ML20126F689

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

ML20126F689
Person / Time
Issue date:	12/16/1992
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Grossman N ENERGY, DEPT. OF
References
PROJECT-674A NUDOCS 9212310064
Download: ML20126F689 (59)

Text

{{#Wiki_filter:_ A ! m mac k o,, UNITED STATES , j NUCLEAR REGULATORY COMMISSION n 0* t WASHINGTON, D. C. 20$55 ~ 'g *...,o December 16, 1992 Project No. 674 i Mr. Nicholas Grossman, Director Division of LMRs j Office of Advanced Reactor Programs I* U.S. Department of Energy NE 452 Washington, DC 20545

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 aapers have been provided to the Commission: (1) a draft paper providing tie 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. The proposed schedule paper reflects the staff's assessment of its resources and the needs of the preapplicants.: The staff will; continue to try to expedite its reviews and complete the work ahead of-schedule. l i g.100 l I RETUM TO REGU3 TORY CITlA. OS A f W 4 79 r t -s - " " " ~ I [ 9212310064 921216 PDR -PROJ l 674A-PDR. /1 j i< l

6 . g Mr. Nicholas Grossman-December.16, 1992 ~ The proposed positions on policy issues have' not _ been reviewed by the Commission, and, therefore, do not represent agency _ positions. Your comments concerning these issues should be sent to the project manager, Stephen Sands. Sincerely,- Original signed by: -Dennis M. Crutchfield, Associate Director . for Advanced Reactors and-License Renewal Office-of Nuclear Reactor Regulation-

Enclosures:

1. Draft Commission Paper 2. SECY-92-393 cc w/ enclosures: See next page l Distribution:

Central File NRC POR i

PDAR R/F DMCrutchfield i-WDTravers MMSlosson i THCox i-EDThrom l SPSands 0GC 15/B/18 L ACRS (10) P 315 i-OPA i LLuther PIUS R/F i h D: ADAR .A Rt1 L-LA:PDLR:ADAR PM:PDAR. DAR SC:PDAR: ADAR -SE:PDAR:ADAR-LLuther SPSands:sa- -THCox % . EDThrom7 t- - HM on DMCrutchfield' l .JJp//9) 4 /r//4/92- <2/ty/92 k2 /ty/92 -{'h l 92 ja_//g/92 0FFICIAL ECORD COPY j - Document Name:- PRISSECY.LTR i. b ] 1.

)

e i j . Mr. Nicholas' Grossman - December 16.-1992

s-

' PRISM-. Project No. 674-o cc: Robert Berglund Dr. John Herczeg Advance Nuclear Technology.-

Division of LMRs General Electric. Company Office of Advanced Reactor Programs-4 6835 Via Del Oro

-U.S. Department of Energy P.O. Box 530954: NE 452. San Jo'se, California-. 95153-5354-Washington, DC -20545 4 L.N.- Salerno, Manager ALMR Program Control Dr.-Lee ~Schroeder Advance Nuclear-Technology Assistant Director for Physical" General Electric Company __ Sciences and Engineering-68SS Via Del Oro-Office of Science and Technology. Policy P.O. Box 530954 Executive Office of the President. San Jose. California _- 95153-5354. 'Old Executive Office Building - Room.432 1/2 -Dr..Geza Gyorey:- Washington, DC ;20506. Advance. Nuclear Technology 1 11 General Electric Company - 6835 Via Del.Oro P.O. Box 530954 San Jose, California-.95153-5354' Richard Hardy Advance _ Nuclear Technology General-. Electric company-6835 Via Del: Oro. P.O. Box 530954 San Jose, California ;95153-5354- [ Dr. Sol Rosen, Director -Office of Advanced Reactor Programs U.S. Department of--Energy. -NE'45 Washington, DC 20545-Steve Goldberg, Budget Examiner Office of Management and Budget- -725 17th Street, NW-Washington, DC 20503__ f 4-l ~ ~

i p i } V 1. DRAFT-i e i 4 i l Lor: The Comissioners F from: James M. Taylor s Executive Director'for Operations e }' Subiect: ISSUES PERTAINING TO THE ADVANCED REACTOR (PRISM, MHTGR, AND-PIUS) AND CANDU 3 DESIGNS AND:THEIR. RELATIONSHIP TO CURRENT. 3 i REGULATORY REQUIREMENTS' ) Puroose: 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 j reactor and CANDU 3 designs.: i Backaround: The Advanced Reactor Policy Statement-(51 FR 24643) and [ NUREG-1226, Development and Utilization of the NRC Policy i Statement on the; Regulation of Advanced Nuclear Power Plants," define advanced reactors as those with innovative designs for which licensir.g: requirements will be signif-icantly different from the existing. light-water reactor (LWR) requirements. These documents-also provide guidance for the development of new re - support the advanced designs.gulatory requirements to .1 Staff reviews of these j advanced reactor designs-should utilize existing-regulations - to the maximum extent practicable. - When new requirements are necessary,_the staff should move towards-performance standard regulations and--away. from prescriptive regulations. Each designer is encouraged to propose new criteria and 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 1 assessment of-the potential of--that design to-meet those criteria. CONTACTS:- -M.M. Slosson = =504-1111 ~ b1 l .aw,,- ,e,,m-- rw.m++ w....,s ,e r-, ,.as->,e--,,-,,o-..,,,..,, ,,,,,,,,e- . -wen.,.,wwa,e-,,-,e .,,avwr.w,.,,m ,,v.,e

m., y ^ - = F -.- f41 l.' The Comissioners y i V -l The staff is conducting preapplication. reviews of the l following four designs: 3 l General Atomics (GA) 350-Wt Modular High Temperature Gas-Cooled Reactor-(MHTGR) design sponsored by the'U.S. !L Department of Energy (DOE) Gas Cooled Reactor Program 1 l General Electric (GE) 471-MWt Power Reactor Innovative Small Module (PRISM) reactor # sign sponsored by the DOE j Advanced Liquid Metal: Reactor (ALMR) Program i Atomic Energy of Canada, Limited. Technologies (AECLT) 1378-MWtCanadianDeuteriumNatural-Uranium (CANDU3) t reactor design L Asea Brown Boveri-Combustion Engineering (ABB-CE) 2000-MWt Process inherent Ultimate Safety (PIUS) reactor 1 i design-provides a listing of-pertinent Comission i papers and reference NUREG documsnts for these preap-plication designs.- Some information in the original l documents may be superseded by more recent preapplicant submittals. A sumary of the current' designs is provided as L. l L In response to Comission: staff 1 requirements memorandum (SRMs)- in SECY-91-202,' " Departures From Current Regulatory j L Requirements in Conducting Advanced Reactor Reviews," the staff comitted to identify. issues during the preapp11 cation L review that. require Comission policy guidance or staff i l' technical resolution for design certification, including ) situations in-which advanced reactor designs significantly deviate from current regulatory requirements. I Policy issues for evolutionary and passive LWRs have been [ identified in the following Commission papers:: l SECY-90-016,:" Evolutionary Light Water Reactor (LWR) . Certification Issues and Their: Relationship to Current p Regulatory Requirements"' L Draft SECY (distributed for cumments on February 27, 1992), " Issues: Pertaining to Evolutionary and Passive Light Water Reactors and Their Relationship to Current; Regulatory Requirements"- m,. .~a~ a.,,, n n %.,, n+i,4-,,, ni-,--,,-, .,, - einA:..,,., ,,. ' ~,. - 4 %,.,,,L., w ~

i .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 preapplicants identified how their design complied with the current LWR licensing requirements and, where it did not, provided alternative 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 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 recomendations for Comission approval, are provided in Enclosure 1. The recomendations for Comission approval were develosed by the staff with inputs from the preapplicants, the pualic, and the ACRS. The staff considered the preapplients' 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 recomendation to be applied to all applicable advanced reactor designs. In some instances, the staff recomends that current ~ regulations continue to be applied to the advanced reactor designs despite preapplicant proposals to do otherwise. Where deviations are recomended, 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 guidance at design certification. Some issues are closely related. Accident evaluation and source term provide a basis for containment performance and emergency planning. Approaches taken for residual heat removal and ' reactivity. control are intended to be. consistent with the accident evaluation categories and consequences.

r p i N: {.t

J The Comissioners 4-f A

4 Thestaffproposestotreat'theMNTGR, PRISM lththepolicy [ and PIUS designs as advanced reactors in accordance w i 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. l Therefore, the staff has concluded that a prototype CANDU 3 is not required for design certification. This position is i: consistent with staff conclusions in SECY-89-350, ' Canadian i CANDU 3 Design Certification,"- and SEtY-90-133, ' Prototype l Requirement for CANDU 3 Design.' -The preapplicant..AECLT, has stated that a CANDU 3 reference plant-is a key element 4 in their plan for standard design certification.- If AFCLT - holds to that position,- the regulatory review and con-1 struction in Canada would lead:the NRC's design certi-i fication review. The staff believes that this regulatory l review and construction in Canada would greatly benefit our [ 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 maces no claim of 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 structure,~it does not conform to some current NRC regulations. The staff proposes to apply preapp11 cation review criteria to the CANDU 3 reactor that are consistent with ELWR review requirements. i The staff. intends to use the Comission's guidance 'on these i recomendations to. conduct preapp11 cation reviews of the I conceptual designs. Guidance for review of prototype i requirements for advanced reactors will follow SECY-91-074, ' Prototype Decisions for Advanced Reactor Designs.' Consistent with the requirements of Title 10 of the Code of-Federal' Regulations -(CFR):Section 52.47(b)(2). novel' safety i-features of the-advanced reactors and CANDU 3 will be required _to be demonstrated through analysis, test programs, i L experience, or a combination of these methods.-. Feedback 1 from the review process will be factored into recomended F revisions to the-policy guidance and recomendations for the development of licensing criteria and regulations will t 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 j-identified -in su) sequent Comission papers. i !~ 1 ---*.4y-+,3 s 1V7 y p7_,,.,y,- y-- ,.,,,3 i,,,,..-9.c,-,-w,,,.---w w w-w --+y y,.mi-y-.w,,,-umm,1v D' p's,wwir.w-u w,e uiyw e,vw wr e,dr ywwww i sa * %=rw w r*'w-w- .W "$ ~* uirW 8vD'N" 8$DD=R4'"#Nve'--DW

. i The Comissioners ' 4 l In an SRM dated May 8,1992, the Comission requested the staff-to prioritize the issues for Comission review. The staff recomends that the priority for review be consistent with the PSER issuance schedules and requests that direction be provided in sufficient time to allow the staff to incor-porate Comission decisions into the final PSERs. Since the PRISM design is scheduled as the first preapplication review, Comission attention is recuested on a highest priority for those items identifiet in the enclosure as applicable to the PRISM design.

== Conclusions:== The staff requests approval of, or alternate guidance on, these proposed positions to be taken in the preapplication 4 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 coments. i Recomendations: That the Comission Approve the staff recomendations in Enclosure 1 for conduct of the preapplication reviews. Approve of the staff's conclusion that, based on the losition that the CANDU 3 design is an evolutionary leavy-water design deriving from CANDU designs operating in Canada and elsewhere, a prototype CANDU 3 is not required for design certification. Note that positions which change as preapplication review experience is obtained will be comunicated to the Comission and that as the staff identifies new - issues it will inform the Comission, t 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 l information on the advanced reactor and CANDU 3 designs, I and the preliminary nature of the staff's preapplication f l l

i 6-The Commissioners J reviews, the staff does not-recommend proceeding with generic rulemaking on any of the policy issues 4 identified in this paper. The staff will consider generic rulemaking, as appropriate, as tie reviews progress and the staff gains greater confidence in the final design information. James M. Taylor Executive Director for Operations Enclostres: 1. Policy issue Analysis-4 2. Design Summaries 3. List - Reference Documents I k 1 e l i I r --m -~ ~ v ..-w-- + ow e e n a,

. = - = POLICY IS$UES 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: i 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. Emergency Planning X X X

• 9 E. Reactivity Control X

F. Operator Staffing X X X X G. Residual Heat Removal X X X H. Positive Void Reactivity X X i Category I. Control Room Design X X X X l 2 J. Safety Classification X \\ l Discussions of these issues are on the following pages, including a brief l summary of the issue, current LWR regulations, preapplicant positions, I discussion of staff considerations and a proposed recommendation for staff l action. The staff considered the preapplicant's proposals in light of L applicable Commission policy statements. . ORAFT l

At this preliminary review stage, the staff has limited the scope of the j 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 - i i and the staff's advanced reactor design recommendation was essentially the same, it was not repeated ~1n this paper. In those cases where the 4 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. i i 4 9 3 i I l i l i a 7 4 4 2-DRAFT w ~ n-w ---e,,, =-.,4., .9 .eg,, ya, ,w ,g.,m,,. e--,,,,,.wn,ru-w w e-vs' r' -r m r v= s e-v

} c l A.- ACCIDENT EVALUATION 1 4 Identify appropriate event categories, associated frequency ranges, and-evaluation criteria for events that will be used to assess the safety of the [ proposed designs. 1 2 Current Reaulations i i Genera 1' Design Criterion.(GDC) 4 requires the consideration of accidents in F the design basis. Also,,10 CFR 52.47 requires the consideration of conse-quences for both severe accidents (through the required probabilistic risk assessment) and design basis accidents (DBA) for: designs which differ signif-l icantly from evolutionary designs or uttlize passive or other innovative means ( to accomplish safety functions. l Preacolicants' Aceroach j All three advanced: reactor preapplicants proposed to analyze accidents signif-i icantly less probable than the present design basis range.and to assure. through'their design that these accidents had acceptable consequences limited to specific dose. levels-to the public. All chose to utilize the Environmental Protection Agency's (EPA) lower level Protective Action Guidelines-(PAG) of [ 1 rem whole body and 5 rem thyroid as their limits for a significant portion 2 - of their accident spectrum. The MHTGR accident guidelines invo level PAG dose limit for all sequences more probable than 5x10'pe the lower-per reactor-probable than-10,uidelines invoke the PAGs for accident sequences more year. The PIUS p accident-sequences more probable than' 10 PRISM guidelines: invoke the PAGs for per reactor-year. -The per reactor-year. The-PRISM-probable than 10',on guidelines also limit consequences from any-sequenct more accident evaluati per reactor-yar to the 10 CFR Part 100 dose limits.

[

Guidelines for onsite consequences and offsite consequences from operational transients for al1~ vendors are consistent with or more conservative than present LWR regulations as contained in 10 CFR Part 100. e 1 The CANDU 3 preapplicant, in their current safet'y analyses, has excluded analyses of-the consequences of events with trequencies of less.than 10 per. i year from the safety evaluation. Events which culd be excluded from i consideration, based on the LANDU'3 design characteristics and system [ reliabilities, would include anticipated transient withoutl scram (ATWS), unscrammed loss-of-coolant accidents -(LOCAs), delayed scram events, and other events which could affect reactivity insertion (for example, from control l-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. p . DRAFT i d _i l: _ m < L -- .,.... _ _. _...., ~., ....a. x;. A___,.__,. .L.. .L,, .. w

i Discussion 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 i this enclosure. As discussed in Section D, the staff is not ready to make a recomendation on whether the Comission should consider a reduction in the emergency planning requirements. The CANDU 3 approach which limits the scope-of severe accidents examined appears to be inconsistent with the provisions of 10 CFR 52.47. The accident evaluation scheme envisioned by the staff examines challenging sysnts to thi designs to provide information for a later decision on emergency planning requirements for advanced reactors and includes 4 -consideration of the potential consequences of severe accidents. Addi-tionally, for the multi-module designs (PRISM and MHTGR), the impact of 4 specific events on other reactor modules for the multi-module sites must be assessed. The staff's approach is intended to be structured conservatively so that 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 and reliability data for passive safety features, lack of final design information, unverified analytical tools used to predict plant response, limited supporting technology and research, limited construction and operating i experience, and incomplete quality control information on new fuel manufacturing processes. Later, during the design certification process, some 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 approcch 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 j from probabilistic risk assessments of the specific designs. Categories of events will be established based on expected frequency of l occurrence. The selected range of events will encompass events of a lower likelihood than traditional LWR design basis accidents. Consequence acce)tance limits for fore damage and onsite/offsite releases will be establis ted for each category to be consistent with Commission policy guidance with appropriate conservatisms factored in to account for uncertainties. Methodologies and evaluation assumptions will be developed for analyzing each category of events consistent with existing LWR practices, l ,, 4 - DRAFT l l

~ Source term determination will be performed as approved by the Commission-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. I 4 f i. p 5- - ORAFT-

8. SOURCE TERM-c hint 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 evaluated.

It is developed u. sing best estimate phenomenological models of the transport of the fission produc",s from the fuel through the reactor coolant system, through all holdup volumes ~and barriers, taking into account mitigation features, and finally, into the environs. Current Reaulations Aspendix'1 to 10 CFR Part 50 (ALARA), 10-CFR-Part 100-(Reactor Site Criteria, w1ich 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.- Prenoolicants'-Anoroach PRISM 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 PRISM fuel to-quantify the. fission product release-fractionstor 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 P1U5 designer ha's proposed using-a mechanistic LWR source term.. -Information has been provided in the Preliminary Safety -Information Document (PSID).for fission product concentrations in both li effluents. It is expected that PIUS ~ designers will quid cnd gaseous-adopt the results of the ongoing EPRI/NRC effort. to revise the TID-14844 source-term previously used for LWRs. - DRAFT 4

4 The CANDU 3 designer 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 from LWRs, a method-for calculating postulated radionuclide releases (source terms) needs to be developed. In a June 26, 1990, staff requirements memorandum'(SRM) related to SECY-90-016, the Commission requested the staff to submit a paper describing the status of efforts to develop'an updated source term that 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-1465, ' Accident Source Terms for Light-Water Nuclear Power Plants,' draft report for comment, June 1992). The differences between the LWR designs and the_MHTGR-and PRISM designs warrant a separate evaluation of source-terms. LThe CANDU 3 will also be different from LWR designs-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 i 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 seismil: criteria factors'are also' included in tie 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 compara against applicable safety criteria. . E.g. commendation Advanced reactor and CANDU 3-source terms should be based-upon mechanistic analyses, provided that: -T- -DRAFT i l q e ____.____________.__________._..__..___.___m_.

1. The performance of the reactor and fuel under normal and off-normal conditions is sufficiently well understood to permit a mechanistic analysis. Sufficient data should exist on the reactor and fuel performance through the research, development, ar,d testing programs to provide adequate confidence in the mechanistic approach. 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. 3. The events considered in the analyses to develop the set of scurce terms for each design are selected to bound credible severe accidsnts and design-dependent uncertainties. The design specific source terms for each accident category would constitute ~ one component for evaluating the acceptibility of the design. e-DRAFT 1

C. CONTAINMENT isnt Should advanced reactor designs be allowed to employ alternative approaches to traditional " essentially leak-tight' containment structures to provide for the control of fission product release to the environment? Current Reoulationi General Design Criterion (GDC) ht barrier against the uncontrolled release of 16 requires that LWR reactor containments provide an essentially leak-tig 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. Prescolicants' ADoroach The MITGR 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 stesm 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 environment, preventing building overpressure. passage of gases to the 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 da credit for plateout on the building walls. y building leak rate, and take PIUS, 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 tle ABWR and SBWR. Below g,'ade, 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 is 5 percent volume per day at the desi;n

• essure of(used in safety analyses) he approximately 30 psig. T structure is designed for a test-aceWante 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

1 1 1 i. i The PRI$N 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 4 steel dome. It differs from light-water reactor containments functionally in the following respects. The containment is specifically designed to mitigate the radioact've release consequences of severe events.. The PRISM containment i i volume is markedly smaller than is typical of LWR containments; there is little separation between the reactor vessel and the containment boundary; and i I no safety grade containment coolers or spray systems are provided. The entire i containment structure is located below grade within the reactor building. ..i Qiscussion i i Each of the advanced designs and CANDU 3 maintair an accident mitigation approach in which containment of fission products is a part. Two of the i j advanced reactor designt PRISM and MHTGR grade, providing protectio (n from external) hazards. place the reactor building below j Generall the advanced designsfocusmoreattentionthandoLWRsonprotectingtheylantbyproviding p passive means of reactor shutdown and decay heat removal j designers proposed less stringent containment design requ(DHR). As:a result, trements. The staff recognizes that reactor designs \\ cant departure from past practi withoutLtraditional containment i structures or systems, represent a signif LWRs, and that existing LWR containment structures have proven to be an j effective com>onent of our defense-in-depth approach to regulation. However, the Advanced Teactor 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 i related features or reduced administrative requirements.- New reactor designs that deviate from current practice need to be extensively reviewed to. assure a-L level of safety at least equivalent to that:of current generation LWRs is l l provided, and that uncertainties in the design and performance are taken into account. The staff believes that new reactor designs with limited o>erational I experience require a containment' system that provides a su)stantial-level of i accident mitigation for defense-in-depth against unforeseen events, including-core damage accidents. This requirement may not necessarily result in a-s 1 high-pressure, low-leakage structure that meets all of the current LWR ) requirements for containment,Lbut 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 i experience base in evaluating their performance. For evolutionary LWRs, the staff, in $ECY-go-016, proposed to use a conditional containment failure )robability-(CCFP) or deterministic containment performance goal to ensure a salance between accident t evolutionary LWR reviews, prevention and consequence mitigation. During the-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-t {0 - DRAFT; m. .. u. .._._.m_,_. .a

I. 1i. 1 in the analysis and evaluation of severe accidents which would lead to jr significant difficulty and uncertainty in assessing a CCFP. For this reason the staff recommends that the deterministic containment performance goal be, c j adopted for the advanced designs and the CANDU 3. The staff proposes to j postulate a core damage accident as a containment challenge event and require j that containment integrity is maintained for a period of approximately 24 hours after the onset of core damage. This approach is used because the i preliminary nature of the advanced desifins precludes a reliable assessment of j-the failure probability of accident mit'gation systems and, therefore, of containment failure probability. Further, the CCT' is grounded in a firm i understanding of LWR tafety. R and advanced reactor technoloflies and their. systems and accident progression. Intrinsic differences exist denU L approaches to the. JWe between accident prevention and n' tigation. A 1 quantitative level v.i understanding of new technologies and systems comparable to that of LWRs is not yet available. Thus, the use of a performance based i criterion rather than a quantitative-one appears to be more appropriate for j advanced reactor and CANDU 3 preapplication review fliven the current level of knowledge of advanced reactor and CANDU 3 risk and ' ts prevention / mitigation elements. Recommendation The staff proposes to utilize a standard based upon containment functional performance to evaluate the acceptability of proposed desifins rather than to i rely exclusively on prescriptive containment design criter< a. The staff l intends to approach this by comparing containment performance with the i accident evaluation criteria. J Containment unsigns must be adequate to meet the onsite and offsite radionuclide release limits-for the event categories to be developed as t 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 resu14s in no greater than the limiting containment leak rate used in evaluation of the event categories, i and structural stresses-are maintained within acce) table-limits (i.e., ASME Level C requirements-or equivalent) leases of-radioactivity.- t After t11s period, the-containment must prevent uncontrolled re 1 l 1 4+ II - DRAFT u .._.-i..__._-_,. _. -~,a;_-...._-,,,._ a., ,....a-m.

D. EMERGENCYPLANNING(EP) r lint should advanced reactors with passive design safety features be able to reduce emergency planning zones and requirements? Current Reoulations Although emergency plans are not necessary for the issuance of a design certification under 10 CFR Part 52, 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. Atthelowerprodecteddoses,protectiveactions should be considered. Atthehigherprojecteddoses,protectiveactionsare warranted. Preaeolicants' Accroach The proposed PRISH approach to emergency planning is significantly different from that of previous LWR applications, particularly in the area of offsite EP. A design objective of PRISM 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 radiat< on, 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 is 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 criticality 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) ORAFT

during accidents. The prea>plicant asserts that the public around the would be outside the area t1at needs to be sheltered or evacuated and, plant further, there is ample time to notify and move the public during an event. i With regard to PlVS, 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 that they intend to limit offsite doses to the PAGs. Discussion Theadvancedreactordesignershaveobjectivesofachievingverylow 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 thost of existing LWRs, are sufficiently safe that the EPZ radius can be reduced to the site boundary, and that detailed planning and exercising of offsite response capabilities need not be required by NRC regulation. 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, includinc offsite exercises, provisions for periodic emergency should be develo)ed. These actions are required by existing NRC regulations which include t1e required establishment of an offsite emergency planning zone (EPZ). Consistent with the current regulatory approach, the staff views the inclusion of emergency preparedness b advanced reactor licensees as an added conservatism to NRC's ' defense-in-de th" philosophy. Briefly stated, this philosophy: (1) requires high quality i the design, construction, and operation of nuclear plants to reduce the likelihood of malfunctions in the first instance; that equipment can fall and operators can make mistakes, there(2) recognizes fore 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 these precautions, serious fuel damage accidents can happen, therefore requiring containment structures and other safety features to prevent the DRAFT

) l ^ release of fission products offsite. The added feature of emergency planning l to the defense-in-depth philosophy provides that, even in the unlikely event of an offsite fission product release, there is reasonable assurance that 1 emergency protective actions can be taken to protect the population around nuclear power plants. l 1 Information obtained from accident evaluations conducted as outlined in Section A of this enclosure will provide input to the Emergency Planning i requirements for advanced reactor des 1<pns. -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, l notificationrequirements,sizeofEPZ,andfreguencyofexercises. This i evaluation will take into account the results o passive LWR emergency planning policy decisions. l ) i .14 DRAFT j e , ~,.,. ~,---.,n ,..y.%., 3 ..y, ,,_y_,,, ,s .-,e....~... y_,:,. em.,, ,,,,op..w-.~-.4

E. REACTIVITY CONTROL SYSTEM i Luna Should the NRC accept a reactivity control system design that has no control rods? Current Reoulations General Design Criterion (GDC control systems be provided. )One of the systems shall use control rods 26 require preferably using a positive means for insertion. The other system shall be capable of controlling planned reactivity changes to assure fuel limits are not exceeded. Preseolicants' Position The PIUS design does not have control rods. However, the preapplicant proposes that the design complies 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 reactor overpower event (with forced flow) could initiate borated water flow 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 burnabla poisons for power sha core size for control of xenon oscillations for slow, ping, and limitations in large, and small reactivity changes. For rapid changes, the design relies on the highly negative moderator temperature coefficient of reactivity. 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 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 density locks would not only )revent a scram, but would interrupt the only safety-grade core cooling mecianism. Discussion The existing LWR regulations provide prescriptive design guidance for one reactivity control system to contain rods. Of the three advanced reactor designs, only PlVS does not have the capability to control reactivity with control rods. lhe PIUS design does have, however, three ways to introduce DRAFT

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 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. Recommendation The staff concludes that a reactivity control system without control rods should not necessarily disqualify a reactor design. A design without control rods may be acceptable but the applicant must provide sufficient information 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 function b. suppression of oscillations c. control of power distribution d. shutdown margin e. operational control t ~~ ,,16 - DRAFT i lo i i i ~ - - - ~ - , - ~, -

F. OPERATOR STAFFING AND FUNCTION hku Should advanced reactor designs be ' allowed to operate with a staffing complement that is less than that currently required by the LWR regulations. (urrent Reaulations 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 ali times and a licensed operator or senior operator must be present at the controls of a fueled nuclear power unit. 50.34(m)(2)(i) 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). preacolicants' Position The HHTGR plant is presently four reactor modules with two modules feeding a single steam su) ply 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 roomThisresultsinthreel and f roving non-licensed operators. 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 SR0 shift supervisor, one SRO assistant supervisor, one RO per power block (three modules) in the control room, and three plant R0s. This results in a minimum of eight licensed operators for nine reactor modules. 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 reactor. The designers of advanced reactors have stated that the highly automated operating systems, the passive design of s_afety DRAFT s e

features and the large heat capacity results in reactor designs that respond totranslentsinamannerthatdemandslessoftheo>eratort,1andothe J current operating plants or evolutionary designs. T,1e preapplicants assert that the passive safety feat *Jres and, in some cases, large coolant inventory of the PRISM, HHTGR, and PIUS 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,d concern that tie man-machinedraft policy p reactors. In that paper, the staff expresse 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. Recommendalian 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 evalu: tion 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 eventual progression to a safe shutdown condition under each of the following conditions: (1)acomplete 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. . DRAFT

i G. RESIDUAL HEAT REMOVAL I g Should advanced reactor designs that rely on a single completely passive, safety-related Residual Heat Removal (RHR) system be acceptable? Current Reaulations 3 General Design Criteria (GDC) 34 requires the RHR function to be accomplished ~ using only safety-grade s stems assuming a loss of either onsite or offsite power, and assuming a sin le fallure within the safety system. Re ulatory Guide 1.139 (issued in dr ft for comment), augmenting the GDC, sta es that the 4 RHR function must be performed to reach a safe shutdown condition within i 36 hours of reactor shutdown. Branch Technical Position (BTP) R$8 5-1 states i that the RHR function must be performed.in a reasonable period of time j following r_eactor shutdown, i j Prenonlicants' Position l Th'e 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 j containment-vessel outer surface. RHR is taen 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 j 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 l 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 [ remainsavailable,thenthenon-safety-gradeauxiliarycoolingsystem(ACS) supplements RVACS. The ACS operates through natural circulat'on air cooling of.the steam generator. The RVACS design basis anal sis erformed by the designer) results in high temperature conditions (wi hin sign-limits for an However,useo)fthe I extended period of time if no other system is operated. ACS system in con unction with RVACS can limit peak coolant temperature for decay heat remova 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 ~ ports.: Redundancy,is provided by~these separate ports and a-cross-connected header-that surrounds the reactor vessel- (i.e., anypanel_canbefedfromanyinletandcandischargeto.anyoutlet). 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. i There are no active components in the RCCS. The system is always_.in DRAFT w -m-.vy1--- ,_cw .,c.-+ w, ,w,r-%,r,m ..,f -...~w..,m-c.--- ,.v w.r,....--e

c)eration. The RCCS is relied upon when the heat trans sort system (HTS) and tle shutdown cooling subsystem (SCS) are inoperable. T,1e HTS utilizes the steam generators and non-safety-grade feed system and condensers and is used during normal operations, startup/ shutdown and refueling. The SCS is a 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,useofthenon-safety-gradebackupRHRsystemsredur.es the frequency, magnitude and duration of high temperature challenges to the reactor vessel. The PlVS design uses a safety-grade passive closed cooling system (PCCS) for residual heat removal from the reactor pool. The system consists of eight 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 tle 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 PRISM 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 end 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 Cair). Elevated temperatures (above normal operating values) may ex4st in the vessel and internal structures for long periods of time. Particularly in the high temperature reactors, the PRISM and HHTGR, creep damage may be more likely as the result of these high-temperature transients. - DRAFT

Recomendation As a result of the unique design features of the PRISM, MHTGR, and P!US designs, the staff believes that reliance on a single, ~~npletely passive, safety-related RHR system may be acceptable. In carrying out its future detailed design evaluation the staff will assure that NRC regulatory treatmentofnon-safety-relatedbackupRHRsystemsisconsistentwith Comission decisions on passive light-water reactor design requiremen's. 1 ? j i l 5 e v = DRAFT _

H. POSITIVE VOID REACTIVITY COEff!Cl[NT hild i 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)the power ope:ating range the net effect of 11 requires that the reactor core and coolant system be designed so that in prompt inherent nuclear feedback characteristics tend to compensate for rapid increases in reactivity. Prenoolitants' Position In the PRISM design the maximum sodium void worth, according to the preapplicant, assumlng only driver fuel and internal blanket assembites 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 constdown, the reactor could experience a severe power excursion and core disruption. The aredicted 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 )ositive throughout the fuel core lifetime. The total core void worth is >etween $1 and $2. T. positive void coefficient is not a concern during normal operation, but, during a large LOCA at specific locations, void reactivity 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 inse)rtion could lead 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 severe, 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 l 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 - {2 - DRAFT

5 1 either the PRISM or CANDU 3 designs may not necessarily lead to a large scale release of the radionuclide inventory to the atmosphere due to their respective mitigative designs. In the CANDU 3 reactor, multiple redundant, diverse fast acting scram systems n're provided to address the positive coefficients. Attempts to modify the designs to reduce the effects of these positive 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 unscrammed loss of heat sink ] events without core damage. B1 commendation The staff concludes that a positive void coefficient should not necessarily disqualify a reactor design. The staff is proposing to require that the PRISH and CANDU 3 preapplicants analyze the consequences of events (such as ATWS, unscrammed LOCAs, delayed scrams, and transients affecting reactivity control) 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 prea)plicants will be required to consider changes in the designs to mitigate tie consequences of these accidents will de)end on the estimated probability of the accidents as well as the severity of tie consequences.

• • DRAFT

d I. CONTROL R00H AND REMOTE SHUTDOWN AREA DESIGN s 11.Ulf + Can current recuirements for a seismic Category 1/ Class lE control room and alternate shutcown panel be fulfilled by a Remote Shutdown Area, and a non-seismic Category 1, non Class IE control room? Grrent Reaulations 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 (CDC) 19 requires that a control room with ad:quate 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 the control room. GDC 17 requires that the electrical system for the control room and remote shutdown equipment meet the requirements for quality and independence. These requirements are defined as Class IE in the supporting 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 functional during a safe shutdown earthquake. Preapolicants' Position The control room for PRISM contains the instrumentation and controls for all nine reactor modules 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 lE controls and instrumentation to protect the core and has the required habitability control system. The HHTGR 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 chutdown area are Class IE. The remote shutdown area does not contain safety-related equipment, nor does it include a 4 ventilation system for operator habitability, or a safety-related manual scram. This is based on the )reapplicant's position that accidents do not require operator response. Tie 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. DRATT

1 i i 1ayouts, and capabilities present in the Main Control Room. The secondary control area is seismically qualified and is electrically isolated from the i i 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 roon j to the secondary control area is to be qualified to %e extent necessary to prevent route blockage,ing ac,cident conditions.CANDU 3 ha' specified requiremen fire or flood. } assure habitability dur c f The central control room for the PIUS design is a seismic Category 1 1 structure. However, the safety-related systems within this structure are for I monitoring only to assure that the core is protected. Although the operator i could take actions, these actions would be with the use of non-safety-grade i controls. The two remote-shutdown areas are housed in separate compartments e i at the bottom of the reactor building in protected seismic 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 i i 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. l 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 j c)erators managing off-normal operations. The control room is the space in t1e plant where operators are most familiar with the surroundings and normally-i manage plant activities. The staff is reluctant to approve any design that 2 would increase the frequency of evacuation of the control room during. design l 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, j seismic Category I and electrical Class IE control room is appropriate. J j The staff also believes that any remote shutdown area should be designed to-complement the main control room. Sufficient Class IE 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 I operations to the greatest extent possible. LA 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 i l l-t- P DRAFT ~ l ? p, e---- .--n,-,e-~,,,,-r, nr. e,n.,+,n v.,.~,.,n, ~.n., --n ,,-,,m...-- ,,, --- w. - -m -n c.,---+-n .e,,---., -+.--vw~+,c,-

paper defined positions on common mode failures in digital systems and on annunciator reliability. Staff requirements for advanced reactor designs will y be consistent with passive LWR policy guidance for these issues, once ",he guidance is finalized. Recommendation The staff recommends that untti passive LWR policy for design requirements of control rooms and remote shutdown facilities is finalized, the staff will apply current LWR regulations and guidance to the review of advanced reactor designs. This will ensure that plant controls and the operators will be idequately situations. protected so that safe shutdown can be assured in accident t 4 4 DRAFT 1

J. SAFETY CLASSlFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS hin What criteria should the NRC apply to the advanced reactor designs to identify the safety-related structures, systems, and components? Current Recuh11gni A)pendix A.Vlf a)(1) of 10 CFR Part 100 list the following c)r(1) and the curre Title 10 of the Code of Federal Regulation Section 50.49(b iteria to define tie safety-related 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) to update criteria used in decisions regarding reactor siting and design for future nuclear power plants, including the advanced LWR designs. These proposed revisions include the temporary relocation of the dose considerations for reactor siting i.e., the current Part 100 guidelines) from Part 100 to Part 50 until su(ch time as more specific requirements are developed regarding accident source terms and severe accident insights. Prenoolicants' 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 movirig 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, systems, and components needed to mitigate the dose consequences at the site 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 RCPBisnotentirelysafetyrelated.(2)nosafety-relatedequipmentis)used l to pressurize and depressurize the RCPB, (3) the coolant moisture monitor is not safety related, and (4) neither the control room or remote shutdown area ORAFT

l i i 2 'd I are safety related, and reactorprotectionormon(5)nosafety-relatedinstrumentationproviding itoring functions are available in the control room- )i or remote shutdown area. i Discussion 4 i The NRC LWR criteria are intended to require defense in depth; the advanced reactor designs include high quality, non-safety-related active systems to provide defense-in-depth capabilities for reactor coolant makeup and decay I seat removal. These would be the first line of defense in the event of i transients or { the designers, plant upsets.The non safety-related systems are a not required for mitigation of design basis evenIs,ccording to .but do i 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 i i the proposed resolution to this issue would be provided later. The staff - plans to treat non-safety-related systems consistent with the eventual j position for-passive LWRs. Recomendations l The staff intends to apply the LWR criteria for identification of safety-1 related structures, systems, and components to the MHTGR design. Requirements for non-safety-related systems will be consistent with the NRC position for

)assive LWRs. We have noted that LWR criteria may be restructured within

' arts 50 and 100, and our expectation is that the criteria in Part 50 will apply to the standard design certification. P 4 DRAFT _,_..__..--_,-__-.,.__a,.. -_,s.

4 e A. CANADIAN DEUTERIUM URANIUM (CANDU) 3 REACTOR DESIGN Develooment 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, Limited, Technologies (AECLT) informed the NRC of their intent to submit the CANDU 3 reactor design for standard design certification in accordance with Part $2. 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 it also pursuing standard design certification of the CANDU 3 with the NRC's Canadian counterpart, the Atomic Energy Control Board of Canada. AECLT's c9rrent plans are to submit a standard design certification application for CANDU 3 in the 1995-1996 time frame, Desian Descrintion l The CANDU 3 is a 450 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 calandria tube. The pressure tubes form part of the reactor co'olant system pressure boundary. Because of the low excess reactivity associated with a natural uranium core, 1 l .1 - DRAFT l

J c : 1! ;' the CANDU design must be fueled on a continuous basis during power operation i b an automatic fueling machine. On-line fueling is the primary means of j c anging reactivity in the CANDU 3 L 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 withasinglefuelingmachine. The primary system operating pressure outlet headers. psig)CANDU'3 light-water secondary system is similar to thatis mai (nominally 1435 j The of a PWR. The fuel channel assemblies are enclosed in a horizontal, cylindrical vessel called a calandria that contains the low-temperature (140 F, low- >rtssure, Thecalandriavessel,inconjunction)withtieintegral heavy-water moderator. i 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 l removal system ensures that the moderator remains subcooled, i 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 shuttino down j the reactor maintaining cooling of the fuel and providing plant monitoring capability In the event that the other group,of systems is unavailable. l Group 1 systems are those primarily dedicated to normal plant power pro-duction. The Group 2 systems include four special safety systems and other i saf:ty-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. i CANDU 3 employs two fast-acting, redundant, and diverse Group-2 shutdown I systems, se arate from the Group 1 reactor regulating; system... Shutdown System No~. 1 (5051 consists of 24 vertically inserted control rods. Shutdown System - No.2(5052 consists of six horizontal. nozzles through which a ga6olinium 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-coolante accident.

dach of_the four' safety systems is required to. ' demonstrate during operation, a dormant unavailability _ of less than 10'8 or about 8 hours per year,-- and be physically and functionally separate from the normal process systems and from one-anoiner. The CANDU 3 shutdown cooling system is designed.to_ remove heat from the HTS at nominal. operating temperature and pressure. .2 - DRAFT .a. .f

B. MHTGR (MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR) Develooment History The Modular High Temperature Gas-cooled Reactor (MHTGR) was proposed to NRC by the U.S. Department of Energy (DOE) in 1986 in response to the Commission's AdvancedReactorPolicyStatement(51FR24643). A preliminary safety information document (PSID)ID and 10 amendments were reviewed by the Office of and twelve amendments were submitted from October 1986 to March 1992. The PS 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 appiteation by September 30, 1996. Commercial gas-cooled reactors began with the graphite-moderated, carbon dioxide-cooled Magnox reactors 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 H1GRs. 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. 00E 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 estabitshed with Japan, which is butiding 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) proposed greater fuel integrity, 4 with a 6 x 10 fraction of failed fuel assumec for the MHTGR, and (4) lower enriched uranium fuel. Desian Descriotion The standard MHTGR plant is four reactor-steam generator modules and two steam turbine-generator sets. Each 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

The low-power density (5.g watts /cc) reactor core is helium cooled and traphite moderated, and uses ceramic coated (he fuel.four major layers) nicrospheres f ' n an organic bonded cylindrical compact as t The core design is intended to provide a large negative doppler coefficient to shutdown the reactor with heatup. The microsphere fuel design is stated to allow fuel temperatures as hijih as 2g00 '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 iner".. The MHTGR has a below-grade, safety-related reactor building, containing the 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 g25 p i, so water would leak into the coolant with a steam generator tube leak or 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 outsioe into the panels. The system (1) is-entirely passivewithnomovingcomponents,(2)isalwaysoperating,(3) automatically-responds to rising temperatures through thermal radiation and natural circu-lation, and (4) has flow path redundancy to the cooling panels through.a cross-connected header. In addition, there are two ot wr non-safety-related, active heat removal systemst (1)the-shutdowncoolingsysteminthebottomof the reactor vessel, and (2) the main circulator / steam generator in the primary 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

( t main steam and feedwater inlet piping. It will not retain the gases from a i rapid RCPB depressurization, but is designed to have a leak rate of less than 100 percent / day after initial depressurization. T e. 4 4 e-5- DRAFT.

~ 5 C. P!US(PROCESSINHERENTULTIMATESAFETY) ((1storical Develooment The Process inherent Ultimate Safety (PIUS) teactor is being designed by ASEA l BrownBoveriAtom(ABBAtom). The concept evolved in the early 1980's from an extension of then ABB-Atom's low temperature district heating design. In October 1989, ABB requested a licensability review of the PIUS design in accordance with NUREG-1226, and in May 1990, ABB submitted the P!US preliminarysafetyinformationdocument(PSID)igninthe for staff review. ABB plans to apply for design certification of the PIUS des 1994-1995 time frame, assuming a favorable preapplication review. The PIUS design concept has already undergone tests related to the design principles. ABB has completed testing using the MAGNE Test Rig to simulate slVS parameters such as diffusion and mixing across the primary loop boundary with consideration for effects of turbulence, stratificatio/ pool n, migration of boron, and others. Large scale tests of the PIUS design principles, such as flow and density lock operation, were done at the ATLE Test Rig. These tests were used to validate the RIGEL code to calculate the design's safety and transient performance. ATLE was a full height simulation of the PlVS pool. Other tests of the P!US design principles have been carried out at MIT and TVA, and other additional large scale tests and a larger test rig are planned to be started this year for the pur optimization, as well as special component testing. pose of design It is planned that this larger test rig will serve as the basic test facility for developing data for the detailed design and verification. Desion Descriotion l PlUS is a 640 MWe advanced pressurized water reachr (PWR) design with four loops. It relies on thermal hydraulic effects to accomplish the control and safety functions that are usually performed by mechanical means. The safety-grade reactor heat removal system for the PIUS design is completely >assive and is always in operation. The PIUS design consists of a vertical 1011ow cylinder, the reactor module, which contains the reactor core. The reactor module is submerged in a large concrete reactor vessel containing l 3,300 H (870,000 gallons) of highly borated water. The reactor module is 5 open to the borated pool at the bottom and at the top of the reactor module. l At these two openings, density locks keep the borated pool water from the reactor module during normal 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 pumped back into the bottom of the reactor module, bypassing both the top and bot. tom density locks. There is no shysical flow barrier in the density locks between the primary loop and the > orated pool, however, the difference in density _between the primary loop l 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 i 1 . DRAFT , - -, =

i 1 I i and the borated water flows into the core and shuts down the reactor. A j natural circulation flow path is then established from the borated pool i through the lower density lock, up.through the core, and back into the borated i pool through the upper density lock for long term shutdown cooling. Unlike j most reactors. P!US does not employ mechanical control rods for regulating reactivity. Reactivity is controlled by the boron concentration and temperature of the primary loop reactor water. 1 Anactivereactorprotectionsystem(RPS),withassociatedinstrumentationand 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 PIUS design are similar to the passive LWRs being considered by the staff (AP-600 and the SBWR). Although P!US is a PWR, its 2 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 value is expected to be 0.5 percent at design pressure. 7-DRAFT ---,n ~.,..'.,,,,,--yw.,m,,. + .c.w,~7--,c.-. '.y,.m.e,w.,_..,,,.. ,._...,.,,.-.E.,.w__,--.,--rn. -,y m._, _.e, n

• -r---w,,

j O. PRISM (POWER REACTOR INNOVATIVE SMALL MODULE) Develcoment History The U.S. Department of Energy (DOE) selected the Power Reactor Innovative Small Module (PRISM) design as the advanced liquid metal reactor (ALMR) design to sponsor for NRC design certification. The conceptual design for PRISM was developed by General Electric (GE) Company in conjunction with an industrial team of commercial engineering firms. Research and development support is being supplied by the Argonne National Laboratory, Energy Technology Engineering Center, Hanford Engineering Development Laboratory, and 0;k Ridge National Laboratory. In addition, a steering group of utility representatives was involved in the PRISM design effort. DOE chose to sponsor the PRISH design as part of its National Energy Strategy because of the design's potential for enhanced safety through the use of passive safety systams 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 been proposed in the current application, DOE has supported studies evaluating 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 components design, reliability data, and safety assessments. This experience consists of operation of a number of facilities such as E8R-II, Phenix, the Fast Flux Test Facility (FFTF), the Joyo reactor in Japan, and others. The PRISM Preliminary Safety Information Document (PSID) was submitted to the NRC for review in November 1986, and the results of an early NRC staff review was the draft PSER (NUREG-1368) issued in September 1989. In order to obtain NRC approval of its planned prototype, DOE plans to apply for preliminary design approval in 1995. The DOE also plans to apply for standard design certification in 2003 after a prototype demonstration. These plans are based on the current DOE goals to demonstrate the commercial potential for the ALHR by 2010, as called for in the Energy Policy Act of 1992. Elant Descriotion The PRISM plant design consists of three separate power blocks each made up of three reactor modules. Each module has a thermal 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 one or two power blocks are possible. PRISM operates at much higher temperatures than current LWRs which will require a rigorous evaluation of the-effects of creep and creep rupture on reactor vessel and systems. The PRISM design also relies on a highly automated and complex control system utilizing i digital processing. l l l DRAFT l

The reactor module consists of the containment system, the reactor vessel, the core, and the reactor's internal components. The reactor vessel encloses and-supports the core, the primary sodium coolant system, the intermediate coolant system heat exchangers (IHXs), and other internal components. The vessel is located just inside the containment vessel, which is located below grade in the reactor silo. The reactor vessel is penetrated only in the closure head. The head is supported by the floor structure, and the floor structure is supported by seismic isolator bearings to reduce horizontal movement during i seismic events. The upper head of the reactor vessel is the closure head. The closure head also supports the intermediate heat exchangers (IHX) and the electromagnetic (EM) pumps. The main components of the Nuclear Steam Supply System (NSSS) in PRISM are the reactor module, primary sodium loop, EM pumps, IHX, intermediate sodium loop and steam generators (SG). The primary sodium loop is contained completely within the reactor vessel, which is hermetically sealed to prevent leakage of the primary coolant. The EM pumps provide the primary sodium circulation. Synchronous machines provide flow coastdown capability to the EM pumps. Flow coastdown is very important for preventing sodium boiling during a loss of.EM-pump power without reactor scram. Reactor-generated heat in the primary loop 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 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 differential across the IHX from the IHTS to the PHTS is maintained. A sodium-water reaction protection system mitigates the effects of reactions between lHTS sodium and water in the SG. The reference fuel for the ALMR is a uranium-plutonium-zirconium (U-Pu-Zr) alloy. The ferritic alloy HT9 is used for cladding and channels to minimize 1 swelling caused by high burnups. The PRISH core is a heterogeneous arrangement of driver fuel and blankets. The PRISH core design is such that the net power reactivity feedback is negative in all ranges of operation, in all transients, and in all accidents not involving voiding. For certain very low probability accident scenarios involving sodium boiling, a 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 and thermal expansion of the core and related structures that dominates the positive moderator density coefficient. The net negative temperature coefficient is so large that analyses predict a11 non-boiling transients and accidents to be terminated by the temperature 'eedback reactivity at temperatures low enough to not threaten fuel or vessel integrity. This passive shutdown function allows the reactor to sustain all non-boiling transient scenarios without damage, even with a failure to scram. . DRAFT

i There are six control rods in the main reactivity control and shutdown system. n Inserting any one of the six will shut the core down..The control rods can be inserted using (1)de reactor protection system (RPS) for. rapid insertion.and; the plant control system (PCS) for normal insertion.- t..' the safety-gra )gravitydropintothecore. If both the normal and safety-grade; systems. l hl, the operator can activate the ultimate shutdown system (USS boron balls into the central location of. the core causing shutdown)-which sends independ - ently of the control rods. The PRISM design also includes passive mechanisms-for controlling reactivity: three gas expansion modules (GEMS) consistin tubes, closed at the top and_ open at the bottom, and filled with helium. g of-- If ? the pumps are running, the static pressure is high, causing the sodium level to rise to a high point in-the GEN.- However, with the pumps off, the static pressure and sodium level drop, which increases neutron leakage..The reac-l tivity change provided by the GEMS between these two states is about l' -70 cents. L Normal shutdown cooling is achieved with the non-safety-grade condenser. If i the condenser becomes unavailable, the safety-grade reactor vessel auxiliary cooling system (RVACS) is used for RHR. The RVACS circulation air cooling of the containment vessel. provides natural The-design-basis RVACS event assumes:that the normal' and auxiliary heat removal _ systems, as.well as the Intermediate Heat Transpert 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-i structures within the reactor vessel under American Society of Mechanical Engineers (ASME)iliary cooling system 1The PRISM design also contains the non : Level C conditions. safety-grade aux (ACS) to assist the RVACS. The ACS uscs. natural circulation within the steam generator-(SG) to remove heat indirectly from the reactor vessel, and-natural 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:- (1)the l favorable combination of viscosity, thermal conductivity, and vapor. pressure-associated with the use of sodium to remove heat, (2) the ability to operate l l 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 l associated with a single phase coolant. l l ( i: 1 \\ - DRAFT-t u 4. _,e-..w_.- ,,,___.m. e-. -,,,~,-,,.,.o .,or_m,- w.

2 SECY-86-368, "NRC Activities Related to the Commission's Policy on the Regulation of Advanced Nuclear Power Plants," December 10, 1986 SECY-89-350, ' Canadian CANDU 3 Design Certification," November 21, 1989 SECV-90-055, ' PIUS Design Review," February 20, 1990 NUREG-1338, ' Draft Preapplication SER for the MHTGR" NUREG-1368, ' Draft Preapplication SER for PRISH" NUREG/CR-5261, " Safety Evaluation of MHTGR Licensing Basis Accident - Scenarios" i NUREG/CR-5364, ' Summary of Advanced LMR Evaluations-PRISM and SAFR" NUREG/CR-5514, "Modeling and Performance of the MHTGR Reactor Cavity Cooling System" NUREG/CR-5647, " Fission Product Plateout in the MHTGR Primary System" NUREG/CR-5815. " Evaluation of 1990 PRISM Design Revisions" t 1 ORAFT l j

,~ f* "*% e...+ POLICY ISSUE j November 23, 1992 (lnfOrrnatlOn) srcY-92-393 l ) [qt: The Commissioners Em: James M. Taylor Executive Director for Operations lub.iect: UPDATED PLANS AND SCHEDULES FOR THE PREAPPLICATION REVIEWS OF THE ADVANCED REACTOR (MHTGR, PRISM, AND PIUS) AND CANDU 3 { DESIGNS i 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. l Backaround: In SECY-91-161, " Schedules for the Advanced Reactor Reviews i and Regulatory Guidance Revisions," the staff informed the l Commission of the following estimates for completion of the j preapplication reviews: PRISM November 1992 I MHTGR December 1992 i CANDU 3 June 1993 PIUS July 1993 i The staff based these dates on broad planning assumptions i including the prea cation schedules, pplicants' design certification appli-availability of the Office of Nuclear Reactor Regulation (NRR) resources to conduct the reviews, i timely receipt of information to support the reviews, and a scope of review consistent with the previously issued draft l preapplication safety evaluation reports (PSERs) for the i PRISM and MHTGR designs. \\ Contacts: NOTE: TO BE MADE PUBLICLY AVAILABLE Thomas H. Cox, NRR IN 10 WORKING DAYS FROM THE 504-1109 DATE OF THIS PAPER Brian W. Sheron, RES ) 492-3500 i .4 v

~ _ ___ ___ _ a 4. ja* The Comissioners 5 l 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 application schedules or modified their pro >osed designs. Most-staff technical review resources have men redirected to higher priority operating reactor and design certifi-cation reviews. Additionally, implementation of the Fee i Recovery Rule has resulted in preapplicants desiring a preapp11 cation review scope limited to key certification / licensing issues. . In February 1992, the staff issued letters to all four preapplicants requesting confirmation of their plans for i design certification and, in some cases, a schedule for i submitting additional preapplication review infomation. 1 All responses were' received by May 1992. 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 l would be dedicated to conduct-the preasplication 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 i 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 4 schedules for additional submittals. The staff also informed the preapplicants of key policy issues that the staff is planning to forward to the Comission for guidance. During these meetings,'the preapplicants and the staff-j agreed on a smaller, more focused scope for conducting the preapplication reviews. Previous planning assumed that the 4 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 l Evaluation for the Modular High-Temperature Gas-Cooled Reactor.' Discussion: The staff considered several factors in developing a revised -l schedule for conducting the preapplication' reviews. The -enclosure to this paper provides a design-specific:sumary i A . ~.. - - ~ - - ..,I

[-

,o-The Comissioners.

of this information and the staff's rationale for sequencing - i the reviews. Based on this rationale and consideration of-available resources,- the staff has identified the following L revised estimates for completing the preapplication reviews:- I PRISM-December 1993 CANDU 3 December 1994

• PIUS April 1995 MHTGR-December 1995 The staff expects to follow the same review process for 4

approval of the PSERs as-is-beino used for the safety 4 evaluation reports on the evolutionary light-water reactor i designs and the Electric-Power Research Institute Utilities Requirements Documents. _Approximately 6 months before-i completing the review, the staff will submit a. draft final-PSER to tie Comission.- With;Comission consent, the_ staff i will forward the draft final PSER to the-preapplicant, the i Advisory Comittee on Reactor Safeguards 1(ACRS), and the NRC-l Public Document Room. -After considering unput during public i meetings with ACRS and_ the preapplicant, a final' PSER will-be forwarded to the Comission for approval.- L The staff intends to conduct most of the preapplication' i reviews with staff..from the NRR Associate Directorate for i 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 Assessment (ADT) will,.in general, not j participate in the preapplication review. ADT technical staff resources are currently required for higher priority. i 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-i this approach appropriate since the preapplication review i considers the conceptual design, and final technical. decisions on safety will not be made until the design-certification review when.the ADT technical staff will be-i involved.. 1 \\ -The staff believes that the changes to_the preapplication review scope-and schedule, and the _ approach for conducting NRR technical. review, provide the most effective use of NRC resources, sThe proposed schedules will allow the staff to provide a timely response to the preapplicants in important-4 areas _regarding their design certification application plans. By emphasizing the key policy issues for the kl u = a

' I' ' - 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. Epig: There has been Congressional interest in this matter and the Chairman previously advised Senator Johnston and Congresswoman Lloyd of the schedules outlined in 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. / h mes xecutive Director .for Operations

Enclosure:

Sumary of Input for i Schedule Revision 9 DISTRIBUTION: Commissioners OGC OCAA 4 OIG OCA OPA i OPP j EDO ACRS ASLBP SECY l

.c 1

!.1 ms In March 1989. the NRC issued NUREG-1368 " Draft Preapplication Safety Evaluation Report for Power Reactor Inherently Safe Module Liquid Metal 4 Reactor," in which it sumarized the results of its-review of the Preliminary PSID) submitted in 1986. In March 1990, the U.S. Safety Information Document-(bmitted Appendix G to the PSID, " Responses to-Department of Energy (DOE)hich it proposed several significant changes to the su Issues-in Draft SE," nw PRISM design. These changes included increasing the power, adding an ultimate shutdown system and containment dome, redesigning the reactor to add gas-i expansion modules (GEMS),- and changing to a single-wall-tube helical coil l steam generator. Brookhaven National Laboratory (BNL): reviewed the revised l design and published its findings.in.NUREG/CR-5815, " Evaluations of 1990 PRISM r 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-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 followingL schedule: Preliminary-Design Approval Application CY 1995 Prototype-Final Safety Assessment Report 1997 ] Standard Design Certification-Application 2003 i (Afterprototypetesting) l 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 demonstration by September 30, 1998. i In June 1992,.the National-Research 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 P government support key reactor designs 1The Council recommended that the. 1 PRISM design lbe the only preapplication design to receive government funding because of.its unique ability as a breeder reactor.- At a public meeting on July 1,1992,- the'NRC staff notified DOE that it would need to delay _ issuing the PSER'beyond the originally scheduled date _ of-4 November 1992. ' DOE noted that it had' submitted all requested information to the NRC to support the preapplication review and requested that the.NRC-a T complete the PSER as'soon as possible to support DOE'in planning for design certification. -DOE recently notified the NRC of problems found with the: 1: [ ,! l 1 I e .-s-- u-n,n~, - - - - - --.~,,n ~--.-n n, ,,<ae .~~---,--,nw-n--~,. ,-,----.~..J.~..wmn ,-----n-- ,-n i

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 )lans to conduct the PRISM preapplication review first to capitalize on the wor ( 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 approval 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. MHIGB 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 alternatives, 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 DOENewProductionReactor(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 to support resolution of the key MHTGR policy issues identified in NUREG-133is. 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 National 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-established 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.

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-approval stage-of design. certification. It stated that this power increase will not affect the key policy issues for 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. The staff plans to conduct the MHTGR review as the final preapplication. review in the series of four projects, because of the uncertainties in the DOE schedule and design.to be proposed for design certification application. When DOE submits NHTGR design certification schedules the-staff will reconsider the: preapplication-review plans. DOE is most interested:in obtaining feedback.on-the implementation of the-key policy-issues for_ the MHTGR-design. Continued-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 On May 25, 1989, Atomic Energy of Canada, Limited, Technologies (AECLT) i informed the NRC of its intent to submit the CANDU 3 reactor design for standard design certification. AECLT, a. wholly _ owned U.S. subsidiary of-preapplication-reviewbysubmittIng(AECL)inCanada,-has' supported.theCANDU3 Atomic Energy of Canada, Limited a Technical Description,' Conceptual-Safety Report, Coe..:eptual 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 i standard design certification-application in 1995 or 1996 if the NRCL eompleted L its preapplication review of the CANDU 3 by June 1993. - Dn 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 its design-certification plans-in the U.S. if Canadian cons,truction! plans did;not'materi-- alize. The National Research Council report identified the CANDU 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 difference.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 started some preapplication review on the CANDU'3 design. NRR H is conducting two projects at DOE laboratories:. a. study of the CANDU~3 'j positive void reactivity coefficient and a. review of the operation of the on-line refueling machine. RES has completed a: systems study to_ identify candidate' event secuences for required safety _ analysis, and it has projects-to assess data base acequacies,- to-perform preliminary transient calculations. L l + -.,.-., ~,.. -, -.. - -. - - -.. - - ~n. .,-~_.-a_. ., u.a

using Canadian codes, to identify code needs for future independent analyses, to initiate severe accident analyses with NRC codes, and to prepare for source term determination. RES.will also provide in-house analytical capabilities for itself and NRR for the CANDU 3 design. To better understand the CANDU 3 containment performance and radiological 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 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: performed this analysis for CANDU 3, and has supplied little directly relevant information on the event and its consequences. AECLT is having problems getting proprietary information released from Canada 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 work is now on the critical path for completing the preapplication review, and the lack of timely submittals of other proprietary information could further delay the review schedule. In a letter dated September 23, 1992, the staff 4 informed AECLT that an inability to transfer proprietary material to the U.S. 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 review because the design and experimental data base are already sufficiently developed to support the review. The June 1994 PSER issuance assumes prompt resolution of the present problems releasing proprietary information required for the review from Canada to the United States. f_1.E 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 direct representative of ABB Atom in the U.S., and is the official preappli-cant of record. In May 1990, ABB/CE submitted a five vol'ume Preliminary Safety Information Document (PSID) to support its request for a preapplication review. NRR has started a project with BNL to support core neutronics modeling. RES has completed a systems study to identify candidate event sequences for required safety analysis, and it has projects to assess data base adequacies, to 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-nation. RES will also provide in-house analytical capabilities for _itself and NRR 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-plication safety evaluation report (PSER) by April 1994 that does not require - 4 _.

• • w t

4N-t significant design changes to the PIUS design, and (2) the commercial environ-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. l The National Research Council report concluded that the PIUS design would not. l likely be ready for commercial operation within the next 20 years and had a low priority for DOE support. The 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 preap)11 cation review because the design is presently at the conceptual stage and tie 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.- Continued 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 i design more fully and respond to staff questions without impacting the preap-plication review schedule. l l l i .}}