to 100 g for each 30 gal drum and 200 g for each 55 gal drum. Pu-239 Equivalent Activity: Packages shall contain no more than 1000 PE- Ci (Plutonium -~ Equivalent curies). Labeling: Each package shall be uniquely identified with a permanently attached number at least 2" high. 88 Data Package: Certified data provided prior to shipment shall include: Package identification number; Dated certification statement that waste content and packaging are in accord with the WIPP WAC, and the waste is unclassified; Waste generation site; Date of packaging; Maximum Surface Dose Rate; Weight; Container type; Physical description of waste form; Assay information, including PE-Ci and Pu-239 fissile gram equivalent contents; Hazardous materials (non~radionuclide) content: identification and quantity; Measured or calculated thermal power; Date of shipment; Carrier identification; Other information considered significant by the shipper. C+.2. WIPP CERTIFICATION COMPLIANCE REQUIREMENTS Source: "Draft A" of WIPP-DOE-158 dated August 1984: For unclassified, RH, TRU waste, both newly-generated and retrieved-from—storage; all other applicable DOE orders must continue to be met. Container: May be the RH container itself, or it may be the container inside an overpack. It must meet design conditions for "Type A" packaging, 49 CFR 173.412(b), specified by DOT. The container (and labeling) shall be certified for a design life of at least 20 years from date of certification. Package Size and Handling: Maximum size 26" 0.D. (0.66 m) and 121" long (3.1 m) including the pintle. Must have an axial pintle, of a design acceptable to WIPP, and no other lifting devices. 89 Immobilization: Required if more than 1 wt % is in form of particles <10 microns, or if more than 15 wt % is <200 microns in diameter. Par- ticulate desiccants (e.g., Portland cement) are exempted. Liquid Waste: Free liquids are not allowed; minor amounts in cans or bottles are acceptable. Sludges: Shall be packaged such that internal corrosion of the container does not occur. Pyrophoric Material: No more than 1 wt % may be pyrophoric forms of radionuclide metals, and these shall be generally dispersed in the waste. Pyrophoric materials other than radionuclides shall be ren-— dered safe by mixing with stable materials (e.g., glass or concrete) . Explosives and Compressed Gas: Not greater than trace quantities of explosive compounds or no vessels capable of being pressurized greater than 7 psig. Radioactive Mixed Waste: No hazardous waste is permitted unless it is also co-contaminated with TRU waste. Reactive material shall be identified. Corrosive materials must be neutralized, rendered non- corrosive, or packaged in a manner to ensure container adequacy through the design lifetime. Waste Package Weight: Shall not exceed 8000 lbs (3636 kg). Nuclear Criticality: Fissile content no more than 50 g/ft3, averaged over 5 ft3 volume and 50% void space. If uniform distribution can- not be assured, the container is limited to 240 g total (in Pu-239 fissile equivalents). 90 Pu-239 Equivalent TRU Activity: Shall not exceed a value (in Curies) TBD. The Pu-239 equivalent is based on MPC values, DOE Order 5480.1A, Chapter XI. Surface Dose Rate: Not greater than 100 Rem/hr at any point. Neutron levels shall be limited to 75 mRem/hr at the container surface(!). Surface Contamination: Smearable contamination no greater than 50 X 10712 Curies per 100 cm? for alpha and 450 x 10712 per 100 cm? for beta- gamma. Thermal Power: Not greater than 300 watts per package. Combustibility: Any combustible TRU waste shall be packaged in a noncom- bustible container. Gas Generation: All RH TRU waste canisters shall be vented, through a Rocky Flats—type carbon filter or equivalent. Color Coding: No criterion specified. Labeling: Each package shall be uniquely identified with a permanently attached number at least 2" high. Documentation/Compliance: The shipper shall have operating procedures with methods for determination of all required data. The data package shall be transmitted to WIPP prior to shipment, in a format acceptable to WIPP., A certificate of compliance shall be provided by a designated individual from the certifying facility_and be main- tained by the shipper. C.3. WIPP: GENERAL COMMENTS Source: (Excerpted from WIPP-DOE-069) 91 The facility is not yet authorized to be permanent. Therefore, retrieval must be allowed for up to 20 years under present criteria. Some experimental packages will be allowed. Out of 170 acres, 10 acres is for experimental purposes. RH TRU has a surface dose rate >200 mRem/hr, but may not be greater than Pu-239 Fissile Gram Equivalent is based on Ky ff, assuming an optimally moderated infinite array. Pu-239 Equivalent Activity (expressed in PE-Ci) is characterized by: k i=1 where there are k isotopes, A; is maximum activity of isotope i, CFy is the MPC correction factor for isotope i, obtained by multiplying the MPC in 10 CFR 20, Appendix B, Table 1, column 1 for soluble materials by 5 x 101! ml/Ci, to normalize relative to Pu-239. TRU Waste is defined as defense waste contaminated with certain alpha- emitting isotopes of atomic number greater than 92 and half-lives greater than 20 yrs, in concentrations greater than 100 nCi/g. Waste Container is the disposable containment intended for emplacement at WIPP, including any integral liner or shielding. Package is the container and contents. Waste volume percent is the material volume, excluding trapped void space, of that form compared to the total waste volume, but not com— pared to the package volume. 93 Appendix D. ANALYSIS OF MSRE CELL PENETRATIONS Prepared by D. Macdonald and A. C. Williamson Martin Marietta Energy Systems, Inc., Engineering August 21, 1985 94 D.1 INTRODUCTION The objective of this study is to review and comment on one aspect of the integrity of the MSRE facility to continue to safely store the fuel and flush salts presently located in the drain tank cell of Building 7503. This study is concerned with penetrations through the cell walls. These penetrations were evaluated in the context of overlapping containment eanvelopes composed of various cell groupings. The more than 100 penetrations into the reactor, drain tank, and fuel processing cells (Fig. D.1) associated with each proposed envelope are described by location, type, current condition, and potential for future use. The identification numbers shown on Fig. D.l indicate cell penetrations and refer to identification numbers shown on facility drawings; they are also referenced tb Table D.1 of this study. Possible closure options are proposed for those penetrations not needed for faci- lity maintenance or for future fuel transfer. These evaluations of the condition of the penetrations are incomplete at present and warrant further study to establish the requirements for any recommended course of action with regard to long-range plans for safe storage of the fuel and eventual final decommissioning of the facility. D.2 CONTAINMENT ENVELOPES While the MSRE was in operation, the fuel salt was circulated in the reactor, drain tank, and fuel processing cells. The fuel and flush salts have been stored in three critically safe storage tanks in the drain tank cell following facility shutdown in 1969. Since the three cells are interconnected to varying degrees, this study was organized in terms of three containment envelopes (Fig. D.2). Envelope 1 is composed of the reactor, drain tank, and fuel processing cells. These cells are grouped together because they contained most of the radioactivity while the MSRE was operating. Envelope 2 is made up of the reactor and drain tank cells, which are still connected via an open penetration (the penetration to the processing cell is sealed). Also, the processing cell is more accessible and more likely to be utilized in the future REF. DWG. X3E-12598-001 v P PPV o a0 Ty pe AP @0, K P oM gyt 1-V v & y Fv v a o | - T S ov.n by &N P 37-42 A REACTOR CELL FUEL_PROCESSING | I CELL Brib-’////L — DRAIN TANY CELL 43-53 I-18 Fig. D-1. Identification and location of cell penetrations at the MSRE. <6 96 REF. DWG. X3E-12598-001 N ENvELOPE T 1| | N\=r7 EnveLore 11 N EnveLore 1L Fig. D-2. Containment envelope options. 97 than is the reactor cell. Once the processing cell is accessed, the penetrations associated with Envelope 2 (between the fuel processing and drain tank cells) can be examined and sealed. Finally, Envelope 3 is the drain tank cell itself, where the fuel and flush salts are actually stored. Once the reactor cell has been opened, the large penetration between the drain tank cell and the reactor cell can be sealed. D.3 DESCRIPTION OF PENETRATIONS During reactor operation, the molten fuel salt had to be maintained at a temperature above 840°F (the melting point of the salt) in order to prevent a plug of solid salt from forming. This was achieved by surrounding salt transfer lines with electrical heaters, insulation, thermocouples, pressure reading ins;ruments, and redundant instrumen-— tation. This complex array of electrical and instrumentation lines necessitated numerous penetrations through the cell walls. 1In addition to the cell wall penetrations, there are two large cell-to—cell ope- nings: a 36—in. opening between the reactor and drain tank cells (Fig. D.3), and a 14-in. opening between the drain tank and reprocessing cells (Fig. D.4). These two openings carry insulated, heated lines for salt transfer. D.4 REACTOR CELL PENETRATIONS Typical reactor cell penetrations (see Figs. D.5 through D.9) are electrical, thermocouple, water, off-gas, and salt transfer lines. All of these penetrations pass through the sand and water—-filled annulus., In addition, the reactor cell penetrations are provided with some mecha-—- nism to allow movement relative to the cell walls. The 12-ft radius indicates the reactor containmment cell wall, which was constructed of stainless steel plate ranging from l-1/4 to 4-in. in thickness. The 15-ft radius indicates the reactor tank liner, which was constructed of 3/8-in.~thick stainless steel plate. The lines shown inside the larger penetrations are presently connected or were used as spares and left capped. ' 242 REF. DWG. X3E-12598-002 36°0D. 3 WALL THK. /HEATER { INSULATION CELL SIDE CEL o & 2] T % A a1 e Vi it WO L. N = 36 CONCRETE=H] |=— 30" =% - / S WAL /) JED ONCGRETE VY| 13 PLATE i 1 WAL Y I5°R /15 SCH. 40 PIPE ReacTor-Drain Tank INTERCONNECT Fig. D-3. \ 3?75 0D, 2 WALL THK. A~ 2-12'R Existing interconnecting penetration between reactor cell and drain tank cell. 86 REF. DWG. X3E-12598-003 g BARYTES BLOCK gfLELL LINER [ T 20 CONCRETE-—=1"T5 b9 s 7 WALL - 2 v, ;’;glfi |4 SCH. 80 PIPE b v il Y o v v v 2 »oa & 1o v D :'P ‘ L ELANGE (BOTH ENDS) ta Fig. D-4. Existing sait transfer 11ine penetration between drain tank cell and fuel processing cell. 66 REF. DWG. X3E-12598-E002 e - N o~ vl N ANNULU S 2" SCH.40 PIPE 30 DIA ; o 25 MIN. OFFSET CELL SIDE X t 4 PLATE 'R 3 PLATE Fig. D-5. Typical existing electrical, thermocouple and instrumentation pentrations into reactor cell. 00T 101 REF. DWG. X3E-12598-002 48 CELL SIDE 6 SCH. 160 PIPE 15 SCH. 40 PIPE 2R SN Ao o U g p o) / r? A » L6 SCH. 40 PIPE 5 S <" CCoNCRETE ( 2 3 pATE 8 AND § WA ANNULUS 2’ PLATE 2 I'PIPE BRACES 5 (60" APART) Sameier anp Sampier Orfons LiNes Fig. D-6. reactor cell. Existing sampler and sampler offgas line penetrations into _CELLSIDE £ SCH. 40 HMPE {TVP) Fig. D-7. REF. DWG. X3E-12598-002 39 SEAL PLATE (BOTH ENDSY) 4" PLATE ' 3 PLATE | _ 8" SCH. 40 PIPE WATER LiNes ano OFroas Lines Existing water and offgas penetrations into reactor cell. ¢0T TrsTRUMENTATION ELEcTRICAL Lines PLUG IF NOT USED ; 3'/3 TYp I i { Eet : : £ v & :,‘:’;::9 Rop g b Che e e B b = DV e T T e e & aba F b- <1 a T -',.&».‘--'b-.v". e ¥oopi: : AR 36" CONCRETE WALL sl s B segee BT LA v R v 8 - g 8"R et /% SCH. 40 PIPE & T g Sra aArtl e o Kl e e - LT o T - b X AN <5 t CEL\_ L\NER | | ¥, THREADED Dran Tank Ceul Fig. D-8. COUPLINGS REF. DWG. X3E-12598-003 Cooiant AR\ Cover (5AS LINES L pLATE * A a-. ~F :I 7T :1/_ g SCH. BO PIPE B ',?.'k."ff;f-': jf,'_-‘-‘,\ A T SN S TN S J" | ot T e e H N \ [ N N \ 36' CONCRETE WALL \\ "&& :: l % \g PLATE ‘. -?." '-,'b‘ T A"i“,.-;'p" -1 { e T A gTLer v < A e SIS ) . [‘;\_s CELL LINER \6 SCH 80 P\PE (SSTY Drain Tank CeLL :3" 'SCH. 40 PIPE Typical existing penetrations into drain tank cell. €0T REF. DWG. X3E-12598-003 7 / ' SCH.40 PIPE S L iJg SCH. 40 PIPE l<— 2 CELL LINER (5ST) Deay ANk CELL 36" CONCRETE WALL - HeLium\ WATER LINES Fig. D-9. Typical water and helium penetrations into drain tank cell. 01 105 D.5 DRAIN TANK AND FUEL PROCESSING CELL PENETRATIONS Figures D.8 and D.9 show typical electrical, cover gas, coolant air, helium, and water transfer lines. All of these lines, except for those interconnecting the two cells, pass through the concrete cell wall into access areas. As in the reactor cell, the lines are either con- nected or were used as spares and left capped. D.6 TABULATION OF PERTINENT INFORMATION Much information about the cell penetrations has been collected and tabulated (Table D.1). This table has been divided into three sections, one for each of the main cells (reactor, drain tank, and fuel pro- cessing). Table headings are: (1) penetration identification number, (2) type/usage, (3) present coundition, (4) location in the cell, (5) access area, (6) size of the major and interior penetratioms, (7) deter- mination as to whether the penetration is needed now or in the future, (8) closure options, and (9) envelope involved. The present condition of the penetrations (connected, capped, or empty) is yet to be deter-— mined; however, a detailed on—site inspection would provide most of this information. Some of the penetrations used in the facility, and their need for maintenance or for future salt transfer, were not readily available at the time this study was completed. D.7 CLOSURE OPTIONS Numerous penetrations are currently needed for facility surveillance maintenance. The solid salt is heated annually to recombine fluorine, in addition to the periodic measurements of temperature and pressure. Besides the penetrations needed for surveillance and maintenance, the facility must maintain the capability for transferring the fuel and flush salts out of the drain tank cell in the event this may be required in the future. Several closure options are available for those penetra- tions that are not needed now or in the future. Penetrations with interior lines can be sealed in one of two ways (Fig. D.10). Each interior penetration can be capped and left in place, or the end seal- plate can be removed, the interior lines cut inside the larger penetra- tion, and the seal-plate replaced with a new, solid plate. The option Table D-1. TYPE/USAGE MSRE reactor and drain tank cell penetration 1ist. 1.D. PRESENT CONDITION CELL LOCATION ACCESS AREA SIZE NEEDED CLOSURE OPTIONS ENVELOPF Drain Tank Cell 1 STEAM FROM DOMES (1) South Wall South ESA 3" in &" 1 1 2 STEAM FROM DOMES (1) South Wall South ESA 3" in 4" 1 1 3 WATFR TO STEAM DOMES (1) South Watl South ESA " 1 1 4 WATER TO STEAM DDMES (1) South Wall South ESA " 1 1 5 WATER (1) South Wall South ESA " 1 1 6 WATER (1) South Wall South ESA 1" ] 1 7 SPARE (1) South Wall South ESA " NO 1 i A SPARE (1} South Wall South ESA 1" NO 1 1 9 HELIUM (1) South Wall South ESA 1/2" in 1" 1 1 10 HELTIM (1) South Wal)l South ESA 1/2" in I" 1 1 11 HELTUM (1) South Wall South ESA 172" in 1" 1 1 12 HELTUM (1) South Wall South ESA 172" in 1" 1 1 13 HELIUM (1) South Wall South ESA 1/2" 4n 1" 1 1 14 HELIIM (1) South Wall South ESA 172" in 1" 1 1 15 HELIUM (1) South Wall South ESA 1/2% in 1" 1 1 16 HELIUM (1) South Wall South ESA 172" in I" 1 ] 17 SLMP DISCHARGE (1) South wWall Waste Cell 3/4" 1 1 ig SUMP DISCHARGE (1) South Wall Waste Cell 3/4% 1 OR 2 1 19 AIR TO SIMP (1) West Wall West of Bldg. /4" 1 OR 2 l 20 AIR TO SUMP (1) West Wall West of Bldgp. 3/4" 1 OR 2 1 21 North Wall Fuel Proc. Cell 1-1/2" 1 OR 2 2 22 North Wall Fuel Proc. Cell 1-1/2" I OR 2 2 23 North wall Fuel Proc. Cell 1-1/2" 1 OR 2 2 24 North wWall Fuel Proc. Cell 1-1/2" 1 OR 2 2 25 North Wall Fuel Proc. Cell !-1/2" 1 OR 2 2 26 North Wall Fuel Proc. Cell 4" I OR 2 2 27 North Wall Fuel Proc. Cell 1-1/2" 1 OR 2 2 28 North wWall Fuel Proc. Cell 1-1/2" 1 OR 2 2 29 North Wall Fuel Proc. Cell 1-1/2" 1 OR 2 2 30 North wWall Fuel Proc. Cell 14" p 2 A~-36 INSTRUMENTATION (29) East Wall North ESA 374" 3 1 B-36 TINSTRUMENTATION (36) Fast Wall North ESA 374" 3 1 C-36 THERMOCOUPLES (36) East Wall North ESA 3/4" 3 1 D-36 THERMOCOUPLES (36) East Wall North ESA 3/4" 3 1 E-36 THERMCCOUPLES (36) East Wall North ESA 3/4" 3 1 A-24 ELECTRICAL (24) East Wall North ESA 3/16" in 3/a" 3 1 B-24 ELFCTRICAL (24) East Wall North FSA 3/16" in 3/4" 3 1 90T Table D-1, continued. e e e e e e o o S st C-24 ELECTRICAL {(24) East Wall North ESA 3/16" in 3/4" 3 1 D-24 ELECTRICAL (24) . East Wall North ESA 3/16" in 3/4" 3 1 E-24 ELECTRICAL (24) East Wall North ESA 3/16" in 3/4" 3 1 F-24 ELECTRICAL (24) East Wall North ESA 3/16" in 3/4" 3 1 G SPARE (1) East Wall North ESA 6" NO 1 OR 2 1 H SPARE (1) East Wall North ESA 6" NO 1 OR 2 1 I COVER GAS (3) East Wall North ESA 6" (3 1/2") 1 OR 2 1 J SPARE (1) East Wall North ESA 6" NO 1 OR 2 1 K SPARE (1) East Wall North ESA 6" NO 1 oR 2 1 L SPARE (1) East Wall North ESA 6" NO 1 OR 2 1 M COMPONENT COOLANT AIR (3) East Wall North ESA 6" (3 3/4") 1 OR 2 1 N COUMPONENT COOLANT AIR (3) East Wall North ESA 6" (3 I/4") 1 OR 2 ] o COMPONENT COOLANT AIR (1) East Wall North ESA 6" (ONE 1 1/2') 1 OR 2 1 P DP CELL (4) East Wall North ESA 6" (4 1/2™) 1 OR 2 1 Q LEAK DETECTOR (12) East Wall South ESA 6" (12 1/4") 1 OR 2 1 R LEAK DETECTOR (12) East Wall South ESA 6" (12 1/4™ 1 OR 2 1 Reactor Cell Ft-in., Degrees I REACTOR LEAK DETECTORS (60) 836 15 South ESA 24" (60 1/4™) NO 2 0R 3 1 Il ELECTRICAL (44) 834 30 South ESA 24" (6 1";38 3/4") NO 2 0rR 3 1 111 ELECTRICAL (44) B36 45 South ESA 24" (6 1™;38 3/4") NO 2 OR 3 ) v THERMOCOUPLES (60) 834 60 South ESA 24" (7 3/4";28 1/2";25 3/8'")HNO 2 OR 3 1 v INSTRUMENTATTON (60Q) 836 75 South ESA 24" (60 3/8™) NO 2 0R 3 1 VI SAMPLER OFFGAS 847 110 High Bay 4" (2 1/2") NO 2 1 V1l SAMPLER 847 115 High Bay 6" NO 2 1 VIII FUEL PUMP AUX. PIPING (8) 836-9 125 Service Tumnel 18" (3 1/2";4 374";1 1) NO ? 1 X NEUTRON INSTRUMENT TUBE 834-5 145 High Bay 36" (10 ION CHAMBER GUIDES) NO ? 1 X FUEL PUMP LIQ. LEVEL (6) 844-6 155 Special Eq. Rm. 4" (6 1/a™) NO 2 ] X1 FUEL PUMP AUX. PIPING (13) 836-9 160 Special Eq. Rm. 18" (2 1";11 1/2") NO ? 1 X1t COMPONENT COOLANT AIR (1) 829-10 165 Special Eq. Rm. 6" NO ? ] XIIT COOLANT SALT TO HX (1) 840-10 170 Coolant Cell 24" NO ? i X1V WATER LINES (2) 839-9 185 Coolant Cell 8" (2 1" NO 1 OR 2 1 XV SPARE (2) 839-9 200 Coolant Cell 8" (2 2") NO 1 OR 2 1 XV1 WATER LINES (2) 839-9 205 Coolant Cell 8" (2 2") NO 1 0R 2 1 XVII WATER LINES (2) 839-9 210 Coolant Cell 8" (2 2™ NO 1 OR 2 1 XVIII WATER LINES (2) 819-9 220 Coolant Cell 8" (2 2™ NO 1 OR 2 1 XIX COOL. SALT TO RADIATOR (1) 837 220 Coolant Cell 24" NO ? 1 XX OFFGAS (3) 839-9 225 Coolant Cell 6" (3 1™ NO 1 OR 2 } XXI OFFGAS (2) . 839-9 230 Coolant Cell 6" (TWO 1 1/4") NO . 1 OR 2 1 XX11 CELL EXHAUST DUCT (1) 824-10 245 CDT Tunnel 30" NO 2 1 XX1IT THERMOCOUPLE (38) 816 325 West Tunnel 24" (38 3/8™) NO 2 0R 1 1 XX1v DRAIN TANK CEL1. INTERCON. 825-2 330 Drain Tank Cell 36" NO 2 3 XXv SLMpP Bottom Center NO ? 1 L0T Table D-1, continued. et P A FUEL PROCESSING CELL 31 32 33 34 35 36 37 a8 9 40 4] 42 43 Q4 45 46 47 48 49 50 51 52 53 54 DRAIN STEAM LINE STEAM JET DISCHARGE STEAM LINE STEAM LINE DRAIN DRAIN ULTRASONTC PROBE INSTRUMENTATION INSTRUMENTATION INSTRIMENTATION SALT ADDITION LINE Tt AL bt e, gk e - e o NORTH WALL NORTH WALL NORTH WALL NORTH WALL NORTH WALL NORTH WALL EAST WALL EAST WALL EAST WALL EAST WALL EAST WALL EAST WALL WEST WALL WEST WALL WEST WALL WEST WALL WEST WALL WEST WALL WEST WALL WEST WALL WEST WALL WEST WALL WEST WALL ROOF DECON. CELL DECON. CELL DECON. CELL DRAIN TANK CELL DECON., CELL DECCON. CFLL SPARE CELL SPARE CELL SPARE. CELL SPARE CELL SPARE CELL SPARE CELL HIGH BAY ABSOR. ABSOR. ABSOR. ? ? ? INSTR. INSTR. ABSOR. INSTR. CUBE. CUBE. CUBE. CUBE. CUBE. CUBE. CUBF.. HIGH BAY 6" 3" 374" 34" 374" 3/&" 6“ 12“ 12" 12" 6" 1 172" YES YES YES YES YES YES YES Gt g bt gt Gt e i et et e et Bt et et Gt T e e ed b e OR OR OR OR OR OR OR CR OR OR OR OR OR OR OR OR OR OR OR OR OR OR OR B R R RN NN RN NRNNRNRRNRNNRNNLDNRNDN s it (b Wi bt Bt e S e e e ems Mme i e ot A mae R e s e 80T Needed 1ine (connected Unneeded Tines (capped)— . / f 01d seal plate - ] / ! N “1= . -.'"\ GR.EL.840-0 le 3 TR H~ s \ . b? L r "L )--() f'.-.’_‘;.);.\ f 'y Al vt g :-".-_{. 3N tdy ~ A - , ?l 7 " A 3‘ i o DRAIN TANK CELL tfl ' ~N 9 . &= 2g onte - e T“ ~,‘| * ” ?”\ - .~ i ‘ - ‘.J V-f‘ . 4' J’.'—Q J‘)_ CRma = ~ * i -1 A Sl o2 R n s N £ 23 ey L Tl e 1 e ) - AP Fig. E-3. looking north. Section of drain tank cell foundation showing french drain, excerpted from ) o D-KS-19040A8, R4 ! [ . . I TTTTTYTO 7 : ’ | Ly 3 | e \ - VoL Y ELEV. 840l0” . < v = L PF T o — - - — -~ - = —} - _4. At b - . - N q. _ - EONL Y T N L ~ 0 vy ! -7 S g — — .- . ' I L% C j I[ . e ; P t3 conteoL i ! i -k TUNNEL N | "i L L | H.RELEY. B30 LT o T I. \', C - i o r 3 Ll 1" ELev. 816 6% ) || O “'! R "“-_‘,“’-I.':'..‘\Fa'.'-'~i,ln ';"..\_'L__! 1 . L N ,“\.| » "o"-"n. - S ""' r'.v". - Ecase e - ‘."'-. e T g ..“:\ v ‘,»‘ -’-’,’_.\", ' ’ Fig. E-4. Section of reactor cell foundations showing french drain, looking north. (ART configuration) 0ZT excerpted from .t Sm— =TT e e P AREPb b & I -CD"' D-KS-19040A8, R4 /S ELEV. B85 ...\i_-. e ... , - u_ rmfi#mfiwsifiythfiJan!!wfihmufmm -l ll, e * i l.-"rlilfllvl.nll-!.. - -4 P 7 PR ST AT T Wty 4..“..,..1.:..,1.‘.;?..»!!.!!!1 R I, O S N e e ey e e 121 - eyt © oy 41 B2 - N w VTR M ENT £ s o 2 A T " P )" b . v - * 7 g » 2 - - v G [ —— | o , - = b - - v - 1 - b s e ety - s :ma .Tlsw..s. s e ey -1 o s ———— 1 e e e ' —\A N o e e e - — e e - Mo L P 1 o LB _...l._,-,.r Lo ? L ¥ L. i | - 2 ) ¥ e e R e e e 2 W LB & 7 1 T * _l ] 5“.‘ . 4 .‘U. %. —_ L - . | _ &L ol ) i 0 7 1 |- _ > ——= 4 w = Aoy . Nmu.r.ll-l > a | Clm e o e e s < N ! \ N \ “fly v# ¢ , m . | . i e A ) (ART configuration Section of south building foundations showing sump and french drains, looking north. Fig. E-5. 122 Several lines entering the pump room flow into a 55 gal. drum rather than entering the sump directly. The drum is used to collect drainage from areas which might have contained small levels of activity, and allows for monitoring prior to release. The "pit pump" could be used to send drainage colleted in the drum to the radioactive liquid waste tank in the event that activity is detected. Water from the sump is pumped up to a catch basin southwest of the building, and flows out into a field through a 12 in. concrete line. As shown in Fig. E-1, the 3 in. pump discharge line branches into an 8 in. cell annulus drain line before circling around the charéoal absorber pit en route to the catch basin. The reactor cell annulus can be emptied by pumping to the catch basin via line 331 using the pit pump. This operation would remove a major reservior of water adjacent to the cell in which the salts are now stored. Caution in disposal of this water would be dictated by the presence of chromate rust inhibitors, as well as the possibility of induced radioactivity. Storm Water Dispersal Collection of storm water is accomplished with a standard roof drain piping system. Water flows by gravity to a catch basin, separate from the one serving the building sump, from which it flows out another 12 in. concrete line (Fig. E-1). Site grading and ditches are used to carry storm water from around the perimiter of the building. 123 Sanitary Sewage Collection and Disposal A standard sanitary sewage collection system was installed in the building. Sewage flows by gravity to a septic tank located west of the building. The water leaving the septic tank then flows into a drain field west of the septic tank. .y Appendix F. 125 GEOLOGIC INVESTIGATIONS RELATIVE TO MOLTEN SALT REACTOR DECOMMISSIONING Dr. D. W. Byerly, Department of Geology University of Tennessee, Knoxville, Tennessee December 1984 126 Geologic Investigations Relative to Molten Salt Reactor Decommissioning Dr. D. W. Byerly, Department of Geology University of Tennessee, Knoxville, Tennessee The Molten Salt Reactor Experiment (MSRE) facility is located in the Melton Valley area of the Oak Ridge National Laboratory (ORNL). The valley is bordered by Haw Ridge (elevation 317 m) on the northwest and Copper Ridge (elevation 400 m) on the southeast. Bulldings housing the reactor and other facilities related to the reactor are situated on the northwest side of the valley near the base of Haw Ridge. Topographically, the valley is not flat, but instead consists of relatively steep-sided, irregularly-shaped, low hills with summit elevations averaging about 275 m. Relief in the valley ranges up to 43 m. The buildings are at a ground elevation of about 259 m within a slight swale (partially resulting from construction back-fill) near the head of a draw which is tributary to Melton Branch that flows southwesterly through the main valley. All drainage from Melton Valley ultimately enters the Clinch River below Melton Hill Dam southwest of the confluence of Melton Branch at White Oak Creek. The geological conditions of portions of Melton Valley have been interpreted as favorable for storage/disposal of low-level radioactive wastes. The facilities in Melton Valley used for hosting the ORNL- generated low—-level radiocactive wastes include: burial grounds, waste pits, treatment operations, and hydraulic fracturing. Bedrock forming the foundation for the MSRE consists of the lower portion of a "package"” of rocks referred to as the Conasauga Group. Rocks of this group grade compositionally downward into the Rome Forma-~ tion underlying Haw Ridge and upward into the Knox Group which forms Copper Ridge. Variegated shales, commonly calcareous, comprise most of the lower Conasauga; however, intercalations of silty to pure limestone may be present as lenses. The upper portion of the Conasauga, on the southeastern side of the valley where gradation with the Knox Group occurs, generally contains more limestone. 127 The knobby topography in Melton Valley is typical of weathered and eroded shale. The knobs or hills are aligned in rows paralleling the regional structure, forming above the shaly bedrock, whereas the interven- ing low areas are ordinarily underlain by the carbonate rocks. Regolith above the bedrock is variable in thickness. Generally, it is thickest (up to 9 m) on hill crests and thinnest (less than 1.5 m) in the flat, low-lying areas. The total thickness of the Conasauga Group is about 600 m. The trace of the Copper Creek fault, a major thrust fault in the Ridge and Valley province of the southern Appalachian orogen, strikes NE- SW along the northwest slope of Haw Ridge. The fault dips southeast projecting below Melton Valley and Copper Ridge at considerable depth. Sedimentary strata of the Conasauga Group forming Melton Valley also strike northeast and dip to the southeast, forming the hanging wall of the Copper Creek fault. The fault and related deformation of the rocks in the valley resulted from tectonism that occurred over 200 million years ago. Subsequent deformation of these rocks is not evident. Considering the present tectonic setting of North America and the fact that the fault as well as structural elements have not been active, at least during the most recent 70 to 80 million years, the area can be considered to have a low seismic risk (seismic risk zone 2, Modified Mercalli intensity less than V or VI). The nearest significant earthquake epicenters historically are Charleston, South Carolina, and New Madrid, Missouri; however, both are considered too distant to have or have had serious detrimental impacts on this area. Joints or fractures within the Conasauga that are associated with past tectonism have not appreciably affected the strength of the bedrock, but have resulted in a secondary effective porosity which influences the geohydrology of the valley. In Melton Valley, the zone of ground water saturation generally occurs at greater depths beneath summits than in the low-lying bottoms. In low areas such as adjacent to perennial streams like Melton Branch or to ephemeral streams in the gullies dissecting the low hills in Melton 128 Valley, the water table may be found within l.5 m below the land surface. On hill tops, the depth to the water table may exceed 6 m. The general pattern of ground water movement is laterally along bedding strike (NE-SW), but jointing often modifies this pattern. An average rate of ground water movement within the shales of the Conasauga has been deter- mined to be 15 cm/day. Ground water behavior at the site is a key factor which must be addressed prior to deciding the disposition of the radioactive materials contained in the MSRE facilities. A hydrologic monitoring program should be developed to establish baseline qualitative and quantitative water data. Water is an important factor for at least two reasons. One, it can potentially deteriorate the waste containers, especially if the geochem- istry of the water is reactive, and two, water is the most likely vehicle for transmitting the radioactivity to the biosphere should containment fail, Al though the MSRE facility is apparently sound structurally, it must be kept in mind that the facility was not designed for long—-term entomb- ment of radioactive wastes. Multilevels of barriers, including engineered as well as the natural geologic setting, are necessary for safe, long-term storage/disposal of radioactive wastes. Therefore, several geological aspects of the facility site need more assessment of a site-specific nature before on-site storage/disposal is considered a feasible alterna- tive. To be acceptable, the site should be capable of safely containing the wastes without institutional controls (i.e., sump pumping of infiltrated water, etc.). Among the geologic aspects that should be included in further evaluation of the site include: ® Characterization of all earth materials on the site (anthropogeneous material, bedrock, etc.) ® Site design — slopes, placement of fill soils, underdrains, founda- tions, etc. ® Site hydrology — water inventory including quantity and quality, ground water flow nets, piezometric levels, etc. ' Site-specific investigation of the above should be considered as the very minimum of any program plan for the evaluation of the site for on-site entombment of the MSRE radioactive materials. 44, 45"49. 50_540 55-56. 129 ORNL/TM-9756 INTERNAL DISTRIBUTION Je. F. Alexander 26, F. J. Peretz T. W. Burwinkle 27. L. P. Pugh K. W. Cook 28-32. T. H. Row B. L. Corbett 33. J. H. Swanks Ne. W. Durfee 34, R. E. Thoma Je R. Engel 35. W. W. Thompson D. E. Ferguson 36. J. R. Trabalka P. N. Haubenreich 37. V. C. A, Vaughen F. J. Homan - 38, R. G. Wymer J. M. Kennerly 39. Central Research Library L. E. McNeese 40. Document Reference Section T. E. Myrick 41. Laboratory Records K. J. Notz 42, Laboratory Records - RC P. T. Owen 43, ORNL Patent Section EXTERNAL DISTRIBUTION Office of Assistant Manager for Energy Research and Development, DOE-ORO, P.0O. Box E, 0Oak Ridge, TN 37831 R. N. Coy, Office of Surplus Facilities Management, UNC Nuclear Industries, Post Office Box 490, Richland, WA 99352 C. E. Miller, Program Manager, Surplus Facilities Management Program, Richland Operations Office, U.S. Department of Energy, Post Office Box 550, Richland, WA 99352 Technical Information Center, Post Office Box 62, Oak Ridge, TN 37831 - A B c D E : ' 60# STEAM HEADER IN VENTILATION STACK FILTER HOUSE —- TO SEWER IN WEST TUNNEL v FILTER PIT i ' L {— CV3I0 3-40-5— (UNDERGROUND ) TO ~ 50N v AIR DUCT FLoorR—(352> | VENT . CATCH BASIN RADIATGR STACK r—— LANT CELL —=—=—=—====——— o ==|-a= o - 8-40-5—% gggc.,,!lts E_@mcfié W% TO L~-566 BLOWER HOUS : c\ “{ (§3D-—FROM CCPIEASER BLOWER HOUSE RAMP=—L 338 )t [ CHARCOAL @IB-FROM G C WEST TUNNEL ol : BED CELL ; . 1w e . : ! SERVICE RCOM- (3580 . SERVICE TUNNEL— 353 )——m .- REACTOR CELL 1% SERVICE TUNNEL——C350 A BN ANNULUS e REACTOR CELL——38I 8 0 . CELL h DF FRENCH ciiaR, BED CELL—! suMP TO 935 = P YvonaXyv 0 SE 7503 vire V905 vi25 2:40-5 17 CONDENSATE VENT HOUSE VALVE—(365 < o~ COLLECTION PiT DRAIN ol wm| wm] 2 o] ] |owe TANK ™| (CCT-t w [ > > cv337| o V329 a \1o T /‘,* x \1:995/\ 3-40-$ S,f) Fl A el 1] R 2 33l sl3lale y330 CvY330 4-40-5~ \. L —2-a0-ss : el ml=]n " SUMP PUMP DATA VIS T CAPACITY 75 GPM Cv3zs . Y HEAD a0 NT 2_40-55—" MOTOR 1 HP o o rfsrsw SUPPLY @ = NORTH OF | BLDG. 7509 $ ‘ AN | R . TYYY SPECTROMETEJR ROOM V3LIA TS 3 ) — N SUMP 40# STEAM SO R | V339 TO NON-EXISTENT SUMP {N COCLANT DRAIN CELL ‘ T CVZ;ZI RMPC SUMP Vg2t 321 I > FUEL PROCESSING CELL SUMP ] I 3PAKE CELL SUMP | v3178 ) , LIQUID WASTE STORAGE TANK LD LCCAL- IN 3 (WT) RMP CELL VENT TO DUCT 940 IN hie EQUIPMENT STORAGE [ _ (LASC) CELL . EQUIPMENT STORAGE - a2 CELL SUMP ) THIS DRAWING REF! v3iz e T 401 &S @'!flf’:{ : ORI OF 3120 CMTJELLC ‘?gUBBER MATERIAL 304 STANLESS STEEL Wi SRT T | SiZ€ W' x (6 HIGH : CHANGES TRANSMITTER (FUEL PROCESSING CEL CAPACITY 1000 GAL ROCM v348 | HEAT LINE g28 - : pate_12:18:6& s - oot ‘ iIN SPARE CELL i LiQUID WASTE CELL " Lvies ‘ v3a1a PANEL , NP ARE A i SEEETESEY IAGESTERSNY EUEIRNSESEE SUMMARETE $ TENCELEEANY RTINS GRONTRANE R E——— LIA , WTC 15 PSIG STEAM A B " c . o £’ ‘___-T.‘\ v 4‘.\:‘:’--"' *‘..;:.E._..,, ey i s itk ‘_‘J_M — e e e e o __‘ - e o e e A,,;_ ;ai : I i . | P l — Sow i A.'."';._; - it AR s e e Y 5 F G H 1 J o TRAP PTO EXLIGT is + N 2" DRAIN [ 1-40-33 ———{— Pi Xv3osc xV‘OGD - i CREEK b I x wr-:sr TRANSMITTER i TUNNEL V9658 1-a0-5s I v DECONTAMINATION | l NG TANK s . o ? WEST vl : @ - 57"1.D. X 16" 0" HIGH . OF CELL i [ . YN i REACTOR: CELL SUMP ____l ‘ 20M i, : OM REACTOR AND : v96358. MANIFOLD ! AIN TANK CELL PRV'S i l ELECTRICAL S CE voesB cv342 WATER ROCM L . . ‘ OM HOT SINKS semmeny DRAIN TANK V9668 - ¢ -~ - 60" STEAM _ Y . cy——i~4D-SS-‘———d> I | REMOTE MAINTENANCE PRACTICE CELL WEST | V346 @ wALL — . LLEJ.E) I ; OUTSID : s @D 180 CFM— - TO OFF GAS : F . FILTERS 3 ‘,g * V3458 J WASTE BLOWER i " § SAMPLE g : 8OMB (w) | - o OF CELL I Gos> : TO CENTRAL i t—2-4o-ss WASTE 1 STATION © . i WASTE FIiLTER DATA v-306E v i 4 @ v3058 | SIZE 36" ¢ o v3c0 x o FILTER MEDIA SAND ZZ > . WASTE PUMP DATA 5 DESIGN FLOW 30 GPM . i max. Ap s PSt TYPE CANNEgP :QTOR i o . o5 CAPACITY |4 ) _ = HE AD 80 V307 (g : — 2 e : » (m) WASTE FILTER J‘f Wl FUEL PROCESSING SYSTEM D-AA-A-30887 * COOLING WATER SYSTEM D-AA-A-40889 g4 4 3_ 40-S5 LIqUiC WASTE SYSTEM INST. APPLICATION DIA. |D-Aa-B-4050e |® i 7 | FUEL SYSTEM DAA-A-40880. i @ > [CRAIN TANK SYSTEM. D AA-A-40882 : * of GAS SYSTEM B CONTAINMENT VENT. = 1D-AA-A-40883 | %pr 2% - 40-55— .t ' © REFERESCE DRAWINGS DWG. NO. R ' | ) pren. gort: e e St 1o, OAK RIDGE NATIONAL LABORATORY : i .‘.’fi /V/'(““ At L 258 CTHERWIST . % 3418 6 |SEE DCN 3418 SiG. %é&&% i ?5L 63—"'%{’ FROJICT INSR, BEEiCH TRGINIER ToLEmARCE uaL LlQUlD WASTE SYSTEM . ' ' 3306 5 | SEE DCN 3306 T 58 - acnon e s 3102 D [SEE OCN 3102 B-63pim £ e 7 rescrons PROCESS FLOW SHEET 2930 | C |SEE DCN 2930 ikl /‘L,éé..u 77 e = M.S.R.E. 266! B [SEE DCN 2861 ”,;g.‘ (ff.’ T : AAA-40888 249} A |SEE OCN 249i L : . #, : NONE JoB 433-1.0 D-AA-A~ ’ : ~ EArAE: | | Butle 8Ly [Py SEALE? MO THIS DESIGM A8 BETURN FPINLT TG THE PROPERTY OF UNION CARBIDE NUCLEAR COMPANY — DIVISION OF UNION CARBIDE CORPORATION e i o, 4 k. iy a - e v A ){‘ S e g e o i A P50 7T94RA 023D G ", R e N R . | B c. C . : - - . . . i ran i B e e Ak T