ML20126F672

Forwards Two Papers Re Policy Issues & Schedules Concerning Preapplication Reviews of Advanced Reactor & Candu 3 Designs

ML20126F672
Person / Time
Issue date:	12/16/1992
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Erin Kennedy ASEA BROWN BOVERI, INC.
References
PROJECT-680A NUDOCS 9212310051
Download: ML20126F672 (60)

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g WASHINGTON, D. C. 20556 k...* p December 16, 1992 Project No. 680 Mr. Ernie H. Kennedy, Manager Nuclear Systems Licensing ABB Combustion Engineering Nuclear Power 1000 Prospect Hill Road P.O. Box 500-Windsor, Connecticut 06095-0500

SUBJECT:

COMMISSION PAPERS ON POLICY ISSUES AND SCHEDULES CONCERNING THE PREAPPLICATION REVIEWS OF ADVANCED REACTOR AND CANDU 3 DESIGNS Enclosed are two papers which should be valuable _to your continuing regulatory efforts. In mid-1992 the-staff discussed with you its' intent to -identify key policy issues and the projected schedules to complete preapplication reviews of these designs. The_ staff noted that these policy issues and the projected-schedules would be addressed in separate _ papers to the Commission. As a result of the staff's reviews, an assessment of projected resources, and the meetings held with'you and the other preapplicants for the: advanced reactor and CANDU 3 designs, the enclosed papers have been provided to the Commission: (I) a draft paper providing the-staff's positions on 10 policy issues, and (2) a final paper, SECY-92-393, " Updated' Plans and Schedules for the Preapplication Reviews of the Advanced Reactor (MHTGR, PRISM, and PIUS) and CANDU 3 Designs," on the staff's proposed schedules for the preapplication reviews. The paper.on the policy issues is a draft because the staff has not yet-obtained Commission approval on these issues. The-staff will be meeting with the Advisory-Committee on Reactor-Safeguards (ACRS) to discuss these' issues in the near future. The staff will include the views of the ACRS and-document its final recommendations in a revised paper before seeking the Commission's approval. Any-comments you may wish to offer will be considered as we prepare our final positions. _ Please submit any comments by January 25, 1993. 1 The proposed schedule paper reflects the staff's assessment of its resources and the needs of the preapplicants. The staffxwill continue to try to expedite its reviews and ~ complete the work' ahead of, schedule. 4 i i -300o( C E 0 EG E 0M N L E 3 gh 4 ~ '9212310051 921216 k PDR PROJ p g 680A PDR

_--.- -.-~..- 4 l i Mr. Ernie H. Kennedy December 16, 1992-i The proposed positions on policy issues have not been reviewed by'the i Commission, and, therefore, do not represent agency. positions. Your comments concerning these issues should be sent to the project manager, Dino Scaletti. Sincerely, i Original signed by: Dennis M. Crutchfield, Associate Director j' for Advanced Reactors -_and -License Renewal Office of Nuclear Reactor Regulation-c }

Enclosures:

1. Draft Commission Paper 2. SECY-92-393 l cc w/ enclosures: See next page i Distribution: Central File NRC PDR PDAR R/F DMCrutchfield l WDTravers MMSlosson THCox l EDThrom DCScaletti-1 OGC. 15/8/18 ACRS (10) P 315 i-OPA L LLuther PIUS R/F. 4 s LA:PDLR:ADAR PM: R-SC:PDAR: ADAR SE:PDAR:ADAR-D: AR-DThrom % MM DMCrutchfield- -LLuthe. D5eW16tti:sa THCox M<- ~ p,2 //y/92 $ $92-- A//6/92

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I Mr. Ernie H. Kennedy December 16, 1992 PIUS Project No. 6B0 cc: Regis Matzie, Director 9354-0401 Ulf Lindelow, Section Chief Advanced Water Reactor Projects PWR Core Design, Fuel Division ABB Combustion Engineering ACB Atom, ABB Nuclear Power S-721 63 Vasteras 1000 Prospect Hill Road Sweden Windsor, Connecticut 06095-0500 John F. Carew Mark Crump, Manager 9341-0407 Building 475 B Reactor Engineering Brookhaven National Laboratory ABB Combustion Engineering Upton, New York 11973-5000 Nuclear Power 1000 Prespect Hill Road Gregory J. Van Tuyle Windsor, Connecticut 06095-0500 Building 475 B Brookhaven National Laboratory Chuck Molnar 9377-0407 Upton, New York 11973-5000 Nuclear Systems Licensing ABB Combustion Engineering Johann Lindner, Vice President Nuclear Power Technology and Engineering 1000 Prospect Hill Road ABB Atom, Inc. Windsor, Connecticut 06095-0500 200 Great Pond Drive Windsor, Connecticut 06095 Lars Nilsson, Technical Manager PIUS Reactor Division ABB Atom AB S-721 63 Vasteras Sweden Harold D. Thornburg ABB Atom, Inc. 901 South Warfield Drive Mt. Airy, Maryland 21771 Daniel F. Giessing, Director Office of LWR Safety and Technology Mail Stop NE-42 U.S. Department of Energy Washington, DC 205B5 Steve Goldberg, Budget Examiner Office of henagement and Budget 725 17th Street, NW 1 Washington, DC 20503

F l i Endlosure1-- l i _ DRAFT-j j; 4 l [ 1j' 19.t: The Commissioners j Ltg.m: James M. Taylor Executive Director for Operations l Subiect: ISSUE $ PERTAINING TO THE ADVANCED REACTOR (PRISM, MHTGR, AND [ PIUS) AND CANDU 3 DESIGNS AND THEIR-RELATIONSHIP T0 CURRENT i REGULATORY REQUIREMENTS-i ~ Purcose: To request Commission guidance for those areas.where the l staff is proposing to depart from current regulatory requirements in the preapplication. review of the advanced-l reactor and CANDU 3 designs. J [ Backaround: The Advanced Reactor Policy S+4tement (51.FR 24643)'and' i NUREG-1226, " Development and cilization of the NRC Policy Statement on the Regulation / Advanced Nuclear Power. 4- - Plants," define advarced re. tors as those with innovative ~ designs for which licensing requirements will be signif-- icantly different-from the existing light-water reactor-u (LWR) requirements. These documents-also provide guidance for_ the development of-new regulatory requirements;to support the advanced designs.- Staff reviews of_these-l advanced reactor designs.should utilize existing regulations l - to the maximum extent practicable.- When new requirements are necessary, the staff should move towards performance-j' standard regulations and away: from prescriptive-. regulations.- r Each designer is encouraged to propose new criteria.and i novel approaches >for evaluation of their designs,'.and an objective of:early-designer-staff--interaction should be to develop guidance on licensing. criteria for the advanced . reactor design and to make a preliminary assessment of the potential.of that-design to meet those criteria.- k CONTACTS:- M.M. Slosson a '504-1111 i T.H. Cox- '504-1109 ,,s n.,, --e -,-",,s,,, ,--,,.,,,,-,,.i..,--wa ,,,,,,,.,,,,w-,,,, n wrr,....v,n-n,--~ m,, a + c\\

5 i i The Comissioners 2-l The staff is conducting preapplication reviews of the following four designs: l General Atomics (GA) 350-MWt Modular High Temperature l Gas-Cooled Reactor (MHTGR) design sponsored by the U.S. Department of Energy (DOE) Gas Cooled Reactor Program General Electric-(GE) 471-MWt Power Reactor Innovative Small Module (PRISM) reactor design sponsored by the DOE j Advanced Liqutd Metal Reactor (ALMR) Program Atomic Energy of Canada, Limited, Technologies (AECLT) i 1378-MWt Canadian Deuterium Natural-Uranium (CANDU 3) reactor design 4 Asea Brown Bovert-Combustion Engineering (ABB-CE) 2000-MWt Process Inherent Ultimate Safety (PIUS) reactor-design , provides a listing of pertinent Comission papers and reference NUREG documents for these preap-i plication designs. Some.information in the original documents may be superseded by more recent preapplicant submittals. A sumary of the current designs is provided-as i. In response to Commission staff requirements memorandum (SRMs), in SECY-91-202,1" Departures From Current Regulatory Requirements in Conducting Advanced Reactor Reviews," the staff comitted to identify issues during the preapplication review that require Comission policy guidance or staff technical resolution for design certification, including n situations in which advanced reactor designs significantly l deviate from current regulatory requirements. Policy issues for evolutionary and passive LWRs have been 4 identified in the following Comission papers: SECY-90-016, " Evolutionary Light Water Reactor-(LWR) 4 Certification Issues and Their Relationship to Current - ] _ Regulatory Requirements"

• . Draft SECY (distributed for coments on February 27,'

1992), " Issues Pertaining to Evolutionary and Passive -Light Water Reactors and Their Relationship to Current-Regulatory Requirements" f ,-m ,_,c.- y ,s.m.-c.. -ce o.,. g m,-,g...,.

) i The Comissioners, Draft SECY (distributed for coments on June 25,1992), " Design Certification and Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light'- Water Reactor Designs" Discussion: As part of their submittals, the preapplic nts identified how their design complied with the current LWR licensing requirements and, where it did not, provided alternative i criteria for evaluating their designs. The staff has conducted a preliminary review of the four preapplication designs using existing LWR regulations and the evolutionary light-water reactor (ELWR) and advanced light-water reactor (ALWR) policy guidance. This initial review identified 2 10 issues that require policy direction from the Comission for proposed deviations from existing regulations. These are instances where either existing regulations do not apply to the design or preapplicants' proposed criteria are sig-nificantly different from the current regulations. These issues, background information on current requirements, pre-applicants' proposed approaches, and staff recommendations for Commission approval, are provided in Enclosure 1. The recomendations for Comission approval were developed by the staff with inputs from the preapplicants, the public, and the ACRS. The staff considered the preapplicants' proposals in light of the Comission's policy statements and guidance on severe accidents, advanced reactors, and safety goals to develop a single consistent policy recommendation to be applied to all applicable advanced reactor designs. In some instances, the staff recommends that current ' regulations continue to be applied to the advanced reactor designs despite preappitcant proposals to do otherwise. Where deviations are recommended, the staff proposes more conservative alternatives to the preapplicants' proposals to account for uncertainties associated with the conceptual design, which should ensure that conclusions made during the preapplication review will provide a-reasonable basis for the detailed design being found acceptable at design certi-fication. It is intended that the safety level standards - for these designs will be consistent with Comission i guidance at design certification. Some issues are closely related. Accident evaluation and source term provide a basis for containment performance and - l emergency planning. Approaches taken for residual heat-removal and reactivity control are intended to be~ consistent with-the accident evaluation categories and consequences. = v-w - - = -e wv w

w V e- .The Comissioners - The staff proposes to_ treat the MHTGR, PRISM and PIUS-designsasadvancedreactorsinaccordancewIththepolicy l statement. The CANDU 3 design is considered to be an-i evolutionary heavy-water design deriving from the larger CANDU reactor designs operating in Canada and elsewhere. Therefore, the staff has concluded that a prototype CANDU 3 - is not required for design certification.. This position is consistent with staff conclusions-in SECY-89-350, ' Canadian CANDU 3. Design Certification,' and SECY-90-133,. _' Prototype j Requirement for_CANDU 3 Design." The.preapplicant, AECLT, has stated that a CANDU 3 reference plant is a. key element-i in their plan for standard design certification. If-AECLT i holds to that position, the regulatory review and con-struction in Canada would lead.the NRC's design certi fication review. -The staff believes that this regulatory review and construction in Canada would greatly benefit.our 4 j -review of CANDU 3. During the preapplication review, the staff _ intends to utilize.the' foreign operating experience and accident-analysis to aid in predicting the expected [ behavior of the CANDU 3 design. - AECLT makes no claim of i-passive shutdown or decay heat removal; capabilities, i However, because of its unique heavy-water, pressure-tube reactor design and evolution under a different regulatory 1-structure, it does not conform to some current NRC l regulations. The staff proposes to apply-preapplication 1 review criteria to the CANDU 3 reactor that are consistent i with ELWR review requirements. l The staff intends to use the Commission's guidance on these recomendations to conduct preapplication: reviews of the i conceptual designs. Guidance for review of prototype L requirements for advanced reactors will. follow SECY-91-074, " Prototype Decisions.for Advanced Reactor Designs." 1 i Consistent with the requirements of Title 10.of the Code of H Federal Regulations -(CFR) Section 52.47(b)(2), novel safety - 4 features of the-advanced reactors and CANDU 3 will.be [ required to be ciemonstrated-through analysis,- test programs,- i experience, or a combination of-these methods.. Feedback: from the review process will be factored into recomended revisions to the policy guidance and recomendations for the-~ development of licensing criteria and regulations will be made after the preapplication safety evaluation reports (PSER) are issued. Additional issues may be developed during the preap>11 cation review process;-they will be identified in su) sequent Comission papers. 4 L i' i .m

l c. e The Comissioners g i l In an SRM dated May 8,1992, the Comission requested the i staff to prioritize the issues for Comission review. The 2 staff recomends that the priority.for review-be consistent

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with the PSER issuance schedules and' requests that direction be provided in sufficient time to allow the: staff to incor-l porate Comission decisions into the final PSERs. Since the PRISM design is scheduled as the first preapplication review, Comission attention is requested on a highest i priority-for those items identified.in the enclosure as applicable to the PRISM design. [onclusions: The staff requests approval 'of, or alternate guidance on, i these proposed positions to be taken in the preapplication review of the advanced reactor and CANDU 3 designs. coordination: The Office of the General Counsel-has reviewed this paper and has no legal objection. The staff has forwarded a draft of this paper to the ACRS for its-review and comments. Recommendations: That the Comission l Approve the staff recomendations in Enclosure 1 for conduct of the preapplication reviews. - Approve of the staff's conclusion that, based on the >osition that the CANDU 3 design is an evolutionary teavy-water design deriving from CANDU designs operating in Canada and elsewhere, a prototy required for design certification.pe CANDU 3 is not Note that positions which change as preapplication p i review experience is obtained will-be comunicated-to the Comission and-that as the staff identifies new- [ issues it will. inform the Comission. i Note that the Comission-is requested'to provide highest priority _ attention-to those issues identified in the - enclosure as being applicable to the PRISM design. Note that due to the preliminary nature of the design information on the advanced reactor-and CANDU 3 designs, _ and the preliminary nature of the_ staff's preapplication r l w c- .,.-.---%-,-y .... -. ~ - m.-.. =-.-"--s.. , *.. ~.,, -. ~ =,,

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nJ--..a -c y 4 m wJ M-m ..E a m+s a 4 J .u a-. A i' The Commissioners 6-reviews, the staff does not recommend proceeding with generic rulemaking on any of the policy issues identified in this paper. The staff will consider generic rulemaking, as appropriate, as the reviews progress and the staff gains greater confidence in the final design information. i e James M. Taylor 3 i Executive Director l for Operations i

Enclosures:

l. Policy Issue Analysis 2. Design Summaries 3. List - Reference Documents i 4 i I l

1 POLICY ISSUES ANALYSIS AND RECOMMENDATIONS As part of a preliminary review of the PRISM, CANDU 3, MHTGR,'and PIUS -designs, the staff has identified 10 instances where either the staff or the-preapplicants have proposed to deviate from current light-water reactor (LWR) guidance for the review of the designs. This occurred when existing regulations were not applicable to the technology or when the staff identified new departures from existing regulations that are considered warranted based on the preapplicants' design and proposed criteria. -The staff has grouped the issues into two categories: (1) those issues for which the staff agrees that departures from current regulations should be considered; and (2) those issues which the staff does not believe a departure from current regulations is warranted at this time. The following is a matrix of the issues identifying the plant applicability: CATEGORY ISSUES PRISM MHTGR CANDU PIUS A. Accident Evaluation X X X X B. Source Term X X X C. Containment Performance X X X X D. Emorgency Planning X X X "$9 E. Reactivity Control X F. Operator Staffing X X X X t G. Residual Heat Removal X X X H. Positive Vold Reactivity X X Category I. Control-Room Design X X X X 2 J. Safety Classification X i Discussions of these issues are on the following pages, including a brief summary of the issue, current LWR regulations, preapplicant positions, discussion of staff-considerations and a proposed recommendation for staff { action. The staff considered the preapplicant's proposals in light of applicable Commission policy statements. l V DRAFT r l i [

4 At this preliminary review stage, the staff has limited the scope of the issues to those which could affect the licensability of the proposed design. Additionally, if a similar issue had already been raised for the LWR designs and the staff's advanced reactor design recommendation was essentially the same, it was not repeated'in this paper. In those cases where the preapplicants proposed different considerations from the evolutionary or passive LWRs, the issue is treated in this paper in light of the work done in the advanced light-water reactor policy papers. J a 1 3 l l l I d e DRAFT ,.,......,.,..a,,

i [' A. ACCIDENT EVALUATION lillut Identify appropriate event categories, associated frequency ranges, and evaluation criteria for' events that will be used to assess.the safety of the l proposed designs. j Current Reaulations i General Design Criterion (GDC) 4 requires the consideration of accidents in the design basis. ' Also,10 CFR 52.47 requires the consideration of conse-quences for both severe accidents (throu h the required probabilistic risk assessment) and design basis accidents ( BA)Lfor designs which differ signif-icantly from evolutionary designs or uti ize passive or other innovative means to accomplish safety functions. l Preapolicants' Acoroach I All three advanced reactor preapplicants proposed to analyze accidents-signif-icantly less probable than the present design basis range and to assure i through their design that these accidents had.. acceptable consequences limited to specific dose levels to the public. All chose to utilize the Environmental-4 Protection Agency's-(EPA) lower level Protective Action Guidelines (PAG) of-l 1 rem whole body and 5= rem thyroid as.their limits for a significant portion 1evel PAG dose limit for all' sequences more probable than 5x10'pe the lower-of their accident spectrum. The MHTGRl accident guidelines invo per reactor-l- probable than-10,uidelines invoke the PAGs for accident year. The PIUS p per reactor-year. The per reactor-year. The PRISM-probable than 10',on guidelines also limit consequences from any sequence more accident evaluati per reactor-year to the 10.CFR Part 100 dose limits. Guidelines for onsite consequences and offsite consequences from operational 3 transients for all vendors are consistent with or more conservative than f present LWR regulations as contained in -10 CFR Part 100. The CANDU 3 preapplicant, in their current safety analyses, has excluded' per analyses of the consequences of events with frequencies of less than 10' l-year from the safety evaluation. Events which would be excluded from consideration, based on the CANDU 3 design characteristics and system i re11 abilities, would' include anticipated transient without scram (ATWS), i unscrammed loss-of-coolant accidents (LOCAs),Jdelayed scram events, and other i= - events which could affect reactivity insertion (for example, from control-e system failures). As.a result of. the positive void reactivity coefficient-l associated with the CANDU design, events involving even.a relatively short scram delay could result-in a core disruption accident. f t DRAFT: l o

. ___ _ _ _. ~ _ _ _ _ 1 p.. 1-Discussion L The structure proposed by the PRISM, MHTGR, and PIUS preapplicants for-selecting accidents to be evaluated was developed to support their positions for reduction of emergency planning requirements as described in Section D of this enclosure. As d'scussed in Section D, the staff is not ready to make a i recommendation on whether the Commission-should consider a reduction in the l emergency planning requirements.;.The CANDU 3~ approach which limits the scope of severe accidents-examined appears to be inconsistent with the provisions of. 1 10 CFR 52.47. The accident evaluation scheme envisioned by the staff examines. challenging events to the designs to provide information for a later decision on emergency planning requirements for advanced reactors and includes consideration of the potential consequences of severe accidents. Addi-l tionally, for the multi-module designs (PRISM and MHTGR), the impact of specific events on other reactor modules-for the multi-module sites must~ be i assessed. d l The staff's approach'is intended to-be structured conservatively so that i positive conclusions made on the licensability of the conceptual designs during the preapplication review will provide a reasonable basis for acceptability of the' design at design certification. Several sources of uncertainty exist with the conceptual designs including limited performance l _. and reliability data for passive safety features,;1ack of final design j' information, unverified analytical tools. used to predict plant response, l limited supporting technology and research,' limited construction and operating. experience, and incomplete quality control information on new fuel manufacturing processes. Later, during the design certification process, some i of the conservatism could be removed based on improved understanding of the design and analytical tools through completed-research. Recommendation The staff proposes to develop a single approach for accident evaluation to be applied to all advanced reactor designs during the preapplication review. The approach will have the following characteristics: Events will be selected deterministically and supplemented with insights ( from probabilistic risk assessments of the specific designs. Categories of events will be established based on expected frequency of occurrence. Tu selected range of events-will encompass -events of a lower - likelihood than +,raditional LWR design. basis accidents. Consequence acceptance limits for core damage and onsite/offsite releases will be established for each category to be consistent with Commission-policy guidance with appropriate conservatisms: factored in to account-for j uncertainties. L . Methodologies and~ evaluation assumptions will be developed for analyzing L each category of-events consistent with existing LWR practices.

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-. _ ~ _ _. _ i Source term determination will be performed as approved by the Commission l in Section B of this enclosure, A set of events will be selecte'd deterministically to assess the safety margins of the proposed designs, determine scenarios to mechanistically determine a source term and to identify a containment challenge scenario. External events will be chosen deterministically on a basis consistent with that used for LWRs. Evaluations of multi-module reactor designs will consider whether specific events apply to some or all reactors onsite for the given scenario of operations permitted by proposed operating practices. 1 i E 9 3 4 9 i i. i 3 e 5-DRAFT r-,, ,-r-----,r- +-,r w - e s- -A n--s.-en-u ar*~ww ra,

B. SOURCE TERM 1111t Should mechanistic source terms be developed in order to evaluate the advanced reactor and CANDU 3 designs? A mechanistic source term is the result of an analysis of fission product release based on the amount of cladding damage, fuel damage, and core damage resulting from the specific accident sequences being svaluated. It is developed using best et.timate phenomenological models of the transport of the fission products from the fuel through the reactor coolant system, through all holdup volumes and barriers, taking into account mitigation features, and finally, into the environs. {Etent Reaulations A)pendix ! to 10 CFR Part 50 (ALARA),10 CFR Part 100 (Reactor Site Criteria, w11ch references the Technical Information Document (TID) 14844 source term), and 10 CFR Part 20 (Standards for Protection Against Radiation) all have limitations on releases related to power plant source terms. GDC 60 requires that the design include means to control suitably the release of radioactive materials in liquid and gaseous effluents and to handle waste produced during operations including anticipated operational occurrences. Prespolicants' Acoroach PRISH designers have proposed the calculation of a source term different from that done for LWRs. They have proposed siting source terms to bound the release from accidents considered in the design; the magnitude of these source terms is less than the TID-14844 LWR assumed source term. Additionally, at this time there is insufficient experimental data on the PRISH fuel to quantify the fission product release fractions or the behavior of those fission products migrating from the metal fuel through the sodium coolant. MHTGR designers have proposed siting source terms for accidents based on the expected fuel integrity. The coated microsphere fuel particles in the core are predicted by the preapplicant to contain all the fission products except for that released from the small number of failed particles resulting from in-service particle failures and added particle failures during accidents. Insufficient data currently exists to determine whether the MHTGR fuel performance will meet these expectations. The PlUS designer has proposed using a mechanistic LWR source term. Information has been provided in the Preliminary Safety Information Document (PSID) for fission product concentrations in both liquid and gaseous effluents. It is expected that PlVS designers will adopt the results of the ongoing EPRl/NRC effort to revise the TID-14844 source term previously used for LWRs. DRAFT t

The CANDU 3 designir uses a source term for each scenario. Each accident is evaluated and fission product release and transport is determined individually for each scenario. The staff has not, at this time, evaluated the CANDU 3 codes and methods. Discussion in order to evaluate the safety characteristics of advanced reactor designs that are significantly different fro:. LWRs, a method for calculating postulated radionuclide releases (source terms) needs to be developed. In a June 26 1990 staff requirements memorandum ($RM) related to SECY-90-016, the Commission req,uested the staff to submit a ) aper describing the status of efforts to develop an updated source term t1st takes into account 'best available estimates

• and current knowledge on the subject.

Based on this direction, the staff is now developing for LWRs a revision to the TID-14844 source term (NUREG-1455, ' Accident Source Terms for Light-Water Nuclear Power Plants,'draftreportforcomment, June 1992). The differences between the LWR designs and the HHTGR and PRISH designs warrant a separate evaluation of source terms. The CANDU 3 will also be different from LWR desiCns in certain respects. The coolant contains significant amounts of tritium. Following failure of a pressure tube there is no heavy-walled reactor vessel to contain releases (there are large volumes of water in two concentric low-pressure tanks; moderator and shield water). Consequently, the timing of releases is expected to be different from LWRs. Therefore, CANDU 3 also warrants a separate evaluation of source terms. The NRC staff is currently developing revisions to 10 CFR Part 50 and 10 CFR Part 100 to separate siting from source term dose calculations. The revisions to Part 100 being considered by the staff replace the present individual dose criteria with a population density standard. A fixed minimum exclusion area radius of 0.4 miles is specified. Other criteria regarding po)ulation protection and seismic criteria factors are also included in tle source term Part 100 revision. The staff's recommendations for the preapplication review are intended to be compatible with the proposed revisions. The staff's recommendations envision developing a set of scenario-specific source terms for each of the advanced reactors and CANDU 3 to allow a judgment as to whether the release from each specific sequence meets the accident evaluation criteria for sequences of that event category. Also, a source term may be developed mechanistically for core damage sequences to compare against applicable safety criteria. Recommendation Advanced reactor and CANDU 3 source terms should be based upon mechanistic analyses, provided that: . DRAFT

l 1. The performance of the reactor and fuel under normal and off-normal 4 conditions is sufficiently well understood to permit a mechanistic analysis. Sufficient data should exist on the reactor and fuel performance through the research, development, and testing programs to provide adequate cunfidence in the mechanistic approach. i 2. The transport of fission products can be adequately modeled for all barriers and pathways, including specific consideration of containment design to the environs. The calculations should be as realistic as possible so that the values and limitations of any mechanism or barrier are not obscured. I 3. The events considered in the analyses to develop the set of source terms for each design are selected to bound credible :evere accidents and design-dependent uncertainties. The design specific source terms for each accident category would constitute one component for evaluating the acceptability of the design. i 1 i l i a 4 S-DRAFT -+rw-wy r w g>-i== pw3>m, ~-T-

C. CONTAINMENT 1112t Should advanced reactor designs be allowed to employ alternative approaches to traditional

• essentially lea (-tight' containment structures to provide for the control of fission product release to the environment?

Current Reoulations General Design Criterion (GDC) 16 requires that LWR reactor containments provide an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment, and that containment-associated systems-assure that containment design conditions important to safety are not exceeded for as long as postulated accident conditions require. GDC 38-40 set requirements for containment heat removal, GDC 41-43 for containment atmosphere cleanup, and GDC 50-57 for containment design, testing, inspection, and integrity. Requirements for LWR containment leakage testing are established in 10 CFR Part 50, Appendix J. Preacolicants' Acoroach The MHTGR is not designed with a leak-tight containment barrier. The design relies upon high integrity fuel particles to minimize radionuclide release, and on a below-grade, safety-related concrete reactor building to provide retention and holdup of any radioactive releases. The reactor vessel and the steam generator vessel are in separate cavities within the concrete structure, in the event of a reactor coolant pressure boundary (RCPB) rupture, louvers in the reactor building are designed to allow the passage of gases to the environment, preventing building overpressure. The building design does not include containment isolation valves for the ventilation line from the building and has an open path to the environment via a drain line in the reactor cavity cooling system (RCCS) panels. Accident dose calculations assume a constant 100 percent volume per day building leak rate, and take credit for plateout on the building walls. PlVS, above grade, is designed with a low-leakage containment based on a pressure-sup)ression scheme that is integral with the reactor building, similar to tie ABWR and SBWR. Below grade, the concrete pool wall and floor, which is the reactor pressure boundary, and the containment are contiguous, separated only by a steel membrane. CANDU 3 is designed with a large, dry, steel-lined, concrete containment, without containment spray. The maximum leak rate (used in safety analyses) is 5 percent volume per day at the design pressure of approximately 30 psig. The structure is designed for a test-acceptance leak rate of 2 percent per day at the design pressure. These leak rates are significantly higher than those of a typical U.S. PWR containment. . DRAFT .n

The PRISM containment design is a high strength steel, low leakage, pressure-retaining boundary, consisting of two components, the upper containment dome and lower containment vessel. The upper containment is a steel dome. It differs from light-water reactor containments functionally in the following respects. The containment is specifically designed to mitigate the radioactive release consequences of severe events. The PRISM containment volume is markedly smaller than is typical of LWR containments; there is little separation between the reactor vessel and the containment boundary; and no safety-grade containment coolers or spray systems are provided. The entire containment structure is located below grade within the reactor building. Discussion Each of the advanced designs and CANDU 3 maintain an accident mitigation approach in which containment of fission products is a part. Two of the advanced reactor designs (PRISM and MHTGR) place the reactor building below grade, providing protection from external hazards. Generally, the advanced designs focus more attention than do LWRs on protecting the plant by providing passive means of reactor shutdown and decay heat removal (DHt). As a result, designers proposed less stringent containment design requirements. The staff recognizes that reactor designs, without traditional containment structures or systems, represent a significant departure from past practice on LWRs, and that existing LWR containment structures have proven to be an effective component of our defense-in-depth approach to regulation.

However, the Advanced Reactor Policy Statement recognizes that to encoura e incorporation of enhanced safety margins (such as in fuel design in advanced reactor designs, the Commission would look favorably on desirabl design related features or reduced administrative requirements. New reactor designs that deviate from current practice need to be extensively reviewed to assure a level of safety at least equivalent to that of current generation LWRs is provided, and that uncertainties in the design and performance are taken into account.

The staff believes that new reactor designs with limited o)erational experience require a containment system that provides a su)stantial level of accident mitigation for defense-in-depth against unforeseen events, including core damage accidents. This requirement may not necessarily result in a high-pressure, low-leakage structure that meets all of the current LWR requirements for containment, but it should be an independent barrier to fission product release. The proposed criteria will need to provide an appropriate level of protection of the public and the environment considering both the safety advantages of the advanced designs and the lack of an experience base in evaluating their performance. For evolutionary LWRs, the staff, in SECY-90-016, proposed to use a conditional containment failure >robability (CCFP) or deterministic containment performance goal to ensure a >alance between accident prevention and consequence mitigation. During the evolutionary LWR reviews, a great deal of careful review was necessary to assure that a probabilistic CCFP would not be used in a way that could detract from a balanced approach of severe accident prevention and consequence mitigation. For advanced designs and the CANDU 3, limited experience exists -{0-DRAFT l

in the analysis and evaluation of severe accidents which would lead to significant difficulty and uncertainty in assessing a CCFP. For this reason, the staff recommends that the deterministic containment performance goal be adopted for the advanced designs and the CANDU 3. The staff proposes to postulate a core damage accident as a containment challenge event and require that containment integrity is maintained for a period of approximately 24 hours after the onset of core damage. This approach is used because the preliminary nature of the advanced designs precludes a reliable assessment of the failure probability of accident mitigation systems and, therefore, of containment failure probability. Further, the CCFP is grounded in a firm understanding of LWR safety systems and accident progression. Intrinsic differences exist between LWR and advanced reactor technologies and their approaches to the balance between accident prevention and mitigation. A quantitative level of understanding of new technologies and systems comparable to that of lWRs is not yet available. Thus, the use of a performance based criterion rather than a quantitative one appears to be more appropriate for advanced reactor and CANDU 3 preapplication review given the current level of knowledge of advanced reactor and CANDU 3 risk and its prevention / mitigation elements. Recommendation The staff proposes to utilize a standard based upon containment functional performance to evaluate the acceptability of proposed designs rather than to rely exclusively on prescriptive containment design criteria. The staff intends to approach this by comparing containment performance with the accident evaluation criteria. Containment designs must be adequate to meet the onsite and offsite radionuclide release limits for the event categories to be developed as described in Section A to this enclosure within their design envelope. For a period of approximately 24 hours following the onset of core damage, the specified containment challenge event results in no greater than the limiting containment leak rate used in evaluation of the event categories, and structural stresses are maintained within acce) table limits (i.e., ASME level C requirements or equivalent) leases of radioactivity. After t11s period, the containment must prevent uncontrolled re - II - DRAFT

D. EMERGENCYPLANNING(EP) but Should advanced reactors with passive design safety features be able to reduce emergency planning zones and requirements? Current Reaulations Although emergency plans are not necessary for the issuance of a design certification under 10 CFR Part $2, they would be necessary for the issuance of a combined license under Part 52 or a license issued under 10 CFR Part 50. 10 CFR 50.47 requires that no operating license be issued unless a finding is made by the NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Currently, offsite protective actions are recommended when an accident occurs that could lead to offsite doses in excess of the Environmental Protection Agency's (EPA) Protective Action Guidelines (PAG), which are 1-5 rem whole body and 5-25 rem thyroid. At the lower prodected doses, protective actions should be considered. At the higher projected doses, protective actions are warranted. Prescolicants' Aceroach The proposed PRISM approach to emergency planning is significantly different from that of previous LWR applications, particularly in the area of offsite EP. A design objective of PRISH is to meet the lower level PAG criteria such that formal offsite emergency planning involving early notification, detailed evacuation planning, and provisions for exercise of the plan would not be required. In order to attain this objective, the PRISM design emphasizes accident prevention, long response times (36 hours) between the initiation of an accident and the release of any radiatton, and containment of accidents if they should occur. MHTGR proposed reduced offsite emergency planning for similar reasons as those proposed for PRISM. There would be an emergency plan for.an MHTGR and the plan would include any agency that could become involved in a radiological emergency (i.e., sheltering and evacuating the public and controlling the food supply). The differences and reductions from a typical plan for LWRs are that the MHTGR plan would have the exclusion area boundary (EAB) of 10 CFR Part 100 as the boundary of the emergency planning zone (EPZ), as may be allowed by Appendix E of 10 CFR Part 50 for gas-cooled reactors; and that there would be no rapid notification or annual drills for offsite agencies. This te based on the preapplicants' assertion that (1) the predicted dose consequences estimated at the EAB/EPZ for accidents are below the lower-level EPA sheltering PAGs and the public can be excluded from the EAB, (2) the significantly long time expected for the core to return to cr'ticality after being shut down by the doppler coefficient without the reactor protection system functioning (i.e., about 37 hours), and (3) the long time for the fuel and reactor vessel to reach maximum temperatures (i.e., about 100 hours) DRAFT

during accidents. The preasplicant asserts that the public around the plant would be outside the area t1at needs to be sheltered or evacuated and, further, there is ample time to notify and move the public during an event. With regard to PIUS, ABB expects that due to the passive safety features of the PlVS design, onsite and offsite emergency planning will be considerably simplified in comparison with current day LWRs. ABB/CE asserts that there appears to be no credible accident sequences that would lead to severe core damage. Offsite dose for the large break LOCA is claimed to be below the lower level EPA PAGs at 500 meters distance from the containment. No specific information on emergency planning was provided by the preapplicant for review beyond the general assertion thtt they intend to limit offsite doses to the PAGs. Discussion The advanced reactor designers have objectives of achieving very low probabilities (<l.0 x 10' per year) of exceeding the EPA lower-level PAGs. The vendors claim that these advanced reactors, with their passive reactor shutdown and cooling systems, and with core heatup times much-longer than those of existing LWRs, are sufficiently safe that the EPZ radius can be reduced to the site boundary, and that detailed planning and exercisin offsite response capabilities need not be required by NRC regulation. g of The preapplicant's state that this does not mean that there would be no offsite emergency plan developed, but rather that such a plan could have reduced details concerning movement of people, and need not contain provisions for early notification of the general public or periodic exercises of the offsite plan on the scale of present reactors. A similar policy issue was identified for the passive LWR design, but remains open. EPRI is currently working with the NRC staff to define a process for addressing simplification of emergency planning. The results of this effort should be applicable to advanced reactor designs. Recommendation The staff proposes that advanced reactor licensees be required to develop offsite emergency plans. Additionally, exercises, including offsite exercises, provisions for periodic emergency should be develo)ed. These actions are required by existing NRC regulations which include tie required establishment of an offsite emergency planning zone (EPZ). Consistent with the current regulatory approach, the staff views the inclusion of emergency preparedness by advanced reactor licensees as an added conservatism to NRC's ' defense-in-depth" philosophy. Briefly stated, this philosophy: 1 high quality in the design, construction, and operation of nuclear (p) ants torequires l reduce the likelihood of malfunctions in the first instance; (2) recognizes that equipment can fail and operators can make mistakes, therefore requiring safety systems to reduce the chances that malfunctions will lead to accidents that release fission products from the fuel; and (3) recognizes that, in spite of thess precautions, serious fuel damage accidents can happen, therefore requiring containment structures and other safety features to preven.t the DRAFI

release of fission products offsite. The added feature of emergency planning to the defense-in-depth philosophy provides that, even in the unlikely event of an offsite fission product release, ken to protect the population around there is reasonable assurance that emergency protective actions can be ta nucient power plants. Information obtained from accident evaluations conducted as outlined in Section A of this enclosure will provide input to the Emergency Planning requirements for advanced reactor designs. Based in part upon these accident evaluations, the staff will consider w1 ether some relaxation from current requirements may be appropriate for advanced reactor offsite emergency plans. The relaxations to be evaluated will include, but not be limited to, notification requirements, size of EPZ, and frequency of exercises. This evaluation will take into account the results of passive LWR emergency planning policy decisions. s 9 e DRAFT

I E. REACTIVITY CONTROL SYSTEM Issues Should the NRC accept a reactivity control system design that has no control rods? Current Reaulations General Design Criterion (GDC) 26 requires that two independent reactivity control systems be provided. One of the systems shall use control rods preferably using a positive means for insertion. The other system shall be capable of controlling planned reactivity changes to assure fuel limits are l not exceeded. PreaDolicants' Position The PlVS design does not have control rods. _ However, the preapplicant proposes that the design compiles with-the intent of General Design Criterion 26 by having two independent liquid boron reactivity control systems.- The normal reactivity control system pumps boron into the primary coolant loop to control reactor power or effect a reactor shutdown; this system is only safety-grade within the bounds of the containment isolation valves. The fully safety-grade reactivity control system relies on the ingress of highly borated water through the density lock from the reactor pressure vessel to scram the reactor. This ingress occurs when the equilibrium conditions across the thermal barrier of the density locks are disturbed by an imbalance between the thermal core heat generation and removal rates. Either a trip of as few as one of the four reactor coolant pumps or a i reactor overpower event (with forced flow) could initiate borated water flow i into the core. The reactor protection system initiates the scram function by tripping a single reactor coolant pump. Other reactivity control features of the design are in-core burnable poisons for power shaping, and limitations in l core size fer control of xenon oscillations for slow. large and small l reactivity changes. For rapid changes, the design relies on the highly l negative moderator temperature coefficient of reactivity, i L The density locks, essentially bundles of open, parallel tubes about 3 inches in diameter, have no moving parts. They are of. safety-grade construction and intended to be highly reliable. However, their function must be-demonstrated l and the potential for blockage and high cycle thermal fatigue cracking, and the effects of blockage and fatigue must be evaluated. A-failure of the l j. density locks would not only prevent a scram,- but'would interrupt-the only. safety-grade core cooling mechanism.- Discussion The existing LWR regulations provide prescriptive design guidance for one reactivity control system to contain-rods. Of the three advanced reactor-designs, only PIUS does not have the capability to control. reactivity with control rods. The PIUS design does have, however, three ways to introduce DRAFI m ,e-3 .ww<--r. -,--,s------ri--ww-- t g --.--wv --e.e-. ~ ~ ---r----- ,,,e r -y -ew

i 4 i i liquid boron into the core to control and shutdown the reactor. Two of the three rely on flow through the density lock from a common supply of borated j 2001 water. The other system is-the normal reactivity control system which ins a separate boron tank and is used for normal shutdown. The latter system is only safety grade within the bounds of the containment isolation valves. l Recommendation j The staff concludes that a reactivity control system without control-rods e should not necessarily disqualify a reactor design. A design without control rods may be acceptable but the appilcant must provide suff'cient information d to justify that there Is an equivalent level of safety in reactor control and protection as compared to a traditional rodded system. This information must include the areas of: a. reliability and efficacy of scram fun: tion 3 b, suppression of oscillations c. control of power distribution i j d. shutdown margin I e. operational cc.itrol } l 4 l r 4 k l a 4 Y l f d ,16 - ORAFT

i F. OPERATOR STAFFING AND TUNCTION 111M Should advanced reactor designs be ' allowed to operate with a staffinc complement that is less than that currently required by the LWR regulations. Current Reculations The NRC has established the requirements for control room staffing in 10 CFR 50.54 control room (m)(2)(111) which states a senior operator must be present in the at all times and a licensed operator or senior operator must be present at the controls of a fueled nuclear power unit. 50.34(m)(2)(1 provides a table identifying the minimum staffing requirements for an ) operating reactor. Standard Review Plan 13.1.2, Section !!.C states that at any time a licensed nuclear unit is being operated in modes other than cold shutdown, the minimum shift crew shall include two licensed senior reactor operators (SRO), one of whom shall be designated as the shift supervisor, two licensed reactor operators.(RO), and two unlicensed auxiliary operators (AO). Prescolicants' Position The MHTGR plant is presently four reactor modules with two modules feeding a single steam supply system. The design includes a shift-staffing level of eight persons w1o would be dedicated to plant operations; a senior licensed shift supervisor, two licensed reactor operators in the control room, and five roving non-licensed operators. This results in three licensed and five non-licensed operators for four reactor modules. The PRISM control room would contain the instrumentation and controls for all nine reactor modules and their power conversion systems. The objective for the minimum number of operating staff would include: one SRO shift supervisor, one SRO assistant supervisor, one R0 per )ower block (three modules) licensed operators for nine reactor modules. in the control room, and three plant R0s. T11s results tn a minimum of eight During normal plant operations the PIUS main control room would be manned by two R0s and a SR0 shift supervisor. The shift supervisor would not be required to be in the control room at all times. The CANDU 3 prea)plicant has not proposed a specific number of licensed operators, but tie staff's expectation is that CANDU 3 will meet the current LWR staffing requirements. Discussion Present day LWRs would require a minimum of one shift supervisor, one SRO, and two operators per reector. The designers of advanced reactors have stated that the highly automated operating systems, the passive design of safety l l DRAFT

features, and the large heat capacity results in reactor designs that respond to transients in a manner that demands less of the o>erator tl1an do the current operating plants or evolutionary designs. Tie preapplicants assert that the passive safety features and, in some cases, large coolant inventory of the pRISH, MHTGR, and PlVS designs may not require an operator to act or intervene for several days following an accident. These designs also automate systems that start up, shut down, and control these reactors. The vendors of these reactors have suggested that they could be operated with fewer licensed operators and believe that this would reduce significantly the training and operating costs to licensees. A similar policy issue, Role of the Operator in a Passive Plant Control Room, was identified in the staff's June 25, 1992, draft policy pa)er on passive reactors. In that paper, the staff expressed concern that tie man-machine interface for the passive reactors had not been sufficiently addressed and that actual testing needed to be done on a control room prototype. The staff believes that position is also applicable to advanced reactors. Recommendation The staff believes that operator staffing may be design dependent and intends to review the justification for a smaller crew size for the advanced reactor designs by evaluating the function and task analyses for normal operation and accident management. The function and task analyses must demonstrate and confirm through test and evaluation the following: Smaller operating crews can provide effective response to a worst case array of power maneuvers, refueling and maintenance activities, and accident conditions. An accident on a single unit can be mitigated with the proposed number of licensed operators, less one, while all other units could be taken to a cold shutdown condition from a variety of potential operating conditions including a fire in one unit. The units can be safely shut down with evcntual progression to a safe shutdown condition under each of the following conditions: (1) a complete loss of computer control capability, (2) a complete station blackout, or (3) a design basis seismic event. The adequacy of these analyses shall be tested and demonstrated on an actual control room prototype. l DRAFT l

G. RESIDUAL HEAT REMOVAL liin Should advanced reactor designs that rely on a single completely passive, safety-related Residual Heat Removal (RHR) system be acceptable? Current Reoulations General Design Criteria (GDC) 34 requires the RHR function to be accomplished using only safety-grade systems assuming a loss of either onsite or offsite power,andassumingasinglefallurewithinthesafetysystem. Regulatory Guide 1.139 (issued in draft for comment), augmenting the GDC, ion within states that the RHR function must be serformed to reach a safe shutdown condit 36 hours of reactor slutdown. Branch Technical Position (BTP) RSB 5-1 states that the RHR function must be performed in a reasonable period of time following reactor shutdown. frescolicants' Position The PRISM design uses the reactor vessel auxiliary cooling system (RVACS) as the safety-grade system for residual heat removal from the reactor core. Reactor generated heat is transferred through the reactor vessel to the containment vessel outer surface. RHR is then accomplished through natural circulation heat transfer to the atmosphere. Cooler air flows downward into the below grade reactor silo, where it is turned inward and upward to be heated by the containment vessel outer surface and a special collector cylinder. This heated air then flows out of the silo and is released to the atmosphere. The RVACS is completely passive and always in operation. The RVACS is proposed as a backup to normal non-safety-grade cooling through the intermediate heat transport system, the steam generator, and condenser. If the condenser is not available for cooling but the intermediate sodium loop remains available, then the non-safety-grade auxiliary cooling system (ACS) supplements RVACS. The ACS operates through natural circulation air cooling of the steam generator. The RVACS design basis analysis (performed by the designer) results in high temperature conditions (within design limits) for an extended period of time if no u her system is operated. However, use of the ACS system in conjunction with RVACS can limit peak coolant temperature for decay heat removal to about 15 *C above normal operating temperatures. The MHTGR is designed with only one safety-grade system for removing residual heat from the core, the reactor cavity cooling system (RCCS). The RCCS consists of panels within the reactor cavity and ducts connecting the RCCS panels to four inlet / outlet ) orts. Redundancy,is provided by these separate ports and a cross-connected leader that surrounds the reactor vessel (i.e., any panel can be fed from any inlet and can discharge to any outlet). The RCCS operates by absorbing radiant heat from the reactor vessel to the panels which surround the reactor vessel and transferring the heat by convection to the air flowing.by natural circulation in the panels. As the heated air rises, cooler, atmospheric air is drawn to the panels through the inlet ports. There are no active components in the RCCS. The system is always in DRAFT

o e o>eration. tie shutdown cooling subsystem (SCS) are inoperable.The RCCS is relied upon when th sort system (HTS)he and Tie HTS utilizes t steam generators and non-safety-grade feed system and condensers and is used during normal operations, startup/ shutdown and refueling. The SCS is c non-safety-grade backup to the HTS. The SCS system uses an alternate helium circulator for core cooling and an additional heat sink, the shutdown cooling heat exchanger. Again, use of the non-safety-grade backup RHR systems reduces the frequency, magnitude and duration of high temperature challenges to the reactor vessel. The PIUS design uses a safety-grade passive closed cooling system (PCCS)ht for residual heat removal from the reactor pool. The system consists of eig independent parallel loops located in four separate compartments that are physically separated from each other.~ Heat is dissi)ated through four (4) natural draft cooling towers located on the top of tie reactor building. One cooling tower is in each quadrant of the reactor building. The reactor pool water can be maintained at 95 'C with one loop out of service. The system is always in operation. Reactor residual heat can be removed with the condenser during startup/ shutdown and refueling conditions. If the condenser is not available, a non-safety-grade diesel-backed pump system can cool the pool water. Discussion Similar issues were identified for the RHR system of the passive LWR designs. In a draft Commission paper issued for comment on February 27, 1992, the staff identified issues relating to the ability of passive systems to reach safe shutdown, definition of a passive failure, and treatment of non-safety systems which reduce challenges to the passive systems. These issues remain open and the staff will propose recommendations in the future for resolution, in the case of advanced reactors the safety-grade RHR systems are completely passive and are in continuous operation. Continuous performance monitoring of the passive systems is one advantage of the constant operation. The high heat capacity of PRISH and MHTGR lead to longer time periods before exceeding temperature limits. PRISM and MHTGR use the natural circulation of air to remove residual heat. PlVS uses natural circulation of water through natural draft cooling towers for its RHR system. The lack of check and squib valves, the continuous operation and use of a single phase fluid in the system appear to offer increased reliability over the passive LWR systems. However, reliance only on passive systems may lead to high temperature challenges to the reactor vessel and reactor internal structures since higher heat removal rates in passive cooling situations require larger temperature differences between the reactor and cooling medium (air). Elevated temperatures (above normal operating values) may exist in the vessel and internal structures for long periods of time. Particularly in the high temperature reactors, the PRISM and MHTGR, creep damage may be more likely as the result of these high-temperature transients. ' DRAFT l e

i 1 i i Recomendation As a result of the unique design features of the PRISM, MHTGR, and PIUS designs, the staff believes that reliance on a single,ing out its future completely passive, safety-related RHR system may be acceptable. In carry detailed design evaluation, the staff will assure that NRC regulatory treatment of non-safety-related backup RHR systems is consistent with Comission decisions on passive light-water reactor design-requirements. P i 9 4-21 - - DRAFT ,w. r-- m-w -.-%--.y mm,,y ww-- ..-...-=-*% ,,,.-.--y ,..-.-has-. e aw, e.-,--g

H. POSITIVE YOID REACTIVITY COEFFICIENT liin Should a design in which the overall inherent reactivity tends to increase under specific conditions or accidents be acceptable? Current Reaulations General Design Criterion (GDC) 11 requires that the reactor core and coolant system be designed so that in the power operating range the net effect of prompt inherent nuclear feedback characteristics tehd'to compensate for rapid increases in reactivity. Prescolitants' Position In the PRISM design, the maximum sodium void worth, according to the preapplicant, assuming only driver fuel and internal blanket assemblies void, is nominally $5.50. If radial blanket assemblies are included, the sodium void worth is nominally $5.26 which does not include the -70 cents from gas expansion modules (GEM). Should sodium boiling begin, on a core-wide basis under failure to scram conditions with a total loss of flow without coastdown, the reactor could experience a severe power excursion and core disruption. The )redicted temperature reactivity feedback is approximately -80 cents prior to tie onset of sodium voiding. This mitigates to some extent the positive reactivity addition. For sodium voiding to occur, multiple failures of redundant and diverse safety-grade systems would be required. Although the overall power coefficient for a CANDU 3 reactor is claimed to be slightly negative and very close to zero, the coolant void reactivity is aositive throughout the fuel core lifetime. The total core void wo.h h aetween $1 and $2. The positive void coefficient is not a concerr Luring normal operation, but, during a large LOCA at specific locations, void r9 activity increases dramatically. If CANDU 3 were to experience a large-break LOCA (guillotine rupture of an inlet header) with a failure of both shutdown systems, the positive void reactivity insertion could 1 cad to a power excursion followed by core melting, The CANDU 3 design is intended to prevent an unscrammed event from occurring through the use of two separate shutdown systems each to be independent, redundant, diverse, and safety grade. Discussion The staff considers the existence of positive coolant void coefficients, or any reactivity effect that tends to make a postulated accident more se.ere, a significant concern. As a result of a positive void reactivity coefficient, events involving even a relatively short scram delay could result in a core disruption accident. The staff intends to require the preapplicant to analyze the consequences of events (such as ATWS, unscrammed LOCAs, delayed scram events, and transients which affect reactivity control) that could lead to core damage as a result of the positive void coefficient, taking into account the overall risk perspective of the designs. A core disruption accident in D7!ATT

i

l. -

aI; j either the PRISM or CANDU 3 designs may not necessarily lead to a large scale i release of the radionuclide inventory to the atmosphere due to their i respective mitigative designs. In the CANDU 3 reactor, multiple redundant, 2 diverse fast acting scram systems s're provided to address the positive j coefficients. j Attempts to modify the designs to reduce the effects of these positive j coefficients may result in other consequences potentially as serious. For example, in the PRISM design, the positive void coefficient seems to result from the design objectives of maintaining a passive shutdown capability and of minimizing the reactivity swing over core life. Attempts to reduce the PRISM void worth might have the effect of increasing the severity of rod withdrawal accidents or reducing the ability to withstand an unscramed loss of heat sink events without core damage. l Recommendation l The staff concludes that a positive void coefficient should not necessarily i disqualify a reactor design. The staff is proposing to require that the PRISM unscramed LOCAs, delayed scrams, and transients affecting rea(such as ATWS,c and CANDU 3 preapplicants analyze the consequences of events i j ~i that could lead to core damage as a result of the positive void coefficients. The staff's review of these analyses will take into account the overall risk perspective of the designs. Whether the preasplicants will be required to j consider changes in the designs to mitigate tie consequences of these j accidents will de)end on the estimated probability of the accidents as well as l the severity of tie consequences. 1 l l l l l i i h-e F i l DRAFT. I + ,.~.i,..r,,..,

l 1 1. CONTROL ROOH AND REMOTE SHUTDOWN AREA DESIGN t h12t Can current recuirements for a seismic Category 1/ Class IE control room and I alternate shutcown panel be fulfilled by a Remote Shutdown Area, end a non-seismic Category I, non-Class IE control room? Current Reculations The current LWR requirements for control room and remote shutdown area design are provided in 10 CFR Part 50, Appendix A, and 10 CFR Part 100. General Design Criterion (GDC) 19 requires that a control room with adequate radiation protection be provided to operate the plant safely under normal and accident conditions and that there be an ability to shut down the plant from outside i the control room. GDC 17 requires that the electrical system for the control room and remote shutdown equipment meet the requirements for quality and j independence. These requirements are defined as Class IE in the supporting i IEEE standards. GDC 2 and 10 CFR Part 100 require that structures and systems important to safety be designed to seismic Category I standards to remain 4 functional during a safe shutdown earthquake. i Prenoolicants' position The control room for PRISH contains the instrumentation and controls for all nine reactor modeles and their power conversion systems. The control room structure is not considered safety related and, therefore, the room is not designed to seismic Category I design requirements. Additionally, no equip-ment in the control room is safety grade. A separate alternate shutdown console is located in the protected area of the reactor service building. The alternate shutdown console is within a seismic Category I structure and is equipped with the necessary Class IE controls and instrumentation to protect the core and has the required habitability control system. The MHTGR design has, for the four modules, a non-safety-related central con-trol room to operate the plant and a seismic Category I remote shutdown area from which to respond to accidents if necessary. Neither the equipment in the-control room nor the remote shutdown area are Class IE. The remote shutdown area does not contain safety-related equipment, nor does it include a ventilation system for operator habitability, or a safety-related manual _ scram. This is based on the areapplicant's position that accidents do not require operator response. Tae only manual scrams are non-safety-related and are located in the remote shutdown area, not the main control room. The CANDU 3 design utilizes a main control room to perform all monitoring and control functions for normal operation and all accident-conditions, except those events for which the control room becomes unavailable. :The main control-room is not designed to be operable.following an earthquake, tornado, fire, or loss of Group 1 (non-essential) electrical power, but the operator must remain available to proceed to the secondary control area. The secondary control-- area duplicates, to the fullest extent possible, the control locations, 24 - DRAFT .m- ..yw. y r v -.,y

Iayouts, and capabilities present in the Main Control Room. The secondary control area is seismically qualified and is electrically isolated from the main control room so that failures occurring in the Group 1 area will not interfere with control and monitoring of safety systems from the secondary control area. All equipment located in the route from the main control room to the secondary control area is to be qualified to the extent necessary to prevent route blockage, fire, or flood. CANDU 3 has specified requirements to assure habitability during accident conditions. The central. control room for the PIUS design is a seismic Category I structure. However, the safety-related systems within this structure are for monitoring only to assure that the core is protected. Although the operator could take actions, these actions would be with the use of non-safety-grade controls. The two remote shutdown areas are housed in separate compartments at the bottom of the reactor building in protected selsmic Category I areas. Each remote area contains one half of the safety-grade control equipment, e.g., the reactor trip and interlock system, control of certain isolation valves, and safety-grade monitoring systems. The manual reactor trip system is a push-button control of the main reactor coolant pumps. Both the main control room and the emergency shutdown areas are serviced by a safety-grade ventilation system to assure habitability during accidents. Discussion The staff believes that the operators remain a critical element in ensuring reactor plant safety and that no increased burden should be placed on operators managing off-normal operations. The control room is the space in the plant where operators are most familiar with the surroundings and normally manage plant activities. The staff is reluctant to approve any design that would increase the frequency of evacuation of the control room during design basis accident conditions or hamper the control or monitoring of upset conditions as the event sequence progresses. The staff believes human performance will still play a large role in the safety of the advanced plants and CANDU 3 and that the quality of support provided by the safety-related, seismic Category I and electrical Class IE control room is appropriate. l The staff also believes that any remote shutdown area should be designed to complement the main control room. Sufficient Class 1E instrumentation and controls should be available to effectively manage anticipated accidents that would result in a loss of the control room functions. The location and structure of the remote shutdown areas should also ensure continuity of operations to the greatest extent possible. A related policy issue was identified in the staff's February 27, 1992, draft paper on policy issues for the passive LWRs where EPRI proposed less conservative control room habitability requirements and that analyses of control room habitability be limited to 72 hours instead of the accident duration. The staff disagreed with the proposed EPRI guidance and offered different criteria. Similarly, the staff in its June 25, 1992, draft policy DRAFT

4 ] paper defined positions on common mode failures in digital systems and on annunciator reliability. Staff

• Suirements for advanced reactor designs will 3

be consistent with passive LWR p... y guidance for these issues, once the guidance is finalized. j Recommendation l The staff recommends that until passive LWR policy for design requirements of control rooms and remote shutdown facilities is finalized, the staff will i apply current LWR regulations and guidance to the review of advanced reactor i designs. This will ensure that plant controls and the operators will be adequately protected so that safe shutdown can be assured in accident i situations. e i I l 4 i . DRAFT

i. i l

J. SAFETY CLASSlFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS hing What criteria should the NRC apply to the advanced reactor designs to identify the safety-related structures, systems, and components? j Current Reaulations Title 10 of the Code of Federal Regulation Section 50.49 A>pendix A.VICa)(1) of 10 CFR Part 100 list the following(b)(1) and the current criteria to define tle safety-re'ated structures, systems and components: a. those needed to maintain the integrity of the reactor coolant pressure boundary (RCPB); ) b. those needed to shut down the reactor and maintain it in a safe condition; and c. those needed to prevent or mitigate the consequences of accidents that could result in doses comparable to Part 100 guidelines. Amendments to Parts 50 and 100 have been proposed (57 FR 47862)for fut to update criteria used in decisions regarding reactor siting and design 1 nuclear power plants, including the advanced LWR designs. These proposed revisions include the temporary relocation of the dose considerations for l reactor siting (i.e., the current Part-100 guidelines) from Part 100 to Part 50 until such time as more specific requirements are developed regarding accident source terms and severe accident insights. i Prenonlicants' Position The advanced reactor designs rely on a limited number of safety-related systems to-protect the core-and the public. Some of these systems are entirely passive, with no moving components and do not require operator action. The vendors believe that this reduction in safety-related equipment results in simpler plant designs with lower costs. This also results in many structures, systems, and components, which are considered as safety related in LWR designs, being classified as non-safety-related in the advanced reactor designs. Of the advanced reactor designs, only the MHTGR design is not using the current LWR criteria above for safety classification. For the MHTGR design, the only criterion for safety-grade classification is those structures, i systems, and components needed to mitigate the dose consequences at the site i boundary from accidents or events to below the guidelines in the current -10 CFR Part 100. Several major issues with safety classification were identified previously by the staff in the Draft PSER (NUREG-1338 : (1) the RCPB is not entirely safety related (2) no safety-related equipm)ent is used H to pressurize and depressurize the RCPB notsafetyrelated,and(4)neitherthe,co(3)thecoolantmoisturemonitoris ntrol room or remote shutdown area 27 - DRAFT .v., e.

are safety related, and (5) no safety-related instrumentation providing reactor protection or monitoring functions are available in the control room or remote shutdown area. Discussion The NRC LWR criteria are intended to require defense in depth; the advanced reactor designs include high quality, non-safety-related active systems to >rovide defense-in-depth capabilities for reactor coolant makeup and decay 1 eat removal. These would be the first line of defense in the event of transients or plant upsets. The non-safety-related systems are, according to the designers, not required for mitigation of design basis events, but do provide alternate mitigation capabil<ty. In a recent draft SECY paper covering the passive A.WRs, the NRC staff stated that it was still evaluating the issue of treatment of non-safety-related systems for the passive ALWRs and the proposed resolution to this issue would be provided later. The staff plans to treat non-safety-related-systems consistent with the eventual position for passive LWRs. Retomendations The staff intends to apply the LWR criteria for identification of safety-related structures, systems, and components to the MHTGR design. Requirements for non-safety-related systems will-be consistent with the NRC position for passive LWRs. We have noted that LWR criteria may be restructured within Parts 50 and 100,dard design certification. and our expectation is that the criteria in Part 50 will apply to the stan 9 V DRAFT ~ j

A. CANADIAN DEUTERIUM URANIUM (CANDU) 3 REACTOR DESIGN DeveloomtDt History The CANDU 3 is the latest version of the pressurized heavy-water reactor (PHWR)systemdevelopedinCanada. The CANDU 3 design evolved from other CANDU PHWRs, most notably the CANDU 6 design. The CANDU 3 is a generic standard design that has retained many key components (steam generators, coolant pumps, pressure tubes, fuel, on-line refueling machines, instrumentation, etc.) that have been proven in service on operating CANDU power reactors. Currently, there are 25 CANDU reactors in operation in 6 different countries and 19 under construction. The first CANDU reactor was placed in service in 1968. CANDU experience to date amounts to over 175-years of effective full power operation. On May 25, 1989, Atomic Energy of Canada}t the CANDU 3 reactor design for Limited. Technologies (AECLT) informed the NRC of their intent to subm standard design certification in accordance with Part 52. AECLT of Rockville, Maryland, is a wholly-owned subsidiary of Atomic Energy of Canada, Limited (AECL) (a crown corporation of Canada), and is the preapplicant for the CANDU 3 design. AECL in Canada is also pursuing standard design certification of the CANDU 3 with the NRC's Canadian counterpart, the Atomic Energy Control Board of Canada. AECLT's current plans are to submit a standard design certification application for CANDU 3 in the 1995-1996 time frame. Desian Descriotion The CANDU 3 is a 45C MWe heavy-water-cooled and -moderated, horizontal pressure tube reactor that evolved from the CANDU 6 design. The CANDU 3 uses deuterium oxide (heavy water) as a moderator because its small thermal neutron capture cross section allows the use of natural uranium as fuel. However, because the moderation properties of heavy water are not as good as light water, the volume ratio of moderator to fuel is five to eight times that of an LWR. Thus, the CANDU core is larger than an LWR core generating the same power. This results in a lower core power density for CANDU 3. In addition, the CANDU 3 core is neutronically loosely coupled which results in xenon induced flux tilts that requires a relatively complicated computer operated spatial flux control system. As in LWRs, CANDU 3 fuel elements consist of pressed and sintered uranium dioxide pellets enclosed in a zirconium cladding. Each CANDU 3 fuel bundle is about 20 inches long, consists of 37 fuel compacts and is loaded into each of the 232 horizontal fuel channels. Each of the 232 horizontal fuel channels consists of a pressure tube concentrically placed inside a calan @ia tube. The pressure tubes form part of the reactor coolant system pressure boundary. Because of the low excess reactivity associated with a natural uranium core. DRAFT J

the CANDU design must be fueled on a continuous basis during power operation by an automatic fueling machine. On-line fueling is the primary means of changing reactivity in the CANDU 3. For the CANDU 3 design, heavy water coolant flow through the core is uni-directional, thereby facilitating on-line fueling from one end of the reactor with a single fueling machine. The primary system operating pressure (nominally 1435 outlet headers. psig) is maintained by a pressurizer connected to one of the The CANDU 3 light-water secondary system is similar to that of a PWR, The f e1 channel assemblies are enclosed in'a horizontal, cylindrical vessel called a calandria that contains the low-temperature (140 'F),ith tie integral low->ressure, heavy-water moderator. The calandria vessel, in conjunction w end shields, supports the horizontal fuel channel assemblies and the vertical and horizontal reactivity control unit components. The CANDU 3 utilizes four reactivity control systems for reactor control and shutdown during normal operation, and two redundant and diverse safety-grade shutdown systems are used for reactor shutdown following a transient. A separate moderator heat removal system ensures that the moderator remains subcooled. All systems in the CANDU 3 design are assigned to one of two groups - either Group 1 or Group 2. The systems of each group are capable of shutting down the reactor, maintaining cooling of the fuel and providing plant monitoring capability in the event that the other group,of systems is unavailable. Group 1 systems are those primarily dedicated to normal plant power pro-duction. The Group 2 systems include four special safety systems and other safety-related systems. These maintain plant safety in the event of a loss or partial loss of Group 1 systems, and mitigate the effects of accidents, including the design basis earthquake. The Group 1 and Group 2 systems are, to the greatest extent possible, located in separate areas of the plant. CANDU 3 employs two fast-acting, redundant, and diverse Group 2 shutdown systems, separate from the Group 1 reactor regulating system. Shutdown System No. 1 (5051 consists of 24 vertically inserted control rods. Shutdown System No.2(5D52 consists of six horizontal nozzles through which a gadolinium nitrate sol tion is injected. Both shutdown systems inject into the low-pressure moderator, precluding a rod ejection accident. In addition to the two shutdown systems, the remaining special safety systems include containment and emergency core cooling system (ECCS). The CANDU 3 containment system includes a reinforced concrete containment structure with a reinforced concrete dome and an internal steel liner. The containment is designed with a test acceptance leakage rate of 2 percent per day. ECCS supplies light-water coolant to the' reactor in the event of a loss-of-coolant accident. Each of the four safety systems is required to demonstrate during operation, a dormant unavailability of less than 10'3 or about 8 hours per year, and be physically and functionally separate from the normal process systems and from one another. The CANDU 3 shutdown cooling system is designed to remove heat from the HTS at nominal operating temperature and pressure. DRAFT L _ _ - _ _ _ _ _]

8. MHTGR (MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR) Development History The Modular High Temperature Gas-cooled Reactor (MHTGR) was )roposed to NRC by the U.S. Department of Energy (DOE) in 1986 in response to tle Commission's AdvancedReactorPolicyStatement(51FR24643). A preliminary safety information document (PSID) and twelve amendments were submitted from October 1986 to March 1992. The PSID and 10 amendments were reviewed by the Office of Nuclear Regulatory Research and a draft preapplication safety evaluation report (PS:R) was issued by NRC in March 1989. DOE recently advised the NRC that the MHTGR design certification application schedule will be established in August 1993, when a DOE decision on the gas-cooled new production reactor funding will be_made. The Energy Policy Act of 1992 requires DOE to submit a preliminary design approval application by September 30, 1996. Commercial gas-cooled reactors began with the graphite-moderated, carbon dioxide-cooled Magnox re;ctors developed in the early 1950's in the United kingdom and France, in the United States, gas reactor development resulted in the 40 MWe Peach Bottom 1, which operated from 1967 to 1974, and the 330 MWe Fort St. Vrain, which operated from 1976 to 1989. There have been about 50 gas-cooled reactors in the world totaling about 1000 reactor-years of operation. In this total, there has been about 50 reactor-years of experience with the HTGRs. The BISO and TRISO (trade names) multi-layered microsphere fuel form is used in HTGRs. The BISO fuel form, a fuel kernel with two major layers, was used in Peach Bottom 1; and the TRISO fuel form, a fuel kernel with four major layers (including a silicon carbide layer), was used at Fort St. Vrain. The TRISO fuel form provides higher fuel integrity requirements than the BISO fuel and is the reference fuel for the MHTGR. DOE maintains agreements with Germany and France for the exchange of technical information concerning the integrity of the reference MHTGR fuel, and experiments will be conducted in France. As part of DOE's Technology Development Program for the MHTGR, post irradiation testing of development fuel at Oak Ridge National Laboratory is being performed, and a technical information exchange agreement was established with Japan, which is building an experimental HTGR. Major trends in recent HTGR designs, including the MHTGR, are the following: (1) increased system pressure, (2) steel pressure vessels for the smaller HTGRs, including the MHTGR, versus the prestressed concrete reactor vessel for larger HTGR designs as fort St. Vrain, (3) pro)osed greater fuel integrity, 4 with a 6 x 10 fraction of failed fuel assumec for the MHTGR, and (4) lower enriched uranium fuel. Desion Descriotion The standard MHTGR plant is four reactor-steam generator modules and two steam turbine-generator sets. Eacn module is designed for a thermal output of 350 MWt. Two reactor modules are coupled to a steam turbine-generator set to produce a total plant electrical output of 540 MWe. DRAFT

Thelow-powerdensity(5.9 watts /cc)reactorcoreisheliumcooledand graphite moderated, and uses ceramic coated (four major layers) microspheres in an organic bonded c lindrical compact as the fuel. The core design is intended to provide a arge negativa doppler coefficient to shutdown the reactor with heatu. The microsphere fuel design is stated to allow fuel temperatures as hi h as 2900 'F without significant fission product release. The compacts are p aced in small vertical holes in the hexagonal graphite block fuel assemblies. The fuel assemblies are cooled through passages in the blocks. There are about 660 graphite blocks in the 66-column annular core region between the inner and outer reflector regions. The helium is a single phase coolant chemically and neutronically inert. The MHTGR has a below-grade, safety-related reactor building, containing the 1 reactor and steam generator vessels. The core is in a steel vessel located, with the steam generator, in the reactor building below ground to reduce seismic loads. The reactor vessel is above the steam generator vessel to prevent natural circulation and connected to this vessel by a horizontal crossduct vessel. The reactor and steam generator vessels are in separate cavities. The secondary side water is superheated in the steam generator. The core outlet helium temperature is about-1300 *F and the steam outlet temperature is about 1005 'F. The secondary side pressure is higher (about 2500 psig) than that on the primary side (about 925 ps , so water would leak into the coolant with a steam generator tube leak or f ure. Reactor protection is provided by two safety-related reactor protection systems (control rods and boron carbide balls), which are diverse and redundant, and one non-safety-related system (control rods). The non-safety-related system is independent from and redundant to the safety-grade systems. The equilibrium shutdown core temperature would be approximately 250 'F, the design temperature for refueling. The safety-related RCCS is a set of panels surrounding the reactor vessel with a header connection to four inlet and exhaust ports above ground. This allows hot air to rise thus removing heat transferred from the reactor vessel while cold air is drawn from outside into the panels. The system 1) is entirely passive with no moving components, (2) is always operating,

3) automatically responds to rising temperatures through thermal radiation an natural circu-lation, and (4) has flow path redundancy to the cooling panels through a cross-connected header.

In addition, there are two other non-safety-related, active heat removal systems: (1 the shutdown cooling system in the bottom of the reactor vessel, and (2) the m)ain circulator / steam generator in the primary 4 cooling loop. The non-safety-related systems are not relied upon for accident safety analyses. The multiple barriers to fission product release are the coated fuel microspheres, the graphite blocks, the ASME Code reactor coolant pressure boundary (RCPB), and the containment. The containment is the reactor building below ground with containment isolation valves on the steam generator DRAFT

main steam and feedwater inlet piping. It will not retain the gases from a rapid ACPB depressurization, but is designed to have a leak rate of less than 100 percent / day after initial depressurization. i l I 1 I 4 t I l t I 5-DRAFT l i l c g y ~ ' + g, ~ ~x., --w.-~v.. ...,..--v,v-,,--,v-,v,.,v,v ,-,y-, ~, -+-, v w -e e-- - ~

C. PIUS (PROCESS INHERENT ULTIMATE SAFETY) Historical Development The Process inherent Ultimate Safety (PIUS) 7eactor is being designed by ASEA Brown Sveri Atom (ABB-atom). The concept evolved in the early 1980's from an extent an of then ABB-Atom's low temperature district heating design. In October 1989, ABB requested a licensability review of the PIUS design in I accordance with NUREG-1226, and in May 1990, ABB submitted the PIUS preliminary safety information document (PSID) for staff review. ABB plans to apply for design certification of the PIUS design in the 1994-1995 time frame, assuming a favorable preapplication review.

• 21US design concept has already undergone tests related to the design Mes. ABB has completed testing using the MAGNE Test Rig to simulate o r m :ters such as diffusion and mixing across the primary loop / pool lth consideration for effects of turbulence, stratification, m, rat n of boron, and others. Large scale tests of the PlVS design y

p r W. '. s, such as flow and density lock operation, were done at the ATLE o Tm-These tests were used to validate the RIGEL code to calculate the 3 t w. w safety and transient performace. ATLE was a full = height simulation et v MS pool. Other tests of the PIUS design principles have been carried ] out at M and TVA, and other additional large scale tests and a larger test rig are 3 1anned to be started this year for the purpose of design i ptimization, as well as 4pecial component testing. It is planned that this larger test rig will serve as the basic test facility for developing data for the detailed design and verification, y Desian Descriotion h PlVS is t 640 MWe advanced pressurized water ro r tor (PWR) design with four loops. It relies e thermal hydraulic effect' accomplish the control and safety functions im ~re usually performed L, . chanical means. The safety-grade reactor heat removal system for the PlVS design is completely passive and is always in operation. The PIUS design consists of a vertical bollow cylinder, the reactor module, which contains the reactor core. The raactor module is subinerged in a large concrete reactor vessel containing 3 3,300 H (870,000 gallons) of highly borated water. The reactor module is open to the ber*ted pool at the bottom and at the top of the reactor module. At these two upr Ugs, density locks keep the borated pool water from the rmtor module during normtl operation. Under normal operations, the primary loop reactor water flows up through the core, out of the top of the reactor module to the steam generators, and is pum;.ed back into the bottom of the reactor module, bypassing both the top and bottom density locks. There is no physical flow barrier in the density locks between the primary loop and the borated pool, however, the difference in density between the primary loop reactor water and the cooler borated pool water provides a relatively stationary interface. When sufficiently upset during transient conditions, such as loss of flow or a power mismatch, the density difference is overcome - DRAFT ~^ ~

d and the borated water flows into the core and shuts down the reactor. A natural circulation flow path is then established from the borated pool through the lower density lock, up.through the core, and back into the borated pool through the upper density lock for long term shutdown cooling. Unlike most reactors, PIUS does not employ mechanical control rods for regulating reactivity. Reactivity is controlled by the boron concentration and temperature of the primary loop reactor water. An active reactor protection system (RPS), with associated instrumentation and actuation systems, is also provided in PIUS. The RPS and the associated systems have the task of detecting departures from acceptable operating conditions and initiating coolant pump trip to cause density lock flow and a reactor scram. Other aspects of the P1U5 design are similar to the passive LWRs being considered by the staff (AP-600 and the SBWR). Although PIUS is a PWR, its operating pressure (1,305 psi) is close to that of a BWR, The proposed containment for the PlVS design is integral with the reactor building, similar to the ABWR and SBWR, Leak-rate has been defined as not to exceed 1 volume percent per day at a design pressure of 26 psig. The acceptance leakage VC.Je is expected to be 0.5 percent at design pressure. h 1 i 4 4 4 4 .. DRAFT wv e-

1 {., " i D._. PRISM (POWER REACTOR INNOVATIVE SMALL MODULE)_ Development History The U.S. Deoartment of Energy (DOE) selected the Power Reactor Innovativt _. 3 {. Small Module (PRISM) design as the advanced liquid metal reactor-(ALMR) cesign - to sponsor for NRC design certification. The conceptual design for PRISM was l developed by General Electric (GE) Company-in conjunction with an industrial' i team of commercial engineering firms. Research and development support is-being supplied by the Argonne National Laboratory, Energy _ Technology l Engineering Center, Hanford Engineering Development Laboratory, and Oak Ridge ' National Laboratory. -In addition, a steering group of utility representatives l was involved in the-PRISM design-effort. l DOE chose to spor.sor the PRISM design as part of its National Energy Strategy because of the design's potential for enhanced safety through the use of- ?- passive safety systems-and greater safety margins, reduced cost through. modular design and construction, and possible future development of an actinide recycling capability. -Although this:last alternative has not yet i been proposed.in the current application, DOE has supported studies evaluating i the use of. actinides separated from-spent fuel in.an advanced liquid-metal-reactor (ALMR) fast-flux core. The PRISM design has considered liquid: metal reactor :(LMR) experience to date developed both nationally and-internationally in terms of systems and 4 components design, reliability-data, and-safety assessments. This experience-consists of operation of a number of facilities such as,_EBR-II,-Phenix, the- - Fast Flux Test Facility (FFTF), the Joyo. reactor in' Japan, and others.: I T% PRISM Preliminary Safety-Information~ Document (PSID) was ' submitted to' the l' 'M for review in November 1986, and the results of an early NRC. staff review 3 was'the draft PSER (NUREG-1368) issued in' September 1989. 'In order to obtain NRC approval of its planned. prototype, DOELplans to apply for preliminary-design approval--in 1995. - The DOE also plans to apply for standard design L certification in 2003 after a prototype demonstration. -These plans are based on the' current DOE goals to demonstrate the commercial potential-for:the ALMR-i by 2010, as called for in the Energy-Policy Act of 1992, plant Description F The PRISM plant design consists of three se>arate power blocks-each made up of -three reactor modules. Each module has a t termal. output of 471-MWt-and. an electric output of 155 MWe for a total (plant)' output of 1395 MWe. The PRISM design contains-three: turbines, each supplied.from a power block., Options:for 1 ( one or two power blocks are possible. PRISM operates at much higher { temperatures:than current LWRs which will require a-rigorous evaluation of-the. 4 = effects of creep _and creep rupture on reactor vessel and systems. The: PRISM design also relies on a highly automated and complex control: system utuizing digital processing. - -- 8-- - DRAF.T. . J, -..... ~ m . 4 ,-___m am,,_._,J-,m._..m.,m,,_m.m-_mm.,...__,.... i m wm.m L

j z.- The reactor module consists of-the containment system, the reactor vessel, the core, and the reactor's internal components. - The reactor vessel encloses and i supports the core, the primary sodium coolant system, the intermediate coolant i system heat exchangers (IHXs), and other internal components. The vessel is l located just inside the containment vessel, whichtis located below grade in: i the reactor silo. The reactor vessel is penetrated onlysin the closure head.- L The head is supported by the floor structure, and the floor structure is supported by seismic isolator bearings to reduce horizontal movement ~ during-l seismic events. The upper head of the reactor vessel is the closure head. i The closure head also. supports the intermediate heat-exchangers (IHX) and the electromagn. etic (EM) pumps. 1he main components of the Nuclear Steam Supply System (NSSS) in PRISM are the i reactor module, primary sodium loop, EM pumpr, -lHX, intermediate sodium loop 1 and steam generators (SG. The primary sodium loop is con _tained completely. within the reactor VM se, which is hermetically sealed to prevent leakage of n the >rimary coolant. The EM pumps provide the primary sodium circulation. l Synetronous machir,es provide flow coastdown capabil.ity to-the EM pumps. Flow coastdown is very important for preventing sodium boiling during a loss of EM i pump power without reactor scram Reactor-generated heat in=the primary. loop L is transferred through the IHX to the: intermediate heat transfer system (IHTS). IHTS sodium is circulated by a centrifugal pump. The IHTS operates at a higher pressure than-the primary loop so that,:in case of a tube rupture i in the IHX,..the = sodium would not flow out of the reactor vessel. A pressure of approximately 15 psig-is used to assure a minimum 10 psi positive pressure l differential across the IHX from the IHTS to the:PHTS is maintained. A' sodium-water reaction protection system mitigates the effects of reactions i betweenLIHTS sodium and water.in the SG. 3 The reference fuel for the ALMR is a uranium-plutonium-zirconium'(U-Pu-Zr). j. alloy. The ferritic alloy HT9-is used for cladding and channels to minimize i swelling caused by high burnups. The PRISM. core is a heterogeneous l: arrangement of driver fuel-and blankets. [ The PRISM core design is such that the net power reactivity feedback is i negative in all ranges uf operation,1in all. transients,1and in all' accident:- not involving voiding. For certain.very low probability accident scenarios involving sodium boiling, a i-positive feedback can' occur. positive. void coefficient dominates and a net In all other situations without extensive:.- voiding,;an increase in temperatures produces negative feedbacks from Doppler L and thermal expansion of the: core and related' structures that dominates the i positive moderator density coefficient. The: net negative temperature coefficient is so large that-analyses predict-all non-boiling transients and accidents to be terminated by the temperature. feedback reactivity at-temperatures low enough to not threaten fuel: or! vessel 1 integrity. ' This-L passive shutdown function. allows the reactor. to sustain-all: non-boiling - l - transient scenarios _without damage, even with a failure to scram, f 1:

- DRAFT i

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There are six control rods in the main reactivity control and shutdown system. Inserting any one of the six will shut the core down. The control rods can be inserted using (1) the plant control system (PCS) for normal insertion, (2) the safety-grade reactor protection system (RPS) for rapid insertion, and (3) gravity drop into the core. If both the normal and safety-grade systems fail, the operator can activate the ultimate shutdown system (USS) which sends boron balls into the central location of the core causing shutdown independ-ently of the control roris. The PRISM design also includes passive mechanisms for controlling reactivity: three gas expansion modules (GEMS) consisting of tubes, closed at the top and open at the bottom, and filled with helium. If the pumps are running, the static pressure is high, causing the sodium level to rise to a high point in the GEM. However, with the pumps off, the static pressure and sodium level drop, which increases neutron leakage. The reac-tivity change provided by the GEMS between these two states is about -70 cents. Normal shutdown cooling is achieved with the non-safety-grade condenser. If the condenser becomes unavailable, the safety-grade reactor vessel auxiliary cooling system (RVACS) is used for RHR. The RVACS provides natural circulation air cooling of the containment vessel. The design-basis RVACS event assumes that the normal and auxiliary heat removal systems, as well as the Intermediate Heat Transport System (IHTS) sodium, are lost immediately following reactor and primary EM pump trips. The preapplicants' analysis has shown that the RVACS heat removal rate is sufficient to maintain fuel temperatures within acceptable limits, and temperatures of the internal structures within the reactor vessel under American Society of Mer.hanical Engineers (ASME) level C conditions. The PRISM design also contains the non-safety-grade auxiliary cooling system (ACS) to assist the RVACS. The ACS uses natural circulation within the steam generator (SG) to remove heat indirectly from the reactor vessel, and natur:0 circulation air cooling of the SG, with heat rejection directly to the atmosphere. The ACS can be used in combination with the RVACS to reduce the cooldown time. Some of the inherent safety characteristics of the PRISM design with respect to RHR are: (I) the favorable combination of viscosity, thermal conductivity, and vapor pressure associated with the use of sodium to remove heat, (?) the ability to operate at essentially ambient pressure, thus reducing the pressure exerted on the coolant system boundaries, and (3) operation far below the sodium boiling temperature, thus obtaining the operational and analytical simplicity associated with a single phase coolant. 1 DRAFT

4 SECY-86-368, "NRC Activities Related to the Commission's Policy on the Regulation of_ Advanced Nuclear Power Plants," Decem'xt 10, 1986 4 SECY-89-350, " Canadian CAND'U 3 Design Certification," November 21, i 1989 .SECY-90-055, ' PIUS Design Review," February 20, 1990 t NUREG-1338, ' Draft Preapplication SER for the MHTGR" i NUREG-1368, " Draft Preapplication SER for PRISM" i NUREG/CR-5261, " Safety Evaluation of NHTGR Licensing Basis Accident Scenarios" NUREG/CR-5364, " Summary of Advanced LMR Evaluations-PRISM and SAFR" NUREG/CR-5514, "Modeling and Performance of the MHTGR Reactor Cavity l Cooling System" NUREG/CR-5647, " Fission Product Plateout in the MHTGR Primary System" NUREG/CR-5815. " Evaluation of 1990 PRISM Design Revisions

• i 4

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, Enclosure 2 p** "*% i POLICY ISSUE November 23, 1992 (Inf0TmatlOn) SECY-92-393 EDI: The Commissioners From: James M. Taylor Executive Director for Operations Subiect: UPDATED PLANS AND SCHEDULES FOR THE PREAPPLICATION REVIEWS OF THE ADVANCED REACTOR (MHTGR, PRISH, AND PIUS) AND CANDU 3 DESIGNS Puroose: To inform the Commission of the staff's current plans and schedules for conducting preapplication reviews of the advanced reactor (MHTGR, PRISM, and PIUS) and CANDU 3 designs. Backaround: In SECY-91-161,

• Schedules for the Advanced Reactor Reviews and Regulatory Guidance Revisions," the staff informed the Commission of the following estimates for completion of the preapplication reviews:

P3154 November 1992 MITGR December 1992 CANDU 3 June 1993 PIUS July 1993 The staff based these dates on broad planning assumptiuns including the preapplicants' design certification appli-cation schedules, availability of the Office of Nuclear Reactor Regulation (NRR) resources to conduct the reviews, timely receipt of information to support the reviews, and a scope of review consistent with the previously issued draft preapplication safety evaluation reports (PSERs) for the PRISM and MHTGR designs. Contacts: NOTE: TO BE MADE PUBLICLY AVAILABLE Thomas H. Cox, NRR 504-1109 IN 10 WORKING DAYS FROM THE DATE OF THIS PAPER Brian W. Sheron, RES 492-3500 ~

-.. - ~ i s e i h' The Comissioners, p i Subsequent to issuing the estimates in SECY-91-161,.a number = i of factors have necessitated a revision to the schedules for ? PRISM, MHTGR, CANDU 3, and PIUS preapplication reviews.- Some preapplicants have extended their design

• certification-i application schedules or modified their pro >osed designs.

j-Most staff technical review resources have >een redirected l to higher priority operating reactor and design certifi-l cation reviews. -- Additionally, implementation of the Fee Recovery Rule has resulted in preapplicants desiring a [ preapplication review scope limited to key l certifi ntion/ licensing issues. l In February 1992,- the staff issued letters to all four 1 preapplicants requesting confirmation of their plans for design' certification and, in some cases, a schedule for i submitting additional preapplication review information. All responses were received by May 1992. l During the April 21, 1992, briefing on advanced reactor reviews, the staff advised the Comission that due to other higher priority work, a relatively small~ number of staff-would be dedicated to conduct the prea> plication reviews. The staff proposed _to concentrate on tiose key policy issues requiring Comission guidance and revise the preapplication review content and schedule to more effectively use the-reduced NRR technical review resources. In June and July 1992, the staff held public meetings with each preapplicant to discuss the relevant scheduling infor-mation, the desired scope of preapplication review,-and schedules for additional submittals. The staff also informed the preapplicants of key policy issues that the L staff is planning to forward to the Comission for guidance. During these meetings, the preapplicants and the. staff agreed on-a smaller, more focused scope for conducting the i preapplication reviews. Previous planning assumed that the l scope of all preapplication reviews would be similar to the. scope of the PRISM and MHTGR preapplication reviews i-documented as NUREG-1368, " Draft Preapplication. Safety - - Evaluation for Power Reactor Inherently Safe Module Liquid Metal' Reactor," and NUREG-1338, " Draft Preapplication Safety- ' Evaluation for the Modular High-Temperature Gas-Cooled Reactor." Discussion: The staff considered-several factors in developing a revised I schedule for conducting the preapplication reviews. The j enclosure to this paper-provides a design-specific sumary 4 .. ~..,, -, -, - *,.,,,,, = -,, - y,.m5 ,..,.,,-,,.y y,,w ,~v,y - n .w., .,,.,-w..,m,wr-- v-q - y--,+ r, y -y 4 -,-em

I' The Comissioners ~ of this information and the staff's rationale for sequencing the reviews.- Based on this rationale and consideration of available resources,.the staff has identified the following revi m v timates for completing the preapplication reviews: PRISM December 1993 CANDU 3 December 1994 PIUS April 1995

• MHTGR December 1995 The staff expects to follow the same review process for approval of the PSERs as:is being used for the. safety evaluation reports:on.the evolutionary light-water reactor designs and the Electric Power _ Research: Institute Utilities Requirements Documents.' Approximately.6 months before:

completing the review,- the staff will: submit a draft final-PSER to the Comission. -With Comission consent, the staff will forward the draft final; PSER to the preapplicant, the Advisory Comittee-on Reactor Safeguards _ (ACRS), and the NRC Public Document Room. After considering input-during public meetings with ACRS and the preapplicant,.a final PSER will be forwarded to.the Commission for' approval. -The staff intends to conduct most of the preapplication reviews with staff from the NRR Associate Directorate for Advanced Reactors and License Renewal-(ADAR), national ' laboratory technical assistance, and support from the Office l of' Nuclear Regulatory.Research'(RES). Due to. limited staff-resources each design-PSER will be developed in-a sequential-order. NRR technical staff within the Associate Directorate for-Technical Assessaert (ADT) will.ein gear 4i, t.ut participate:in the preapplication-review. -ADT technical staff resources are currently required for higher. priority operating reactor technical reviews and light-water reactor - (LWR)-design certification. However, once the draft final PSER has been developed,- ADT: management will-review the. report for its policy implications. The staff considers this approach appropriate-since the. preapplication review -considers.the conceptual l design, and final technical decisions-on safety will not be made'until. the design certification review when 'the ADT technical staff will be. L involved. L L The staff believes-that the changes to the preapplication l NRR technical revie.hedule, and the approach for conducting' review scope and.sc w, provide the most effective use of NRC- . resources.---The proposed schedules will allow the staff to provide a timely response to the_preapplicants in important. areas-regarding their. design certification application o L plans. By emphasizing the key. policy. issues for the: L -.--.-,,,...-....,..c

The Comissioners '- advanced reactor designs, the staff will address the preapplicants' most significant questions about NRC's licensing requirements. Resolution of these issues in the preapplication reviews is expected to allow the preapplicants to reduce the current uncertainty regarding design and design certification schedules. The staff will continue its assessment of the schedular and resource implications of the reviews for these advanced designs. NRC preapplication review schedules may be altered to support the recent Energy Policy Act of 1992 goals. The status of the preapplicants' plans and the staff's reviews will be provided to the Comission as appropriate. The staff will, within the next few weeks, submit to the Comission a draft paper on key policy issues affecting the advanced reactor and CANDU 3 designs. Comission guidance on these issues could significantly affect the preappli-cants' planning milestones. These issues include proposals by the preapplicants for significant departures from existing regulations and regulatory guidance.. Role: There has been Congressional interest in this matter and the Chairman previously advised Senator Johnston and Congresswoman Lloyd of the schedules outlined in 1 SECY-91-161. Therefore, the staff plans to submit this paper to the appropriate subcomittees, the Office of Management and Budget, and the Department of Energy. _ / mes or xecutive irector for Operations

Enclosure:

Sumary of Input for i Schedule Revision l DISTRIBUTION: Commissioners OGC OCAA OIG OCA OPA ~ OPP EDO ACRS l ASLBP l SECY

.. o [MM In March 1989, the NRC issued NUREG-1368, " Draft Preapplication Safety Evaluation Report for Power Reactor Inherently Safe Module Liquid Metal Reactor," in which it summarized the results of its review of the Preliminary Safety Information Document (PSID) submitted in 1986. In March 1990, the U.S. Department of Energy (DOE) submitted Appendix G to the PSID, " Responses to ) Issues in Draft SER," in which it proposed several significant changes to the 4 PRISM design. These changes included increasing the power, adding an ultimate shutdown system and containment dome, redesigning the reactor to add gas expansion modules (GEMS), and changing to a single-wall-tube helical coil steam generator. Brookhaven National Laboratory (BNL) reviewed the revised design and published its findings in NUREG/CR-5815 " Evaluations of 1990 PRISM Design Revisions." BNL is also reviewing the performance of the GEMS and the consequences of a hypothetical core disruptive accident. The Office of-Nuclear Reactor Regulation (NRR) staff is continuing to review the preap-1 plicant's submittals and is writing the final preapplication safety evaluation report (PSER). The Office of Nuclear Regulatory Research (RES) performed the early part of NRC's review and continues to support NRR's work with projects to provide formal documentation for reference in the PSER, to update code validation, to investigate behavior of the new metal fuel, to assess reac-tivity feedback, and to prepare for source term determination. -In a letter of March 12, 1992, DOE submitted the following schedule: Preliminary Design Approval Application CY 1995 = Prototype Final Safety Assessment Report 1997 = Standard Design Certification Application 2003 (After prototype testing) This schedule appears consistent with the Energy Policy Act of 1992 which established DOE goals for the advanced liquid-metal reactor program to submit a preliminary design approval application to the NRC by September 30, 1996, and to make a decision on prototype des nstration by M tember 30, 1998. In June 1992, the National Reseanch Council of the National Academy of Sciences published a report, " Nuclear Power: Technical and Institutional Options for the Future," in which-it discussed prerequisites needed to preserve the U.S. nuclear power option and recommended that the Federal government support key reactor designs. The Council recommended that the PRISM design be the only preapplication design to receive government funding because of its unique ability as a breeder reactor. At a public meeting nn July 1,1992, the NRC staff notified DOE that it would need to delay issuing the PSER beyond the originally scheduled date of November 1992. DOE noted that it had submitted all requested information to the NRC to support the preapplication review and requested that the NRC

omplete the PSER as soon as possible to support DOE in planning for design certification. DOE recently notified the NRC of problems found with the o

o reference fuel during testing at the Argonne National Laboratory. DOE indicated it may need to redesign the fuel. The staff does not know how this decision will impact the PRISM design certification schedule. The staff plans to conduct the PRISM preapplication review first to capitalize on the work already completed and currently in )rogress. DOE has provided all necessary submittals to support the review and 1as been responsive to staff questions during the review. DOE's plans for preliminary design anproval application in CY 1995 are supported by the National Research Council's recom-mendation of PRISM as highest priority for DOE support of the four designs in preapplication review. The staff intends to treat the PRISM fuel problem as an open issue in the PSER. MHTGR In March 1989, the NRC issued NUREG-1338, " Draft Preapplication Safety Evaluation Report for the Modular High-Temperature Gas-Cooled Reactor," in which it summarized its review of the PSID submitted by DOE in 1986 and 10 subsequent amendments. Responding to the draft PSER, DOE /GA submitted Amendments 11, 12, and 13 to the NRC. These recent amendments provide additional information for NRC's review of the originally proposed design. NRR is reviewing the submittals on the fuel design and fission product transport analyses at the DOE laboratories. RES performed the early part of NRC's review and continues to support NRR's work with projects to provide formal documentation for reference in the PSER, to evaluate DOE's containment design alterratives, to investigate moisture ingress events, to assess data base adequacies, and to prepare for source term determination. In its May 18, 1992, response to the NRC letter of February 18, 1992, DOE stated that it would not establish the schedule for the MHTGR design cert-ification until August 1993 when it expected to select the technology for the DOE New Production Reactor (NPR). In September 1992, DOE and the U.S. Department of Defense agreed to defer the NPR program and close the design efforts. The MHTGR schedule depends primarily on the gas-cooled NPR program which technically supports much of the MHTGR design. DOE wants the final PSER issued by April 1993 t1 support resolution of the key MHTGR policy issues identified in NUREG-1338. DOE asserted that its submittals include all the information needed by the NRC to complete its review for the final PSER. DOE believes that the final PSER is needed in 1993 for the nuclear power industry to understand that the MHTGR is a viable power reactor concept. The Naticnal Research Council report recommended that the commercial MHTGR program be given a low priority for DOE funding because its U.S. market potential was judged to be low. However, the Energy Policy Act of 1992 estab?ished DOE goals for the MHTGR program to submit a preliminary design approval application to the NRC by September 30, 1996, and to make a decision on prototype demonstration by September 30, 1998. These new goals may result in a DOE schedule that would require earlier preapplication review of the MHTGR design and resequencing of other design reviews. . s

i.p,c o, l/k DOE has discussed plans to revise the MHTGR design to increase the power of the modules from the current-350 MWt to 450 MWt at the preliminary design 4 approval stage of design. certification. -It stated -that this power increase - will.not affect the key policy issuestfor the design. DOE recently informed- -) NRR of problems in testing the reference MHTGR fuel. ' The preliminary failure - rate for the latest test of the MHTGR fuel-design <is significantly higher than that needed to meet the MHTGR design criteria. DOE expects to complete the-post-irradiation examinations in May 1993 at the earliest. l l The staff plans to conduct-the MHTGR review as the final preapplication review L in the series of four projects, because of the uncertainties 11n the DOE schedule and design to be proposed for design certification application. - When DOE submits MHTGR design co tification schedules the staff will reconsider the j preapplication review plans. DOE is most interested in obtaining' feedback on the implementation of the key policy issues.for the MHTGR design.- Continued i emphasis by the staff in obtaining Commission guidance for resolution of the key policy issues will provide DOE-valuable feedback on their proposed l approach for the MHTGR design in advance of the final.PSER. CANDU 3-j' On May 25, 1989 Atomic Tr.srgy of Canada, Limited,' Technologies (AECLT) F informed the NRC-of its intent to submit the CANDU 3' reactor design for L - standard design certification.. AECLT, a wholly _ owned U.S.x subsidiary of l Atomic Energy of Canada,-Limited, (AECL);in Canada. has: supported the CANDU 3 preapplication review by submitting a Technical Description,-- Conceptual Safety i Report, Conceptual.Probabilistic Safety Assessment, and several technology transfer reports describing the CANDU design. + In a letter of March 18,1992,- AECLT informed the NRC that'it could support a standard design certification application in~1995 or.1996 if' the NRC completed its preapplication review of the CANDU 3 by June-1993. 'On June 29, 1992, AECLT gave the staff a schedule of submittals to support the preapplication review. AECL has completed much.of the-final-design work for the CANDU 3 reactor and is negotiating to start construction-in a Canadian province which - could serve as= a prototype for the CANDU 3 design' certification in the U.S. In' September 1992,- AECL acknowledged that-it would re-evaluate lits design certification plans -in the U.S. if Canadian construction plans did not materi- . alize. L The National Research Council report identified theLCANDU 3 design as a mature design that could be licensed this century. The report noted that the licensing process could:be lengthy because of the differen'ce in regulatory requirements between the_U.S. and Canada. The Council did not find sufficient advantages with.the design to justify DOE _ support for design certification. The-staff-has.s' tarted some preapplication review on the CANDU 3: design. ; NRR L is conducting two projects at-DOE laboratories: a study of thr NDU 3: I positive void reactivity coefficient-and a review of the operat + r of the- - on-line refueling machine. RES has completed a systems study tc .entify - candidate event sequences-for required safety analysis, and it has projects to s - assess data base adequacies, to perform preliminary transient calculations 3-r,-- 4 --vr W - n w-wh - _m,-wf. .v m .r ,,,c u -y _m_.r. m.,rs., ,,.,r_...r-.yem,q-r,,

be 4 i [g 1 h* using Canadian codes, to identify code needs for future independent analyses, .to initiate severe accident analyses with NRC codes, and to prepare for source i term determination. RES.will also provide in-house analytical capabilities L for itself and NRR for the CANDU 3 design, i i To better understand the CANDU 3 containment performance and radiological i releases, NRR is reviewing-the consequences-of a large break loss-of-coolant accident (LOCA) with a-failure to shut down. NRR is performing this work to i support the Commission's decision on a key policy issue: the acceptability-of. a design with a dominant positive void coefficient. The preapplicant has not j performed this analysis for CANDU 3, and has supplied little directly relevant 3 l information on the event and its consequences. j-AECLT is having problems getting.propriett y information released from Canada i to the U.S. This has delayed the staff in obtaining Canadian codes thus interrupting RES's work to use these codes for preliminary calculations. Code-1 l work is now on the critical path for completing the preapplication review, and i-the lack of timely submittals of other proprietary infomation could further delay the review schedule. In a lettsr dated September 23, 1992, the staff informed AECLT that an inability to transfer proprietary material to the U.S. i i may affect the proposed CANDU 3 preapplication review schedule. - AECLT is now - pursuing transfer of-proprietary material directly from AECL to the NRC. The staff plans to conduct the CANDU 3 review as:the second preapplication j review because the design and experimental. data base are already. sufficiently developed to su) port the review. The June-1994_PSER issuance assumes: prompt resolution of tie present problems releasing proprietary information required for the review from Canada to the. United' States. i ele In October 1989, Asea Brown Boveri (ABB) Atom requested that the NRC perform a licensability review of its Process-Inherent Ultimate Safety (PIUS) plant design. ABB Combustion Engineering (ABB/CE) of Windsor,-Connecticut,-.is.the i direct representative of ABB Atom-in the-US., and is the official preappli-- I cant of record. )e In May 1990, ABB/CE submitted.a five volume, Preliminary Safety Information - i-Document-(PSID) to support its request for a preapplication review.' NRR has i started a project with BNL to support core neutronics modeling. 'RES has-F completed a' systems-study to identify candidate event sequences for required safety analysis, and it has projects to assess: data base adequacies.cto perform preliminary transient calculations using the existing TRAC code, to identify code needs for future independent: analyses, to-initiate' severe-accident analyses with NRC codes, and to prepare for' source term determi-f nation. RES will also provide in-house analytical capabilities for itself and- ] NkR for:the PIUS' design. In a letter of April 22, 1992! ABB-stated that it would' submit a-design certification application in 1994 or'1995 if- (1) the NRC issues.a preap - plicdionisafety evaluation report (PSER))by April'1994 that does not require g 4-p [e [,,,,,,_,,-_,...,..m,,, ....-,mf,_,,,,..%-m,, ,.m ,y ,.-.,a.-.,~ m .r.s

9' significant design changes to the PIUS design, and (2) the commercial environ-4 ment at that time is favorable to that decision. ABB is negotiating with the Italian state utility to. support testing of the PIUS design and will give the NRC details of its overall test plan when the basic negotiations are complete. During an August 6, 1992, meeting, ABB informed the staff of a proposed change to the PIUS design. The design change involves adding four " scram valves" and associated piping. The feed lines to these valves take suction from the borated reactor pool water, and the valves discharge to the suction of each of the four reactor coolant pumps. Activating the valves is expected to result in a rapid and uniform insertion of boron by a means redundant and diverse -from the passive scram process. The passive scram through the density locks will still be the ultimate shutdown process. ABB plans to submit the design change in a November 1992 supplement to the PSID. ABB also plans to submit a PRA supplement in early 1993 for the PIUS design. The National Research Council report concluded that the PIUS design would not likely be ready for commercial o>eration within the next 20 years and had a low priority for DOE support. Tie lack of operation and regulatory experience is expected to delay acceptance by utilities of this advanced LWR design. The staff plans to conduct the PIUS review as the third preapplication review because the design is presently at the conceptual stage and the experimental data base for the design is still being developed. ABB is most interested in obtaining feedback on the implementation of the key policy issues for the PIUS design. Continc2d emphasis by the staff in obtaining Commission guidance for resolution of these issues will provide ABB feedback in advance of the final PSER. Conducting the PIUS review third will allow ABB time to develop the design more fully and respond to staff questions without impacting the preap-plication review schedule. e, +, ,er v ,}}