ML20126F627

Forwards Two Papers Which Would Be Valuable to Continuing Regulatory Efforts Re Policy Issues & Schedules Concerning Preapplication Reviews of Advanced Reactor & Candu 3 Designs

ML20126F627
Person / Time
Issue date:	12/16/1992
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Hink A AECL TECHNOLOGIES
References
PROJECT-679A NUDOCS 9212310027
Download: ML20126F627 (48)

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{{#Wiki_filter:_= / y arc g UNITED ST ATES l' ,..y /[ % NUCLE AR REGULATORY COMMICSION ,of w AssinoTon. o. c. rosss 5 "V L E December 16, 1992 e Project No. 679 Mr. A.D. Hink Vice President / General Manager AECL Technologies 9210 Corporate Boulevard, Suite 410 Rockville, Maryland 20850

Dear Mr. Hink:

COMMISSION PAPERS ON POLICY ISSUES AND SCHEDUL PREAPPLICATION REVIEWS OF ADVANCED REACTOR AND

SUBJECT:

Enclosed are two papers which should be valuable to your continuing regulatory In mid-1992 the staff discussed with you its intent to identify key policy issues and the projected schedules to complete preapp efforts. of these designs. schedules would be addressed in separate papers to the Comission. As a result of the staff's reviews, an assessment of projected resources, and the meetings held with you and the other preapplicants (1) a draft paper providing the staff's positions on 10 policy issues, and (2) a final paper, SECY-92-!.93, " Updated Plans and Schedules for Commission: the Preapplication Reviews of the Advanced Reactor (MHTGR, PRISM, and plUS 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 yetThe sta obtained Commission approval cn these issues.the Advisory Comm a

The staff will include the views of the ACRS and document its final recommendations in a revised paper before seeking the Commission's the near future. Any comnunts you may wish to offer will be considered as we prepare approval. our final positions. Please submit any comments-by January 25, 1993. The proposed schedule paper reflects the staff's assessment of its resources The staff will continue to try to and the needs of the preapplicants. expedite its reviews and complete the work ahead of schedule. 300020 Q QV 1Y d M/ fCURG0 LEG!E0RY CE!iTRL f!!ls gi g23m37921216' ' PDR

Mr. A.D. Hink 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, Janet Kennedy. 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-Distribution: Central File NRC PDR PDAR R/F DMCrutchfield WDTravers MMSlosson THCox EDThrom JLKennedy OGC 15/B/18 ACRS (10) P 315 OPA LLuther CANDU R/F LA:PDLR:ADAR PM:PDAR:ADA SC:PDAR: DAR SE:PDAR:ADAR D:P fDAR dRP-LLuther JLKennedy:sa THCo d EDThrom % MMS 1 sson DMCrutchfield $/g/92@,, /A/14/92-i.a/ty/92 V, n /sy/92 Q/ 92 /2/#/92 0FFICIAL R RD COPY Document Name: CANDSECY.LTR l i

Mr. A.D, Hink December 16, 1992 CANDU Project No. 679 cc: Louis N. Rib, Licensing Consultant AECL Technologies 9210 Corporate Boulevard, Suite 410 Rockville, Maryland 20B50 Bernie Ewing, Manager Studies and Codification Division Atomic Energy Control Board P.O. Box 1046, Station B 270 Albert Street Ottawa, Ontario, Canada KlP 5S9 A.M. Mortada Aly, Senior Project Officer Advanced Projects Licensing Group Studies anJ Codification Division Atomic Energy Control Board P.O. Box 1046, Station B 270 Albert Street Ottawa, Ontario, Canada KlP 5S9 Project Director - CANDU-1 AECL CANDU 2251 Speakman Drive Mississaugua, Ontario, Canada L5K 1B2 L. Manning Muntzing Newman & Holtzinger, P.C. i 1615 L Street, N.W., Suite 1000 Washington, DC 20036 Steve Goldberg, Budget Examiner i -c Office of Management and Budget 725 17th Street, NW. Washington, DC 20503 r i i

1 ' + . DRAFT F I BLt: The Comissioners Im: James M. Tay1or-1 Executive Director for Operations Sub_iect : ISSUES PERTAINING'TO~THE ADVANCED REACTOR (PRISM, MHTGR, AND PIUS) AND CANDU 3 DESIGNS AND THEIR RELATIONSHIP TO CURRENT-REGULATORY REQUIREMENTS i Puroese: To request Commission guidance for those areas where the staff is proposintto depart from current regulatory 4 requirements in the preapplication review of the advanced reactor and CANDU'3 designs. l' Backaround: The Advanced Reactor Policy Statement (51 FR 24643) and NUREG-1226, ' Development and Utilization of the NRC Policy Statement on the Regulation of Advanced Nuclear Power Plants," define advanced reactors as.those with innovative designs for which licensing requirements will-be signif-icantly different from the existing._ light-water reactor j' (LWR) requirements.. These_ documents also provide guidance for the development of new regulatory requirements to i support the advanced designs.. Staff reviews of these advanced reactor designs should utilize existing regulations a . to_ the maximum extent practicable. When'new. requirements are necessary, the staff should move towards performance - standard regulations and away from prescriptive regulations. l Each designer is encouraged to propose new criteria-and novel approaches for evaluation of their designs,.and an 1 [_ objective of early designer-staff interaction should be to I develop guidance on licensing criteria for the advanced i reactor design and to make a' preliminary assessent of the potential of that design to meet those criteria, j CONTACTS: H.M. Slosson 504-1111' q E T.H. Cox j SO4-1109 y ,-+.e-w c e-e w ~+.-wn ve,.<.- v N, ~. w w , -,. e.e n - er n, .,ne

_4' 4 1 [ The Comissioners - : b The staff is conducting preapplication reviews of the-following four designs: General Atomics (GA) 350-MWt Modular.High Temperature - + Gas-Cooled Reactor-(MHTGR) design sponsored'by the U.S. Department of Energy (DOE) Gas Cooled Reactor Program- !~ J General Electric (GE) 471-MWt Power. Reactor Innovative i + Small Module (PRISM) reactor design sponsored by the DOE-Advanced Liquid Metal Reactor (ALMR) Program i Atomic Energy of Canada, Limited,. Technologies (AECLT) [ 1378-MWt Canadian Deuterium Natural-Uranium (CANDU 3) reactor design Asea Brown Boveri-Combustion Engineering.(ABB-CE)- l 2000-MWt: Process' Inherent-Ultimate Safety-(PIUS). reactor design

provides a listing of-pertinent Comission papers and reference NUREG documents for these preap-plication designs.1 Some information in the teriginal -

i documents may be superseded by more recent pre:pplicant i submittals. A sumary of the current designs is provided as l In response to Comission staff requirements memorandum (SRMs), in SECY-91-202, " Departures-From Current Regulatory Requirements in Conducting Advanced Reactor Reviews," the staff comitted to identify issues during the preapplication-review that require Commission policy guidance or staff technical resolution' for design certification, including situations'in which advanced reactor designs significantly-deviate from current regulatory. requirements, p Policy issues for evolutionary and passive LWRs have been identified in the following Comission papers: i c 'SECY-90-016, " Evolutionary Light Water Reactor.-(LWR) Certification Issues and Their Relationship-to Current-Regulatory Requirements" Draft SECY (distributed for coments on-February 27, a' [- 1992), " Issues Pertaining to Evolutionary and Passive l' Light Water Reactors and Their Relationship to Current Regulatory Requirements" m ,,.-_,.._m-. ~ _.,.,

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r The Commissioners. I Draft SECY (distributed for coments on June 25, 1992), j ' 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 t i for Commission approval, are provided in Enclosure 1. The recommendations for Comission approval were developed by the staff with inputs from the preapplicants, the public, and the ACRS. The staff considered the preapplicants' l. proposals in light of-the Comission's policy statements and l-guidance on severe accidents, advanced reactors, and safety goals to develop a single consistent policy recomendation i 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 a designs'despite preapplicant proposals to do otherwise. Where deviations are recomended, the staff proposes more . conservative alternatives to the preapplicants' proposals to r account for uncertainties associated with the conceptual-I design, which should ensure that conclusions made during the preapplication review will provide a reasonable basis for l 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 l emergency planning. Approaches taken for residual heat L removal and reactivity control are intended to be~ consistent with the accident evaluation categories and consequences.- l-

[ ^ 31 l The Comissioners. .4-1 i j i I The staff proposes to treat the MHTGR, PRISM, and PIUS. designs as advanced reactors =in accordance with the policy ~ i statement. The CANDU 3 design.is-considered to be an. evolutionary heavy-water design deriving from-the larger-- CANDU reactor designs operating in Canada and elsewhere. 2 Therefore, the staff has-concluded that-a prototype CANDU 3 i is not required.for design certification. This position is consistent with staff-conclusions in SECY-89-350, ' Canadian CANDU 3 Design Certification," and SECY-90-133, ' Prototype Requirement for CANDU 3 Design."E The-preapplicant, AECLT, i has. stated that a CANDU 3 reference plant-is a key element-in their plan for standard design certification. If AECLT i holds to that position, the regulatory review and con-i struction in Canada would lead the NRC's ~ design _certi-. fication review. The staff believes that this regulatory review and construction in Canada would greatly benefit our j review of CANDU 3. During the preapplication review,-the staff intends to utilize the foreign operating experience and accident analysis to aid in predictint the expected i behavior-of the CANDU 3 design. AECLT maces no claim of-passive shutdown or decay heat removal capabilities.- However, because of its unique heavy-water, pressure-tube reactor design and evolution under a different regulatory i structure, it does not conform to some current NRC. I regulations. The staff proposes to apply preapplication i review criteria'to the CANDU 3 reactor that'are consistent with ELWR review requirements. The staff intends to use the Commission's guidance on these recomendations to conduct preipp11 cation reviews of the i conceptual designs. Guidance for review of prototype a i requirements for advanced reactors will-follow SECY-91-074,. " Prototype Decisions for. Advanced Recctor Desigas." Consistent with the requirements-of litle.10- of the Code of Federal: Regulations (CFR) Section 52.47(b)(2),- novel safety i features of the advanced reactors and CANDU 3 wil1< be-l - required to be demonstrated through analysis, test programs, t experience, or.a combination of these methods.. Feedback i from the review. process will be factored into recomended revisions to'the_ policy guidance and recommendations = for i - the development of licensing criteria and regulations will L be made after the preapplication _-safety evaluation reports. (PSER) are issued. : AdditionalLissues may be developed b ~during the preap)11 cation review process; they will be: identified in su) sequent Comission papers.< h.- [ 4 3 r, yv, W eN.- w,w4. j, v --,- h > - h w e., n w.w h,' v- .+. -, a- $m.- .,.~.--.........r,, ,-nw,.,-o.rw..wnwr..u,-.us

-..-... -. - ~ --. - -. - -. - - _ l 4 4-.- l j.* 5-The Commissioners In an SRM' dated May'8, 1992, the Commission requested the-p staff to prioritize the issues for Commission review..The staff recommends that-the priority for review be consistent with the PSER issuance-schedules-and requests that direction = L be _provided in sufficientLtime to allow the staff to incor-porate Commission decisions into-the. final PSERs. -Since:the= L PRISM design 1s scheduled as the first preapplication i. b review. Commission attention-is requested on a-highest = priority for those items identified in the enclosure ts appilcable sto the PRISM design. j The staff requests' approval of, or alternate guidance on, 6-

== Conclusions:== these proposed positions to be taken in the preapp11 cation i review of the advanced reactor and CANDU 3 designs. .The Office of the General Counseli has ' reviewed'this' paper Coordination: and has no legal objection.,The: staff has Torwarded a draft of this paper to the ACRS for its review and comments.. Recommendations: That-the Commission? q-Approve the-staff recommendations in Enclosure 11for conduct of-the preapplication reviews.. Approve of the staff's conclusion'that, based on the I position that.the CANDU.3 design is an evolutionary-heavy-water design deriving from CANDU designs-operating L L in Canada and elsewhere,'a prototype CANDU 3 is not-L required for design certification.- L Note that-positions which change as preapplication' [ ' __a review experience is obtained-will be communicated ato' the-Commission and that as the staff: identifies new . issues it will 1nform the Commission. F Note-that'the Commission ~is requested to1 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 thL design information on the advanced reactorLand CANDU 3 designs, and the preliminary nature of.the staff's preapplication 2 -s L j j _r w =r v -v-e t t-1 , E w v w-w -

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J'. 6-The Commissioners 4 reviews, the staff does not recommend proceeding with . generic rulemaking on any of the policy-issues identified in this paper.1 The staff will consider generic rulemaking, as appropriate..as the reviews progress and the staff gains greater confidence in the final design information. James M. Taylor Executive Director for Operations

Enclosures:

1. Policy Issue Analysis 2. Design Summaries 3. List - Reference Documents a L i _.., _,. - -, s.,..

- -= ~ POLICY ISSUES ANALYSIS AND RECOMMENDATIONS 4 As part of a preliminary review of the PRISM, CANDU 3, MHTGR, and PIUS designs, the staff has identified 10 instances where either the staff or the-preapplicants have proposed to deviate from current light-water reactor (LWR) guidance for the review of the designs. This. occurred when existing regulations were not applicable to the technology or when the staff identified. new departures from existing regulations that are considered warranted based on the preapplicants' design and proposed criteria. The staff has grouped the issues into two categories: (1) those issues for which the staff agrees that departures from current regulations should be considered; and (2) those issues which the staff does not believe a departure from current regulations is warranted at this time. The following is a matrix of the issues identifying the plant applicability: CATEGORY ISSUES PRISH 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 1 H. Positive Void Reactivity X X Ste;;ry

1. Control Room Design X

X X X 2 J. Safety Classification X Discussions of these issues are on the following pages, including a brief 2 a summary of the issue, current LWR regulations, preapplicant positions, discussion of staff considerations and a proposed-recommendation for staff action. The staff considered the-preapplicant's proposals in light of applicable Commission policy statements. t ll 4 = DRAFT

w At this preliminary review stage, the staff has limited the scope of the issues to those which could affect the licensability of the proposed design. Additionally, if a similar issue had already been raised for the LWR designs and the staff's advanced reactor design recomendation was. essentially the same, it was not repeated in this paper. In those cases where the preapplicants proposed different considerations from the evolutionary or passive LWRs, the issue is treated in this paper in light of the work done in the advanced light-water reactor policy papers. s 0 + ll i f 1 e j- . DRAFT .... < -, -.,. ~.., -.a 4 ,~

4 A. ACCIDENT EVALUATION 11 8!1 Identify appropriate event categories, associated frequency ranges, and evaluation criteria for events that will be used to assess the safety of the proposed designs. Current Reaulations General Design Criterion (GDC) 4 requires the consideration of accidents in the design basis. Also,10 CFR 52.47 requires the cosideration of conse-quences for both severe accidents (through the required probabilistic risk assessment) and design basis accidents (DBA)-for designs which differ signif-icantly from evolutionary designs or utilize passive or other innovative means to accomplish safety functions. Preapolicants' Accroach All three advanced reactor preapplicants proposed to analyze accidents signif-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 I rem whole body and 5 rem thyroid as their limits for a significant portion level PAG dose limit for all sequences more probable than 5x10'pe the lower-of their accident spectrum. The MHTGR accident guidelines invo per reactor-probable than 10,uidelines invoke the PAGs for accident se year. The PlVS p per-reactor-year. The per reactor-year. The PRISM probable than 10',cn guidelines also limit consequences from any sequence more accident evaluati per reactor-year to the 10 CFR Part 100 dose limits. Guidelines for onsite consequences and offsite consequences from operational transients for all vendors are consistent with or more conservative than present LWR regulations as contained in 10 CFR Part 100. The CANDU 3 preapplicant, in their current safety analyses, has excluded analyses of the consequences of events with frequencies of less than 10 per year from the safety evaluation.. Events which would be excluded from-consideration, based on the CANDU 3 design characteristics and system reliabilities, would include anticipated transient without scram (ATWS), unscranned loss-of-coolant accidents (LOCAs), delayed scram events, and other events which could affect reactivity insertion (for example, from control systemfailures). As a result of the positive void reactivity coefficient associated with the CANDU design, events involving even a relatively short scram delay could result in a core disruption accident. - ORAFT

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 1 this enclosure. As discussed in Section D, the staff is not ready to make a recommendation on whether the Commission 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 4 10 CFR 52.47. The accident evaluation scheme envisioned by the staff examines challenging events to the designs to provide information for a later decision I on emergency planning requirements for advanced reactors and includes consideration of the potential consequences of severe accidents. Addi-tionally, for the multi-module designs (PRISH and MHTGR), the impact of 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 reliabilit/ 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 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 Tf.e staff proposes to develop a single approach for accident evaluation to be applied to all advanced reactor designs during the preapplication review. The approach will have the following characteristics: Events will be selected deterministically and supplemented with insights from probabilistic risk assessments of the specific designs. Categories of events will be established based on expected frequency of occurrence. The selected range of events will encompass events of a lower likelihood than traditional LWR design basis accidents. Consequence acceptance limits for core damage and onsite/offsite releases will be established for each category to be consistent with Comission policy guidance with appropriate conservatisms factored in to account for uncertainties. Methodolegies and evaluation assumptions will be developed for analyzing each category of events consistent with existing LWR practices., ., 4 - DRAFT

1 ~ Source term determination will be performed as 2pproved 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 j 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 l i s 1 0 i u l i l (. [ 5-DRAFT l .,u r., e -,r,- a,.,.

,s B. SOURCE TERM lEin 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 using best estimate phenomenological models of the transport of the fission products from the fuel through the reactor coolant system, through all holdup volumes and barriers, taking into account mitigation features, and finally, into the environs. Current Reaulations Appendix I to 10 CFR Part 50 (ALARA), 10 CFR Part 100 (Reactor Site Criteria, ) which 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. Preapolicants' Acoroach 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 releatu fractions or the behavior of those fission products migrating from the netal fuel through the sodium coolant. MHTGR designers have proposed siting source terms for accidents based on the expected fuel integrity. The cohted microsphere fuel particles in the core are predicted by the preapplicant to contain all the fission products except for that releaseo 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 PIUS designer has proposed using a mechanistic LWR source term. Information has been provided in the Preliminary Safety Information Document (PSID) for fission product concentrations-in both liquid and gaseous effluents. It is expected that PIUS designers will adopt the results of the ongoing EPRI/NRC effort to revise the TID-14844 source term previcusly used for LWRs. ,, DRAFT

l 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. The 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 to Part 100 being considered by the staff replace the present individual dose criteria with a population density standard. A fixed minimum exclusion area radius of 0.4 miles is specified. Other criteria regarding po)ulation protection and seismic criteria factors are also included in 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 compare against applicable safety criteria. Recommendation Advanced reactor and CANDU 3 source terms should be based upon mechanistic analyses, provided that: - DRAFT

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, and 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 source terms for each design are selected to bound credible severe accidents and design-dependent uncertainties. The design specific source terms for each accident category would constitute one component for evaluating the acceptability of the design. 9 o DRAFT -_a

4 4 p, C. CONTAINMENT i Issue I Should advanced reactor designs be allowed to employ alternative approaches to p traditional ' essentially leak-tight" ' containment _ structures to provide for the control of fission product releaselto the environment? Current Reaulations General Design Criterion (GDC) 16 requires that LWR reactor containments provide an essentially leak-tight barrier.against the uncontrolled release of i radioactivity to the. environment, and that containment-associated systems assure that containmentLdesign conditions important to safety are not ex:eeded-2 for as long as postulated accident conditions require. -GDC.38-40 set: i requirements for containment heat removal,- GDC 41-43 for containment i 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.: I Preapplicants' Aporoach i The MHTGR is not designed with a leak-tight containment barrier. The design relies upon high integrity fuel = particles to minimize radionuclide release, and on a below-grade, safety-related concrete-reactor building to provide-retention and holdup of any radioactive releases. The reactor vessel and the steam generator vessel are in separate cavities within-the concrete structure. I In the event of a reactor coolant pressure boundary (RCPB) rupture, _ louvers in the reactor building are designed to allow the passage of gases to the i environment, preventing building overpressure. _The building design does not include containment isolation valves for the ventilation line from the building and hhs an open path to the environment via a drain line in the reactor cavity cooling system (RCCS) panels. - Accident dose calculations assume a constant 100 percent volume per day building leak rate, and take i' credit for plateout on the building walls, a SIUS, above grade, is designed with a low-leakage containment bated on a L pressure-suppression scheme that is integral with the reactor building, similar to the ABWR and SBWR, -Below grade, the concrete pool wall and floor, which is the reactor pressure boundary, and the containment are contig'uous, separated only by a steel membrane.- CANDU 3 is designed with a large, dry, steel-lined, concrete containment, without containment spray. -The maximum leak rate-(used in~ safety analyses) is 5 percent volume per_ day at the design pressure-of approximately 30 psig.. The structure is designed-for a test-acceptance leak rate of 2 percent-per day. at the design pressure.- These;1eak rates are significantly higher than thoseof i a typical U.S. PWR containment. E e o DRAFT / a

The PRISM containment design is a high strength steel, low leakage, pressure-retaining boundary, consisting of two components, the upper containment dome and lower containment vessel. The upper containment is a steel dome, it differs from light-water reactor containments functionally in the following respects. The containment is specifically designed to mitigate the radioactive release consequences of severe events. The PRISM containment volume is markedly smaller than is typical of LWR containmentsi there is little separation between the reactor vessel and the containment boundary; and no safety-grade containment coolers or :; pray systems are provided. The entire containment structure is located below grade within the reactor building. Discussion Each of the advanced designs and CANDU 3 maintain an accident mitigation approach in which containment of fission products is a part. Two of the advanced reactor designs (PRISM and MHTGR grade, providing protection fram external) hazards. place the reactor building below Generally, the advanced designs focus more attention than do LWRs on protecting the plant by providing passive means of reactor shutdown and decay heat removal (DHR). As a result, designers proposed less stringent containment design requirements. The staff recognizes that reactor designs, without traditional containment structures or systems, represent a significant departure from past practice on LWRs, and that existing LWR containment structures have proven to be an effective component of our defense-in-depth approach to regulation. However, the Advanced Reactor Policy Statement ree.ognizes that to encourage incorporation of enhanced safety margins (such as in fuel design) in advanced reactor designs, the Commission would look favorably on desirable design related features or reduced administrative rsquirements. New reactor designs that deviate from current practice need to be extensively reviewed to assure a level of safety at least equivalent to that of current generation LWRs is provided, and that uncertainties in the design (nd performance are taken into account. The staff believes that new reactor designs with limited operational experience require a containment system that provides a substantial level of a accident mitigation for defense-in-depth against unforeseen events, including core damage accidents. This requirement may not necessarily result in a high-pressure, low-leakage structure that meets all of the current LWR requirements for containment, but it should be an independant barrier to fission product release. The proposed criteria will need to provide an appropriate level of protection of the public and the environment considering both the safety advantages of the advanced designs and the lack of.an experience base in evaluating their performance. For evolutionary LWRs, the staff, in SECY-90-016, proposed to use a conditional containment failure probability (CCFP) or deterministic containment performance goal to ensure a balance between accident prevention and consequence mitigation. During the evolutionary LWR reviews, a great deal of careful review was necessary to assure that a probabilistic CCFP would not be used in a way that could detract from a balanced approach of severe accident prevention and consequence mitigation. For advanced designs and the CANDU 3, limited experienc~e exists -{0-DRAFI

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

t D. EMERGENCYPLANNING(EP) hint 1 Should advanced reactors with passive design safety features be able to reduce emergency planning zones and requirements 7 i j 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 Itcense issued under 10 CFR Part 50, 10 CFR 50.47 requires that no operating license be issued unless a finding is made by the NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Currently, offsite protective actions are recommended when an accident occurs that could lead to offsite doses in excess of the Environmental Protection Agency's (EPA) Protective Action Guidelines (PAG), which are 1-5 rem whole body and 5-25 rem thyroid. At the lower projected doses, protective actions should be considered. At the higher projected doses, protective actions are warranted. Prenvolicants' Acoroach The proposed PRISM approach to emergency planning is significantly different from that of previous LWR applications, particularly in the area of offsite EP. A design objective of 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 radiatton, and containment of accidents if they should occur.,

MHTGR proposed reduced offsite emergency planning for similar reasons as those a 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 cr'ticality after being shut down by the doppler coefficient without the reactor protection system functioning (i.e., about 37 hours), and (3) the long time for the fuel and reactor vessel to reach maximum temperatures (i.e.,.about 100 ho'urs) DRAFT

during accidents. The preasplicant asserts that the public around the plant 4 would be outside the area t1at needs to be sheltered or evacuated and, further, there is ample time to notify and move the public during an event. J 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 i Iower level EPA PAGs at 500 meters distance from the containment. No specific i information on emergency planning was provided by the preapplicant for review beyond the general assertion that they intend to limit offsite doses to the PACS. Discussion The advanced reactor designers have objectives of achieving very low probabilities (<l.0 x 10' per year) of exceeding the EPA lower-level PAGs. The vendors claim that these advanced reactors, with their passive reactor shutdown and cooling systems, and with core heatup times much longer than those of existing LWRs, are sufficiently safe that the EPZ radius can be reduced to the site boundary, and that detailed planning and 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. Pecomendation The staff proposes that advanced reactor licensees be required to develop offsite emergency plans. Additionally, provisions for periodic emergency exercises, including offsite exercises, 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 by advanced reactor licensees as an added conservatism-to NRC's ' defense-in-depth" philosophy. Briefly stated, this philosophy: (1) requires high quality in the design, construction, and operation of nuclear plants to reduce the likelihood of malfunctions in the first instance; (2) recognizes that equipment can fail and operators can make mistakes, therefore requiring safety systems to reduce the chances that malfunctions will lead to accidents that release fission products from the fuel; and (3) recognizes that, in spite of these precautions, serious fuel damage accidents can happen, therefore requiring containment structures and other safety features to preven.t the DRAFT

1 release of fission products offsite. The added feature of emergency planning to the defense-in-depth philosophy provides that, even in the unlikely event of an offsite fission product release, there is reasonable assurance that emergency protective actions can be taken to protect the population around nuclear power plants. Information obtained from accident evaluations conducted as outlined in Section A of this enclosure will provide input to the Emergency Planning requirements for advanced reactor designs. Based in part upon these accident evaluations, the staff will consider whether some relaxation from current requirements may be appropriate for advanced reactor offsite emergency plans. The relaxations to be evaluated will include, but not be limited to, notification requirements, size of EPZ, and frequency of exercises. This evaluation will take into account the results of passive LWR emergency planning policy decisions. 4 he e DRAFI y e

E. RfACTIVITY CONTROL SYSTEM lssues should the NRC accept a reactivity control system design that has no control 1 rods? Current Reaulations General Design Criterion (GDC) 26 requires that two independent reactivity control systems be provided. One of the systems shall use control rods preferably using a positive means for insertion. Theothersystemshallbe capable of controlling planned reactivity changes to assure fuel limits are not exceeded. Eteardlicants' Position The PIUS design does not have control rods. However, the preapplicant proposes that the design complies with the intent of General Design i Criterion 26 by having two independent liquid boron reactivity control systems. The normal reactivity control system pumps boron into the primary 2 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 ingrest 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 burnable poisons for power shaping, and limitations in-core size for control of xenon oscillations for slow 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 arevent a scram, but would interrupt'the only safety-grade core cooling mec1anism. Discussion The existing LWR regulations provide prescriptive design guidance for ons-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. The PlVS design does have, however, three ways-to introduce-DRAFT-

~ i 4 F ) quid boron into the core to control and shutdown the reactor. Two of the t' ne rely on flow through the dentity lock from a common supply of borated pe water. The other system is the normal reactivity control system which i ,(. a separate boron tank and is used for normal shutdown. The latter systesa r, only safety grade within the bounds of the containment isolation valves. Recommendat. ion The staff concludes that a reactivity control system without control rods i 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 i c. control of power distribution d. shutdown margin i e. operational control i __c l i 4 1 ,16 - DRAFT.

1 F. OPERATOR STAFFING AND FUNCTION 1111Lt Should advanced reactor designs be ' allowed to operate with a staffing complement that is less than that currently required by the LWR regulations. Current Reculations The NRC has established the requirements for control room staffing in 10 CFR 50.54(m)(2)(iii) which states a senior operator must be present in the control room at all times and a licensed operator or senior operator must be present at the controls of a fueled nuclear power unit. 50.34(m)(2)(1) provides a table identifying the minimum staffing requirements for an operating reactor. Standard Review Plan 13.1.2, Section II.C states that at any time a licensed nuclear unit is being operated in modes other than cold shutdown, the minimum shif t 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). Prespolicants' Position The MHTGR plant is presently four reactor modules with two modules feeding a single steam supply system. The design includes a shift-staffing level of eight persons who would be dedicated to plant operations; a senior licensed shif t supervisor, two licensed reactor operators in the control room, and five roving non-licensed operators. This results in three licensed and five non-licensed operators for four reactor modules. The PRISM control room would contain the instrumentation and controls for all nine reactor modules and their power conversion systems. The objective for the minimum number of operating staff.would include: one SRO shift supervisor, one SRO assistant supervisor, one 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 PlVS main control room would be manned by two R0s and a SRO shift supervisor. The shift supervisor would not be required to be in the control room at all times. The CANDU 3 preapplicant has not proposed a specific number of licensed operators, but the 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 safety DRAFT 1

i il.*. l i'* i features, and the large heat capacity results in reactor designs that respond to transients in a manner that demands less of the o>erator than do the i current operating plants or evolutionary designs. Tie preapplicants assert j that the passive safety features and, in some cases, large coolant inventory of the PRISH, MHTGR, and PlUS designs may not require an operator to act or i l l intervene for several days following an accident. These designs also automate I The vendors of systems that start up, shut down, and control these reactors. l these reactors have suggested that they could be operated with fewer licensed l operators and believe that this would reduce significantly the training and 1 operating costs to licensees. A similar policy issue, Role of the Operator in a Passive Plant Control Room, i l draft policy pa)er on passive 25, 1992,d concern that tie man-machine i was identified in the staff's June in that paper, the staff expresse I reactors. 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-i believes that position is also applicable-to advanced reactors. i Recommendation The staff believes that operator staffing may be design dependent and intends i 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 3 j confirm through test and evaluation the following: Smaller operating crews can provide effective response to a worst case array of power maneuvers, refueling and maintenance activities, and accident conditions. An accident-on a single unit can be mitigated with the proposed number of less one, while all other units could be taken to a licensed operators cold shutdown condltion 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 a complete loss of computer control capability,_ (2) a complete station bl)cko ~ a 1 (3)adesignbasis.seismicevent. 4 The adequacy of these analyses shall.be tested and demonstrated on an-i e l actual control room prototype. 4 . DRAFT l 0 _.,..m.

i = s 1 a l 1 l G. RESIDUAL HEAT REMOVAL h1M j Should advanced reactor designs that rely on a single completely passive, j safety-related Residual Heat Removal (RHR) system be acceptable? Current Reaulations 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 failure within the safety system. Regulatory Guide 1.139 (issued in dr ft for comment), augmenting the GDC, states that the RHR function must be 1erformed to reach a safe shutdown condition within 36 hours of reactor slutdown. Branch Technical Position (BTP) RSB 5-1 states that the RHR function must be performed in a reasonable period of time following reactor shutdown. Prenonlicants' Position The PRISM design uses the reactor vessel auxiliary cooling system (RVACS) as the safety-grade system for residual heat removal from the reactor core. Reactor generated heat is transferred through the reactor vessel to the containment vessel outer surface. RHR is then accomplished through natural circulation heat transfer to the atmosphere. Cooler air flows downward into the below grade reactor silo, where it is turned inward and upward to be heated by the containment vessel outer surface and a special collector cylinder. This heated air then flows out of the silo and is released to the atmosphere. The RVACS is completely passive and always in operation. The RVACS is proposed as a backup to normal non-safety-grade cooling through the intermediate heat transport system, the steam generator, and condenser. If the condenser is not available for cooling but the intermediate sodium loop remains available, then the non-safety-grade auxiliary cooling system (ACS) supplements RVACS. The ACS operates through natural circulation air cooling of the steam generator. The RVACS design basis analysis (performed by the designer) results in high temperature conditions (within design limits) for an extended period of time if no other system is-operated. However, use of the ACS system in conjunction with RVACS can limit peak coolant temperature for decay heat removal to about 15 *C above normal operating temperatures. The MHTGR is designed with only one safety-grade system for removing residual heat from the core, the reactor cavity cooling system (RCCS). The RCCS consists of panels within the reactor cavity and ducts connecting the RCCS panels to four inlet / outlet 1 orts. Redundancy,is provided by these separate ports and a_ cross-connected leader that surrounds the reactor vessel (i.e., any panel can be fed from any inlet and can discharge to any outlet). The RCCS operates by absorbing radiant heat from the reactor vessel to the panels which surround the reactor vessel and transferring the heat by convection to-the air. flowing by natural circulation in the panels. As the heated air rises, cooler, atmospheric air is drawn to the panels-through the inlet ports. There are no active components in the RCCS. The system-is always in DRAFI n

o>eration. The RCCS is relied upon when the heat trans3 ort system (HTS) and tie shutdown cooling subsystem (SCS) are inoperable. T1e 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, use of the non-safety-grade backup RHR systems reduces the frequency, magnitude and duration of high temperature challenges to the reactor vessel, i 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 4 available, a non-safety-grade diesel-backed pump system can cool _ the pool water. DiscusshD 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 j 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 a remove residual heat. PIUS uses natural circulation of water through natural l draft cooling towers for its RHR system. The lack of check and squib valves, the continuous operation and use of a single phase fluid in the system appear to offer increased reliability over the passive LWR systems. However, reliance only on passive systems may lead to high temperature challenges to the reactor vessel and reactor internal structures since higher heat removal rates in passive cooling situations require larger temperature differences between the reactor and cooling medium (air).. Elevated temperatures-(above normal operating values) may exist in the vessel and internal structures for long periods of time. Particularly in the high temperature reactors, the PRISM and MHTGR, creep damage may be more likely as the result of these high-temperature transients. .. DRAFT-

Recommendation As s result of the unique design features of the PRISM, MHTGR, and PIU$ designs, the staff believes that reliance on a single, completely passive, 4 safety-related RHR system may_ be acceptable. In carrying out its future detailed design evaluation, the staff will-assure that NRC regulatory treatment of non-safety-related backup RHR systems is consistent with 1 Commission decisions on passive light-water reactor design requirements. i ) i a 1 4 .a I 4 . DRAFT -

4 l H. POSITIVE VOID REACTIVITY COEFFICIENT j liin should a design in which the overall inherent reactivity tends to increase under specific conditions or accidents be acceptable? Current Regulations General Design Criterion (GDC) 11 requires that the reactor core and coolant system be designed so that in the >ower operating range the net effect of prompt inherent nuclear feedback ciaracteristics tend to compensate for rapid increases in reactivity. Prespolicants' Position In the PRISM design, the maximum sodium void worth, according to the preapplicant, assuming only driver fuel and internal blanket assemblies void, is nominally 55.50. If radial blanket assemblies are included, the sodium void worth is nominally $5.26 which does not include the -70 cents from gas expansion modules (GEM). Should sodium boiling begin, on a core-wide basis under failure to scram conditions with a total loss of flow without coastdown, the reactor could experience a severe power excursion and core disruption. The predicted temperature reactivity feedback is approximately -80 cents prior to the 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 setween $1 and $2. The 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 insertion could lead to a 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, diverso, 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 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 DRAFT

1 i ' 1. either the PRISM or CANDU 3 designs may not necessarily lead to a large scale release of the radionuclide inventore to the atmosphere due to their respective mitigative designs. In tia CANDU 3 reactor, multiple redundant, 3 dive:',e f ast acting scram systems s're provided to address the positive coefi'cients. Attempts to modify the designs to reduce the effects of these positive coef ticients may result in other con',equences 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 Attempts to reduce the PRISM minimizing the reactivity swing over core life. 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 i events without core damage. Recomendation The staff concludes that a positive void coefficient should not necessarily The staff is proposing to require that the PRISM disqualify a reactor design. and CANDU 3 preapplicants analyze the consequences of events (such as ATWS, unscramed LOCAs, delayed scrams, and transients affectinq reactivity control) that could lead to c(.e damage as a result of the positivs vuld coefficients. The staff's review of these analyses will take into account the overall risk 4 l perspective of the designs. -Whether the prensplicants 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. i a mM g 3 4 e DRAFT

1. CONTROL ROOM AND REMOTE SHUTDOWN AREA DESIGN 111E Can current requirements for a seismic Category !/ Class IE control room and alternate shutdown panel be fulfilled by a Remote Shutdown Area, and a non-seismic Category 1, non-Class IE control room? Current Reoulations The current LWR requirements for control room and remote shutdown area design General are provided in 10 CFR Part 50, Appendix A, and 10 CFR Part 100. Design Criterion (GDC) 19 requires that a control room with adequate radiation protection be provided to operate the plant safely under normal and accident conditions and that there be an ability to shut down the plant from outside GDC 17 requires that the electrical system for the control the control room. 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. Prescolicants' 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 e,ot 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 MHTGR design has, for the four modules, a non-safety-related central con-trol room to operate the plant and a seismic Category I remote shutdown area from which to respond to accidents if necessary. Neither the equipment in the control room nor the remote shutdown area are Class IE. The remote shutdown area does not contain safety-related equipment, nor does it include a ventilation system for operator habitability, or a safety-related manual This is based on the preapplicant's position that accidents do not scram. require operator response. The 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, DRAFT

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

I 9 i 1;- I paper defined positions on common mode failures in digital systems and on annunciator reliability. Staff requirements for advanced reactor designs will be consistent with passive LWR policy outdance for these issues, once the cuidance is finalized. Recommendation i i The staff recommends that until passive LWR policy for design requirements of 1 control rooms and remote shutdown facilities is finalized. ",he staff will 2 apply e trent LWR regulations and guidance to the review of advanced reactor designs. This will ensure that plant controls and the operators will be adequately protected so that safe shutdown can be assured in accident i situations. 4 i ( j h i h l 4 i f 2E - DRAF1- "'*"'S wg 4 9 iry w yoww+ y vr t +-'i 4w1-- = > p ..g-mm .e's-e-44yJ-*y e t-w-vwg,y.-wammsg g g.i q v-- '-eww-e-e-w etep--- pr

e J. SAFE 1Y CLASSlFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS 1111Lt What criteria should the NRC apply to the advanced reactor designs to identify the safety-related structures, systems, and components? Current Reoulations Title 10 of the Code of Federal Regulation Section 50.49(b)(1) and the current A?pendix A.Vl(a)(1) of 10 CFR Part 100 list the following criteria to define t u safety related structures, systems and components: a. those needed to maintain the integrity of the reactor coolant pressure boundary (RCPB) J 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. to update AmendmentstoParts50and-100havebeenproposed(57FR47862)forfuture criteria used in decisions regarding reactor siting and design nuclear power plants, including the advanced LWR 6esigns. - These proposed revisions include the temporary relocation of the dose considerations for the current Part 100 guidelines) from Part 100 to reactorsiting(i.e.imeasmorespecificrequirementsaredevelopedregarding Part 50 until such t accident source terms and severe accident insights. Prearolicants' Position The advanced reactor designs rely on a limited number of safety-related l systems to protect the core and the public. Some of these systems are entirely passive, with no moving components and do not-require operator-i action. The vendors believe that this reduction in safety-related equipment L ~ results in simpler plant designs with lower costs. This also results in many l' 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 destg' ns, only the MHTGR design is not using the current LWR criteria above'for safety classification. For the MHTGR design, i 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):RCPB is not en -(1 l s used l to pressurire and depressurize the RCPB, (3) the coolant moisture monitor is-l- not safety related, and (4) neither the control. room or remote shutdown area DRAFT l' l L

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

A. CANADIAN DEUTERIUM URAN 1UM (CANDU) 3 REACTOR DESIGN Development History The CANDU 3 is the latest version of the pressurized heavy-water reactor (PHWR) system developed in Canada. The CANDH 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. Atomic Energy of Canada, Limited, Technologies (AECLT) Dn May 25, 1989, informed the NRC of their intent to submit the CANDU 3 reactor design for standard design certification in accordance with Part 52. AECLT of Rockville, Maryland, is a wholly-owned subsidiary of Atomic Energy of Canada, Limited (AECL) (a crown corporation of Canada), and is the preapplicant for the AECL in Canada is also pursuing standard design certificatfor, CANDU 3 design. of the CANDU 3 with the NRC's Canadian counterpart, the Atomic Energy Control Board of Canada. AECLT's current plans are to submit a standard design certification application for CANDU 3 in the 1995-1996 time frame. Desian Description 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 This results in a lower core power density for CANDU 3. In addition, power. the CANDU 3 core is neutronically loosely coupled which results in xenon induced flux tilts that requires a relatively complicated computer operated a 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 celandria tube. The pressure tubes form part of the reactor coolant system pressure boundary. Because of the low excess reactivity associated with a natural uranium core, i ~IATT D Enclosure 2

O the CANDU design must be fueled on a continuous basis during power operation by an automatic fueling machine. On-line fueling is the primary means of changing reactivity in the CANDU 3. Fer the CANDU 3 design, heavy water coolant flow through the core is uni-directional, thereby facilitating oreline fueling from one end of the reactor with a single fueling machine. The primary system operating pressure is maintained by a pressurizer connected to one of the (nominally 1435 psig) ANDU 3 light-water secondary system is similar to that outlet headers. The C of a PWR. The fuel channel assemblies are enclosed in a horizontal, cylindrical vessel low-)ressure, called a calandria that contains the low-temperature (140 'F),ith t1e integral The calandria vessel, in conjunction w heavy-water moderator. end shielcr supports the horizontal fuel channel assemblies and the vertical and horizontal reactivity control unit components. The C'NDU 3 utilizes four reactivity control systems for reactor control and shutd vn during normal ope at;on, ar4 two redundant and diverse safety-grade shutdown systems are A separate moderator heat n ec for r u ctor shutdown followinp. a transient. remort.1 system ensures that the moderator remains subcooled. All systems in the CANDU 3 design are assigned to one of two groups - either The systems of each group are capable of shutting down Grrap 1 or Group 2. the reactor, maintaining cooling of the fuel, and providing plant monitoring capability in the event that the other group of systems is unavailable. Group 1 systems are those primarily dedicated to normal plant power pro-duction. The Group 2 systems include four special safety systems and other safety-related systems. These maintain plant safety in the event of a loss or partial loss of Group 1 systems, and mitigate the effects of accidents, including the design basis earthquake. Tie Group 1 and Group 2 systems are, to the greatest extent possible, located in separate areas of the plant. CANDU 3 employs two fast-acting, redundant, ano diverse Group 2 shutdown systems, separate from the Group 1 reactor regulating system. Shutdown System No.1 (SD51) consists of 24 vertically inserted control rods. Shutdown System No. 2 (5052) consists of six horizontal nozzles through which a gadolinium nitrate solution is injected. Both shutdown systems inject into the low-a pressure moderator, precluding a rod ejection accident. In addition to the two shutdown systems, the remaining special safety systems include containment and eurgency core cooling system (ECCS). The CANDU 3 containment systein includes a reinforced concrete containment structure with a reinforced concrete dome and an internal steel liner. The containment is designed with a test acceptance leakage rate of 2 percent per day. ECCS supplies light-water coolant to the' reactor in the event of a loss-of-coolant accident. Each of the four safety systems is required to 3 demonstrate during operation, a dormant unavailability of less than 101 or about B hours per year, and be physically and functionally separate from the normal process systems and from one another. The CANDU 3 shutdown cooling system is designcd to remove heat from the HTS at nominal operating temperature and pressure. - DRAFT'

= -. _ J. B. MHTGR (MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR) Develooment History The Modular High Temperature Gas-cooled Reactor (MHTGR) was sroposed to NRC by the U.S. Department of Energy (DOE) in 1986 in response to tie Commission's Advanced Reactor Policy Statement (51 FR 24643). A preliminary safety information document (PSID) and twelve amendments were submitted from October 1986 to March 1992. The PSiD and 10 amendments were reviewed by the Office of Nuclear Regulatory Resear;h and a draft preapplication safety evaluation report (PSER) was issued by NRC in March 1989. DOE recently advised the NRC that the MH1GR 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 Ac" of 1992 requires DOE to submit a preliminary design approval application by September 30, 1996. a 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 d$velopment resulted in the 40 MWe Peach Bottom 1, which operated from 1967 to 1v74, and the 330 MWe Fort St. Vrain, which operated from 1976 to 1989. There have been about 50 gas-cooled reactors in the world totaling about 1000 reactor-years of operation. In this total, there has been about 50 reactor-years of experience with the HTGRs. The BISO and TRISO (trade names) multi-layered microsphere fuel form is used in H1GRs. The BISO fuel form, a fuel kernel with two major layers, was used in Peach Bottom 1; and the TRISO fuel form, a fuel kernel with four major layers (including a silicon carbide layer), was used at fort St. Vrain. The TRISO fuel form provides higher fuel integrity requirements than the BISO fuel and is the reference fuel for the MHTGR. DOE maintains agreements with Germany and France for the exchange of technical information concerning the integrity of the reference MHTGR fuel, and experiments will be conducted in france. As part of DOE's Technology Development Program for the MHTGR, post irradiation testing of development fuel at Oak Ridge National Laboratory is being performed, and a technical information exchange agreement was a established with Japan, which is building an experimental HTGR. Major trends in recent HTGR designs, including the MHTGR, are the following: (1) increased system pressure, (2) steel pressure vessels for the smaller HTGRs. including the MHTGR, versus the prestressed concrete reactor vessel for larger HTGR designs as fort St. Vrain, (3) pro)osed greater fuel integrity, 4 with a 6 x 10 fraction of failed fuel assumec for the MHTGR, and (4) lower enriched uranium fuel. Desion Descriotion The standard MHTGR plant is four reactor-steam generator modules and two steam turbine-generator sets. -Each module is designed for a thermal output of 350 MWt. Two reactor modules are coupled to a steam turbine-generator set to i produce a-total plant electrical output of 540 MWe, DRAFT

i, The low-power density (5.9 watts /cc) reactor core is helium ccoled and graphite moderated, and uses ceramic coated (four sajor layers) nicrospheres in an organic bonded cylindrical compact as the fuel. 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 high as 2900 'T without significant fission product release. The compacts are placed in small vertical holes in the hexagonal graphite d 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 i j phase coolant chemically and neutronically inert. 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, i 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 2005 'F. The secondary side pressure is higher (about 2500 psig) than that on the primary side (about 92o psig), so water would leak into the coolant with a steam generator tube leak or failure. t i Reactor protection is provided by two safety-related reactor protection systems (control rods and boron carbide balls), which are diverse and redundant, and one non-safety-related system (control rods). The non-safety-related system is independent from and redundant to the-safety-grade systems. The equilibrium shutdown core temperature would be approximately 250 'F, the design temperature for refueling. The safety-related RCCS is a set of panels-surrounding the reactor vessel with a header connection to four inlet and exhaust ports above ground. This allows hot air to rise thus removing heat transferred from the reactor vessel while cold air is drawn from outside into the panels. The' system 1) is entirely passive with no moving components, (2) is always operating, (3) automatically ( a 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 ot1er non-safety-related, active heat removal systems:.(1) the shutdown cooling system in-the bottom of 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 isolatio'n valves on the steam generator . DRAFT i e < ,<n,~,, 9 w r-w- n s,a--

4 main steam and feedwater inlet piping. It will not retain the gases from a rapid RCPB depressurization, but is designed to have a leak rate of less than 100 percent / day after initial depressurization. 4 e=muin. 2 4 l DRAFT ~ [:_ --

t 1 C. P!US(PROCESSINHERENTULTIMATESAFETY) Historical DeyM o m d ) The Process inherent Ultimate Safety (PlUS) reactor is being designed by ASEA extensionofthenABB-Atom'). The concept evolved in the early 1980's from an Brown Boveri Atom (ABB-Atom q s low temperature district heating design. In October 1989, ABB requested a licensability review of the PlVS design in accordance with NUREG-1226, and in May 1990, ABB submitted the PlVS preliminary safety information document (PSID)ign in the for staff review. ABB plans to apply for design certification of the P!US des 1994-1995 time frame, assuming a favorable preapplication review. l The PlVS design concept has already undergono tests related to the design principles. ABB has completed testing using the MAGNE Test Rig to simulate PlVS parameters such as diffusion and mixing across the primary loop / pool c boundary with consideration for effects of turbulence, stratification, migration of boron, and others. Large scale tests of the P!US 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 1 design's safety and transient performance. ATLE was a full height simulation of the PlVS pool. Other tests of the PlUS 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 purpose of design optimization, as well as special component testing. It is planned that this larger test rig will serve as the basic test facility for developing data for the detailed design and verification. Desian Description P!US is a 640 MWe advanced pressurized water reactor (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 PlVS design is completely passive and is-always in operation. The PIUS design consists of a vertical hollow cylinder, the reactor module, which contains the reactor core. The reactor module is submerged in a large concrete reactor vessel containing 3,300 M3(870,000 gallons) of highly borated water. The reactor module is t open to the borated pool at the bottom and at the top of the reactor module. 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 bottom density locks. There is no shysical flow bariier in the density locks between the primary loop and the aorated pool, however, the difference in density between the primary loop-reactor water and the cooler borated pool water provides a relatively stationary interface. - When sufficiently upset during transient conditions, such as loss of flow or a power mismatch,~ the density difference is overcome l ~

DRAFT

and the borated water flows into the core and shuts down the reactor. A natural circulation flow path is then established from the borated pool through the lower density lock, up.through the core, and back into the borated pool through the upper density lock for long term shutdown cooling. Unlike most reactors, PlVS does not employ mechanical control rods for regulating reactivity. Reactivity is controlled by the boron concentration and temperature of the primary loop reactor water. An active reactor protection system (RPS), with associated instrumentation and actuation systems, is also provided in P1US. 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 PlVS design are similar to the passive LWRs being considered by the staff (AP-600 and the SBWR). Although PIUS is a PWR, its operating pressure (1,305 psi) is close to that of a BWR. The proposed containment for the PlVS design is integral with the reactor building, similar to the ABWR and SBWR. Leak-rate has been defined as not to exceed 1 vo19me percent per day at a design pressure of 26 psig. The acceptance leakage value is expected to be 0.5 percent at design pressure. hm e DRAFT

l. i D. PRISM (POWER REACTOR INNOVATIVE SMALL MODULE) 4 4 Develoneent History 4 i 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 l team of commercial engineering firms. Research and development support is J being supplied by the Argonne National Laboratory Energy Technology Engineering Center, Hanford Engineering Development Laboratory, and Oak Ridge ) In addition, a steering group of utility representatives i National Laboratory. I was involved in the PRISM design effort. DOE chose to sponsor the PRISM design as part of its National Energy Strategy because of the design's potential for enhanced safety through the use of passive safety systems and greater safety margins, reduced cost through 4 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 1 l components design, reliability data, and safety assessments. This experience consists of operation of a number of facilities such as, EBR-II, Phenix, the i i Fast Flux Test Facility _(FFTF), the Joyo reactor in Japan, and others. The PRISH Preliminary Safety information Document (PSID) was submitted to the i i NRC for review in November 1986, and the results of an early_NRC staff ~ review was the draf t 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 ALMR by 2010, as called for in the Energy Policy Act of 1992. a plant Dfferiotton l The PRISM plant design consists of three se>arate power blocks each made up of three reactor modules. Each module has-a tiermal output of 471'MWt and an electric output of 155 MWe for_ a total-(plant)from a power-block. output o l' design contains three turbines, each supplied. 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 digital processing. L I DRAFT - - - = -.9.m.--.-,w.., -, -,---.,,,.,,y.w -,wrur---y% n n,.e, r-r er i-,r y-

The reactor module consists of the containment system, the reactor vessel, the core, and the reactor's internal componerts. The reactor vessel encloses and supports the core, the primary sodium coolant system, the intermediate coolant system heat exchangers (IHXs), and other internal c.,mponents. The vessel is located just inside the containment vessel, which is located belnw 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 s>ovement during seismic events. The upper head of the reactor vessel is the closure head. The tiosure 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). fhe primary sodium loop is contained completely within the reactor vessel, which is hermetically sealed to prevent leakage of the arimary coolant. The EH pumps provide the primary sodium circulation. Syncironous machines provide flow coastdown capability to the EH 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). lHTS srdium is circulated by a centrifugal pump. The IHTS operates at a higher pressure than the primary loop so that, in ces6 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 !His 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 swelling caused by high burnups. The PRISM core is a heterogeneous arrangement of driver fuel and blankets. The PRISM core design is st'ch 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 void coefficient dominates and a net positive feedback can occur. In all other situations without extensive voiding, an increase in temperatures produces negative feedbacks from Dop)1er and thermal expansion of the core and related structures that dominates tie positive moderator density coefficient. The net negative temperature coefficient is so large that analyses predict all non-boiling transients and accidents to be terminated by the temperature. feedback reactivity at temperatures low enough to not threaten fuel or vessel integrity. This passive shutdown function allows the reactor to sustain all non-boiling transient scenarios without damage, even with a failure to scram. ' 01ATT

3 s There are six control rods in the main reactivity control and shutdown system. Inserting any one of the six will shut the core down. The control rods can be the plant control system (PCS) for normal insertion, inserted using (1)de reactor protection system (RPS) for rapid insertion, and

2) the safety-gra

3) gravity drop into the core.

If both the normal and safety-grade systems ail, the operator can activate the ultimate shutdown system (USS) which sends boron balls into the central location of the core causing shutdown independ-ently of the control rods. The PRISM design also includes passive mechanisms for controlling reactivity: three gas ex)ansion modules (GEMS) consisting of tubes, closed at the top and open at the )ottom, and filled with helium. If the pumps are running, the static pressure is high, causing the sodium level to rise to a high point in the GEH, However, with the pumps off, the static pressure and sodium level drop, which increases neutron leakage. The reac-tivity change provided by the GEMS between these two states is about -70 cents. Normal shutdown cooling is achieved with the non-safety-grade condenser. If the condenser becomes unavailable, the safety-grade reactor vessel auxiliary cooling system (RVACS) is used for RHR. circulation air cooling of the containme The RVACS The design-basis RVACS event assumes that the normal and auxiliary heat removal systems, as well as the Intermediate Heat Transport System (IHTS) sodium, are lost immediately following reactor and primary EH pump trips. The preappilcants' analysis has shown that the RVACS heat removal rate is :ufficient to maintain fuel temperatures within acceptable limits, and temperatures of the internal structures within the reactor vessel under American Society of Mechanical Engineers (ASME) Level C conditions. The PRISM design also contains the non-safety-grade auxiliary cooling system (ACS) to assist the RVACS. The ACS uses natural circulation within the steam generator (SG) to remove heat indirectly from the reactor vessel, and 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 f avorable combination of viscosity, thermal conductivity, and vapor pressure associated with the use of sodium to remove heat, (2) the ability to operate at essentially ambient pressure, thus reducing the pressure exerted on the coolant system boundaries, and (3) operation far below the sodium boiling temperature, thus obtaining the operational and analytical simplicity associated with a single phase coolant. . DRAFI f

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, 3 1989 ( 1 SECY-90-055, ' PIUS Design Review," February 20, 1990 NUREG-1338, " Draft Preapplication SER for the'MHTGR" NUREG-1368, ' Draft Preapplication SER for PRISM" NUREG/CR-5261, ^ Safety Evaluation of MHTGR Licensing Basis Accident Scenarios

• NUREG/CR-5364, " Summary of Advanced LMR Evaluations-PRISM and SAFR' NUREG/CR-5514, 'Modeling and Performance of the MHTGR Reactor Cavity Cooling System" NUREG/CR-5647, ' Fission Product Plateout in the MHTGR Primary System' NUREG/CR-5815. " Evaluation of 1990 PRISM Design Revisions" i

I 4 i 0 RAFT. . L

s... m r s., o POLICY ISSUE November 23, 1992 (Information) SECY-92-393 For: The Commissioners 4 3 from: James M. Taylor Executive Director for Operations I Subiect: UPDATED PLANS AND SCHEDULES FOR THE PREAPPLICATION REVIEWS i 0F THE ADVANCED REACTOR (MHTGR, PRISM, AND PIUS) AND CANDU 3 l DESIGNS

Purpose:

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. 4 Beckaround: In SECY-91-161, " Schedules for the Advanced Reactor Reviews .and Regulatory Guidance Revisions," the staff informed the l Commission of the following estimates for completion of the l preapplication reviews: l PRISM November 1992 HHTGR December 1992 i CANDU 3 June 1993 PIUS July 1993 i The staff based these dates on broad planning assumptions i a including the prea cation schedules, pplicants' design certification appli-i availability of the Office of Nuclear i Reactor Regulation (NRR) resources to conduct the reviews, timely receipt of information to support the reviews, and a i scope of review consistent with the previously issued draft preapplication safety evaluation reports (PSERs) for the PRISM and MHTGR designs. i Contacts: NOTE: TO BE MADE PUBLICLY AVAILABLE Thomas H. Cox, NRR IN 10 WORKING DAYS FROM THE j 504-1109 DATE OF THIS PAPER Brian W. Sheron, RES 492-3500 Wh,+ vw% Dr r) =

i w !. 7 (v ~ The Comissioners 2-Subsequent to issuing the-estimates in SECY-91-161, a number-of-factors have necessitated a revision to the schedules for j L PRISM, MHTGR, CANDU 3,.and PIUS preapplication reviews.- Some preapplicants have extended their design' certification 'l application schedules or modified their proposed designs. _ 1 l Most staff technical review resources have been redirected l to higher priority operating reactor.and design certifi-cation reviews. Additionally, implementation of the fee Recovery Rule'has resulted in preapplicants-desiring _ a preapplication review scope ' limited to key-certification / licensing issues.-- In February 1992, the. staff issued letters to all_ four i i preapplicants requesting confimation of their plans-for design certification and, in some' cases, a schedule for-submitting additional preapplication review information. All. responses were received by May 1992. During the April 21, 1992, briefing on advanced reactor l reviews, the staff-advised the Comission that due to other higher priority work,' a relatively small number of-staff i would be dedicated to conduct the prea)p11 cation reviews. l The staff proposed to-concentrate on tiose key policy istues requiring-Comission guidance and revise the_ preapp11 cation - review content and schedule to more effectively.use the reduced NRR technical review resources. j, i In June and July 1992, the staff held pub 1_ c meetings with each preapplicant to discuss the relevant scheduling-infor-- mation,-the desired scope of preapplication review,- and. schedules for additional = submittals. ' The-staff also-- informed the preapplicants of key policy issues that the. 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-r 1-l- scope of all preapplication reviews would be similar to-the scope of the' PRISM and MHTGR. preapplication reviews . documented as NUREG-1368, " Draft Preapplication Safety . Evaluation for Power Reactor-Inherently Safe Module Liquid: i Metal' Reactor," and NUREG-1338. " Draft Preapplication. Safety I-Evaluation for the Modular High-Temperature Gas-Cooled = i Reactor." l Discussion: The. staff considered several factors in developing a revised' schedule for conducting the preapplication reviews..The enclosure to this paper provides a design-specific _ sumary [ a . m m.. .=..-.

V,, - 3 .ps ['.' The. Comissioners -. } of this information and the staff's rationale for sequencing l the reviews. Based on this rationale and consideration of: available resources, the staff has identified.the following - revised estimates for completing the preapp11 cation reviews: PRISM -December 1993-CANDU 3 December 1994 PIUS April 1995 MHTGR December 1995 4 The staff expects to follow the same review process for approval of the PSERs as is being used for the safety-evaluation re> orts on the evolutionary light-water reactor designs and tie Electric Power Research Institute Utilities Requirements Documents. Approximately 6 months before i completing the review, the staff will-submit a. draft final l'- PSER to the Comission. With Comission consent the staff will forward the draft-final PSER to the preapplicant, the 1.Jvisory Comittee on Reactor Safeguards-(ACRS), and the NRC-Public. Document Room.. After considering input during public l meetings ~ with ACRS and the preapplicant, a final-PSER will i be forwarded to the Comission for approval. } The staff intends to conduct most of the preapplication reviews with staff from the NRR Associate Directorate for l Advanced Reactors and License Renewal (ADAR), national l laboratory technical assistance,-- and support from the Office l l-of Nuclear Regulatory Research (RES). Due to-limited staff resources each design _ PSER will be developed _in a sequential L order. NRR technical staff within the Associate Directorate for Technical: Assessment (ADT) will, in general, not participate in the preapplication review iADT technical staff resources are currently required for higher priority 4 operating reactor technical: reviews and: light-water reactor a (LWR)-design certification. ' However,' once the draft final l PSER has been developed, ADT-management will: review the L = report for its-policy implications. The staff considersi this approach-appropriate since the preapp11 cation review- _ considers the conceptual design, 'and final technical decisions on-safety will not-be made. until-the designL certification review when the ADT technical staff will be. L L -involved. The staff. believes that the changes'to the preapplication-review; scope and schedule, andlthe approach for conducting 6 'NRR technical review, provide 1the most effective use of NRC. resources. ' The. proposed schedules will allow the staff to provide a: timely response to the-preapplicants in important. - areas regarding their design certification application plans. :By'emphasi. zing _the-key policy issues for the - t-w-- 1 .~, .m ..i my --,M_.. ,.m m, .,J..~.~.J. ...,-.--..,.r..,.w . e. -, M., J

A* j The Commissioners - advanced reactor designs, the staff will address the preapplicants' most significant questions about NRC's j licensing requirements. Resolution of these issues in the preapplication reviews is expected to allow the preapplicants to reduce the current uncertainty regarding i design and design certification schedules, The staff will continue its assessment cf 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 Commission as appropriate. The staff will, within the next few weeks, submit to the Commission a draft paper on key policy issues affecting the advanced reactor and CANDU 3 designs. ' Commission 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. Erla: 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 5- 'mit this paper to the appropriate subcommittees, the L:? ice of Management and Budget, and the Department of Energy. s' $?T xecutive Director for Operations

Enclosure:

Summary of Input for Schedule Revision 9 DISTRIBUTICH: Conunissioners OGC OC1J\\ OIG OCA OPA OPP EDO ACRS ASLBP SECY

f.. -

p., fs / J., g - In March 1989,-the NRC issued NUREG-1368,e* Draft Preapplication Safety-1 i Evaluation Report for Power Reactor Inherently Safe Module Liquid Metal L Reactor,".in which it summarized the results of'its review of the Preliminary Safety Information Document (PSID). submitted in 1986;- In March 1990, the U.S. Department of Energy (DOE) submitted Appendix G to the PSID, " Responses to-Issues in Draft SER," in which-it' proposed several significant changes to the PRISM design. These. changes-included increasing the power, adding an ultimate shutdown system and containment dome.. redesigning the reactor to add gas: i l expansion modules _ (GEMS), and changing to a single-wall-tube helical coil steam generator. Brookhaven National Laboratory (BNL) reviewed the revised design and published its findings in NUREG/CR-5815. " Evaluations _ of 1990 PRISM i Design Revisions."~ BNL is also reviewing the performance of the GEMS and the I 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 i 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. l In a letter of March 12.-1992, DOE submitted the following schedule: l Preliminary Design Approval Application CY 1995-Prototype Final Safety Assessment Report-1997 l Standard: Design-_ Certification Application 2003-l- (After prototype testing) 4 This schedule appears consistent with-the Energy Policy Act.of 1992 which I 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.-- In June 1992, the National Research Council of the National Academy of Sciences published a report, :" Nuclear Power:: Technical and Institutional I" Options for_the Future," in which it discussed prerequisites needed to- } preserve the U.S. nuclear power option and-recommended that the Federal-1 government support key reactor designs.o The Council recommended that the-l - PRISM design be the only preapplication design to receive government funding i_ 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 F _ November 1992. DOE noted that'it had submitted all requested information to L the.NRC to: support the preapplication review and requested that the NRC l' complete the; PSER as soon as possible to support DOE in planning for design j certification. - DOE recently-notified the NRC of-problems found with the [ l'- i=r ..m. Zw-.hA.-i._--,-.-. .~.a~ alm. .-r-~ -L-.i.a--._a.-~~-~..-,.m~~-

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• w reference fuel during testing at the Argonne National Laboratory. DOE indicated it may:need to redesign the fuel. The staff does not know how this decision will-impact the PRISM design certification schedule.

The staff plans to conduct the PRISM preapplication review first.to capitalire' j on the work already completed and currently in )rogress. DOE _has provided all necessary submittals-to support the review and 1as been responsive to staff questions during the review. DOE's plans for preliminary design approval application in CY-1995 are-supported by.the National Research Council's recom-j 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. I tLHIGE j In March 1989, the NRC issued NUREG-1338,1"Draf t 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 i Amendments 11, 12, and-13 to the NRC. These recent amendments provide additional information for NRC's rev.iew of the originally proposed. design. I 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 letier 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 i DOE New Production Reactor (NPR). In September 1992, DOE and the U.S. Department of Defense agreed to defer the NPR program and close the design efforts. --The MHTGR schedule depends primarily on the gas-cooled NPR program . which. technically supports much of the MHTGR design. ' DOE'wants the final PSER issued by April 1993 to support resolution of the key MHTGR policy issues

a identified in NUREG-1338. DOE asserted that itsi submittals include all the i

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 1, to understand 'that. the MHTGR is a~ viable power-reactor. concept. i 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 - i 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-. l l _2_ ,-r ,, I o[*--, .,s, ..~Em ~.--u~---,~,, ~~--.L .4 ,,w.. + m.-...,.-,,

l ,r # -s DOE has discussed plans.to revise the MHTGRl 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 increaseu will not affect.the key policy-issues for"the design. DOE recently. informed; NRR of problems in testing the reference MHTGR fuel. The preliminary failurei 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 MHTGR 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 approach for the MHTGR design in-advance'of the final PSER.- CANDU 3 On May 25, 1989, Atomic Energy of Canada, Limited. Technologies (AECLT) 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 - Atomic Energy of Canada, Limited,-(AECL)-in Canada, has supported the CANDU 3 preapplication review by submitting a Technical: Description, Conceptual Safety F Report, Conceptual-Probabilistic Safety Assessment, and several technology-l transfer reports describing the CANDU design.e 7 In a letter of March-18,1992, AECLT informed the NRC that it could support a standard design certification application in:1995 or-1996.if the'NRC' completed its preapplication review-of the CANDU 3 by June 1993. Dn June 29, 1992 L AECLT gave the staff a schedule of-submittals-to-support the preapplication: I review.- AECL has completed much of the final design-work for the CANDU 3' j-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 Septemtrer 1992, AECL acknowledged that-it would re-ev_aluate. its! design certification plans'.in the-U.S. if Canadian construction plans _did not materi-l a j. alize. i l The National Research Council report identified the CANDU 3 design as a nature design 'that could~ be licensed this century. The report noted that the' l-licensing process could be lengthy because of the difference in regulatory L l requirements between the U.S. and Canada.s The Council did notLfind sufficient ID . advantages with the design to justify DOE support for design certification. The staff has started-some preapplication review on:the CANDU 3 design. =NRR is conducting,two projects at DOE laboratories: a study of the CANDU13.

positive voi6 reactivity coefficient and a review of the operation of-the on-line refueling machine.

RES has completed a systems study to identify k candidate event sequences for required safety-analysis, and it has_ projects-to o assess data base adequacies, to perform preliminary transient-calculations = R i l~ L - s =. = - -

4 .J {? 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 the consequences of a:large break loss-of-coolant. releases, NRR is reviewinglure to shut down. NRR is performing this work to accident (LOCA) with a fa support the Commission's decision on'a: key-policy issue:L 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. i information on the event and its consequences.; + a .AECLT is having problems getting proprietary information-released from Canada-- to the U.S. _ This has delayed the staff in obtaining Canadian' codes thus Code interrupting RES's work to use these-codes for preliminary calculations. work is now on the critical path for completing the preapplication review,:and; I the lack of timely submittals of other proprietary information could further In a letter _ dated: September 23, 1992, the staff-delay the review schedule. : informed AECLT that an inability to' transfer proprietary material to the U.S. 2 l may affect the proposed CANDU 3 preapplication. review: schedule. AECLT:is.now - pursuing transfer of proprietary material directly from AECL to-the NRC. L i 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 I i for the review from Canada to._the United States. l l ELMS i In October.1989, Asea Brown.Boveri (ABB) Atom requested that the NRC-perform a ~~ licensability review of its Process-Inherent Ultimate Safety-(PIUS)' plant-ABB Combustion Engineering (ABB/CE) of Windsor, Connecticut, is the design. direct representative of ABB= Atom inithe U.S.,-and is-the official preappli- . cant of record. In May 1990, ABB/CE submitted a' five volume 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" l~ completed a systems study to-identify candidate event sequences for_ required: safety analysis,.and it has projects to assess data base adequacies,l to perform preliminary transient ~ calculations using the existing TRAC code, to identify code needs~ for future independent analyses, tocinitiate severe accident analyses with NRC codes, and to; prepare for source term determi-RES will also provide in-house analytical capabilities-for itself. and nation.'.the PIUS design. NRR for j 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-ire L plication safety-evaluation report:(PSER).by April 1994 that does not requ ~ 4 - 4'-- [r i k . s

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the commercial environ-significant design changes to the PIUS design, and (2)is negotiating with the ment at that time is favorable to that decision. ABB 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 dischargt to the suction of each of the four reactor coolant pumps. Activating the valves is expected to result in a rapid and uniform insertion of boron by a means redundant and diverse from the passive scram process. The passive scram through the density locks will still be the ultimate shutdown process. ABB plans to submit the design change in a November 1992 supplement to the PSID. ABB also plans to submit a PRA supplement in early 1993 for the PIUS design. The National Research Council report concluded that the PIUS design would not likely be ready for commercial 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 preapplication review because the design is presently at the conceptual stage and the experimental data base for the design is still being developed. ABB is most interested in obtaining feedback on the implementation of the key policy issues for the PlVS Continued emphasis by the staff in obtaining Commission guidance for design. resolution of these issues will provide ABB feedback in advance of the final PSER, Conducting the PlVS review third will allow ABB time to develop the design more fully and respond to staff questions without impacting the preap-plication review schedule. a l .}}