o eS OAK RIDGE NATIONAL LABORATORY WY 7Y MARIETT. OPERATED BY MARTIN MARIETTA ENERGY SYSTEMS, INC. FOR THE UNITED STATES DEPARTMENT OF ENERGY 2 ORNL/TM-9780/V2 Nuclear Power Options Viability Study Volume I, Reactor Concepts, Descriptions., and Assessments D. B. Trauger D. White . T. Bell S. Booth |. Bowers C. Cleveland . G. Delene ri Gat . C. Hampson J J R H J J U D Jenkins L. Moses E. Pasqua L. Phung . Spiewak E T D P D | R. E. Taylor APPLIED TECHNOLOGY Any further distribution by any holder of this document or of the data therein to third parties representing foreign interests, foreign governments, foreign companies and subsidiaries, or foreign divisions of U.S. companies should be coordinated with the Deputy Assistant Secretary for Reactor Systems, Development, and Technology, U.S. Department of Energy. Releasad far anntuncement in ATF. Gietribution fimiled to partisigants in the LFFBR program. Others request from R34T, DOE Printed in the United States of America. Available from the U.S. Department of Energy Technical Information Center P.O. Box 62, Oak Ridge, Tennessee 37830 This report was prepared as an account ot work sponsored by an agency of the United States Government Nesther 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. Reterence 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 thereof & £, -y 5\ - LGS Rige® ORNL/TM--9780/V2 Department of Energy | Technical Information Center TI87 025578 P.O. Box 62 Qak Ridge, Tennessee 37830 MA-28:WDM To Addressees HANDLING OF APPLIED TECHNOLOGY REPORTS The purpose of this memorandum is to reiterate the necessity to protect Applied Technology (AT) reports in the UC-79 (Liquid Metal Fast Breeder Reactors), UC-83 (Nuclear Converter Reactor Fuel Cycle Technology), and UC-86 (Consolidated Fuel Reprocessing Program) categories from unauthorized release in order to preserve their trading value vis a vis international exchange agreements. 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Access to AT reports within a lab- oratory or contractor facility should be controlled so as not to vitiate the intent of the above policy. This is particularly important in those cases where foreign nationals are visiting, assigned, or empleyed at a facility. The TIC official AT Standard Distribution Lists that have been approved by Headquarters are considered to be the sole distribution for AT reports, with the -2- exception of internal recipients (not subcontractors or outside program participants); when an AT report is originated by an organization, internal distribution within that organization may be made directly. Where external distribution outside that organization is involved, either foreign or domestic, only the TIC lists may be used for an AT report. Any exceptions to this sit- vation will require written approval of the responsible Headquarters Program Office. You are also reminded that AT reports are not to be presented, referenced, or form the basis of presentations of information in technical society meetings or journals, meetings with foreign interests, or other means without Headquarters Program Office approval. B Wil William D. Matheny Chief, Control Branch Document Control and Evaluation Division ORNL/TM-9780/V2 Distribution Category UC-79T (Applied Technology) NUCLEAR POWER OPTIONS VIABILITY STUDY VOLUME 1II, REACTOR CONCEPTS, DESCRIPTIONS, AND ASSESSMENTS - "y ights. Refer- , recom- ts use would not infringe privately owned ri D. B. Trauger, Editor Je Do White J. T. Bell R. S. Booth H. I. Bowers Je Ce Cleveland J+ G. Delene U. Gat D. C. Hampson T. Jenkinsl D. L. Moses P. E. Pasqua2 D. L. Phung3 I. Spiewak" R. E. Taylor1 y legal liability or responsi- tion, apparatus, product, or y thereof. The views ponsored by an agency of the United States ent nor any agency thereof, nor any of their » process, or service by trade name, trademark necessarily constitute or imply its endorsement d States Government or any agenc herein do not necessarily state or reflect those of the DISCLAIMER agency thereof. express or implied, or assumes an completeness, or usefulness of any informa lTennessee Valley Authority 2The University of Tennessee 3professional Analysis, Inc. “Consultant , or represents that i ence herein to any specific commercial product manufacturer, or otherwise does not Date Published -~ September 1986 Government. Neither the United States Governm This report was prepared as an account of work s employees, makes any warranty, bility for the accuracy, process disclosed mendation, or favoring by the Unite and opinions of authors expressed United States Government or any Prepared for the Office of the Assistant Secretary for Nuclear Energy U.S. Department of Energy Prepared by the OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37831 operated by !4 MARTIN MARIETTA ENERGY SYSTEMS, INC. ffij \ £ }gp F for the U.S. DEPARTMENT OF ENERGY under Contract No. DE-AC05-840R21400 Released for annoum:fme?l In ATE. Bigtribution !imned_ E; pariicipants in he Lmir R pragran. Others request 110 asot, DOE, kfi} L PREFACE The Nuclear Power Options Viability Study (NPOVS) was initiated at the beginning of calendar year 1984, The objective of NPOVS was to ex-— plore the viabilities of several nuclear options for this country for electric power generation after the year 2000. The study emphasized technical issues but also considered institutional problems. Innovative reactor concepts were identified which may be marketable at the time when studies show that the demand for new electrical energy capacity will dincrease significantly. These concepts were considered with em-— phasis on cost, safety features, operability, and regulation as well as research needs. The study 1s reported in four volumes. Volume I is an executive summary. This report, Volume II, provides descriptions and assessments with respect to criteria established in the study of potential nuclear power plants which could be deployed early in the next century. Volume III, Nuclear Discipline Topics, provides supporting analyses; and Volume IV is a bibliography containing approximately 550 entries, A detailed outline covering all four volumes is given in Appendix A, The study was initiated by Oak Ridge National Laboratory (ORNL), which, recognizing the need for a broad base of knowledge and exper-— ience, engaged the Tennessee Valley Authority (TVA) and The University of Tennessee to participate as partners. TVA concentrated its efforts on evaluation of the concepts and on licensing. The University of Tennessee assisted in the evaluation of construction costs and public opinion issues. Both institutions contributed extensively to the evaluation of issues and in review of reports. In conducting the study, the authors extensively contacted segments of the nuclear industry for current information concerning the concepts studied and for other valued assistance, Many of the problems encountered by the nuclear industry are insti- tutional in nature and are related to the way the utility companies, designers, constructors, and regulators are organized and function. Although this study attempted to identify those institutional factors, it has not addressed them in all aspects. It was observed that the in- stitutional problems derive in some measure from technical aspects, which, in turn, originate at least in part from the large size, com- plexity, and exacting requirements for existing nuclear plants. Emphasis in the study was placed on technical aspects that have poten- tial merit and on improved design concepts that may help or have promise of helping to alleviate institutional problems. TInstitutional factors related to market acceptance have also been surveyed and studied. Con- sideration of additional institutional factors is thought to be desir- able, perhaps necessary, but is beyond the scope of this study. The study emphasized criteria by which nuclear power reactors can be judged and which are thought to be appropriate, at least in part, for iii judging future commercial viability. Other design or operational needs that are important but are more difficult to quantify are presented as either essential or desirable characteristics. Several innovative reactor concepts are described and evaluated with respect to these measures. Related and generic information on construction, economics, regulation, safety and economic risk, waste transportation and disposal, and market acceptance which supplements the evaluation is included in Volume III. ' This study differs in several respects from other studies concern- ing the future of nuclear power in the United States. The first is the time frame of interest. The NPOVS effort was focused on a time frame a little later than most studies, the years 2000 through 2010. For the near term, existing Light-Water Reactor (LWR) designs, or evolutionary modifications to them, would be the most likely nuclear choices if there is a sufficient demand for increased electrical generating capacity. Projections by the electric industry indicate that new base load capac- ity will be needed before the year 2000. Therefore, it is probable that decisions to order baseload capacity will be made by 2000-2010 and, furthermore, that the reactor concepts discussed in this report have the potential for competing with existing IWR designs or coal-fired plants at that time. For the more distant future, nuclear plant concepts incorporating more innovative, if not revolutionary, features could be the best choices. A second aspect making this study different is the level of tech- nical detail in the evaluation of the specific designs. Significantly more design information was generated by all the nuclear designers in- volved with innovative concepts in the last three years, and much of this information was made available to NPOVS. Recognition has been given to the special features of each concept and thus to the role that each may achieve in a mature nuclear economy. Systematic development of the information presented in this report was completed in September 1985. Delays in funding and review have pre- vented timely publication. An attempt has been made to include new information where substantial changes 1in programs or designs have occurred, but it has not been possible to bring the report fully up to date. Subsequent developments and events, particularly the Chernobyl accident, may alter some of the findings. iv ABSTRACT SUMMARY OF FINDINGS 1. INTRODUCTION BACKGROUND 1.1 CONTENTS e O 6 5 0 s s 0 b s b e PO USSP OS S e sSSP0 NSRS S ®® 8P SN BSOSO E PSP OSSR P YRR Na 0 0 & 0O P eSO SR NES SO EP NSRS NNPESEO eSS 1.2 REPORT ORGANIZATION 84 for a single Power Pak. PRISM High commercial grade to reduce cost. The reactor module and refueling equip- ment are of nuclear safety grade. Refueling using a mobile refueling machine which moves from one module to the next. Reprocessing and refabrication may take place either on—- or off- site. Reactor modules are shop fabricated and assembled and are rail shippablie. In ad- dition, the intermediate sodium loop, the steam generator, and other BOP systems will be modularized and factory produced. Each module has its own containment. 32.5 88 estimated; 80 used in economic assessments, LT-¢ Table 3.1. selected for initial investigations (continued) Design and construction characteristics of the IMR designs Design and construction characteristics LSPB Loop™ SAFR PRISM Plant Lifetime (yrs) 40 60 40 (but reactor modules can be replaced at relative low cost) Core Design Characteristics a) Type Heterogeneous Heterogeneous Homogeneous b) Height (meters) 4.83 3.25 1.76 ¢) Diameter (meters) 5.71 4,04 1.93 d) Resident time in 3 (4 for advanced core) 4 4 core (yrs) e) Refueling Intervals 1 1 1 (yrs) f) Cover gas Ar He He Reactor Vessel a) 1ID (meters) 14.6 11.9 5.8 {(containment) b) Height (meters) 19 14.5 19.5 Burnup (MWd/kg) 109 158 107 Breeding Capability Doubling time of 25 years for breeder core reload. Breeding is not required for the initial core. System will need only a feed of U-238 since Pu~239 needs will be supplied by conversion. Breeding ratio of 1.04 for oxide fuel and 1.22 for metal fuel. *The LSPB pool concept has similar characteristics but offers a higher power level and plant efficiency It also will require a larger vessel (19 m ID, 21 m H). [1350 MW(e) and 38.5%] and improved shutdown heat removal. 8¢t 3-29 design which is a rectangular, steel-lined, concrete building with roof hatches for construction and maintenance. Adjacent nuclear island buildings are integral with the containment thereby providing cost- effective containment and confinement capabilities. The extent to which the LSPB design has achieved lower capital costs is suggested by comparison to the CRBRP design. The LSPB plant, while producing four times the net electrical power of the CRBRP design, occupies a nuclear island which is physically smaller than that of CRBRP. Finally an option has been maintained to use a fuel designed for cost performance by reducing the breeding specification. These design studies indicated that fuel cycle costs could be reduced by about 3 mills/kWh. Design modifications under consideration could enhance inherent protection for failure-to-scram events through temperature-induced ex- pansion of control rod drive-lines or temperature-induced control rod releases or other Self Actuated Safe Shutdown (SASS) type devices. Calculations are being conducted to identify the design measures needed to assure no boiling for a loss of flow with trip failure. 1In addition, the LSPB decay heat removal capability is enhanced by incorporating the capability for natural circulation in the normal heat transport sys- tems. These design features should increase the licensability and ac- ceptance of the plant by the public and the utilities. Additional design features associated with safety and licensability include the use of two independent and diverse reactor shutdown systems and the use of two independent and diverse, safety-grade, decay heat removal systems. One of these decay heat removal systems consists of two, forced-circulation loops and the other is a passive, natural circ- ulation loop. Both decay heat removal systems use the outside air as their ultimate heat sink and sodium in the reactor vessel as the heat source. The LSPB also utilizes a heterogeneous core design. Because of these enhanced safety features, the LSPB balance of plant (BOP) design has been downgraded from safety grade to commercial codes to obtain cost reductions and enhance constructibility. 3e4.3.2 Sodium Advanced Fast Reactor (SAFR) This plant, being designed by the team of Rockwell International, Bechtel, and Combustion Engineering, for the U.S. Department of Energy, consists of one or more independent power generating units called Power Paks, as illustrated in Figure 3.5.13-16 The utilization of multiple units at one site permits cost savings through sharing certain facili- ties and services. These shared facilities include the control build- ing, the plant service building, the nuclear island maintenance build- ing, and the fuel cycle facility if colocated with the power plant. A major goal for the initial SAFR design effort was to determine the Power Pak power level, and therefore size, which is the optimum trade-—-off of cost, passive safety, utility acceptance, licensability, and constructi- bility. Factors which influenced the selection of the 350-MW(e) size included short construction times, low investment risk, economy of scale, and moderate energy costs. For the basic design configuration of each Power Pak, Rockwell made effective use of their previous IMR design 3-30 ORNL—-DWG 86—4051 ETD Rockwell Inter- Source: "SAFR Discussions at ORNL, Power pak elevation for SAFR. Fig! 3!5! national, 1985." January 11, Rocketdyne Division, 3-31 experience, particularly that associated with the Large Pool Plant (LPP).17 Advanced LMR technology and enhanced passive safety features introduced into the design are listed as follows: (1) metal fuel and its associated reprocessing innovations have been retained as an option, (2) redundant and passive decay heat removal systems have been employed, (3) a relatively high primary system temperature was selected with the use of an advanced material, 9 Cr—l Mo, for the entire intermediate loop, (4) a backup, self-actuated shutdown system has been included, and (5) heterogeneous core designs with self-regulating characteristics have been incorporated with the objectives of limiting the potential effect of hypothetical accidents. As indicated in Table 3.1, each 350-MW(e) Power Pak consists of a reactor vessel, primary and intermediate heat transport systems, a steam generator system, and a turbine generator. Safety-related systems and components are minimized and localized in the design such that nuclear safety is decoupled from the BOP and Intermediate Heat Transfer System (IHTS). The reactor assembly is factory built and barge shippable. It contains the primary system and a spent—fuel storage rack. Fuel trans- fer is by a hoist mechanism and rotating plug which is part of the ves- sel head closure. Included in the primary system are the reactor, two inducer-type primary pumps, and four intermediate heat exchangers (IHXs). In each of the two independent, intermediate loops, non-radio- active sodium is circulated through the IHXs and a booster-tube, hockey- stick steam generator operating in the once-through mode. The super- heated steam from the two steam generators (one for each loop) is directed to the turbine generator. The reactor containment building for each Power Pak encloses the reactor vessel and the in-containment, conventional (A-frame) fuel handling system. This building is a rec- tangular, reinforced concrete structure with a flat roof. Hatches are provided in the roof to facilitate handling of components, if necessary, thus limiting the building size and hence the coanstruction commodities required. The reactor guard vessel constitutes part of the containment envelope. The non-safety-grade, steam generator building for each Power Pak is a conventional building mounted on the base mat. The normal mode of decay heat removal uses natural circulation of sodium through the heat transport systems of the Power Pak. 1In addi- tion, two independent, natural circulation, backup systems are pro- vided. The first is a direct reactor auxiliary cooling system (DRACS) which transfers heat from the primary pool to the outside air using a sodium-to—air heat exchanger. The second is a passive, safety-related, reactor air cooling system (RACS) which operates with natural circula- tion to provide the ultimate decay heat removal capability through cool- ing of the reactor guard vessel. The RACS also provides passive cooling of the reactor cavity. The diverse and redundant shutdown system con- sists of both primary and secondary control rods as well as a self- actuated inherent shutdown system which responds to sodium overtempera- tures. The site construction time for a Power Pak unit, from ground breaking to initial power operation, 1is estimated to be thirty-three months., The 1licensing plan for SAFR stresses standardization and a 3-32 prelicensed Power Pak so that only site-related licensing considerations are required after obtaining a Final Design Approval. 3.4.3.3 Power Reactor-Inherently Safe Module (PRISM) The PRISM concept of General Electric is being designed under con- tract for the U.S. Department of Energy.l8 21 A gsimplified drawing of the concept is shown in Figure 3.6. A major design emphasis of the PRISM concept is incorporation of passive safety through use of: (1) a relatively low power reactor core of 133 MW(e), (2) a pool design with relatively low primary sodium temperatures, (3) a safety-grade passive decay heat removal system, and (4) large negative temperature reactivity feedback in the core design with the iatent of limiting potential core disruptive accidents to the initiating stage. Another major emphasis of the PRISM concept is licensing by demonstration of plant safety through tests conducted with at least the primary system of a prototype reactor module at a test facility. This safety-grade reactor module is the basic power-producing unit in the PRISM design. The low-pressure, primary system of each module is a pool-type design with the reactor core, four cartridge-type, electro- magnetic primary pumps, and four cartridge-type intermediate heat ex-— changers all contained within the reactor vessel. The intermediate sys- tem associated with each module consists of a single loop which trans- fers heat energy from a common header, fed by the four intermediate heat exchangers, to a single steam generator. Thus, the primary loops and the single, intermediate loop associated with each reactor module are independent of those of other modules. The common tie between the reactor modules occurs on the turbine side of the steam drums. Steam from three steam generators drives a single turbine. Therefore, the PRISM design, 1like the HTR concept considered by NPOVS, has multiple reactors and their associated heat transport system supplying steam to a single turbine. This power unit, or segment, containing three reactors, three steam generators and one turbine produces about 415-MW(e) of power. A power station, in turn, would consist of one or more of these segments. The reference PRISM design produces 1245-MW(e) and has three segments for a total of three turbines and nine reactors. FEach segment is functionally independent of the others. The homogeneous reactor core is fueled with U-Pu oxide. Through- the-head refueling will occur once each year using a mobile refueling machine. The residence time of the fuel is 4 years. The radial blank- ets containing UO2 contribute to a breeding ratio of about 1.03, de- signed to compensate for losses during recycle. The diverse and redun- dant control and shutdown system contains six primary control rods and two secondary control rods. 3-33 ORNL—DWG 86--4052 ETD — RVYACS AIR OUTLET - ey RYACS AIR INLET - ". | _~CONTAINMENT VESSEL HODULAR THX (8| :/(//,///-(1910 x 54'H) v s 10 | ,~.///,»{M PUMP (4) t : | ; | [/;/ ol ] A - ! | R . .L' 1 ‘ . M | Fig. 3.6. The below grade modular concept for PRISM, October 1984. Source: General Electric Company. 3-34 The containment vessel is 5.79 meters in diameter and 19.5 meters high. The vessel is shop fabricated and assembled and rail shippable. It is installed below grade to facilitate ground-level refueling, to re- duce building costs, and to provide a natural barrier to missiles. A sodium containment vessel surrounds the reactor vessel and is sized so that the reactor core will always remain covered by sodium even if the reactor vessel should develop a leak. Details of the containment/con- finement design are still under consideration. The primary pumps and intermediate heat exchangers can be removed easily through the top head for maintenance. The reactor vessel and containment vessel are important components of the safety grade, shutdown heat removal system. Normally this resid- ual heat would be removed by the non-safety grade, secondary heat trans- port loop associated with each reactor module. If this normal heat path is not available, the safety grade Reactor Vessel Auxiliary Cooling Sys- tem (RVACS) would provide this safety function. The RVACS is a passive, natural circulation system that is always in operation. Radiative heat is transferred from the reactor vessel to the containment vessel. This heat is removed to the atmosphere by natural circulation of outside air past the outside surface of the containment vessel. Calculations by GE indicate that this system can accommodate decay heat removal require- ments after loss of normal heat removal capability concurrent with a reactor scram. For this case, the peak in the primary sodium tempera- ture would be about 600°C and would occur several hours after the start of the event. An important aspect of the RVACS system is that its heat removal capability increases substantially with increasing primary sodium temperature. The passive safety features of PRISM are further indicated by its response to the very severe and unlikely accident where the loss of primary coolant pumping power, the loss of normal heat sink, and a fail- ure to scram all occur at the same time. The GE analysis of this hypo- thetical event with no operator intervention predicted that, after some initial oscillations in core reactivity and temperature, an equilibrium situation would be reached within about ten hours without exceeding allowable temperatures. At this equilibrium state, the heat generation rate of the critical core would be matched by the heat rejection rate of the RVACS system with a system temperature of about 630°C. Factory fabrication and assembly, standardization, and a reduction in systems required to be safety grade have been stressed in the PRISM design as a means of offsetting a perceived diseconomy of scale for small units. Advantages projected for this construction technique include more efficient use of site labor, a much shorter construction time of three years from start of construction to full power operation, "learning curve” benefits due to replication, and a closer potential match of a utility's power production capabilities to its load. Since the reactor module is the only nuclear qualified component, the balance of plant can be constructed economically to high quality industrial standards. 3-35 The PRISM licensing plan calls for prelicensing of a prototypic reactor module so that only site specific issues need be addressed for licensing a plant. This prelicensing would be accomplished through a design and safety test program during which the basic safety and economic claims for the concept would be demonstrated by prototypic, full scale tests. 3.4.4 Advantages and Disadvantages of the IMR Concepts with Regard to the NPOVS Criteria and Essential Characteristics 3e4e4.1 General Overviews Commercialization and marketing of an IMR in the anticipated market between now and around the year 2010 may be difficult to accomplish. Not only do IMRs have the same negative market factors as other con- cepts, including an uncertainty in the need for power, licensing chal- lenges, and financial uncertainties, but IMRs must also overcome addi- tional concerns such as higher capital costs associated with traditional designs, their perceived role only as breeders, a lack of wutility experience with IMRs, and uncertainties associated with an adequate and cost competitive fuel cycle. In fact, one could argue that IMRs will penetrate this market only if they have a unique and very important advantage over other power generating concepts. Such an advantage may arise from the innovative ILMR designs evalu- ated here. Their strong emphasis on cost reduction, passive safety, rapid construction, licensability, and low economic risk are certainly appropriate to meet the challenges of future markets. 1In the discussion which follows, the advantages and disadvantages, or challenges, outlined above will be discussed in the same order as the NPOVS criteria, essen- tial characteristics, and desirable characteristics presented in Section 2+.2.1+ Many of these comments apply to all of the IMR concepts and they will be presented first. These will be followed by comments specific to a particular concept. 3.4.4.2 Advantages of the IMR Concepts 1. Public Risk: A significant feature of IMRs is the passive safety which may be incorporated into their designs.7 Among the passive features 1is the tendency for sodium to provide natural convection cooling, the high thermal conductivity of sodium, the large heat capacity of the reactor system (which affords long grace periods for problem diagnosis and correc-— tive action), low-pressure design, and operating temperatures far below the boiling point of sodium. The usefulness and ef- fectiveness of these features were successfully demonstrated in tests at several plants including the Prototype Fast Reac- tor (PFR), Phenix, and the FFTF. They are utilized in all three of the designs considered here. Because of passive 3-36 safety features, these designs require fewer engineered (active) safety systems and less emergency power than conven- tional IWRs. One caution is that, since these designs rely on outside air as the final heat transfer medium/sink for decay heat removal, they might be susceptible to common cause external degradation events, such as fires or dust storms .22 The three designs considered here incorporate a reactor shutdown system similar to the CRBRP concept. They may en- hance the passive safety of their system with respect to Hypo- thetical Core Disruptive Accidents (HCDAs) through a passive control-rod release mechanism which will be activated by high sodium temperature. This feature may terminate any over-heat- ing event before sodium boiling occurs. The designs also limit the total amount and rate of reactivity insertion possible in the event of a control-rod withdrawal accident. In additionm, the assurance of decay heat removal capabilities is provided by both active and passive systems which incorporate signif- icant redundancy and diversity. Finally, for the PRISM and SAFR designs in particular, the core and control drive Ilines are being designed so that many HCDA initiating events will be terminated by feedback responses from temperature increases and resulting thermal expansion before core degradation initiates. In our judgment, IMR designs can meet and probably sig- nificantly exceed the goals of NPOVS Criterion 1. For exam-— ple, the PRA study completed for the CRBRP calculated a core damage frequency for an HCDA to be 3.6 x 10~ °/year, with seismic events being the major initiator.23 The frequency for internal initiators was about a factor of ten less. Another independent study for the SNR-300 plant in Germany concluded that, "both the frequency of major accidents and the extent of damage associated with such accidents are smaller than those estimated in the German Risk Study for the PWR-1300."2% These designs achieve 1low HCDA probabilities to a great extent because of the reliability of active safety systems, partic- ularly the diverse and redundant reactor shutdown systems. Credit for inherent or passive responses of the core which could result 1in early termination of the event are incor- porated into the calculations in a conservative manner. Investment Risk: In our judgment, the IMR designs can meet and probably significantly exceed the goal of NPOVS criterion 2. The emphasis on simplicity of design, the use of fewer complex safety systems, and the incorporation of passive design features, discussed under criterion 1, would all contribute to low investment risk. In addition, extensive reliability studies and PRA evolutions are planned for each design. 3-37 Economic Competitiveness: The capability of breeding signifi- cantly more fuel than is consumed in producing power is a major long-term advantage of IMRs. This breeding capability, coupled with a complete fuel cycle, would enable LMRs to ex- tract between 60 and 80 times the energy from a given quantity of natural uranium than can be done using non-breeders.’ In addition, IMR operating costs need not be as sensitive to fuel costs as non~breeders. The IMR designs can offer breeding as a design option to be implemented by a relatively easy and in- expensive core modification when it becomes economically attractive to do so. Comparative evaluations reported in Chapter 3, Volume III, of this report indicate a potential competitiveness with both the best IWR experience and with coal-fired plants. The LSPB concept has perceived economy-of-scale advan- tages and has incorporated significant cost reduction features and a short construction schedule into the design. The abil- ity to add plants in smaller power increments, thereby better matching utility needs, is a potential advantage of the PRISM and SAFR designs. Their lower capital risk achieved by modular construction and very short construction times is also attractive. However, it is not clear how costs for the fac- tories to build these modules will be assessed and costs for the fuel cycle will be incorporated. This may increase the cost of the first several plants, and it may be difficult to justify the high initial costs for factory automation which would improve manufacturing efficiency. SAFR plans are to use existing facilities with increased automation for vessel assembly production up to a few units per year. Probability of Cost/Schedule Overruns: All three concepts have stressed constructibility and simplicity, and a complete design before construction. They utilized modular construc- tion of major components im a factory and shipment to the site, and non-safety grade construction at the site for the BOP. These approaches should minimize delays and cost over-— runs attributable to quality assurance problems and large con- struction crews. There is a lack of U.S. industry experience in IMR construction. However, recent documentation of con- struction experience indicates that construction problems are more a function of the management and construction team and their interaction with the NRC than the reactor type.23 The concept of learning by experience should apply to SAFR and PRISM if additional modules and Power Paks can be built by the same team without delay after completion of the first plant segment. This can be done while the first segment is pro- ducing power, but care must be taken to avoid jeopardy to the operating unit by the construction activities where close proximity is required such as in the control building. 6. 3-38 Licensability. Assurance of licensability before construction is emphasized by these IMR commercialization plans. Each stresses early approval by NRC of a standard plant design. Thus, only site-specific NRC concerns would need to be addressed for licensing of subsequent plants. The 1licens- ability of the LSPB should be relatively high because the design basis accident analysis and many key safety design features are based on the CRBRP licensing experience. The first choice for PRISM licensing, and an alternative for SAFR, calls for demonstration of the plant's passive pro- tection against traditional HCDA initiators through tests of a prototypic reactor module. This concept of licensing by test has attractive features. Chief among these are a possible re- duction of analyses, validation of computer codes, and demon- stration of safety claims to the public, potential investors, and the NRC. Some precedence has been established for such tests through the extensive program at the Southwest Experimental Fast Oxide Reactor (SEFOR) which demonstrated the effect of the Doppler coefficient on power excursion,l!? and the recent tests at Raposdiel® and EBR-II where loss-of-flow HCDAs were initiated and subsequently terminated by passive feedback of the core. The SAFR designers indicate that a possibly more cost ef- fective approach involves resolving the main licensing issues by extrapolation of test results from FFIF and EBR-1I. Then a plant installed on a utility grid would be the vehicle for ob- taining a standard plant FDA with rulemaking to apply to sub- sequent plants of the same design. Demonstration of Readiness: Furope and Japan, which have less abundant natural supplies of fissile material, perceive a need for breeders sooner than the United States. For this reason these countries are pursuing a vigorous program of demonstra- tion and commercialization of the entire LMFBR fuel cycle. One can estimate from projects now in place that 50 plant-years of operation could be compiled by IMR demonstration plants by the year 2000.8 From past experience, acceptable performance is expected from these plants. For example, since 1973 the French, 250-MW(e) Phenix prototype plant has operated with an overall capacity factor of 60%.8 This experience base will be relevant to the requirement for a successful demonstration plant. A strict interpretation of Criterion 6 requires that demonstration plants for the specific ILMR plant concepts be built and operated in the United States before a utility decision to buy 1is made. To accomplish this task within the NPOVS time frame is a significant challenge. Nevertheless, 3.4.4.3 1. 3-39 our judgment from evaluations of the marketing and commercial- ization plans for LSPB, SAFR, and PRISM is that implementation of any of these plans with a dedicated effort could result in satisfying this criterion. Owner Competence: There are many similarities in the oper- ation of LMRs and LWRs, particularly with regard to reactor control and BOP functioning. Thus a significant fraction of LWR operator training and experience would be relevant to LMRs. In addition, worldwide experience indicates that LMRs are relatively easy to operate and maintain. Personnel spe- cifically trained in the operation of sodium systems within the United States are at national laboratories, industrial test facilities, and at the U.S. operating IMRs, EBR-II and FFTF. Essential Characteristics: The IMR concept designers have stressed shop fabrication, minimizing nuclear grade compo- nents, standardization, long plant lifetime, ease of construc- tion, and passive safety features. The PRISM and SAFR designs offer a variety of plant sizes to match load growth and, as explained in Chapter 3 of Volume III dealing with economics, some availability advantages may result from smaller, multiple reactor cores. Desirable Characteristics: Relatively high thermal efficien- cies (=40%) have been achieved with IMR designs and very low radiation exposures to workers (on the order of a few man-rems per year) have been experienced in demonstration plants. En- hanced diversion and proliferation resistance is possible with on-site fuel recycle and with the metal fuel option. Fuel elements can be retained in the core for several years, there- by yielding burnup values >100 MWd/kg. Disadvantages of the IMR Concepts Public Risk: Unlike IWRs which are designed to maximize k. ¢¢, an ILMR under normal operating conditions is not in its most reactive configuration. Thus, loss of sodium coolant from the core or core compaction could result in a reactivity increase. The designs considered here provide protection against loss of sodium inventory due to leaks and have substantial mitigating features — which are amenable to demonstration — for accommo- dating hypothetical accidents even beyond the design basis. Nevertheless, the way in which traditional licensing concerns associated with hypothetical accidents are addressed will need to be fully developed. Investment Risk: In addition to the comments made under public risk, some concern still exists about the performance and reliability of IMR steam generators. Data which could verify the performance of current designs should be available 3-40 within the NPOVS time period from component testing programs and foreign plant experience. Economic Competitiveness: Evaluations of prototype LMR designs and foreign construction experience indicates that the capital costs for LMR commercial plants, based on traditional designs of the 1980s, could be substantially higher than present LWRs. This higher capital cost, resulting in part from the need for an intermediate loop, could be compensated by lower fuel costs and higher efficiencies for LMRs. Higher efficiencies for LMRs have indeed been realized; the Phenix plant, for example, has a gross efficiency of 44%.l1%® But, as indicated below, it is not clear that the potential fuel-cycle cost advantage for LMRs will be realized within the NPOVS time constraints. Longer core 1lifetimes are being studied in future plans. In summary, cost competitiveness can not be claimed for operating LMR demonstration plants and, assuming no dramatic changes in fuel costs within the NPOVS time frame, competitiveness of commercial LMR plaunts can best be achieved by significant reductions in capital costs. Probability of Cost/Schedule Overruns: No specific disadvan- tage identified except that these are new design concepts with no direct base of experience. Licensability: The merits of licensing by prototypic tests have been discussed previously. There are, however, some lim- itations of this approach. Not all safety claims or hypothet- ical accident sequences can be demonstrated, and analysis of accident sequences may still be required. 1In addition, this could be an expensive test program even if the module can subsequently be used commercially since the test program could last several years and analyses of pre— and post-test results could be a significant effort. On the other hand, the PRISM designers believe this demonstration to be relatively 1less expensive for a small reactor when compared to the potential costs and risks associated with licensing a large reactor. An alternative would be to use the demonstration facility not only as a test of the PRISM and/or SAFR designs but also as an advanced research and development facility for general LMR passive safety features tests. It could demonstrate reac- tivity feedback effects as well as provide data for code verification. Perhaps alternate cores, metal and/or carbide, could be designed for the same facility. Passive shutdown systems and decay heat removal systems could be demonstrated as well. However, its utility for some of these purposes should be evaluated with respect to the FFTF and EBR-II capabilities., In addition to licensing by test, other LMR 1licensing igssues would still need to be considered for the standard plant designs. Prominent among these issues will be the 3-41 site-suitability source term, safety functions and design decisions associated with containment, passive features which accommodate HCDA concerns, and the need for redundancy and/or diversity within and in addition to safety systems which are passive. Although wuseful experience was gained through FFTF and CRBRP interactions and 1licensing activities with the NRC, licensing rules, guidelines, and procedures are not as well established for LMRs as for LWRs. However, preliminary discussions have been initiated with NRC for the IMR concepts. Demonstration of Readiness: Providing funding for an IMR demonstration plant will be a significant challenge. Owner Competence: Even though a large number of utilities participated to varying degrees in the CRBRP, experience in LMR operation does not currently exist within the U.S. utility organizations, and the FFTF and EBR II afford only part of the requirement. Perhaps a more pertinent question is whether the owner/ operator could be convinced to purchase a new reactor concept for which utility experience is limited. This latter need is perhaps most clearly evident when one considers aspects of the LMR fuel cycle. 1In short, each fuel cycle option appears to have some significant difficulties. To provide unique IMR advantages associated with breeding, such as relative freedom from concerns about a reliable fuel supply, a complete fuel cycle should be utilized. This means that proven and reliable on-site or off-site reprocessing, refabrication, and waste handling of suitable scale mst be available to the owner/ operator at a reasonable cost. The basic technology required for IMR fuel cycles has been developed in the United States and demonstrated overseas, and the first few LMRs could be supported by small-scale development facilities. However, if one assumes that this capability will be provided on-site, then uncertainties associated with available trained person- nel, cost, safeguards, reliability, licensability, and public and utility acceptance are envisioned. (See also Appendix E). It is not difficult to conclude, for example, that costs savings or other incentives must be significant and proven by experience before a wutility would choose to purchase and operate a reprocessing plant. Technical and organizational options making this concept more attractive include a less complex fuel cycle, the IFR concept for example,26 or the option that some other institution (not the utility) operate all facilities except (or including) the power station. These, and perhaps other options, could improve the viability, 3-42 but acceptance of this concept by a utility and its implemen- tation and demonstration in the NPOVS time frame seems unlikely. 1f, on the other hand, one assumes that off-site, cen- tral, reprocessing facilities would be used to complete the fuel cycle, it is difficult to envision the economic need for commercial facilities of this type much before the middle of the 21st century. Thus, off-site reprocessing may not be available in the United States within the NPOVS time frame. Still another option for closing the fuel cycle is to rely on other countries to provide this service. Difficulties associated with this option include problems associated with Pu shipments between countries, adverse balance of payments, and the assumption that such a commercial industry will in fact be available to the United States. If counting on a commercial fuel reprocessing industry is imprudent, another option is to consider a once-through cycle, including the possibility of spent fuel storage until commer- cial reprocessing/refabrication facilities are available. Difficulties associated with this choice are economic (tradi- tional IMRs with once-through fuel cycles would have fuel costs about twice those with Plutonium recyc1e27) and institu- tional. Once—-through cycles may need to use 235y enriched to 20 to 30% which are levels beyond present production for com— mercial use. The once~through option could likely be enhanced by the incorporation of low-power density, heterogeneous, long-lived (10 years or more) core designs. Essential Characteristics: Maintenance requirements and oper- ating staffs for PRISM, and to a lesser extent SAFR, may ex- ceed those for plants with a single reactor. Security staff requirements for PRISM can be small because of underground lo- cation and inaccessibility of key safety features during oper- ation. On the otherhand, regularly scheduled refuelling and maintenance reduces the need for extra manpower peaks at an-— nual refuelling in a monolithic plant. 1In addition, design of the control system for PRISM must accommodate multiple reactor cores providing the main source of energy to a single tur- bine. Licensing requirements, particularly those associated with the option of reprocessing and refabrication of fuel on- site, are not completely defined. 1If the overall nuclear in- dustry, including government support, continues to decline, the availability of qualified vendors may be in question. Desirable Characteristics: On-line refueling, though consid- ered, has not been incorporated into any designs. The PRISM plant, however, does have the capability to generate electric power continuously while a single module is being refueled. Completion of the fuel cycle, important for freedom from fuel 3-43 supply concerns and accomplished in foreign programs, has not been accepted in the United States because of economic and in- stitutional considerations. 3.4.5 Research and Development Needs for the LMR Concepts 3.4.5.1 Introduction Two different perspectives are presented in connection with IMR re- search and development (R&D) needs. First, the viewpoint of the plant designer is reflected through a collation of design—specific R&D re- quirements for the three LMR concepts considered in this report. Then consideration 1s given to general goals for the U.S. IMR R&D program which could contribute to a healthy and competitive industry considering the worldwide marketplace. 3.4.5.2 Design—-Specific R&D Requirements Each designer of the three LMR concepts considered by NPOVS recent- ly completed an assessment of specific R&D needs and reported conclu- sions, 2830 Appendix D presents summaries of these needs, where in sev- eral instances, similar needs have been combined. These design-specific needs can be classified as follows: (1) advanced core design tasks which include developing improved neutron counting channels, evaluating shielding designs, testing self-actuated shutdown systems, performing PRA assessments, and evaluating responses to accidents; (2) shutdown heat removal experiments and analyses to evaluate design effectiveness, design margins, and immunity to external events; (3) fuel related activ- ities such as evaluations of metal fuel cycles, high burnup tests of oxide fuels, and performance of these two fuels during upsets or when breached; and (4) system— and component-related studies emphasizing operating plant experience, scale model flow and temperature tests, in- corporation of advanced instrumentation and control technologies, and improving steam generator performance. A large base of test experience exists for the oxide fuel but that for metal fuel is limited. It is anticipated that an extensive fuel testing program would be required for metal fuel before proceeding to commercial use. 1In the French qualification of oxide fuels for LMFBR use, the testing program included an extended operation with refabri- cated fuel from the reprocessing demonstration. A similar effort for metal fuel may be prudent. Reprocessing and refabrication are discussed more extensively in Appendix E. 3.4.5.3 General R&D Goals for the U.S. National IMR Program A necessary but perhaps not sufficient list of goals for IMR R&D includes the following: 1. Develop an IMR design which has a clear, unique, and significant advantage in the marketplace over other concepts. The current de- sign studies are judged to be consistent with this goal. However, 2. 3-44 a small or debatable advantage for IMRs may not be adequate for penetration of a market dominated by LWR designs. Present programs are appropriately directed toward the innovative design of a cost competitive, modern (i.e., incorporating new technologies), and in- herently safe IMR. Licensability advantages as well as public and utility acceptance also are important reasons for this goal to be achieved. Maintain the option for rapid incorporation of breeders and of a complete fuel cycle into the future marketplace. The potential long-term market for breeders is assured unless nuclear fission energy 1is to have only a transitory role. Also, there exists a possibility for substantially increased shorter term demand if, for example, increased burning of coal should be found unacceptable. Complement the IMR R&D being performed by Europe and Japan so that the United States will be in a strong negotiating position to ex- change our accomplishments for experience from their more acceler- ated programs of demonstration and commercialization. Programs which typify contributions to this goal include advanced computer code development, materials research, licensing reform, advanced designs, metal fuel research, advanced instrumentation, contrel and simulation, and development of double-wall steam generators. Contribute to a reduction in licensing concerns, costly design mar- gins, and special systems resulting from the potential for core- disruptive accidents. Each of the IMR designs counsidered by NPOVS have already contributed to this goal. Advances in the future should stress demonstration of passive safety features, computer code validation, and experimental verification of specific reac-— tivity feedback effects incorporated into designs. Demonstrate, test, and utilize to the fullest extent possible ad- vanced technologies, components, and design concepts. Implementa- tion of R&D to satisfy this goal would increase the available de- sign options,31 thereby increasing the likelihood of optimizing the design to accomplish a larger number of desirable objectives and specifications. These advanced technologies could include automa- tion, research resulting in higher plant operating temperatures and efficiencies, use of artificial intelligence, and increased use of computers for control and simulation, surveillance and diagnostics, data display and verification, and maintenance functions. Automa- tion may be very important to the licensing and economic operation of multiple modules which feed a common steam system. Study and develop containment concepts which both simplify the overall nuclear system and ensure protection against both internal and external events, which may be judged credible. This work must be coupled closely with source term evaluations. 3-45 7. Investigate IMR core designs which might be competitive using a once—-through fuel cycle. These studies should include the poten- tial use of Pu obtained from foreign sources. This task will prob- ably require determining an optimum core geometry, power density, core lifetime, and neutron energy. It could contribute signif- icantly to competitive LMRs for a scenario of low energy-growth- rates. One such concept is an ultra long-life core which would require refueling only at major inspection intervals of approxi- mately every ten years. 8. Develop and demonstrate technical solutions to the challenges asso- ciated with the IMR fuel cycle which were identified in the previous section of this report. 3.5. HIGH TEMPERATURE REACTORS (HTRs) The focus of NPOVS HTR evaluation is on the modular HTR with the steam generator and core in separate steel vessels connected by con- centric crossducts in a side-by-side configuration. An extensive amount of information has been derived from the DOE HTR Program.”"‘38 To place the safety and economic features of the modular HTR in perspective, the large HTGR [2240 MW(t), 860 MW(e)], which was the focus of the DOE Pro- gram for several years, is carried by NPOVS as a point of reference. A summary of its advantages, disadvantages, and R&D needs can be found in Refs. 39 and 40 respectively. Appendix F presents the general design features of a large HTGR as a reference for HTR Technology that was originally oriented to that design. 3.5.1. Design Descriptions Modular steel-vessel HTR development began in The Federal Republic of Germany (FRG) in the late 1970s. Concepts have been developed by In- teratom, a subsidiary of Kraftwerk Union (KWU) and by Hochtemperatur Reaktorbau (HRB).*!=*2 Kernforschungsanlage (KFA), the Nuclear Research Center at Jlilich, has also been very active in the FRG program. They have taken advantage of favorable HTR characteristics (e.g. high heat capacity of the core and reflector, high temperature capability of the fuel, large negative temperature coefficient of reactivity) to develop a simpler plant to ease regulatory, construction and financing difficul- ties, as well as minimize development requirements. Both the Interatom and the HRB modular HTR concepts involve small modules of 200 to 250 MW(t) each. The thermal output of several modules can be combined to obtain a larger total plant capacity. This approach obviously re- duces the fission product inventory in any single reactor and reduces the amount of heat which must be removed from a reactor core in the event of an accident, thereby contributing to a high degree of safety. Both concepts utilize pebble fuel as do the two existing German HTRs [the Arbeitsgemeinschaft Versuchs Reaktor (AVR) and the Thorium Hoch Temperatur Reaktor (THTR)]. The Interatom concept places the core and steam generator in separate steel vessels in a side-by-side configura- tion, while in the HRB concept the steam generator is located above the 3-46 core in the same vessel., For both concepts, the reactor vessel is housed in a reinforced concrete cavity for both confinement and bio- logical shielding. A vessel cooling system, mounted on the inside sur- face of the cavity wall, is normally in operation cooling the concrete and is capable of providing decay heat removal by heat radiation from the uninsulated reactor vessel. Design parameters (such as core size and power density) for these modules were judiciously combined with generic HTR features so that in extreme accidents public safety is pro- vided without the operation of active heat removal equipment. Engi- neered systems are employed, but their role is primarily for investment protection. KWU/ Interatom is actively proposing their plant design for near-term commercial generation of electricity and for cogeneration of electricity and process heat. HRB proposes their concept for small electricity users and for process steam application. For the longer term, Interatom and HRB are developing their concepts for advanced pro- cess heat purposes such as the production of syngas through steam re- forming of methane or by steam gasification of coal., For larger plants HRB offers the HTR 500 [500 MW(e)]. An informal but broad survey of US utilities by Gas Cooled Reactor Associates identified a preference for plant sizes in the 200-700 MW(e) range for capacity additions beginning in the mid- to late 1990s.“3 Other more general studies also have indicated interest in smaller plants. In response to these factors, the U.S. HTR Program was re-— aligned in May 1984 to evaluate the potential for small reactor concepts with emphases on plant investment protection and safety. 1In particular, the plant design should be such that there would be no need for emer- gency sheltering or evacuation of the public as a consequence of licens- ing-basis events. Four concepts which resulted from a preliminary screening process were: 1170-MW(t) HIGR Cylindrical Prismatic Core Concept; (Ref. 44) 1260-MW(t) HTGR Annular Core Prismatic Concept; (Ref. 45) 250-MW(t) Pebble Bed Reactor Vertical-In-Line Steel Vessel Concept (4 units of 250 MW(t) each); (Ref. 46) 250-MW(t) Pebble Bed Reactor Side-by-Side Vessel Concept (4 units of 250 MW(t) each). (Ref. 35) A Concept Evaluation Plan3* specified criteria (generally consis- tent with NPOVS criteria) against which these plant concepts were evalu- ated. As a result, the modular HTR in a side-by-side configuration was selected in early 1985 as a preferred concept. Initially, emphasis was placed on the pebble bed core concept; however, a subsequent evaluation 3-47 between pebble and prism fuels led to the selection of a prismatic core in September 1985. The prismatic core obtains higher capacity with power levels up to 350 MW(t) by employing an annular core design. The higher power level reduces the plant capital cost per kW(e) for the prismatic fueled core relative to that for the 250 MW(t) cylindrical- core pebble bed reactor. In addition, the problem of compensating for reactivity insertions due to water ingress is reduced in the annular core design. The current reference modular HTR plant consists of 4 x 350 MW(t) reactor units for a total capacity of approximately 560 MW(e). Through an integrated approach, the modular HTR concept 1is being designed to meet the goals of safe, economical power.4?7 To meet these goals, the design must satisfy the following requirements: 3° 1. Equivalent availability factor of 807 with planned downtime of less than or equal to 10% per year. 2. 50 yr life measured from issuance of construction permit. 3. Have at least 107 economic advantage over the best coal-fueled alternative source of electricity. 4, Capable of start of operation in mid 1990s. 5. Separate the nuclear and non-nuclear portions of plant to mini- mize the number of components and systems which must be pro- cured, installed, operated, and inspected to nuclear standards. 6. Satisfy investment protection goals: a) less than 10% unscheduled unavailability b) provide protection against long outages c) limit the cost of decontamination and decommissioning d) freguency of events leading to plant loss to be less than 1072 per plant year 7 Satisfy HTGR safety goals: a) doses not to exceed EPA Protective Action Guidelines for public evacuation down to an accident frequency of 5 x 10~7 per plant year b) meet NRC interim safety goals The reference modular HTR is shown in Figure 3.7. A plant would consist of four 350 MW(t) reactor modules generating steam for two nominal 300 MW(e) turbine generators to produce a net plant output of 558 MW(e) (other design alternatives using 1 x 560 kW(e) and 4 x 140 MW(e) turbines are also being examined in the DOE program to determine the best approach).%8 3-48 ORNL-DWG 86-7363 CONTROL ROD DRIVE ASSEMBLY REACTOR VESSEL ANNULAR PRISMATIC CORE —— I~ MAIN CIRCULATOR 4 1 MOTOR N i N : ¥ 1 ‘ a1 ifl.—/— MAIN STEAM { — 3 C OUTLET L } . — STEAM GENERATOR SHUTDOWN COOLING VESSEL HEAT EXCHANGER UM | | —STEAM GENERATOR SHUTDOWN COOLING 1 } CIRCULATOR/MOTOR v S FEEDWATER INLET Fig. 3.7. 350 MW(t) annular prismatic HTGR: Primary coolant flow path during normal operation. 3-49 Fach reactor module is housed in a reinforced concrete enclosure (silo) which is fully embedded in the earth. The nuclear island con- gists of four enclosures and adjacent structures which house fuel handling, helium purification, storage, and transfer systems, the rad- waste system, nuclear island cooling water systems, and other essential reactor service systems. A common control room is used to operate all four reactors and the turbine plant. Each 350 MW(t) unit consists of separate reactor and steam gener- ator vessels connected by a horizontal coaxial crossduct. The core, graphite reflector, core support structure, and restraining devices are installed in the reactor vessel. The current core concept uses pris-— matic fuel elements most of which will be geometrically identical to the Fort St. Vrain standard (non-control) elements. The elements contain vertical through-holes for coolant flow and blind holes for fuel rods. The core consists of fuel elements in an annulus between an inner and outer region of hexagonal reflector elements. A number of the elements to be placed adjacent to the inner reflector contain an off-center hole to accommodate the insertion of reserve shutdown materials. Although the internal configuration of these elements differ from those used at Fort St. Vrain, the external geometry is the same. A number of the internal and external reflector elements which bound the core contain off-center holes for control rod insertion. The hexagonal fuel and reflector elements are designed for periodic replacement via the control rod penetrations in the vessel top head. The outermost radial reflector elements are irregular in shape so as to interface with the hexagonally stepped outer boundary of replaceable reflector elements and the lateral core support structure. Gravity-assisted control rod drive mechanisms are positioned above the radial reflector to operate control rods in the channels in the inner and outer reflector. The active core consists of 66 10-block high columns of fueled ele- ments. This makes the annular core configuration three elements wide and gives an average core power density of 5.91 W/cmd3. The fuel ele- ments contain 1.27 em (0.50 in.) diameter by 6.35 cm (2.50 in.) long fuel rods consisting of coated UCO and THO2 particles of low enriched uranium (LEU) fuel (U-235 < 20%) bonded in a graphite matrix. Refueling is accomplished with the reactor shut down and the vessel depressur- ized. The refueling operations are predicated on a three-year fuel residence time whereby half the fuel elements are replaced at the in- tervals of 18 months. The new fuel is placed into alternate columns adjacent to the half-burned fuel. During refueling, all the fuel elements in the core are moved within the vessel in 60 deg sectors at a time; fresh and spent fuel pass through the top head refueling penetrations which are located over the inner-reflector-to-core inter- face. [Each sector is rebuilt with half new and half-burned fuel. At discharge, the spent fuel ©burnup of the equilibrium cycle is 82,460 MWD/tonne. During each refueling, one-sixth of the reflector elements adjacent to the active core is replaced which corresponds to a nine year residence time. An alternate cycle has also been evaluated whereby the entire core is fueled as a batch, with a lifetime of about 3-50 2.7 yr. This cycle is stated to have nearly as favorable costs and to require less frequent shutdown for refueling. Replacement of fuel and reflector elements is performed with the fuel handling machine (FHM) which is placed over the inner penetration corresponding to the sector to be removed. The FHM elevates the spent elements into a fuel transfer cask., The fuel transfer cask, loaded to its maximum with five elements, is used to place the elements in a fuel storage well. Here the elements are dry-cooled before shipment off- site. The reactor plant cooling water system is used to remove heat from the well. Helium flows downward through the core coolant channels to an out- let plenum and then through the central duct of the cross duct to the top of the steam generator. It then flows downward across the once through helical coil steam generator with uphill boiling in the steam generator tubes. Cool helium is drawn from the bottom of the steam gen- erator and flows through an annulus surrounding the steam generator outer shroud to the circulator located on top of the vessel. The circu- lator discharges helium to a plenum from which helium flows through the outer annulus of the cross duct to the reactor vessel. It then flows upward through channels in the outer graphite reflector to a plenum above the top of the core. The reactor internal structures consist of graphite and metallic components. The major graphite components are the outer permanent re- flector, bottom reflector, core support posts, and top reflector. The major metallic components are the core support plate, core barrel lateral support structure, and the hot duct portion of the concentric cross duct. The reactor internals are designed for the full operating life, but are also designed to be inspectable, removable, and replace- able, if necessary. The main circulator, a variable speed, motor-driven single stage centrifugal compressor using gas/magnetic bearings, is mounted verti- cally on top of the steam generator vessel. Design parameters are summarized in Table 3.2. The basic approach has been to judiciously select design parameters and engineered systems so that they combine with inherent HTR features to yield a high degree of passive safety, and to provide investment protection as discussed in the following paragraphs. Two independent, diverse reactivity control/reactor shutdown sys- tems are provided. The primary system utilizes control rods located in the inner and outer replaceable reflector. The second system, the re- serve shutdown system (RSS), consists of boronated graphite pellets in storage hoppers which can be discharged into channels in the innermost row of fuel columns. Reactivity control requirements for basic opera- tions, including cold shutdown, are adequately covered by the reflector rod systems alone, with at-power operations possible without insertion 3-51 Table 3.2, Summary of major design features of modular HTR (side-by—-side configuration) Power per module, MW(t) Core power density, kW/1 Core inlet temperature, °C Core outlet temperature, °C Helium flow rate, kg/sec Helium flow direction Helium pressure, MPa (psia) Active core diameter, m Active core height, m Fuel element Fuel Equilibrium reload, kg:U/Th Average discharge burnup, MWd/kg Radial reflector thickness, m Reactor vessel material Reactor vessel, 0D, m Reactor vessel thickness, cm Reactor vessel height, m Steam condition, pressure MPa (psia) temperature, °C Net thermal efficiency, % 350 5.91 258 687 156.6 downward 6.38 (925) 1.65 inner, 3.5 outer 7.8 prismatic hex—block, 20.78 cm sides x 79.3 cm height LEU/Th 965/881 82.5 1.0 Low alloy steel-Mn-Mo, Sa 533 GrB Class 1 7.44 13.3 21.95 17.3 (2515) 541 39.6 3-52 of the inner-reflector rods. Cold shutdown with maximum positive reac- tivity due to water ingress requires the combined insertion of the reflector rods and the RSS. During a conduction cooldown event, the inner control rods could be damaged because of high temperatures. To avoid damage, although it does not affect safety, a control rod opera- tional strategy has been adopted where the inner rods are normally used for startup to 25% power and for normal cold shutdown. Steam generator tube leaks are detected by a moisture monitor located at the circulator outlet. If excessive moisture is detected, the steam generator is isolated and dumped and the main circulator is stopped. A shutdown cooling system (SCS) is provided to achieve and maintain the reactor thermal conditions required for maintainence in the event of failure of the main heat transport system (HTS) and to help meet the overall plant availability goal. The SCS is located in the bottom of the reactor vessel and consists of a heat exchanger and a circulator with a submerged motor. The reactor cavity is provided with a natural draft air cooling system (RCCS), Fig. 3.8. 1t consists of cooling panels mounted on the cavity wall through which air circulates by natural convection. The de- sign has no valves or active components. The surface of the panels serves as a barrier separating the outside atmosphere from the reactor cavity atmosphere. The system uses four separate inlet/outlet struc- tures to minimize the possibility of flow blockage. 1In addition, the four loops are interconnected by inlet/outlet plenums in the cooling panels. This provides a heat sink sufficient for decay heat removal in the event the main heat transport system (steam generator and main helium circulator) and the shutdown cooling system are not available. Heat transport from the reactor core is by natural processes of conduc- tion and radiation (and convection if the primary system is pressurized) through the core to the vessel wall and by radiation and convection to the cooling panels. The reactor utilizes a confinement equipped with dampers which open on excessive pressure loads resulting from feedwater, main steam, or re- actor coolant line ruptures. Program studies indicate that the fission product releases from the core are small enough that reliance need not be placed on conventional pressure~tight containment or a confinement with a filter system to meet the defined safety criteria. For decay heat removal, under pressurized or depressurized condi- tions, the main cooling loop (consisting of the main circulator and the steam generator) is the first option. If either the main circulator or the steam generator is not operational, then forced circulation using the shutdown cooling system is the next option for either pressurized or depressurized conditions. The next option is to remove decay heat through the vessel wall by radiation to the RCCS. This system is designed to limit the fuel temperatures to 1200°C under pressurized conditions (when there can be a significant redistribution of heat 3-53 ORNL—-DWG 864054 ETD —> | l\\\ \\\) I \.". NN ) 4. ..l Feoanr T TV » ®a .« o e, = . 2%, T avte " v 1.t R a® b LI - ‘.:t.‘.. {2ee. FEATURES : 1. COOLS THROUGH PANEL WALL 2 OPERATES UNDER ALL MODES OF REACTOR OPERATION. 3. SIMPLE TO OPERATE, VIRTUALLY MAINTENANCE FREE. Fig. 3.8. Reactor cavity cooling system. 3-54 within the core by natural convection) and to 1600°C under depressurized conditions. 3.5.2 Claims, Advantages, and Disadvantages Evaluated Against Criteria, Essential and Desirable Characteristics The claims and reported advantages of the modular HTR are discussed briefly as follows in the order of the criteria first and essential and desirable characteristics second. A more detailed evaluation of modular HTR claims has been included in Appendix G. 1. Public Risk: While the calculated risk to the public has not been quantified for the modular HTR, there are important fea- tures which provide the design with a high degree of passive safety, and thereby also provide confidence that the calculated risk to the public due to accidents will be equal to or less than the calculated risk associated with the best modern LWRs. These features are: ® The capability for afterheat removal through the vessel wall by natural heat transport mechanisms (convection, conduction, and radiation). This capability has been demonstrated partially on the smaller experimental AVR which shares features of the modular HTR. Future confirmatory experimentation for more generally applicable data is being considered and may be possible at the AVR subject to the approval of the German authorities. ® Very good retention of fission products within the fuel to high temperatures. This feature has been demonstrated by U.S. and German coated particle fuel systems exXperience and in fuel test programs. ° No need for a fast acting shutdown system for core heatup events, which again has been demonstrated on other HTRs. Other potentially severe accidents, such as major water and/or air ingress events, have been argued to be of such low consequence or low probability by virtue of system design that these types of accidents pose no significant public risk. However, the NRC may require that these accidents be factored into the cost benefit analysis of the use of confinement versus containment. Investment Risk: The probability of loss of investment for the modular HTR is claimed to be less than 107° per vyear. This claim requires independent review, but many of the features which preclude or reduce the effect of incidents on public safety can be argued as being favorable to investment protec- tion. 3-55 Economic Competitiveness: Until the plant design is complete and commodity requirements determined, a firm estimate of cost cannot be made. With regard to meeting the financial goals of the utility, the ability to add capacity in small increments as well as the potential for achieving short construction time through factory fabrication should reduce the utilities' capi- tal investment exposure and investment risk, thereby helping to meet their financial goals. With regard to acceptable busbar costs, factory fabrication of modules coupled with the rela-— tively high burnup achievable with HTR fuel cycles may com- pensate for higher fuel fabrication costs typical of HTRs and potentially higher distributed capital cost usually associated with smaller sized plants. The use of multiple modules may also increase overall availability, although at lower power levels, thereby providing flexibility in scheduling outages. Assumptions about availability for the modular plants play an important role in estimating overall competitiveness with both the coal fired and the better current generation LWR plants. Probability of Cost/Schedule Overruns: The DOE and the in- dustrial proponents recognize the need for complete design before initiating construction. Detailed design and associated studies of construction needs, options and costs still remain to be completed, so that cost and schedule factors cannot be quantified. However, the DOE funded program has produced indepth studies of construction needs, options, and costs for the modular HTR so that uncertainties should be well defined. Licensability: The modular HTR has a draft licensing plan. The DOE and industrial proponents are actively engaged in dialogue with NRC-NRR. Early concurrence on licensing may be essential to meeting the 2000-2010 time frame for commercial- ization. DOE and the industrial proponents plan to secure an NRC final design approval (FDA) by 1996. A preliminary safety information document (PSID) will be submitted in CY 1986. There is also a utility effort led by the Tennessee Valley Authority (TVA) which proposes joint funding of a single plant to demonstrate licensability by test; however, such testing probably could not address all safety questions particularly those beyond design basis accidents such as a major air ingress, acts of sabotage and selsmic events. Demonstration of Readiness: Many features of the modular HTR are or will have been demonstrated in the German AVR plant before the modular HTR is offered commercially. A successful demonstration of high powered, gas/magnetic bearing circulators would represent a significant contribution to the demonstration of readiness since most of the other major component tech- nologies either are borrowed or have evolved from AVR, THTR, and Fort St. Vrain experience. More fuel testing is already planned to support licensing as well as normal operation requirements. 3-56 Owner Competence: The operation of multi-module plants may pose new concerns about interdependence and common mode inter- actions of systems. Such concerns may influence NRC mandates on acceptable control configurations which may in turn be more costly or manpower intensive than currently envisioned; how- ever, the overall technology of the modular HTR appears to be as readily assimilable as IWR technology. The lessons from the Fort St. Vrain HTGR also appear to be clear to a potential owner/operator of an HTR. Some of these lessons are: (1) keep moisture out of the primary system and any other part of the plant where it can cause corrosion [which should be helped in the modular design by incorporating hardware features based on lessons learned to date], (2) maintain excellent secondary coolant chemistry, (3) wmaintain an intensive and extensive surveillance program of (1) and (2) above, (4) ensure quality and confirmatory testing of both original and replacement materials and equipment, and (5) maintain a clean physical plant. The large thermal margins inherent in the HTR fuel sys- tems and the low graphite corrosion rates in the presence of numerous moisture ingress events at Fort St. Vrain could 1lull plant designers and operators into a failure to recognize the significance of operating problems. Some of the observed -oper— ational problems at Fort St. Vrain have included the effect of moisture on leaching and distribution of other corrosive mate- rials (e.g., chlorides), the apparent inability to detect ab- normal control configurations and reactivity anomalies quickly (e.g., confirming subcriticality by excore detectors and detecting dropped control material) and the possible inter- dependence of redundant emergency ac power sSystems. These types of situations should not be repeated with the modular HTR. Essential Characteristics: Many of the essential character- istics are integral requirements for meeting one or more of the criteria and as such are discussed more fully above. However, in general, the modular HTR has promise of achieving many of these characteristics as outlined, in some cases repetitively, below: a. High availability due to use of small-sized turbines and modularity which allows higher availability at reduced power. be Maximum use of shop fabrication of reactor systems Ce A high degree of passive safety d. Potentially no need for developing or demonstrating a plan for evacuation of the public beyond the site boundary e. Potential for demonstrating features important for passive safety The 3-57 f. Low thermal discharge (due to high thermal efficiency) g Low radiocactive effluent as demonstrated by Peach Bottom 1, Dragon, Fort St. Vrain and AVR experience h. Low investment risk to the utility resulting from adding capacity in small increments and from what is intended to be a simpler approach to meeting safety requirements and licensing. Desirable Characteristics: Several of these characteristics are addressed in regard to criteria. The advantageous omnes are listed again as follows. The modular HTR appears to have modest RD&D requirements, relative ease in siting based on pro- jected low source terms for both normal operation (worker expo- sure and effluents) and accident conditions, good fuel utilization (high burnup), high thermal efficiency, high ver- satility in application because of the production of high coolant temperatures, and a low visual profile through full embedment. Full embedment and passive safety should also con- tribute to a high degree of sabotage resistance. The claim is made that the use of low enriched uranium increases resistance to proliferation and diversion and that appears to be the case for the fresh fuel supply. Also, the HTR spent fuel appears to have a high resistance to diversion and proliferation tech- nologies. potential disadvantages are discussed as follows: Public Risk: As alluded to under the discussion of advantages, the resolution of concerns for severe accidents will preferably be handled by probabilistic risk analyses to demonstrate a low contribution to the overall risk to the public. If require- ments such as the use of inerted containment are imposed, over- all costs will increase. Investment Risk: Independent assessments are needed. Economic Competitiveness: The possibility of higher costs for fuel fabrication and plant capital investment are a concern; as is the availability which will be achieved. Independent evalu- ations appear prudent to perform. Probability of Cost/Schedule Overruns;: Since this 1is a new concept, it is particularly important that the design is com— pleted before construction begins. The current approach is based on defining "top-down" requirements from which design data needs and RD&D will proceed using functional analysis. Construction plans and schedules must be coordinated carefully with the availability of design and safety related data. Licensability: As mentioned above, the analysis requirements and expected design needs in response to "beyond design basis 3-58 accidents” must be settled, preferably, as early as possible in the design process. 1In the post-TMI licensing environment, the modular HTR could still face defense-in-depth requirements such as containment, emergency ac power sources and safety grade components in the balance of plant. These can have a severe effect on increasing plant costs if imposed. Seismic consider- ations with regard to the reactor core and the side-by-side connecting pipe must also be addressed for licensing. 6. Demonstration of Readiness: Other than answering questions about needing high availability for overall competitive eco- nomics, the modular HTR would appear to have a lesser require- ment for a demonstration plant because of AVR and THTR experi- ence and the ability to incorporate lessons learned at Fort St. Vrain. 7. Owner Competence: No specific disadvantage identified, how- ever, as indicated under advantages, a potential owner/operator should be thoroughly familiar with details of the engineering and licensing experience at St. Vrain. The lessons learned are positive with respect to avoiding potential pitfalls. 8. Essential Characteristics: The relatively low power per module [~350 MW(t)] does affect the capital cost as a disadvantage. The side-by-side HTR module has also been questioned because of potentially adverse seismic response at the connecting pipe be- tween the reactor and steam generator vessels. Both of these features may be improved through design enhancement and innova- tion. The power of the module may be increased if higher fuel temperatures (>1600°C) become acceptable by further fuels test-— ing and verification. The connecting pipe will require thorough and extensive analysis to show that it can withstand the potential consequences from seismic events. 9. Desirable Characteristics: The use of LEU/Th fuel leads to lower fuel conversion ratios relative to the use of highly en- riched fuels. 3.5.3 Modular HTR Research and Development Needs Evaluated Within the DOE HTR Program, development of a modular HTR Technology Development Plan using the Integrated Approach is under way, but results are not sufficiently complete for incorporation into NPOVS. However, ORNL has prepared a document“? which was presented to the Subcommittee on Energy Research and Production of the U.S. House of Representatives and which discussed the key research and development areas required for modular HTRs. This section presents R&D needs as excerpted from this document with modifications reflecting additional information obtained since that time. The key research and development (R&D) areas are con- sidered in the following categories: A. B. C. 3-59 Base Technology Applied Technology; and Design and Economic Studies. Item A generally refers to basic information needed to establish the feasibility of the reactor concept and to materials data needed for the detailed design; item B refers to R&D needed to assure the practi- cality of components and systems; and item C refers to the effort re- quired to specify the entire reactor plant in sufficient detail to permit reliable economic estimates of plant performance. The key R&D areas which need to be addressed for the modular HTR are shown below: 3.5.3.1 1. 8. 3.5.3.2 1. Base Technology Determination of fission product retention of the fuel coat- ings, graphite and metal surfaces of the primary system and confinement during and subsequent to extreme accident condi- tionse. Process development for fuel fabrication and irradiation test- ing to obtain understanding of the importance of specific pro- cessing parameters on fuel performance. Irradiation testing and examination of fuels produced in com— mercial-scale production equipment. Fission product behavior during normal reactor operation as related to lift-off and the source terms under depressurization accidents. Development of detailed materials properties under conditions of creep, fatigue, corrosion, and radiation necessary for designing and operating components. Obtaining statistical data on graphite properties as a basis for estimating fuel element stresses. Critical experiment testing of LEU/Th cores, including water ingress reactivity effects and temperature coefficients for low enriched uranium fuel with plutonium concentrations representa-— tive of equilibrium burnup. Obtaining experimental data to validate codes applicable to the passive heat removal system. Applied Technology Development, verification, and application of analytical tools for reactor design, safety, and risk analyses, including data bases. 3-60 2. Plant safety and risk analyses. This risk associated with normal operation and design basis accidents needs to be inves- tigated. Also, risk associated with postulated events beyond design basis may need to be investigated to confirm that the total risk from such accidents is small relative to the risk from normal plant operation. 3. Detailed reactor physics analysis including computation of cross sections, power distributions, temperature coefficients, and control rod worth under normal conditions and with water ingress. Also shielding analysis to determine fluence for the design of reactor internal components at various locations. 4. Design and testing of refueling equipment to demonstrate that the reference reactor concept can be refueled on the assumed schedule, 5. Design and testing of prototypic components and systems such as the helium circulator, core support structure, and shutdown cooling heat exchanger. 6. Development of multi-module control system, service systems, and heat exchange systems. 3.5.3.3 Design and Economic Studies The design of the modular HTR plant must be completed in sufficient detail to permit a firm estimate of plant costs, based on features which limit fuel temperatures under accident conditions, facilitate shop fabrication, and reduce balance-of-plant (BOP) costs. Also, a detailed determination of operating and maintenance and fuel cycle costs are re-— quired. 3.6 l. 5. 10. 3-61 REFERENCES FOR CHAPTER 3 K. Hannerz (ASEA-ATOM), Towards Intrinsically Safe Light Water Re- actors, ORAU/IEA-83-2 (M)-Rev. (research memorandum), Institute for Energy Analysis, Oak Ridge Associated Universities, 0Oak Ridge, Ten-— nessee, July 1983. SECURE P: Design Progress Information, ASEA-ATOM, Vasteras, Sweden (April 1984). ASEA~ATOM PROPRIETARY. D. Babala and K. Hannerz, "Pressurized Water Reactor Inherent Core Protection by Primary System Thermohydraulics,” Nuclear Science and Engineering 90(4), 400—410 (August 1985). C. Sundqvist and T. Pederson, "PIUS: The Forgiving Reactor, Safety and Operational Aspects,” full text of paper presented at the 1985 Annual Meeting of the American Nuclear Society (ANS), Boston, Massachusetts, June 9—14, 1985; ASEA-ATOM, Vasteras, Sweden. Je. De Duncan and C. D. Sawyer, "Capitalizing on BWR Simplicity at Lower Power Ratings,” SAE Technical Paper Series 859285 reprinted from p. 164, Proceedings of the 20th Intersociety Energy Conversion Engineering Conference, Miami Beach, Florida, August 18-23, 1985, General Electric Company, San Jose, California. Lyle C. Wilcox, "U.S. Department of Energy Programs on Cost Reduc-— tion,” presented at the Institute of Applied Energy International Symposium on LMFBR Development, Tokyo, Japan (November 7, 1984). Alan E. Walter and Albert B. Reynolds, Fast Breeder Reactors, Per- gamon Press (1981). Transactions of the ANS 1984 Winter Meeting 47, 13-16 (Novem- ber 11-16 1984). Consolidated Management Office for the IMFBR of the Electric Power Research Institute, LSPB Design Descriptions, Vol. 1, CDS 400-8, for the U.S. Department of Energy, September 1984. APPLIED TECHNOLOGY . R. A. Lindley, Large Scale Prototype Breeder Cost Effectiveness Considerations, Consolidated Management Office for the IMFBR of the Electric Power Research TInstitute, August 1984. APPLIED TECHNOLOGY. 11, 12. 13, 14, 15. 16. 17. 18. 19. 20. 21. 22. 23, 24, 3-62 Consolidated Management Office for the LMFBR of the Electric Power Research Institute, LSPB Overall Plant Design Specification, CDS 100-2, Rev. 7, for the U.S. Department of Energy and Electric Power Research Institute, Washington, D.C., February 1984, APPLIED TECHNOLOGY. LSPB Constructibility Report, to be published by the U.S. Depart- ment of Energy. APPLIED TECHNOLOGY. Modular IMFBR Pool Plant Final Report, AI-DOE-13502, Rockwell In- ternational, Atomics International, Canoga, Park, California, September 30, 1984, APPLIED TECHNOLOGY. "Advancing Breeder Reactor Design in the United States,” Engineering International 30(365), 17-20, (February 1985). Nuclear Transactions of the ANS 1984 Winter Meeting 47, 299-300 (November 11-16, 198%4). “SAFR discussions at ORNL, January 11, 1985," a collection of view- graphs presented at this meeting., APPLIED TECHNOLOGY. Large IMFBR Pool Plant, Vol. 1, Design Description, ESG-DOE-13410 Rockwell International, Energy Systems Group, Canoga Park, Cali- fornia, September 1983. APPLIED TECHNOLOGY. PRISM Semiannual Report, April-September, 1984, XL-897-840073/L3, General Electric Co., Nuclear Systems Technology Operation, Sunny- vale, California, October, 1984. APPLIED TECHNOLOGY. Jo S¢ Armijo et al., "General Electric Strategy for Achieving a Low—Cost Liquid Metal Reactor Plant,” Presented at the Institute of Applied Energy International Symposium on ILMFBR Development, Tokyo, Japan, November 7, 1984, "Advancing Breeder Reactor Design in the United States,” Engineering International 30(365), 17-20 (February 1985). Nuclear PRISM Design Requirements, Preliminary Rev B, 23A3071, General Electric Co., Nuclear Systems Technology Operation, Sunnyvale, California, October 1984, APPLIED TECHNOLOGY. Internal Correspondence from G. F. Flanagan to Distribution, "In- herently Safe LMRs,” December 6, 1984, Clinch River Breeder Reactor Plant Probabilistic Risk Assessment, prepared for the U.S. Department of Energy by Technology for Energy Corporation, September 14, 1984, A. Bayer and K. Koberlein, "Risk-Oriented Analysis on the German Prototype Fast Breeder Reactor SNR-300," Nuclear Safety 25(1), 30 (January-February 1984). 25, 26. 27. 28. 29 . 30, 31. 32. 33. 34. 35. 36. 37. 3-63 Transactions of the ANS Winter Meeting 47, 333-338, (November 11-16, 1984). "Looking to the Future with the Integral Fast Reactor,’ Engineering International 30(365), 20 (February 1985). R. Balent and J. Yedidia, DRAFT, Large Scale Prototype Breeder Fuel Cycle Plan, to be published by the U.S. Department of Energy. APPLIED TECHNOLOGY LSPB Research and Development Requirements, CDS 500-6, U.S. Depart-— ment of Energy, Washington, D.C., September 1984, APPLIED TECHNOLOGY. SAFR Requirements for Base Technology Program, 149T1000002, Rock- well International Rocketdyne Division, Canoga Park, California, January 1985, APPLIED TECHNOLOGY. Letter number XL-897-850016 from L. N. Salerns to Francis X# Gavigan, dated January 11, 1985, "WBS2B(O.5- Initial PRISM R&D Re- quirements Statements.” APPLIED TECHNOLOGY. Letter from J. Ray to Dr. Bill Harms, dated January 31, 1983, with the attachment, "LMFBR Safety Philosophy Issues”, Advanced Reactors Subcommi ttee, draft, January 27, 1983. Letter number T-85-053 from J. D. Mangus to D. C. Gibbs, dated July 23, 1985, with attachment, "Long Life Liquid Metal Core Concept.” Utility/Users Design Requirements for Small High Temperature Gas- Cooled Reactors, GCRA 84-011, Gas—Cooled Reactor Associates, San Diego, California, November 1984. APPLIED TECHNOLOGY. HTGR Program Concept Evaluation Plan for Small HTGRs, GCRA 84-009, Gas—-Cooled Reactor Associates, San Diego, Califormia, October 31, 1984+ APPLIED TECHNOLOGY. Preliminary Concept Evaluation Report, 4 x 250 MW(t) HTGR Plant Side~by—-Side Steel Vessel Concept, HTGR-85-005, issued by Bechtel Group, Inc., et al., for Gas—Cooled Reactor Associates, San Diego, California, February 1985. APPLIED TECHNOLOGY. FY 1985 HTGR Summary Level Program Plan, HP-20202-85, Gas-Cooled Reactor Associates, San Diego, California, October 1984, APPLIED TECHNOLOGY. Licensing Plan for the Standard HTGR (Draft), GCRA 85-001, Bechtel Group, Inc., et al., January 1985. 38. 39. 40. 41, 42, 43. 44, 45. 46, 47. 48. 3-64 Preliminary Concept Description Report, 4 x 350 MW(t) HTGR Plant Side-by—~-Side Steel Vessel Prismatic Core Concept, HTGR-85-142, issued by Bechtel Group Inc. for Gas—Cooled Reactor Associates, San Diego, California, October 1985. APPLIED TECHNOLOGY. P. R. Kasten et al., Assessment of the Thorium Fuel Cycle in Power Reactors, ORNL-TM-5565, Oak Ridge National Laboratory, Oak Ridge, Tennessee, January, 1977. P. R. Kasten, "Statement on an Inherently Safe High-Temperature Gas—Cooled Reactor Program,” presented to Subcommittee on Energy Research and Production, U.S. House of Representatives, M. Lloyd, Chairman, February 7, 1984. He Reutler and G. Lohnert, "The Modular High-Temperature Reactor,” Nuclear Technology, 62(1), 22-30, (July 1983). HTR 100-MW(e) Konzeption; Technik, Termine, Kosten, Hochtemperatur Reaktorbau, Hochtempatur Reaktorbau, Mannheim, Federal Republic of Germany. Summary Report on the Utility Industry Questionnaire, GCRA.84-001, Gas Cooled Reactor Associates, San Diego, California, February 1984. Preliminary Concept Evaluation Report, 1170 MW(t) HTGR Plant PCRV Concept, HTGR-85-004, issued by Stone & Webster Engineering Corpor- ation for Gas—Cooled Reactor Associates, San Diego, California, February 1985. APPLIED TECHNOLOGY. Preliminary Concept Evaluation Report, 1260 MW(t) HTGR Plant PCRV Concept, HTGR-85-003, issued by Stone & Webster Engineering Corpor- ation for Gas-Cooled Reactor Associates, San Diego, California, February 1985. APPLIED TECHNOLOGY. Preliminary Concept Evaluation Report, 4 x 250 MW(t) HTGR Plant In- Line Steel Vessel Concept, HTGR-85-006, 1issued by Bechtel Group, Inc., et al., for Gas—-Cooled Reactor Associates, San Diego, Cali- fornia, February 1985. APPLIED TECHNOLOGY. An Integrated Approach to Economical, Reliable, Safe Nuclear Power Production, ALO-1-11, Combustion engineering, Inc., Windsor, Connecticut, June 1982. Turbine Selection Trade Study 4 x 250 MW(t) HTR Plant SBS/SV Con- cept, HTGR-85-075, Stone and Webster Engineering Corporation for Gas-Cooled Reactor Associates, San Diego, California, July 1985. APPLIED TECHNOLOGY. 4. ACKNOWLEDGMENTS The form and scope of this study necessitated the involvement of many individuals and organizations. 1In fact, the numbers are so great and the involvement so often indirect that complete individual recogni- tion is next to impossible. However, the cooperation was extensive and effective; those listed as authors recognize and greatly appreciate this assistance. The institutions and individuals who contributed through interview and/or written reports and, in some cases, through work specific to the study are as follows: Reactor Vendors ASEA-ATOM Babcock and Wilcox Combustion Engineering GA Technologies General Electric Company Rockwell International Westinghouse—~Advanced Energy Systems Division Architect-Engineers Bechtel Sargent and Lundy Stone and Webster United Engineers and Constructors Utility Companies and Associations Baltimore Gas and Electric Central Electricity Generating Board, UK Carolina Power and Light Duke Power Company Flectric Power Research Institute Electric Power Research Institute Counsolidated Management Office Gas-Cooled Reactor Assoclates Houston Power and Lighting Southern California Edison Wisconsin Electric Power Company Laboratories, Institutions, and Universities Argonne National Laboratory Atomic Industrial Forum Institute for Euergy Analysis International Atomic Energy Agency Los Alamos National Laboratory Massachusetts Institute of Technology Nuclear Energy Agency Office of Technology Assessment The University of Tennessee U.S. Nuclear Regulatory Commission 4-1 4=2 Individuals at the three cooperating institutions (ORNL, TVA, and the University of Tennessee) who provided assistance include the follow- ing: Qak Ridge National Laboratory S. J. Ball R. M. Harrington T. E. Cole W. O. Harms R. M. Davis J. E. Kibbe J. C. Ebersole (consultant) 0. H. Klepper J. R. Engel A. E. Levin G. F. Flanagan G. Samuels L. C. Fuller J. We Sims S. R. Greene Tennessee Valley Authority D. T. Bradshaw J. G. Stewart D. L. Lambert R. E. Taylor H. G. O0'Brien S. Vigander J. E. Simmons The University of Tennessee H. L. Dodds, Jr. NPOVS Advisory Committee Members S. Burstein, Vice-Chairman of the Board, Wisconsin Electric Power Company G. F. Dilworth, Director of Engineering and Technical Services (DETS), TVA T. S. Elleman, Vice President of Corporate Nuclear Safety and Research, Carolina Power and Light Company P. R. Kasten (Secretary), Technical Director, Gas-Cooled Reactor Programs, ORNL L. M. Muntzing, Doub and Muntzing, Washington, D.C. D. R. Patterson, Assistant to Manager, Office of Engineering Design and Construction, TVA W. T. Snyder, Dean, College of Engineering, The University of Tennessee J. Taylor, Vice President and Director, Nuclear Power Division, EPRI N. E. Todreas, Department of Nuclear Engineering, Massachusetts Institute of Technology 4-3 NPOVS Advisory Committee Members (continued) J. Taylor, Vice President and Director, MNuclear Power Division, EPRI N. E. Todreas, Department of Nuclear Engineering, Massachusetts Institute of Technology APPENDIX A BASIC OUTLINE FOR NUCLEAR POWER OPTIONS VIABILITY STUDY VOLUME 1 VOLUME II VOLUME TI1I VOLUME IV FINAL REPORT EXECUTIVE SUMMARY REACTOR CONCEPTS, DESCRIPTIONS, AND ASSESSMENTS (see page v) NUCLEAR DISCIPLINE TOPICS ABSTRACT 1. INTRODUCTION 2. CONSTRUCTION 3. ECONOMICS 4. REGULATION 5. SAFETY AND ECONOMIC RISK 6. NUCLEAR WASTE TRANSPORTATION AND DISPOSAL 7. MARKET ACCEPTANCE 8. ACKNOWLEDGMENTS APPENDIX A. INTERVIEW FORMAT FOR THE ISSUE DEFINITION RESEARCH AND OUTLINE OF ISSUES USED FOR THE CASE STUDY INTERVIEWS APPENDIX B. TABLES ON THE SAMPLE USED FOR THE ISSUE IDENTIFICATION RESEARCH BIBLIOGRAPHY ABSTRACT 1. INTRODUCTION 2. ORGANIZATION AND RETRIEVAL 3. KEYWORD LIST 4 KEYWORD INDEX 5. NUCLEAR OPTIONS CITATIONS 6. LIGHT WATER REACTORS CITATIONS 7. LIQUID METAL REACTORS CITATIONS 8. HIGH TEMPERATURE REACTORS CITATIONS 9. ACKNOWLEDGMENTS 10. REFERENCES APPENDIX B THE OUTLOOK FOR ELECTRICITY SUPPLY AND DEMAND* The principal determinants of future electricity demand will prob- ably be the utilities and their regulators. During the past ten years, utilities have been evolving from a supply industry concerned only with meeting electricity requirements to a service-oriented industry con- cerned not only with the supply of electricity but also with controlling and shaping its use through conservation and load management. Future electricity use will depend on how far this evolution proceeds. The approach taken to estimate future energy use involves an anal- ysis and/or estimate of the trend of factors that determine energy use, such as population, persons per household, gross national product (GNP), shifts in the industrial product mix, conservation, etc. The projec- tions made here do not represent anything even approaching the tech- nology limits of energy conservation nor do they come close to the eco- nomic 1limit of conservation as projected by "least cost energy strategies.” They do depend on continued efficiency improvements and, to some extent, on a coutinuation of utilities' aversion to investment in new capacity, which has resulted in conservation and load management programs to limit demand growth. They probably represent a narrow band in the upper part of a rather wide range that could be expected. Table B.}l summarizes the estimates of this study for growth rates of electricity and nonelectrical energy requirements to the year 2000 for the residential, commercial and industrial sectors. The total growth rate for electricity is estimated to range between 1.8 and 2.3%/ year and for nonelectrical energy between 0.1 and 0.5%/year. These rates result in a growth of primary energy requirements of 0.9 to 1.4%/ year, which is equivalent to using between 67.3 and 73.9 quads (exclud- ing tramsportation) in the year 2000. The transportation sector is not analyzed in this study since this sector does not use a significant amount of electricity and, barring a breakthrough in battery technology is expected to use very little electricity for the remainder of the cen- tury. The residential sector projections are based on the following assumptions: (1) a population growth rate (as projected by the Bureau of the Census) of 0.85%/year between 1980 and 2000); (2) a household growth rate of 1.4%/year, which would continue the trend of households growing at a rate about 607 greater than the population; (3) a continu- ation, at a modest rate, of the trend to less energy use per household; and (4) a continuation of the trend to electric space heating. *Taken from G. Samuels, The Outlook for Electricity Supply and Demand, ORNL/TM-9469, 0Oak Ridge National TLaboratory, Oak Ridge, Tennessee, April 1985, Table B.l. residential, Projected energy use for the comnercial, and industrial sectors in the year 2000 1980-2000 End use Primary Sector annual growth energy energy use (%/year) (1013 Btu/year) (1015 Btu/year) Residential Electricity 1.50 to 2.00 3.30 to 3.64 11.29 to 12.45 Nonelectricity -1.50 to -1.00 5.09 to 5.63 5,09 to 5.63 Total primary 16,38 to 18.08 Commercial Electricity 2.00 to 2,50 2.83 to 3.12 9.70 to 10.69 Nonelectricity 0 4,09 4,09 Total primary 13.79 to 14,78 Industrial Electricity 2,00 to 2.50 4,13 to 4.56 14.15 to 15.60 Nonelectricity 0.50 to 1.00 22.99 to 25.39 22,99 to 25.39 Total primary 37.14 to 40.99 U.S. total Electricity 1.83 to 2.33 10,26 to 11.32 35.14 to 38.74 Nonelectricity 0,06 to 0.50 32.17 to 35.11 32.17 to 35.11 Total primary 0.90 to 1.37 67.31 to 73.85 B-3 The commercial sector projections are predicated on a substantial decline in the growth rate of both sectoral employment and floor space—to an annual rate of 1.5%Z. Electricity use per employee or per unit of floor space was assumed to increase at a rate 0.5 to 1.07% greater than employment or floor space. The industrial sector projections are based on a detailed analysis of the manufacturing industries between 1975 and 1980, which examined changes in the energy intensity and output of these industries at the four—-digit Standard Industrial Classification level. Electricity use for these industries is projected to grow at a rate equal to about 80% of the gross national product growth rate, which is expected to be in the range of 2.5 to 3.0%Z for the remainder of the century. Although these estimates are small compared to most projections of several years ago, they are in the range of recent projections and close to current "conventional wisdom.” An examination of past energy use suggests that the rapid growth between 1950 and 1970 was self limiting and that the o0il price shocks of the 1970s were a catalyst that ended this rapid growth. The technologies that led to this growth were avail- able by 1930. However, the Depression and World War II delayed their growth, which resulted in their impact being compressed into a shorter time span and the rapid growth of the 1950's and 1960's. The utilities' projections of future demand and their plans for future generating capacity have declined steadily over the past ten years., Projections for peak demand and electrical energy requirements in 1992 represent a 2.25 and 2.61%/year growth from actual 1980 values. Their projections indicate that average reserve margins for the contiguous United States should be adequate through 1992. Reserve mar- gins are projected to decline slowly from 417 in 1982 to 30% in 1992. Furthermore, based on utility projections, each of the nine regional re- liability councils will have reserve margins of at least 207 in 1992. However, the adequacy of both regional and U.S. electricity supply depends primarily on the validity of the drastically reduced projections of future demand growth and to a lesser extent on the utilities' ability to provide the planned generating capacity. For example, if utilities were to complete only those units now under construction and if demand grows as projected, 1992 reserve margins would be 22 to 23%. However, if demand were to reach that projected in 1980 (a 47 annual growth rate), completion of all currently planned capacity by 1992 would pro- vide only a 6% margin—far too small to maintain service during peak demand periods., The sensitivity of reserve margins to the demand growth rate, com- bined with a long lead time required to add economical capability, has led to concerns about the adequacy of future electricity supply. At the same time consumer resistance to higher electricity prices and the re- sulting pressure on Public Utility Commissions has seriously affected the utilities' ability to finance the capacity now being built. Adding more capacity as insurance for an unexpected increase in demand would be difficult to sell to either consumers or utilities at this time. B-4 Relatively low—cost approaches exist for lessening the probability of future electricity shortages. One approach would be to allow advance siting and permitting and then "banking” of sites so that the lead time would be reduced to that required for constructiom—about half of the current 8- to l2-year lead time. The time for which the construction permit remains valid would have to be increased. A second approach would follow a path now being adopted by a few utilities. This approach would treat conservation and load management as supply options. Utilities would, with the approval of regulators, channel capital into the most economical option to meet future service requirements whether this option be increased capacity or reduced de- mand. Treating demand-reducing options as a supply would permit “"capacity” addition to more closely match increases in demand. Further- more, this option would provide results in less time than that required for adding large central stations. This shorter lead time would also alleviate the debate over including construction work in progress in the rate base. REFERENCES [Used in G. Samuels, The Outlook for Energy Supply and Demand, ORNL/TM-9469, 0Oak Ridge Natioumal Laboratory, Oak Ridge, Tennessee 37831 (April 1985).] A. P. Sanghvi, "Least Cost Energy Strategies for Power System Expan- sion,” Energy Policy 12(1),75-92 (March, 1984). R. H. Williams, G. S. Dutt, and H. S. Geller, "Future Energy Savings in U.S. Housing,” Annual Review of Energy 8, 269-332 (1983). Survey of Utility Load Management and Energy Conservation Projects, EPRI/EM-1606, FElectric Power Research Institute, Palo Alto, Calif., November 1980. Conference Proceedings Utilities and Energy Efficiency; New Opportun- ities and Risks, October 23-24, 1980, CONF-8010146, , Port Chester, N Y. State Energy Data Report, 1960 through 1980, DOE/EIA-0214(80), U.S. Department of Energy, Washington, DC, July 1982. Statistical Abstract of the United States, 1982-83, U.S. Department of Commerce, Bureau of the Census, Washington, DC. Residential Energy Consumption Survey: Consumption and Expenditures April 1980 through March 1981, DOE/EIA-0321/1, U.S. Department of Energy, Washington, DC, September 1982. Residential Energy Consumption Survey: Housing Characteristics 1980, DOE/EIA-0314, U.S. Department of Energy, Washington, DC, July 1982. B-5 Residential Energy Conservation, Volume I, OTA-E-~02, U.S. Congress, Office of Technology Assessment, Washington, DC, July 1979. 1982 Annual Energy Outlook with Projections to 1990, DOE/EIA-0383(92), U.S. Department of Energy, Washington, DC, April 1983. The Future of Electric Power in America: Economic Supply for Economic Growth, DOE/PE-0045, Department of Energy, Washington, DC, June 1983. J. F. Gustaffero, "U.S. Energy For the Rest of the Century,” EPRI Work- shop Proceedings, Palo Alto, Calif., October 25-26, 1983. Economic Report of the President, February 1982. Nonresidential Buildings Energy Consumption Survey: Fuel Character-— istics and Conservation Practices, DOE/EIA-0278, U.S. Department of Energy, Washingtomn, DC, June 1981. Nonresidential Buildings FEnergy Consumption Survey: Building Character- istics, DOE/EIA-0246, U.S. Department of Energy, Washington, DC, March 1981. Nonresidential Buildings Energy Consumption Survey: 1979 Consumption and Expenditures, Part 1l: Natural Gas and Electricity, DOE/EIA-0318/1, U.S. Department of Energy, Washington, DC, March 1983, 1980 Annual Survey of Manufacturers: Fuels and Electric Energy Con- sumed, Industry Groups and Industries, M80(AS)-4.1, Bureau of the Census, U.S. Department of Commerce, Washington, DC, August 1982, Monthly Energy Review, DOE/EIA-0035(93/08), U.S. Department of Energy, Washington, DC, August 1983. Survey of Current Business, 62(7), Bureau of Economic Analysis, U.S. Department of Commerce, Washington, DC, July 1982. Survey of Current Business, 63(7), Bureau of Economic Analyses, U.S. Department of Commerce, Washington, DC, July 1983. G. Samuels, D. P. Vogt, and D. M. Evans, Shifts in Product Mix Versus Energy Intensity as Determinants of Energy Consumption in the Manufac- turing Sector, presented at Electric Power Research Institute Workshop on Forecasting Industrial Structural Change in the U.S.A., October 25- 26, 1983. U.S. Industry Outlook 1977 with Projections to 1985, U.S. Department of Commerce, Washington, DC, January 1977. 1979 U.S. Industrial Outlook with Projections to 1983 for 200 In- dustries, Industry and Trade Administration, U.S. Department of Com— merce, Washington, DC, January 1979. B-6 1983 U.S. Industrial Outlook for 250 Industries with Projections for 1987, Bureau of Industrial Economies, U.S. Department of Commerce, Washington, DC, January 1983. Industrial Energy Use, OTA-E-198, U.S. Congress, Office of Technology Assessment, June 1983. C. C. Burwell, Glassmaking: A Case Study of the Form Value of Elec- tricity Used in Manufacturing, ORAU/IEA-82-9(M), Institute for Energy Analysis, Oak Ridge Associated Universities, Oak Ridge, TN, July 1982, C. C. Burwell, Industrial Electrification: Current Trends, ORAU/IEA-83- 4(M), Institute for Energy Analysis, Oak Ridge Associated Universities, Oak Ridge, TN, February 1983. Electric Power Supply and Demand 1983-1992, North American Electric Re=- liability Council, Princeton, NJ, July 1983. Energy Projections to the Year 2010, DOE/PE-0029/2, U.S. Department of Energy, Washington, DC, October 1983. G. Samuels, Options for Electricity Use and Management during a Petroleum Shortage, ORNL-5918, Oak Ridge National Laboratory, Oak Ridge, TN, January 1983, GADS—Generating Availability Data System, Equipment Availability Report 1972-1981, North American Electric Reliability Council, Princeton, NJ. Monthly Energy Review, DOE/EIA-0035/80, U.S. Department of Energy, Washington, DC, July 1980. 13th Annual Review of Overall Reliability and Adequacy of Bulk Power Supply in the Electric Utility Systems of North America, North American Electric Reliability Council, Princeton, NJ, August 1983. S. Kichen and L. Pittel, "Utilities: Are the Good Times Over?," Forbes (December 5, 1983). “"Around the State Legislatures,” Modern Maturity (December 1983—January 1984). in NPOVS. APPENDIX C DISCUSSION OF CONCEPTS NOT INCLUDED FOR ASSESSMENT Many reactor concepts were proposed and considered for assessment A list of those concepts that were not selected for detailed assessment follows. The exclusion of concepts was based primarily on the ground rules although other considerations contributed to the selec- tion process. LWR APWR - ABWR - CNSS - Explanations are included with each concept. The Advanced PWR by Westinghouse is considered suffic- iently developed to be available now; hence, there is no merit in NPOVS assessment of the concept as a future viable option. Furthermore, safety relies substantially on conventional and engineered systems. The Advanced BWR by General Electric is considered suffi- ciently developed to be available now; hence, there is no merit in NPOVS assessment of the concept as a future viable option. Some of the Advanced BWR features are re-— flected in the small BWR and, thus, are being considered in NPOVS, Safety relies substantially on conventional engineered systems. The consolidated Nuclear Steam Supply System concept by B&W is based on available technology and included 1little emphasis on passive safety. Steam~Cooled LWR - This "Schultz-Edlund” concept has no current ac- W-NUPACK tive vendor promoting it. As a result, it is judged that the concept will not be available as a demonstrated option by 2010. 600 - The small [600 MW(e)], barge-mounted plant offers numerous cost advantages based on the maximum use of fac-— tory quality fabrication, standardization, and modulariza- tion. Westinghouse proposes marketing the plant with an NRC final design approval so that utilities would face primarily only the site suitability issues in licensing. NUPACK will probably incorporate other design simplifica- tion and advanced fuel cycle features of the APWR. Al- though NUPACK relies significantly on passive safety it is more traditional in its approach, primarily employing en- gineered safety features. CE-Realistic Alternative Reactor - This concept calls for a self- pressurizing, single vessel, reactor-steam generator mod- ule. It is similar in many ways to the CNSS, but uses natural circulation for powered operation and does not rely on the use of control rods or soluble poison for con- trol during burnup. Pressure feedback is the control me- chanism under powered operation. Design simplification has been employed to limit the effects of many anticipated transients and traditional design basis events for conven- tional LWRs, but the ultimate safety response would still rely on the intervention of engineered safety features. LMR CANDU - The Canadian heavy water reactors have served their domes- tic needs well and have been deployed in several other countries. Thus it is a wviable option, but there is no U.S. sponsor aand the concept depends on engineered safety features for decay heat removal, A principal rational cost advantage derives from its use of natural uranium. However, this advantage is lost when enrichment exists, as in the United States. A smaller reactor, CANDU 300 has been announced recently which is to have improved features for safety and reliability, but it relies on engineered safety systems and does not meet a passive safety cri- teria. Large Pool — This collective term applies to several concepts that are being demonstrated in other countries and some con- cepts studied in the United States. The concepts reviewed have no active U.S. vendors promoting them and, hence, are not considered available by 2010. However, the EPRI-COMO program recently turned attention to a large pool design. Large Loop — The large loop IMR concepts (other than the LSPB) have no active proponent that would accomplish a demonstration of the concept by 2010. These concepts are designed with active, diverse, and redundant safety and do not emphasize passive safety. The economic approach to the large 1loop LMR is based on the need for the breeder and thus do not meet the economic ground rule with present and near—-term fuel prices. W~Pool - The Westinghouse pool LMR concept was one of the con- tenders for the DOE support of advanced concepts. Orig- inally relying on an integrated fuel cycle with on-site reprocessing and refabrication, it was later changed not to require the integral fuel cycle. Not enough informa- tion and detail have been available to NPOVS to include this concept in the detailed assessment. Hybrid - The Stone & Webster concept is based on two vessels, one for the core and one for components, connected by pipes. Not enough information is available to NPOVS to include this concept in the detailed assessment. Thermal IMR - The moderated core, cooled by liquid metal, has no current sSponsor. It is judged as not available by the year 2010, and not enough information is available for it to be considered in an assessment. There is little infor- mation about its present economic potential or its passive safety features. IFR ~ The Integral Fast Reactor, based on metallic fuel, inte- grated pyrometallurgical reprocessing and on-site fabri- cation, with the emphasis on metallic fuel, is promoted by ANL. The reactor portion of the concept was not developed in sufficient detail for assessment. The 1lack of an active vendor contributed to the concept not being judged GCR MSR HTR - GCFR - AGR - MSR - Other - C-3 available for deployment by 2010. However, features of this concept have been incorporated in the SAFR and PRISM concepts that are included in this report. Also, an analysis of the fuel cycle is presented in Appendix E. This collective name applies to various versions of the High-Temperature Gas-Cooled Reactors. 0f these, the "Side-by~Side"™ prismatic fuel concept was chosen for assessment. Other concepts were not examined in detail because the side-by-side modular concept had been selected for detailed study within the U.S. HTR Program. However, experience from the pebble bed concept now operating in two German demonstration wunits was wutilized in the study. The 860-MWe large HTR has been included as an appended reference since much of the HTR technology devel- opment has been related to this concept and because it has significant passive features, see Appendix F. The Gas-Cooled Fast Reactor has no current active propo- nent and hence is judged not to be available by the year 2010. Also, the available designs for a GCFR do not in- corporate significant passive safety features. This British designed and operated GCR has reached a point of virtual standardization in the Heysham II and Torness single—-cavity PCRV designs. These plants share the large capital investment requirements of the large HTR but at a lower power rating due to lower gas temperatures for the carbon dioxide coolant. Therefore, competitive capital costs in a U.S. market would be very doubtful. Recent tests at Hinkley Point B have shown adequate passive cool- ing of the pressurized core to the PCRV concrete without damaging fuel or 1liner; however, the depressurized core cooling does require forced convection. As at Fort St. Vrain, liner cooling of the PCRV must be maintained to re- tain any released fission products resulting from a de- pressurized loss of normal heat sink. All Molten Salt Reactor versions are excluded from de- tailed assessment since having uno current active pro- ponent, they cannot become available by 2010. Designs for molten salt concepts date back many years. Passive safety is not advertised, although many passive features are evi- dent and some can be considered "inherent” to liquid fuel systems. Economic estimates that were made are all obso- lete and cannot be used for evaluating economic viability. A few other concepts ("exotica") such as the fluidized bed reactor were briefly considered and rejected for lack of design information, lack of a sponsor, and insufficient other information. APPENDIX D R&D GOALS AND SPECIFIC REQUIREMENTS FOR LIQUID METAL REACTOR (LMR) CONCEPTS Table D.l is a detailed presentation of R&D needs judged by the LMR designers as essential or important to the success of their specific power plant designs.l™3 Similar needs have been combined. The table also indicates which R&D needs might apply to other reactor concepts and provides justification for inclusion of each need. It should be emphasized that Table D.l includes only those R&D tasks required to complete a design to meet requirements and specifications. Several challenges were identified for the IMR industry in the sec- tion dealing with advantages and disadvantages of the concepts. Consid- eration is given here to general R&D goals which could help meet these challenges. However, to put this discussion in perspective, two assertions are made and potential goals formulated. First, the IMR has a long-term potential for breeding to extend fuel resources. Therefore, one goal should be to maintain the capability to meet this challenge. We also assert that the worldwide nuclear program will be sustained through the NPOVS time frame, that a significant market for IMR converters and/or breeders eventually will develop, and that U.S. industry will seek a share of this market. Thus, a second goal should be to sustain a competitive IMR industrial potential in the United States for a significant range in growth rates of domestic and foreign power needs. A competitive industry would have an adequate number of properly trained technologists, up—-to-date and appropriate facilities, and a competitive design to sell. These general R&D needs have been organized in the form of a hierarchy in Table D.2 where these goals and programs have been categorized within three major headings. This table, though preliminary, provides a framework for evaluating the importance of various R&D activities. For example, one could determine the rela-— tive importance of the listed and augmented R&D tasks under the scenario of low, modest, and high growth rates for the utility industry. Pre- liminary assessments indicate that the list of R&D tasks is not sensi-— tive to the schedule; only the relative importance of the R&D tasks was scenario dependent, Table D.l. Specific research and development needs identified for the LMR concept LMR concepts for which the R&D needs was Other concepts identified (*) for which the R&D or applicable (X) may be applicable Justification for the R&D need Research and Development (R&D) needs identified by the designer Demonstrates Supports Increases Investor or as essential or important LSPE SAFR PRISM LWR HTR low cost licensing public confidence SAFETY-RELATED REQUIREMENTS Advanced Core Design Evaluate core features which can assure * * X X X a benign core response to core disrup- tive accident initiators,. Demonstrate a low—cost and reliable * * X X X X approach tor a self-actuated shut- down system, Provide experimental verifications of this concept needed for licensing discussions. Develop high temperature, wide range, * * X X X X X fission channels and ion chambers for power monitoring at in-vessel locations, and high sensitivity source-range fission channels for startup monitoring. Perform experiments to demonstrate the * * * X X X X effectiveness of B4C as an in-vessel shield and verify shield design. Perform detailed shielding and flux calculations needed for the design. Perform core critical experiments at * * * X X X ZPPR to provide nuclear parameters and detector requirements, and test loading sequences for all cores considered. Provide analytical verification ot X X * X X benign response of the core to all design basis accidents, Table D.1l. Specific research and development needs identified for the IMR concept (continued) Research and Development (R&D) needs identified by the designer as essential or important Perform seismic analysis and tests to predict the response of the core and reactor assembly to a seismic event. Validate the code used through experimental tests. Collect and apply data on joint failure probabilities of- key components for use in reliability and risk assessment calculations, Develop methodologies for assigning probabilities for accident sequences associated with core disruptive accidents. These probabilities are to be used in Event Trees and PRA studies for core responses and structural responses. Evaluate the reactor system and other system responses to earthquakes to determine seismic event categories, evaluate safety systems reliabilities, and provide inputs to PRA studies, Develop necessary input and perform analyses needed to quantify containment response event trees for accident sequences, LMR concepts for which the R&D needs was identified (%) or applicable (X) Other concepts for which the R&D may be applicable Justification for the R&D need LSPB SAFR PRISM * X * LWR HTR Demonstrates low cost X Supports licensing X Increases Investor or public confidence X ¢ - Table D.l. Specific research and development needs identified for the IMR concept (continued) LMR concepts for which the R&D needs was identified (*) or applicable (X) Other concepts for which the R&D may be applicable Justification for the R&D need Research and Development (R&D) needs identified by the designer as essential or important LSPE SAFR PRISM Develop and verify a 3-D coupled thermo- * X * hydraulic, mechanical, neutronic, tran- sient code used to support the design of inherently safe reactor cores. Develop a thermal-hydraulics core to characterize temperature and flow fields in large cores, Provide friction and wear correlations * to support innovative core holddown designs. Perform testing and analysis to quantify * corrosion and tission product migration and plateout in a sealed vessel without cleanup systems. Develop and validate an analysis code * X X that can calculate deformations of various core components due to creep and swelling and calculate mechanical loads thereby produced. Shutdown Heat Removal Perform experimental simulations of DRACS, * * * RVACS, and RACS systems to evaluate their passive design features, demonstrate their operating principles, and optimize their performance. Review 1984 tests of the CRBRP NDHX system and understand uncertain- ties in the performance of this system at low air flow conditions. Perform 3-D thermal-hydraulic analysis of the RVACs performance. Perform tests of associated flow control devices. LWR HTR Demonstrates low cost Supports licensing X Increases Investor or public confidence X b-Q Table D.l. Specific research and development needs identified for the LfiR concept (continued) LMR concepts for which the R&D needs was identified (%) or applicable (X) Other concepts for which the R&D may be applicable Research and Development (R&D) needs identified by the designer as essential or important LSPB SAFR PRISM Determine and increase, if necessary, X * X .the immunity of the decay heat removal function to sodium fires. Perform tests to verify design margins * X X for creep of bellows in piping systems at elevated temperature, FUEL-RELATED REQUIREMENTS Integral Fast Reactor Design and evaluate the performance of X * * a metal (U-Pu-Zr) core. Test fuel assemblies in to demonstrate per- formance for normal and cff-normal conditions. Evaluate and demonstrate the repro- X * % cessing and refabrication of metal fuel . Validate safety claims associated X * with metal fuel. Long-life core Perform extended burnup tests at * * * FFTF of core materials, and blanket, and control assemblies. Demonstrate RBCB performance at EBR-II. LWR HTR Justification for the R&D need Demonstrates Increases Investor or public confidence ¢-ad Table D,1. Specific research and development needs identified for the IMR concept (continued) LMR concepts for which the R&D needs was identified (*) or applicable (X) Other concepts for which the R&D may be applicable Justification for the R&D need Research and Development (R&D) needs identified by the designer as essential or important LSPE SAFR PRISM LWR HTR Demonstrates low cost Supports licensing Increases Investor or public confidence Perform characterization test of X * * cladding and duct materials at FFTF to determine irradiation effects at prototypic temperatures. Perform transient tests of high- * burnup fuel pins to demonstrate reliable performance under upset condirions. Automated Fuel Fabrication Evaluate approaches and provide conceptual design of fabrication processes and equipment system requirements. SYSTEM—-AND COMPONENT-RELATED REQUIREMENTS Plant Experience Utilize operating reactor experience * * and data to evaluate shielding pre- dictions, core performance predictions, and flux monitor responses, and verify under-sodium—-viewing device performance. Establish and test methods to detect, * * locate, and fix steam generator leaks. Fabricate a prototypic detection and location system. X X 9-a Table D. lc Specific research and development needs identified for the LMR concept (continued) IMR concepts for which the R&D needs was identified (*) or applicable (X) Other concepts for which the R&D may be applicable Justification for the R&D need Research and Development (R&D) needs identified by the designer as essential or important LSPB SAFR PRISM Improve the design of conventional cold * X X traps or develop new designs which are more reliable, thereby improving plant availability. Perform analyses and testing associated * with the PHTS siphon breaker to determine its position, size, erosion/corrosion resistance and reliability, Advanced Plant Technology Perform hydraulic tests, using a scale * * * model, of the temperature and fluid flow of the plenum, IHX, vessel wall, and reactor vessel under normal power and natural circulation conditions. Evaluate various candidate materials as * in-vessel insulation between the closure head and sodium surface, with particular attention given to French designs. Study the effectiveness of redan * as a thermal barrier and pressure seal. Develop an approach for automating main- * tenance functions for a multi-module reactor site. Investigate methods for validation and * verification of software used in reactor control and protection systems. LWR HTR Demonstrates low cost X Supports licensing Increases Investor or public confidence [~a Table D.1. Specific research and development needs identified for the ILMR concept {(continued) LMR concepts for which the R&D needs was identified (*) or applicable (X) Other concepts for which the R&D may be applicable Justification for the R&D need Research and Development (R&D) needs identified by the designer as essential or important LSPB SAFR PRISM LWR HTR Demonstrates low cost Supports licensing Increases Investor or public confidence Investigate advanced instrumentation and * * * control stress autcmation, distributed control multiplexing, improve measurement sensors and systems, simplified maintenance, use of artificial intelligence, operator aids, and human engineering. Steam Generator Performance Conduct steam generator endurance tests X * * to demonstrate long-term integrity. Investigate other options to simplify the overail system, Investigate the performance of booster tubes in steam generators. Recommend or reference system and ideatify necessary key features tests. Verify that existing inspection tech- * X X niques for S$SGs meet code requirements and develop new techniques which might be used at elevated temperatures and in the presence of sodium. Improve computer code predictions of * X X 5G performance under low sodium flow conditions. Improved Materials Obtain code approval for advanced * * * materials and simplify or improve code rules for conventionial materials. X X X 8- Table D.l. Specific research and development needs identified for the IMR concept (continued) ILMR concepts for which the R&D needs was Other concepts identified (*) for which the R&D or applicable (X) may be applicable Research and Development (R&D) needs identified by the designer as essential or important Perform thermal striping tests for various candidate materials for upper internal designs. Evaluate materials which could enhance radiative heat transfer associated with decay heat removal systems. Evaluate purification methods for primary sodium and cover gas in a sealed, vessel during normal operation, and during refueling. Verify the capability of an under sodium viewing system to satisfy in-service inspection requirements and refueling inspection require- ments. Perform component testing and obtain information from the British and French concerning location of failed fuel using a sodium sipper. Develop a high sensitivity, fission channel with remote signal trans- mission capabilities for use as a monitor of initial core loadings. Justification for the R&D need LSPB SAFR PRISM * * LWR HTR Demonstrates low cost X Supports licensing X Increases Investor or public confidence 6—d Table D.1l. Specific research and development needs identified for the LMR concept (continued) Research and Development (R&D) needs identified by the designer as essential or important LMR concepts for which the R&D needs was identified (*) or applicable (X) Other concepts for which the R&D may be applicable Justification for the R&D need LSPB SAFR PRISM Advanced Sodium Component Feature Tests Develop advanced pool-pumps such as a compact, self-cooled electro-— magnetic pump and a shrouded inducer pump. Test contreol rod drive-line designs which provide inherent negative reactivity in response to core excursiaons. Increase confidence in the use of flexible joints in piping system through their testing at EBR-II. Determine flow distributions and investigate vibrations for IHXs through tests of physical models. Develop a conceptual design of an innovative refueling and main- tenance system, and demonstrate key features by testing. Demonstrate the functioning of the core support systems through tests using an engineering scale model., Develop methods for under-sodium, in-service inspection of heat exchanges. X * % X X * X * X X * X * * * Demonstrates Supports Increases Investor or LWR HTR low cost licensing public confidence X X X X X X X 01 Table D.l. Specific research and development needs identified for the IMR concept (continued) Research and Development (R&D) needs identified by the designer as esseuntial or important LMR concepts for which the R&D needs was Other concepts identified (*) for which the R&D or applicable (X) may be applicable Justification for the R&D need Demonstrates Supports LSPB SAFR PRISM LWR HTR low cost licensing Increases Investor or public confidence Perform tests of the reliability * X X of the bearings and seals for the rotating plug of the closure head. Perform test to demonstrate the * X * performance under design and abnermal conditions of centri- fugal pumps, inducer pumps, and electromagnetic pumps. Perform tests to verify that primary and secondary control rod systems satisfy design requirements. X *One or more of the needs of the associated list was specifically identified for this design. XOne or more of the needs of the associated list would be applicable for this design. 11-C Table D.2. A hierarchy of R&D tasks to keep the IMR/Breeder option as healthy and competitive as possible considering a realistic range in future nuclear energy usage Maintain an adequate work force of technologists and appropriate up-to-date facilities Continue to improve IMR designs so that the concepts available will be competitive Continue to solve institutional problems and improve the marketability of IMR concepts Support R&D that increases the design options available for new LMR concepts and significantly improves the technology Materials research for higher operating temperatures and improved efficiencies Steam generator designs to eliminate sodium—water reactions (double-wall concepts) or provide instrumentation for more accurate and reliable detection Improved instrumentation and control to incorporate advances in automation, artificial intelligence, digital control etc. Advanced oxide fuel designs for higher burnup, and metal fuels for safety and reprocessing advantages Improved core designs for a once-through cycle so that reprocessing is not necessary for cost competitiveness Support university research to maintain a continuous supply of technologists Complete and demonstrate technical solutions to long-established design challenges Demonstrate passive safety against core~disruptive accidents Establish the plant size and con- figurations which have the potential for lowest power costs Demonstrate cost competitive off- site and/or on-site reprocessing and refabrication Support standardization of design for improved licensability Demonstrate simpler, passive, decay heat removal concepts Strive for a significantly better reactor design with convincing advantages in cost, public acceptance, licensability, etc. Produce, test, and qualify whole plant designs and/or components such as steam generators, pumps, etc., which can be sold to non-U.S. markets Provide monetary incentives for utilities and industry to build LMR demonstration plants and facilities (for example, license them as R&D facilities Encourage and support R&D that increases consumption of electrical energy within the guidelines of national policy (for example, support storage battery research to make electric cars attractive) Decrease the complexity and shorten the time required for licensing (for example, one step licensing process) Increase utility involvement in IMR technology through personnel exchanges, joint research, etc. Complement non-U.S. R&D and commerciali- zation activities so information exchange with other countires will be mutually beneficial If required, obtain support for LMR tech- nologists and R&D facilities from closely related areas such as defense, space research, etc., so that their skills will be maintained c1-a REFERENCES FOR APPENDIX D 1. 3. LSPB Research and Development Requirements, CDS 500-6, U.S. Depart-— ment of Energy, Washington, D.C., September 1984. APPLIED TECHNOLOGY. SAFR Requirements for Base Technology Program, 149T1000002, Rockwell International Rocketdyne Division, Canoga Park, California, January 1985. APPLIED TECHNOLOGY. Letter number XIL-897-850016 from L. N. Salerno to Francis X Gavigan, dated January 11, 1985, "WBS2BO.5 — Initial PRISM R&D Requirements Statements.,” APPENDIX E LIQUID METAL REACTOR (LMR) FUEL REPROCESSING- REFABRICATION EVALUATION Je Te Bell and D. C. Hampson This evaluation was developed to compare IMR fuel recycle systems for oxide and metal fuels. An adequate data base was not then avail- able, partiecularly for the metal fuel of the Integral Fast Reactor (IFR). An extensive program of study is now in progress at Argonne National Laboratory to develop a metal fuel system for the IMR and thus Fill this wvoid. Although the following evaluation is preliminary, it illustrates the questions that must be resolved to arrive at a final comparigon of the fuel systems. The conclusions should be viewed quali- tatively since the quantitative results are subject to revision as new data are developed by the Argomne study. However, our principal con- cerns for Argonne's estimates of the amount of research and development required and for the project costs, both of which are lower than our analysis indicates, have not been alleviated by work published to date. On the other hand, the scientific quality of the process research reported appears to be excellent. This evaluation classifies fast reactor fuels as either oxide or metal. Reprocessing of oxide fuels is considered only with the Purex process, and the results are based primarily on ORNL experience over three decades. Metal fuels reprocessing is considered for an Argonne National Laboratory (ANL)-developed pyroprocess that includes molten salt and electrochemical techniques. The discussion of Purex processing will relate directly to any fast reactor concept with mixed-oxide (MOX) fuel, while the discussion of metal fuel reprocessing relates directly to the Argonne Integral Fast Reactor (IFR). The total reactor output for each concept is assumed to be 1300 MW(e). Most studies and programs for oxide fuel reprocessing have been for substantially larger plants. The small size here is chosen to match the IFR concept. These two fuel reprocessing schemes will be compared, and the resulting analysis should be generically applicable to other reactor concepts when metal and oxide fuels are considered. The metal fuel could be processed by the Purex route with minor mo- difications. However, this would discard one of the prime benefits of the pyroprocess, which is that the fuel remains essentially in a metal state which is amenable to refabrication steps developed for the metal fuels. The metal fuel refabrication process is somewhat less compli- cated than the pellet pressing process envisioned for the MOX fuels. A cost estimate for the Purex reprocessing of oxide fuels will be more accurate than that for metal fuels because the Purex process, in- cluding management of its wastes, has already been developed to the con- ceptual design stages for the Hot Engineering Facility (HEF) in 1978 and the Breeder Reprocessing Engineering Test (BRET) in 1984. The fused- salt electro-refining process (FSER) proposed for reprocessing metal E-2 fuels is in the development and proof-of-principle stages. Although the FSER process is less developed, our analysis will assume that this pro- cess for metal fuels is valid, in principle, and that design of equip- ment for a Hot Experiment Plant (HEP) could begin in 1986 for the oxide fuel and in 1989 for the metal fuel. It is assumed that an HEP for either process 1is required to provide design data for a commercial demonstration reprocessing plant. This evaluation applies to the 1985-2005 time frame and is based on a commercial demonstration reactor in 2005. The schedule would require a fuel reprocessing demonstration plant about 3 years later. To meet the 2005 goal, we have assumed that certain major facilities are avail- able now. The Fast Flux Test Facility (FFTF) at Hanford would be used for irradiating oxide fuels, and the Fuels and Materials Examination Facility (FMEF), also at Hanford, would be used for an oxide fuel HEP. The Experimental Breeder Reactor No. 2 (EBR II) and associated Fuel Cycle Facility (FCF) at Idaho Falls would be used for irradiating metal fuels and for a molten salt-electrochemical HEP, respectively. In both cases, it is assumed that the existing facilities (FMEF and FCF) can be modified and equipped to provide the functions of the HEP. Each HEP would be used to develop and demonstrate the proof-of-principle of the respective process and would receive irradiated fuels supplied by the associated reactor (FFTF for MOX, and EBR 11 for metal fuels). It should be noted that the FFTF does not have blanket elements, which would be present in a demonstration fast breeder reactor (FBR). The proof—-of-principle demonstrations would provide the technical informa- tion necessary for the design of a demonstration facility. It 1is assumed that these HEPs would contribute sufficiently to design infor- mation to justify a second demonstration plant. However, the latter may be a first-of-a-kind commercial plant. Without these facilities, it is unlikely that either process could be commercially demonstrated in the NPOVS time frame, E.l1 Schedules for Development of Commercial Demonstration The 2000—2010 period has been selected as a feasible objective for demonstration of a selected new power reactor and the associated fuel cycle. Although the necessary time for reprocessing would be 3 to 5 years after the demonstration reactor goes on line, a fully developed fuel cycle would be essential to adoption of the IFR concept. To establish schedules for developing the reprocessing/refabrica- tion systems for oxide and metal fuels, it was presumed that an existing facility could be modified for specific needs of each process. This further implies that the use of the FCF for metal fuels or the FMEF for oxide fuels would be adequate to provide proof-of-principle information for either fuel cycle. However, use of these existing facilities would not provide the hard-schedule financial data for construction that are required for the commercial phase; such data would be a product of the demonstration phase. E-3 The major difference between the two schedules is that the metal fuel reprocessing must be initiated with a process/waste development phase, while this work is not needed for the oxide fuel reprocessing program. Adequate development work has been done on the oxide program to permit immediate initiation of the design activities. The metal fuel recycle program would be divided into three components: ® process/waste development ® proof-of-principle runs in the FCF, and ® design and construction of a demonstration fuel cycle plant at a location to be determined. The oxide fuel recycle program would be divided into two major compo- nents: ® proof-of-principle rumns in the FMEF, and ® design and construction of a demonstration fuel cycle plant at a location to be determined. Three constraints have been incorporated into the schedule for metal fuel processing: 1. The process must be proven on a laboratory scale prior to design of the modifications for the FCF. 2. Cold-testing of the process must be successfully completed in the FCF prior to start of design for the demonstration plant. 3. Hot-testing of both the separation process and the waste process should be completed prior to start of construction of the demon- stration plant. The one constraint that was deemed necessary for oxide fuel processing was that the integrated hot-testing should be completed prior to start of construction of the demonstration plant. Refabrication of both fuels must be included in the respective HEP, and proof-testing should be com- pleted before initial construction of the demonstration facility. The two schedules are based on a commitment to conduct the various activities within the time frames shown. This is critical, and delays in any phases of the program would be reflected in corresponding delays in subsequent phases. As can be seen in the table at the end of this section, there 1is little room for slippage 1in either fuel cycle schedule. These are the most optimistic feasible schedules and are given only to show that a 2005 objective could be achieved. These schedules can be accomplished only if the following are done: E-4 l. The metal-fuel schedule is started immediately. 2. Experimental reactor space for fuel testing is dedicated to this effort. 3. Existing facilities are available and dedicated to the HEPs. (There is no time for construction of new HEPs.) 4. Waste management 1is developed parallel to the chemical pro- cessing. (This will require an additional effort in the metal fuels program since little work has been done for this waste to date. The necessary development may extend past the 1997 date.) 5. Fuel fabrication must be developed simultaneously with the chemical reprocessing. 6. Plutonium must be available for fuel testing and for irradiation in order to have adequate spent fuel available for either HEP in the 1991 time frame. Summary Schedule for Development Metal Oxide Laboratory experimentation started 1985 NR* Complete laboratory experimentation 1987 NR* Start equipment design for FCF or FMEF 1988 1988 Start modifications to FCF or FMEF 1989 1989 Install equipment in FCF or FMEF 1990 1991 Complete base experimental program 1997 1993 Start design demonstration plant 1994 1989 Start construction of demonstration plant 1998 1993 Start design of demonstration equipment 1996 1991 Install demonstration equipment 2000 1996 Complete construction 2002 1997 Start operating the demonstration plant 2004 1999 *NR indicates steps not required. Thus, the oxide schedule could be advanced one year. E.2 The Two Processes The flowsheet for pyrochemical processing of metal fuell and an equivalent Purex flowsheet for oxide fuels? were evaluated. The block- diagram flowsheets described below are based on the conceptual design efforts for commercial~scale fuel cycle facilities to serve 1200 to 1400 MW(e) electricity generating capacity (nine PRISM modules or four SAFR power paks). Refabrication flowsheets follow Argonne information for metal fuels and the mechanical-blending pellet forming process for oxide fuels. E-5 E.2.1 The Pyrochemical Process for Metal Fuels The present IFR concept (1985) includes a reactor-generator system operating in a fuel break-even converter mode but constructed with the potential of operating in a fuel breeder mode. For this analysis, mul- tiple reactors are considered in a 1300-MW(e) unit with associated fuel cycle facilities. A fuel burnup of 117 is proposed without an axial blanket. The break-even operation requires a reprocessing system with a throughput of 7.5 tonnes/year (t/a) (30 kg/d) of core fuel and 8 t/a (33 kg/d) of combined internal and radial blanket fuel.3 However, if the reactor has potential for future fuel breeding, the reprocessing facility must be initially constructed to accommodate the breeder sys- tem. The previously proposed IFR breeder required reprocessing capacities of 7.5 and 18 t/a (30 and 72 kg/d) of core and blanket fuel, respectively. This analysis will consider only the higher—-capacity requirements for the breeder mode. Again, no axial blanket 1is assumed. A block-diagram flowsheet for reprocessing metal fuel is shown in Fig. E.l. A special committee appointed by the University of Chicago has reviewed the chemistry proposed for the reprocessing of metal fuel; further discussion is not presented here. The core and the blanket fuel initially must be processed separately, and the plutonium from the blanket fuel is added to the core fuel. The blanket fuel is dis- assembled, chopped, and dissolved from the cladding by an electrodis- solution process. The fuel dissolved in cadmium is contacted with a salt mixture to oxidize the plutonium to PuCl3, which transfers into the salt solution. Some of the uranium and most of the fission products will also be oxidized and dissolved into the salt. Noble metal fission products will remain predominantly with the unoxidized uranium. After this halide slagging, the salt solution is transferred to the process line for the core fuel and 1is coprocessed through the electrolysis scheme. Some problems with the halide slagging process include waste handling and control of the metal oxidation. The plutonium is oxidized by adding UCl3 to the salt solution. The amount required will depend on the amount of fission products that oxidize before the plutonium oxidizes. The added uranium also replaces that consumed in the blanket. As shown in block form in Fig. E.l1, the fuel bundles are dis- assembled and the individual core fuel pins are punctured to release gaseous fission products and then chopped into short segments. If an axial blanket had been included, that material would be mechanically separated and transferred to the decladding and the halide slagging pro- cess for blanket fuel. The chopped core fuel is dissolved into cadmium; the cladding does not dissolve. The salt solution from the halide slagging of the blanket fuel is added to the cadmium. This results in a two-phase liquid system — a molten salt phase in contact with a molten metal phase. Also, the cladding and the undissolved fuel make up a solid phase in the cadmium liquid metal. 1In electrorefining, the metal phase is used as the anode, and a cathode is placed in the salt phase. The PuCl3 is electro-reduced to plutonium from the salt phase onto the cathode. Plutonium is transferred from the metal to the salt phase by oxidation at the anode to PuCls. The same electrolysis mechanism E-6 ORNL DWG 85-492 CaCl, MgCl, BaCl, SOLIDS *BeD CRUCIBLE 3 WASTE" SKULLS CLAD DECLADDING 4 5.6 9 ME waste [+ stee . stAGemg [ e accounTaeruT | SALT 2 10 1 WASTE @EMBLE FUME | o ACCOUNTABILITY i METAL(s) TRAP BLANKET STORAGE FUME FUEL TRAP Al FABRICATION BLANKET | | SODIUM | WATER FUEL REMOVAL CAUSTIC a EVAPORATOR COMPRESSED GAS, STORAGE CORE SODIUM LOW- o — LEVEL FUME FUEL REMOVAL WASTE 7 TRAP N soLios 1] CORE FUEL WASTE FABRICATION FUME | TRAP A} —w STORAGE SAMPLE ANALYSIS 2 WASTE : 22 METAL () DISASSEMBLE FEED .._-g PREPARATION r FOR 20 — 1 \2 FUME FABRICATION . AL REMOVE TRAP WASTE "] pLENUM - cd STORAGE Ciz 24 15,16 ¥ . PRODUCT 21 13 DISSOLUTION ELECTRO- CONSOLIDATION ENCAPSU- SHEAR CADMIUM REFINING b 23 N REMOVE T 22 19 c AND o 14 ° 7 EL%E'ST?:(-;I-)‘E wetaL | | VITRIFICATION WASTE g L e | | REmOVE PRODUCT CONSOLI~ BASKET DATION SALT l ;' CONVERSION TO OXIDE Figure E.l. The block diagram flowsheet for pyro—electro—-chemical reprocessing fast reactor metal fuel. Numbered blocks indicate space requirements in the containment, and multiple numbers per block indicate duplicate equipment. The decladding step, block 3, has not been defined but is thought to be feasible by an electrochemical process. Any such process will likely require at least three additional space require- ments. The letters a, b, ¢, d indicate a sequence of steps following the electro-refining. During startup of the reactor with enriched uranium, the core uranium cannot be isotopically diluted with blanket uranium, and the blanket fuel may need to pass two times (or more) through the halide slagging step to reduce to uranium content. E~7 applies to the reduction of UCl3 to uranium and of ZrCly to zirconium, and the product is a mixture or an alloy of the three metals. Control of the reducing potential can partially segregate the metals on the cathode. E.2.2 The Purex Process for Oxide Fuels The assumed LMR design, as for the metal fueled system, is a 1300-MW(e) facility. The core and blanket fuels can be processed in the same equipment. A reprocessing facility on the site for this 1300-MW(e) complex would require a capacity of 35 t/a (8 t/a of core and 27 t/a of blanket fuel) or about 140 kg/d at 250 d/a. The Purex process for reprocessing oxide fuels 1is diagrammed in Fig. E.2 and has been previously described.? Both the core and the blanket fuels are processed through the same equipment. The head-end operation involves disassembly of the fuel bundles and chopping of fuel into 1-in. segments. Dissolution includes transfer of fuel from the cladding into nitric acid and removal of cladding hulls from the pro- cess. In the feed preparation step, the chemistry of the dissolved fuel is adjusted for solvent extraction. The details of solvent extraction strongly depend on the character- istics required for the plutonium and uranium products. OQur evaluation will be based on a conventional Purex solvent extraction, in which the solvent extraction step yields pure uranium and plutonium products. The product-conversion step converts aqueous nitrate solutions of plutonium and uranium into the respective oxides. The reference design includes oxalate precipitation followed by thermal decomposition. E.2.3 Fabrication of Metal Fuels for IFR Fabrication of the metal core and blanket fuels will involve separate operations in the initial steps. However, the same equipment can be used during and after the fuel-rod assembly stage. All fabrica- tion steps must be conducted remotely within thick containment shielding because both the recycle uranium from the halide slagging and the repro- cessed plutonium/uranium/zirconium will retain appreciable quantities of fission products. Figure E.3 shows a block diagram for metal fuel fabrication, which is in accord with the fabrication of EBR II fuel. As in the reprocessing, the capacity for fuel fabrication must be capable of quantities for the breeding design basis, so that the IFR will have breeding as a future option. Batch sizes may be 1limited by criticality considerations. The criticality safety analysis for these process steps has not been com- pleted, and therefore the effect of criticality on batch sizes cannot be firmly evaluated. IFR breeder systems with a total output of 1300 MW(e) will require a total processing-refabrication capacity of 25.5 t/a (30 kg/d and 72 kg/d for the core and blanket fuels, respectively, for ORNL DWG B85-493 LOW--LEVEL LIQUID WASTE IRRADIATED | Na FUEL REMOVAL TO HEPA FILTERS AND STACK | FUEL STORAGE | 1" 1 NO. 1 SOLVENT NO. 2 SOLVENT 1 I0DINE CLEANUP TO Law CLEANUP [~ TO LAW CROPPING ' : Ty P 1 » CONDENSERS Ru WASTE _ 9.10 L g2 12 N : I ! 12 3 Ru TRAP : ACCOUNTABILITY st CYCLE || DIGESTION 2nd U CYCLE SHEARING — AND FEED CODECON— 3rd U CYCLE H—={U0,(NO3) DISSOLUTION , TANKS CLARIFICATION TAMINATION U-Pu PARTITION : ; e DISSOLVER | 14 5 RESIDUES , HULLS MONITOR 25 ! | P7] 2nd Pu CYCLE EXCESS WATER TO —TO VOG ‘ 3 LOW-LEVEL WASTE T0 [ 14 METAL r“‘“' 4" 3rd pu cveLE WASTE | RECYCLE WATER ‘24 I 22 TO PROCESS CONDENSER g 15,16 ACID RECYCLE ACID |23 % PLUTONIUM RECOVERY l '| TO PROCESS ] £ K Y LE NITRATE STORAGE 21 ——=T0 VOG J - LAW —— LAW VITRIFICATION - (WASTE} CONCENTRATION OFF—GAS , [~ Rv WASTE Pu FILTRATE PLUTONIUM PRECIPITATION CONDENSER 1 18 26 Pu DISTILLATE | pLuTONIUM - TO FUEL HAW CALCINATION FABRICATION CONCENTRATION —— SOLIDS, AQUEOUS LIQUIDS, 27 l 28 AND GASES HLW VITRIFIED ——— ORGANIC LIQUIDS VITRIFICATION HLW Figure E.2. The block diagram flowsheet for Purex reprocessing of fast reactor oxide fuel. Numbered blocks indicate process steps that require space in the containment facility, and a number common to more than one block indicates that these steps share a space. For example, the solvent extraction steps, #9, are accomplished in an eight-pack of centrifugal contactors. 8- E-9 ORNL DWG 83-523 6 7 20 CAN WELD AND STORE LOADING LEAK TEST 19 S 8 STRAIGHTNESS INSPECT AND BOW TEST 4 9 19 1 DEMOLD BOND TENSILE AND SHEAR TEST TEST h.2.3 I 9 BLANKET INJECTION STRAIGHTEN FUEL CASTING SPADE 21 L‘19 19 WASTE HEXAGONAL [—- ASSEMBLY CAN INSERT CORE INJECTION STRAIGHTEN FUEL CASTING SPADE 13 110, 11,12 18 1 DEMOLD AND BOND SHEAR TEST 14 17 INSPECT BOND 15 16 CAN WELD AND LOADING LEAK TEST Figure E.3. The block diagram flowsheet for fabrication of fast reactor metal fuel. Numbered blocks indicate processes or steps that require space in the containment facility, and multiple numbers per block indicate duplicate equipment. A single number on several blocks indicates that these steps are accomplished in a common space. E-10 250 d/a). Thus, as a result of criticality considerations, the core and the blanket process may each require more than one injection casting unit. The cast fuel will be cooled and the mold removed. The spent mold becomes TRU waste but could, perhaps, be cleaned to low-level status. The fuel will then be inspected and a chemical analysis com- pleted. The fuel that passes inspection will be loaded, along with some sodium, into new cladding. The cladding will be subsequently welded and leak-tested. Approved fuel rods will be bonded into bundles, and this bonding will be inspected. The positioning spade on the bottom of the fuel rods will be straightened, and the rods will be loaded into an assembly. Finally, each assembly will be equipped with the hexagonal can 1insert and welded. Tensile testing will be performed, and the approved assemblies will be examined to assure meeting the requirements for insertion into the reactor. E.2.4 Fabrication of Oxide Fuels Feed material for fabrication of oxide fuels will have relatively low gamma activity because the Purex process removes essentially all of the fission-product elements from the plutonium/uranium products. Therefore, this fabrication can be housed in a low-level containment facility. However, automated remote operations and maintenance will be required because of the long-term increase in the even—-numbered pluton- ium isotopes. The 1300-MW(e) reactor system will require 8, 9, and 18 t/a of core, axial, aund radial blanket fuels, respectively. While the Purex reprocessing must handle all 35 t/a of spent fuel, the fabri- cation facility will prepare only the core fuel (8 t/a); the blanket fuel (27 t/a) will be purchased from an independent vendor. The vendor could be any of those that currently supply LWR fuel. A block-type schematic for fabrication of oxide breeder fuel is shown in Fig. E.4. This flowsheet is in accord with the fabrication flow-sheet for the Secure Automated Fabrication line." The recycled Pu02 and the U02 are first blended in the proper batch quantities. The mixed oxide is milled to achieve uniformity, and a binder material is added. The fine material is then compacted into granules, and a lubri-~- cant 1is added. This mixture is pressed into pellets and loaded into boats for high-temperature treatment, which serves to remove the binding material and sinter the pellets. The sintered pellets undergo several grinding, gauging, cleaning, and inspection steps before they are loaded into the cladding jackets. The loaded jackets are filled with helium and welded closed. Helium leak-testing and ultrasonic testing are car- ried out successively. The fuel rods proceed through fissile assay and physical inspection, and rods that pass all inspections are wrapped with spacer wire and assembled into bundles. A final imnspection of the assembled bundles completes the fabrication process. E-11 ORNL DWG 85-524 1 2 3 4q uo, - MILLING COMPACTION LUBRICANT Pu0, ———={ BATCHING AND AND ADDITION RECYCLE —e BLENDING BINDER ADDITION GRANULATION 11 10 9 7,8 5,6 PELLET PRESSING INSPECTION STORAGE AND SINTERING LOADING 2 | i i CLi‘m')'”G BOAT CLEANUP CHEMICAL GAGING AND INSPECTION ANALYSIS 13 14 15 16 17 COLUMN MAKEUP DECON- HELIUM FILL HELIUM LT i PIN LOAD TAMINATION AND WELD LEAK TEST TESTING 26 25 22,23,24 20, 21 18, 19 ASSEMBLE VISUAL FISSILE ASSAY STORE AND WIRE DIMENSION COMPONENT INSPECT WRAP CHECK PLACEMENT Figure E.4. The block diagram flowsheet for fabrication of fast reactor oxide fuel. Numbered blocks indicate processes or steps that require space in the containment facility, and multiple numbers per block indicate duplicate equipment. E-12 E.3 COSTS OF DEMONSTRATION PLANTS FOR REPROCESSING AND REFABRICATING METAL AND OXIDE FUELS Costs have been estimated for the fuel cycle plants for fast reac- tor facilities with 1300-MW(e) outputs. In Table E.l1, through-puts for such reprocessing facilities are given as tonnes per year (t/a) for metal fuel from an IFR conceptual reactor operating as a breeder and for oxide fuel from a PRISM conceptual breeder reactor. Reprocessing of either type of fuel is conducted in remotely operated, shielded facil- ities. Refabrication of both the core and blanket of the metal fuel is conducted in remotely operated, shielded facilities. However for oxide fuels, only the core is refabricated remotely, the blanket fuel is pur- chased from a commercial vendor. Table E.l. Fuel throughputs (t/a) for reprocessing and refabricating metal and oxide fuels from 1300-MW(e) fast reactors Total for Refabri- remote Core Axial Radial Reprocess cation Purchase operations Metal fuel 7.5 0 18 25.5 25.5 0 51 Oxide fuel 8 9 18 35 8 27 43 The costs of reprocessing and refabrication facilities are highly dependent on the sizes of the containment buildings, because of the shielding and ventilation requirements. Therefore, the relative costs of fuel cycles for oxide or metal FBR fuels can be based on the relative sizes of the required facilities. As a first approximation for compar- ing costs, we have assumed that capital costs are proportional to the number of process steps that require space in the shielded, remotely operated facility. We have numbered such steps in Figs. E.l—E.4. Several studies have calculated and examined the effects of various parameters on reprocessing costs for oxide fuels. Those are summarized in Ref. 5. Those costs and the costs herein are expressed as current- dollar levelized costs (1984 dollars). If the annual levelized fixed charge rate had been based upon constant dollars, the apparent cost would have been lower. Using the same scaling factors (0.5) that are used in those studies and assuming, as did Delene et al.,5 that a facil- ity for reprocessing 150 t/a would cost $1.02 billion, we estimate that the capital cost of an oxide fuel reprocessing facility for a 1300-MW(e) LMR would be $492 million. The validity of this cost figure is sup- ported by the recently published costs for the newest LWR reprocessing plant in the Federal Republic of Germany.® 1In that document, $1.6 bil- lion is the cost reported for a 500-t/a LWR reprocessing plant (exclud- ing costs of the refabrication facility). When scaled downward to the 35-t/a plant, this yields $423 million. Since the costs of reprocessing LMR fuels are generally higher (up to 50%) than those for IWR fuels, the E~-13 two derived costs are comparable, and confidence in their wvalidity is increased. However, preliminary results of an ORNL study currently in progress for a small onsite facility suggest a cost of $270 million com- pared to the above derived $492 million for reprocessing of oxide fuels in a 35 t/a facility.7 We have used these two cost figures to calculate lower and upper boundary costs for reprocessing and refabrication. The oxide fuel reprocessing costs were derived from these values, while the costs for refabricating the oxide fuel and blanket were based on the high end of the values given in Table 2.12 of Ref. 5. The high-end values were used to reflect the cost disadvantage for a small-throughput plant. The corresponding cost estimates for a metal fuel reprocessing and refabricating plant have not been prepared in the same detail as those for the oxide fuel plants. Because of this, the values derived for the metal fuel cycle do not have the same validity as those for the oxide fuel cycle. However, we think the results are generally correct and acceptable as a basis for comparison. A more detailed design study would be required to provide cost estimates of equal validity to those for the oxide fuel plants.* Since the number of major steps required for processing of metal fuels is comparable to that required for oxide fuels (see Figs. E.l1 and E.2), the cost of the oxide reprocessing plant was used as a cost basis for the metal reprocessing plant. However, metal fuel reprocessing equipment may not require the head height that is needed for oxide reprocessing equipment. Therefore, the initial base cost for the metal fuel was reduced by 20% before scaling for capacities was done. The 207% reduction is based on the resulting reduced need for concrete and rein- forcement, relative to the total reinforced concrete requirements for the total cell. This reduced base cost was then adjusted for size based on capacity, using a 0.5 scaling factor. The facility requirements for reprocessing or refabricating metal fuels are similar. Therefore, the costs for refabrication of the metal fuels were based on the relative number of steps required for refabrica- tion (i.e., 20) as compared with the number of steps required for reprocessing of the metal fuel (i.e., 25) to the 0.5 power. Thus, the refabrication capital costs are: (20/25)%9+5 = 0.894 times the metal reprocessing capital costs. The total waste costs were based on the rate of 1 mill/kWh and hence were $9.1 million annually. The cost of new blanket oxide fuel was based on the high range in Table 2.12 of Ref. 5. The high range, $500/kg, was used because of the relatively low quantity to be pur- chased. Hardware costs, estimated at $50,000 per fuel assembly, were *Costs were not available from ANL on the IFR fuel cycle facilities during the period in which this report was being prepared. However, since that time ANL has developed a concept and cost estimates. ANL should be contacted for needed details of this information. E-14 based on experience at the FFTF and the EBR II. The operating costs were scaled using a sizing exponent of 0.7. The cost summary is given in Table E.2. This information suggests that there is no significant economic advantage in a metal-fuel or an oxide—-fuel FBR based on the costs of reprocessing and refabricating the fuel. Calculations were made to provide comparable values for processing oxide fuel in a large fuel cycle facility (1500 t/a), assuming that it existed and that the fuel was both reprocessed and fabricated there. The cost basis was taken from Table 2.13 of Ref. 5. The results showed costs of 4.7 to 6.3 mills/kWh for 80% reactor capacity factors. It is obvious that the costs derived here for a 35-t/a facility are high when converted to the unit cost ($/kg) basis or to mills/kWh, and when compared to costs in a large facility. This is a function of the low capacity requirements for either of the processes. The 25-35 t/a plants are a factor of 50 smaller, and the cost values reflect this; however, the cost comparisons should be more valid than the absolute values, since both the metal and the oxide processes were costed on approximately the same basis. Options other than integral reprocessing have been suggested, as follows:8 l. Store the spent fuel until sufficient quantities are accumulated for large-scale commercial reprocessing. 2. Ship the spent fuel to a large international reprocessing facility set up as a cooperative venture. 3. Reprocess fuel in a small dedicated experimental facility such as BRET, subsidized by research and development funds (for early power plants). 4. Reprocess IMFBR fuel in a joint facility with LWR fuel (the "hybrid” concept). 5. Reprocess IMFBR fuel in existing U.S. reprocessing facilities, utilizing part of their capacity for c¢ivilian purposes, after appropriate modifications are made. E.4 ISSUES THAT NEED ATTENTION The colocation of fuel reprocessing facilities with fast reactors can require that all regulatory concerns for both the reactor and its associated reprocessing facility be addressed before approval of the re- actor is obtained. We have defined several issues in each of six areas of the reactor-reprocessing-refabrication combination and have added a comment on each issue. E-15 Table E.2., Cost summary for reprocessing and refabricating metal and oxide FBR fuel® Fuel type Reprocessing costs ($109) Capital Annual Annua} Total annual capital operating Metal fuel (20-year) 184—336 42—76 13 55—89 25.5 t/a (30~year) 184—336 3156 13 44—69 Oxide fuel (20-~year) 270—492 61—111 16 71127 35 t/a (30~year) 270492 45— 82 16 61— 98 Refabrication annual costs ($106) Capital and . Hardware Purchases Total operating Metal fuel (20-~year) 48--80 10 — 58—90 25.5 t/a (30~-year) 39—61 10 —_— 49—71 Oxide fuel (20~ or 25.6 - 13.5 39 8 t/a 30~year) Combined annual costs ($106) Repro. Refab. Waste Total Metal fuel (20-year) 55—89 58—90 9.1 122—188 (30-year) 44—69 49—71 9.1 102—149 Oxide fuel (20-year) 77127 39 9.1 126176 (30-year) 61— 98 39 9.1 110—147 Busbar costs at 80% reactor capacity (mills/kWh) Metal fuel (20-year) 13.4—20.6 (25.5 t/a) (30~-year) 11.6—16.3 Oxide fuel (20~year) 13.8—19.3 (35 t/a) (30-year) 12, 116.1 Oxide fuel (20-Year) 6.3 1500 t/a (30-year) 4o7 dA constant waste cost of 1 mill/kWh, or $9.1 million per year, is assumed. l. E-16 Waste — One of the major questions that needs to be resolved for the metal fuel processing concept is the means of handling the waste. Before this question can be answered, considerable experi- mental work is required to define where each of the major waste isotopes will reside as a result of these processes. Once the locations and the quantities of the wastes are defined, specific disposal methods can be determined. In the area of waste disposal, there are many questions relative to the specific criteria for the wastes and waste containers for either metal or oxide fuels. Such questions would require clar- ification before final design could be completed. For the metal fuel, considerable development work would be necessary to verify proposed processes, particularly in the area of imperviousness of the disposal product. Environment — The environmental requirements for the total fuel cycle need to be clarified before the experimental work is com-— pleted, particularly for metal fuel, since little experience and no source terms are available. This would require an early dedication to, and funding for, such a project. Again, clarification is needed as to what specific criteria or regulations would apply to the releases from an FBR fuel cycle. The current regulations (40CFR190) apply specifically to the "uranium fuel cycle” for 1light-water reactors. It is not clear whether 40CFR61 was intended to cover fuel reprocessing from non- uranium fuel cycles (Par. IV B). This needs clarification. Safety and Licensing — At present, there are no approved design criteria for IMRs or for the fuel cycle facilities. Can these be approved in time to be included in the design process? At present there are also no regulations (NRC) which define the criteria for fuel cycle facilities. Appendix P to 1OCFR50, which has been proposed (but not approved), contains safety criteria based on aqueous processing. It could be modified to cover pyro- processing; however, this would have to be done well in advance of the final design of a demonstration facility, in order for the re- quirements to be included in the final design. It is not clear that the safeguards requirements for the metal fuel reprocessing and fabrication would be the same as those for similar operations on the oxide fuel. However, the metal fuel would always be highly radioactive and thus less attractive for diversion. For either the oxide or metal fuel cycle, transportation is reduced by using the integral fuel cycle. Would any safety or licensing waivers be required for either FMEF or FCF in order for them to meet DOE requirements, and could these waivers be obtained? E-17 Public Acceptance — Public acceptance has been a major problem for less complex projects than these that are proposed. Could this be- come a major problem? Although the public may accept a reactor or several reactors at a site, the addition of a fuel cycle facility to this "nuclear park,” may require a great deal more education than before. This could be particularly important for the management and disposal of wastes. Industry Acceptance — A major consideration will be the willing- ness of industry to become involved in a complex, total-power pro- gram such as that proposed for the IFR. How would industry propose to handle this? A. In the course of this study, we have observed that utilities would not feel comfortable with staffing and managing the diverse facilities required for a total fuel cycle. Their back- ground and training are related to power generation and do not include the complexities, costs, and risks of these other elements. It may be more practical to have a separate operating organization with the appropriate technical expertise. B. Are the economics of omne central fuel cycle compatible with a limited- (one-) reactor complex during early years when only one or two reactors are in operation? A fuel cycle facility must be originally designed and built to handle the total output from all the reactors at the site. Therefore, the major costs — those of construction — would be incurred at some time before the total capacity of the facility was needed. The introduction of the fuel cycle might be delayed somewhat by storing the fuel after discharge. This would pro- vide an inventory for the future reactors. An economic balance would have to be made in order to define the optimum storage pe- riod. The normal plant life for reprocessing is assumed to be 20 to 30 years, as opposed to an assumed reactor life of 30 to 40 years, however, the optimum lifetimes for both reactors and reprocessing plants require further study and may prove to be much longer. C. How large an analytical complex will be required? Reactors, per se, do not require an extensive analytical capa- bility. The overall complex would require extensive remotely operated facilities for determining chemical, metallurgical, physical properties, and for analytical chemistry measurements. D. Does the operator have sufficient backup technical capabilities available (analytical laboratories, metallurgical caves, etc.) to help resolve day-to-day problems? E-18 The support facilities required to service a reactor fuel cycle complex would be much more extensive than those required for a reactor alone. E. Can reactor(s) be dependent on only one fuel source and one dis- posal method? How do reactors operate in case of a 1- to 2-year shutdown of fuel cycle? All of the facilities involved in the reactor cycle — fuel source, fuel storage and disposal, and waste disposal — are remotely operated and maintained. They would be difficult to duplicate on short schedule in case a major problem developed in any part of the system. Thus, the metal fuel reactors would not have a fall-back option, unless and until several similar facil- ities were available. The latter circumstance would require shipment of spent and refabricated fuel. F. What is the design life of the fuel cycle complex? Can it be refurbished — and how often? The design life of fuel »80% availability appears to be achievable (see below), it cannot be assured at this time. The North American Electric Reliability Council (NERC) Equipment Availability Report for the l0-year period 1974—1983 supports the claim of smaller turbine generators contributing to higher overall plant availability. As shown in Table 3.4 of Volume III, fossil plant turbine—-generator sets in 400 MW(e) and below sizes have distinctly lower forced and scheduled outage rates and higher availabilities than turbine—generator sets in the larger sizes. The advantage is even more significant for nuclear plant turbine-generator sets below 800-MW(e) size, compared with those above 800 MW(e). It also is observed that nuclear plant turbine—-generator sets have a distinct performance advantage over fossil sets in all size ranges., It is speculated that this may be due to the lower steam temperatures and pressures and rotational speeds of nuclear turbine-generator sets, which result in a less severe operating enviroament. G-6 As shown in Table 3.5 of Volume III the area having the greatest potential for LWR plant availability improvement is with the reactor and associated systems. It is obvious that plant availability improve- ment R&D must concentrate on the reactor and its related systems. For example, a 50% reduction in reactor scheduled outage factor plus a 50% reduction in reactor forced outage rate would result in 80% overall plant equivalent availability. The above LWR data provides strong support that the modular HTR goal can be achieved. e. Claime. A four-module plant can be constructed in 36 months from start of site work to commercial operation of the second turbine.? Evaluation. This is an optimistic schedule when judged by U.S. experience, but appears possible. The modular HTR schedule is based on an evaluation of the conceptual design by Bechtel. It should be re- examined after the design has been carried to point of estimating quan- tities of construction materials (structural steel, concrete, piping, and wiring,) and labor manhours. This will provide a firmer basis for estimating elapsed time for placing equipment and materials. The construction schedule should be reevaluated as the design progresses. f« Claim. The plant can be operated and maintained by a staff of 306 (Ref. 2). Evaluation. This estimate is based on a preliminary analysis of staffing requirements. In comparison, typical staffing for current large IWR plants is ~400. The modular HTR staffing estimate is based on the following assumptions: o Regulatory procedures have been stabilized (i.e. no back- fitting by maintenance forces). ° Plant control is highly automated, permitting operatiom of four reactors by one control operator station. ° Plant is designed for maintenance with one module offline re-~ sulting in minimum requirements for peak maintenance forces. ° Plant security is highly automated, requiring minimum security forces. These assumptions, along with the smaller turbine-generator size requir- ing less maintenance personnel, are the principal reasons for the re- duced staffing. Further studies of operation and maintenance staffing should be performed, using a task—analysis approach, especially in the area of control operation of multiple reactors by a single control oper- ation and its relation to safety. e Claim. Adding capacity in small increments 1is/will be a significant financial goal of utilities and results in less financial risk. Evaluation. In a time period of low load growth, high financing and construction costs, and reluctance of public utility commissions to grant rate adjustments, this claim appears intuitively obvious. Two recent draft studies by Los Alamos National Laboratory (LANL) and Applied Decision Analysisg’9 support this claim. Both studies attempted to quantify the additional capital investment cost that utilities could afford to pay for smaller, shorter lead time plants in comparison with larger, longer lead time plants, while continuing to meet their financial goals. LANL found that a reduction from long to medium lead times permits the utility company to pay 40—50% more in overnight construction costs, and a reduction from 1long to short lead times permits a four-fold increase in the overnight construction cost. From a ratepayer viewpoint, Boyd et al.? found that utilities could pay approx- imately $200/kW(e) capital investment cost premium for smaller unit sizes and shorter lead times for utility system sizes 3000 MW(e) and larger, From the shareholder viewpoint, the affordable capital invest- ment cost premium was found to be two to three times higher than from the ratepayer viewpoint. The findings are general in that they apply to any type of power plant, they are also sensitive to a number of parameters f{(e.g., system size, existing generation mix, load growth rate, and financing). However, sensitivity analyses in both studies support the claim. We recommend, however, that the studies in this area be continued and refined to develop a complete understanding of the economics of small nuclear plants. G-8 REFERENCES FOR APPENDIX G 1, 2. 3. 7. 8. K. J. Kruger and G. P. Ivens, “"Safety-related Experiences with the AVR-Reactor,” presented at the IAEA Specialists' Meeting on Safety and Accident Analysis for Gas-Cooled Reactors, 0Oak Ridge, Tennes- see, May 13-15, 1985. Preliminary Concept Description Report, 4 x 250 MW(t) HTGR Plant Side-by-Side Steel Vessel Prismatic Core Concept, HIGR-85-142, issued by Bechtel Group, Inc., et al., for Gas-~Cooled Reactor Associates, San Diego, California, October 1985, APPLIED TECHNOLOGY. HTGR Program Concept Evaluation Plan for Small HTGRs, GCRA 84-009, Gas-Cooled Reactor Associates, San Diego, California, October 31, 1984, Modular HTGR Balance of Plant Design and Cost Status Report, Bechtel Group, Inc., San Francisco, California, September 1982. Modular HTGR System Design and Cost Summary, GCFR-00693, Bechtel Group, Inc., San Francisco, California, September 1983, Constructibility Assessment for Modular High-Temperature Gas-Cooled Reactors, Bechtel Group, Inc., San Francisco, California, July 1982. Phase IV Update of the EEDB, DOE/NE-0051/1, United Engineers and Constructors, Inc., Philadelphia, Pennsylvania, September 1984, Andrew Ford, The Market for New Electric Generating Capacity: A Financial Feasibility Case Study, Los Alamos National Laboratory, Los Alamos, New Mexico, June 1984, D. W. Boyd et al., The Potential Impact of Modularity vs. Utility Generation Investment Decisions, Decision Focus, Inc. and Applied Decision Analysis, Inc., for the Electric Power Research Institute, Palo Alto, California, March 1984. woo~sooon o . o e N O 14, 15-16. 17. 18. 19. 20-24, 25, 26, 101. 102-177. 178-283. INTERNAL DISTRIBUTION Aebischer Anderson Ball Bell Booth Bowers Braid Buchanan Burch Cantor Cleveland Cole Craddick Davis Delene - - - Ll - . . - - . - . - * - - o - o EoOodorgogdEHLAE o OUO Flanagan Forsberg Q=M o] rt M. Haas C. Hampson - UdacoangrnureAGR” S gxom>™ewm 3 Dodds, Jr. 27, 28, 29. 30. 31-47, 48, 49, 50. 51. 52, 53. 54-73, 74. 75. 76=-93, 9. 95. 96. 97. 98-99. 100. ORNL/TM-9780/V2 E. Jones Jr. R. Kasten C., Maienschein P. Malinauskas L R C - . Moses . Mynatt . Oakes Rayner Selby . Trammell . Trauger Uhrig . Wehe . White . Wilbanks Wymer Central Research Library Document Reference Section Laboratory Records Department Laboratory Records (RC) LU RHEEE WHLOO WO @HUOHwnE O mdg = 0 g o EXTERNAL DISTRIBUTION Office of Assistant Manager for Energy Research and Development, ORO, DOE, 0Oak Ridge, TN 37831. Distribution Category UC-79T, Liquid Metal Fast Breeder Reactors: Applied Technology. Nuclear Power Options Viability Study Distribution. wU.S. GOVERNMENT PRINTING OFFICE 1986—631-056/40047