ML20126F641

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

ML20126F641
Person / Time
Issue date:	12/16/1992
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Williams P ENERGY, DEPT. OF
References
PROJECT-672A NUDOCS 9212310033
Download: ML20126F641 (56)

Text

I t

># R80p I + 'o,, UNITED STATES l

! > g NUCLEAR REGULATORY COMMISSION g C WASHINGTON, D. C. 20555 December 16, 1992 k ...+ /

v Project No. 672 Mr. Peter H. Williams, Director HTGR Division Office of Advanced Reactor Programs U.S. Department of Energy, NE-451 Washington, DC 20585

SUBJECT:

COMMISSION PAPERS ON POLICY ISSUES AND SCHEDULES CONCERNING THE PREAPPLICATION REVIEWS OF ADVANCED REACTOR AND CANDU 3 DESIGNS Enclosed are two papers which should be valuable to your continuing regulatory efforts. In mid-1992 the staff discussed with you its intent to identify key policy issues and the projected schedules to complete preapplication reviews of these designs. The staff noted that these policy issues and the projected schedules would be addressed in separate papers to the Commission.

As a result of the staff's reviews, an assessment of projected resources, and the meetings held with you and the other preapplicants for the advanced reactor and CANDU 3 designs, the enclosed papers have been provided to the Commission: (I) a draft paper providing the staff's positions on 10 policy issues, and (2) a final paper, SECY-92-393, " Updated Plans and Schedules for the Preapplication Reviews of the Advanced Reactor (MHTGR, PRISM, and PIUS) and CANDU 3 Designs," on the staff's proposed schedules for the preapplication reviews.

The paper on the policy issues is a draft because the staff has not yet obtained Commission approval on these issues. The staff will be meeting with the Advisory Committee on Reactor Safeguards (ACRS) to discuss these issues in the near future. The staff will include the views of the ACRS and document its final recommendations in a revised paper before seeking the Commission's approval. Any comments you may wish to offer will be considered as we prepare our final positions. Please submit any comments by January 25, 1993.

The proposed schedule paper reflects the staff's assessment of its resources and the needs of the preapplicants. The staff will continue to try to expedite its reviews and complete the work ahead of schedule.

3000K.

g g R{M M ORY C M

-' " " ~

9212310033 921216

- /m)gfV

}Ql PDR PROJ I b72A PDR Ij J

y Mr. Peter M. Williams 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,-Jack Donohew.

Sincerely, 4

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 HMSlosson THCox EDThrom JNDonohew OGC- 15/B/18

'ACRS(10) P-315 ,

OPA LLuther MHTGR R/F

.R (

LA:PDLR:ADAR' PM :ADAR SC:PDAR:ADAR SE:PDAR:ADAR, D:Pl . AR ( 6' Lluther JN ohew:sa THCoxTf-- EDThrom7ftfar HMSlos n D rutchfield

/. /92 g/g92. /A//r/92 /.2//5/92 fpip92 /2//6/92

--OFFICIAL-RECORD COPY-Document Name: MHTGSECY.LTR

1 i

.' 1

]

. 1 Mr. Peter M. Williams December 16 1992 !

MHTGR Project No.672 cc: Jerry D. Griffith, Director Dave A. Dilling-Office of Advanced Reactor Programs Bechtel National, Inc.

U.S. Department of Energy, NE 50 Beale Street Washington, D.C. 20545 P.O. Box 193965 San Francisco, Chlifornia 94119 Sterling Franks Technical Director Office of MHTGR. Walter J. Parker New Production Reactor Program Stone and Webster Engineering U.S. Department of Energy, NP Corporation Washington, D.C. 20545 245 Summer Street Boston, Massachusetts- 02107 J. David Nulton, Associate Deputy .

Assistant Secretary for Reactor Sten Caspersson Systems - Technology Combustion Engineering Office of Advanced Reactor Programs 100 Prospect Hill Road U.S. Department of Energy, NE-40 Windsor, Connecticut 06095 Washington, D.C. 20545 Henry Jones Salma El-Safwany Oak Ridge National Laboratory U.S. Department of Energy P.O. Box 2009 Nuclear Energy Division - Building 9102-1, Mail Stop 8030 1333 Broadway Oak Ridge, Tennessee 37831-8030 Oakland, California 94612 Steve Goldberg-Dr. Daniel L. Hears Budget Examiner =

Gas-Cooled Reactor Associates Office.of Management.and Budget

-10240 Sorrento Valley Road, Suite 300 =725 17th Street, NW San Diego, California 92121-1605 Washington, D.C. 20503

- Lloyd Walker .

Dr. Lee S. Schroeder.

7 Modular HTGR Plant Design Assistant Director for Physical

! Control- Office - West Sciences and Engineering P.O. Box 85608

^

Office of Science and Technology

- 1n Diego, California 92186-9784 OE0B Washington, D.C. 20503 Ray R. Mills

, Modular HTGR Plant Design i Control Office - East 3226 Tower Oaks Boulevard Suite 300

'Rockville, Maryland: 20852 i George-C. Bramblett, Director i MHTGR-NE Project Division General Atomics -

P.O.-Box 85608 San Diego, ; California 92186-9784 I

, Enclosure 1-

- DRAFT-l i

i

! f.o.t: The Commissioners f.rDE: James M. Taylor l l

Executive _ Director for Operations ,

subiect: ISSUEfPERTAININGTOTHEADVANCEDREACTOR(PRISM,'MHTGR,AND PIUS) AND CANDU 3: DESIGNS AND THEIR RELATIONSHIP TO CURRENT >

REGULATORY REQUIREMENTS i' Purnose: To request Commission guidance for those' areas where the staff is proposing to depart from current regulatory requirements in the-preapplication review of the' advanced l reactor and CANDU 3 designs.

<- 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-l icantly different-from the existing light-water reactor j (LWR) requirements. These documents also provide guidance

for the development of new regulatory requirements to support the advanced designs.- Staff reviews of these advanced toactor designs'should ~ utilize. existing regulations

' to-the maximum extent practicable. When new-requirements-

-are necessary, the staff should move towards performance

standard regulations and away.from prescriptive regulations.-

F Each designer is encouraged to propose new criteria and i novel: approaches for evaluation of their designs, and'an objective of early designer-staff interaction should be to-develop-guidance on licensing criteria for the advanced

! reactor design and to make a preliminary assessment of the potential of that.. design:to meet those criteria.

4 CONTACTS:

l b 11 T.H. Cox 1 504-1109- )

,- e.. ,-,s- . - , , , - - - , w+ , , , . , .. , , - ,,e,a - . .-,A , , ,,s, l_--,s- ~m++-

= s,:-~. _ v m s'- ny4 -r e w ,,l

g 9 .*

l, L

i The Comissioners '

i L

L l The staff is conducting preapplication reviews'of the following four designs:

l'

• General-Atomics (GA) 350-MWt Modular High Temperature i Gas-Cooled Reactor (MHTGR) design sponsored by the U.S.

Department of Energy (DOE) Gas Cooled Reactor Program General Electric-(GE) 471-MWt Power Reactor Innovative ,

Small. Module PRISM) reactor design sponsored by the-DOE-Advanced Liqu(d i Metal Reactor (ALMR) Program

! * . Atomic Energy of Canada, Limited, Technologies (AECLT) 1378-MWt Canadian Deuterium Natural-Uranium-(CANDU 3) reactor design .

• Aset Brown Boveri-Combustion Engineering (ABB-CE)

! 2000-MWt Process Inherent-Ultimate Safety _(PIUS)- reactor design j Enclosure 3 provides a listing of pertinent Comission.

papers and referece NUREG documents for these preap-p plication-desigk , Some information in.the original---

documents may be superseded by more recent preapplicant:

, submittals. A sumary of the current designs is provided as Enclosure 2.

~

,. In-response to Commission staff requirements memorandum-(SRMs), in SECY-91-202, " Departures From Current Regulatory Requirements in Conducting Advanced Reactor Reviews," the-

. staff committed to identify issues during the-preapplication review that require Commission policy guidance or staff technical resolution for des'.y certification, including

- situations in which advanced reactor designs significantly-deviate from current regulatory requirements.

Policy issues for evolutionary and passive' LWRs have been.

identified in the following Commission papers: 1

• SECY-90-016, " Evolutionary Light Water Reactor (LWR)

Certification Issues and Their Relationship to Currenti Regulatory Requirements"

• Draft SECY (distributed for coments on February 27,4

-1992), " Issues Pertaining to Evolutionary and: Passive

-Light Water-Reactors and Their Relationship-to Current-Regulatory Requirements"

, ,, ,w - , , - + -.m,- nn._ . . . , . . , , , . , . . ., , , , - - N , . w.. ----. -. ? N:& g- - , +

s .

t The Comissioners

• Draft SECY (distributed for coments on June 25,1992),

' Design Certification and Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light-Water Reactor Designs" Discussion: As part of their submittals, the preapplicants identified how their design complied with the current LWR licensing requirements and,-where it did not, provided alternative criteria for evaluating their designs. The staff has conducted a preliminary review of the four preapplication designs using existing LWR regulations and the evolutionary light-water reactor (ELWR) and advanced light-water reactor (ALWR) policy guidance. This initial review identified 10 issues that require policy direction from the Comission for proposed deviations from existing regulations. These are instances where either existing regulations do not apply to the design or preapplicants' proposed criteria are sig-nificantly different from the current regulations. These-issues, background information on current requirements, pre-applicants' proposed approaches, and si.aff recommendations for Comission approval, are provided in Enclosure 1.

The recomendations for Comission approval were develond by the staff with inputs from the preapplicants, the pu)11c, and the ACRS. The staff considered the preapplicants' proposals in light of the Comission's policy statement and guidance on severe accidents, advanced reactors, and s? y goals to develop a single consistent policy recomendt: i to be applied to all applicable advanced reactor desi In some instances, the staff recomends that current

~ regulations continue to be applied to the advanced reau, r designs despite preapplicant proposals to do otherwise.

Where deviations are recomended, the staff proposes more conservative alternatives to the preapplicants' proposals to account for uncertainties associated with the conceptual design, which should ensure that conclusions made during the.

preapplication review will provide a reasonable basis for the detailed design being found acceptable at design certi-fication. It is intended that the safety level standards for these designs will be consistent with Comission guidance at design certification.

Some issues are closely related. Accident evaluation and source term provide a basis-for containment performance and emergency planning. Approaches taken for residual heat i removal and ~ reactivity control are intended to be' consistent with the accident evaluation categories and consequences.

~

It ,

The Commissioners  !

The staff proposes to treat the MHTGR, PRISM, and PIUS l designs as advanced reactors in accordance with the policy statement. The CANDU:3 design is considered to be an evolutionary heavy-water. design deriving .from the larger 4 CANDU reactor designs operating in Canada and elsewhere.

Therefore, the staff has concluded that a prototype CANDU 3 is not required for design' certification. This position is i

consistent with staff conclusions in SECY-89-350, ' Canadian l CANDU 3 Design certification," and SECY-90-133, ' Prototype i Requirement for CANDU 3 Desi n." -The preapplicant, AECLT,

!- has stated that'a CANDU 3 re.orence plant is a key element

in their plan for standard design certification. If AECt.T 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 i review of CANDU 3. During the preapp11 cation review,L the-staff intends to utilize.the foreign operating experience

and accident analysis to aid in predicting'the expected i behavior of the CANDU 3 design. AECLT-makes 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 F structure, it does not conform to some current NRC regulations. The staff proposes to-apply preapplication

review criteria to the CANDU'3' reactor that are consistent

with ELWR review requirements. -

I The staff intends to use the Commission's guidance.on these i recommendations to conduct preapplication reviews of the i~ conceptual designs. Guidance for review of prototype 1

requirements for advanced reactors will follow SECY-91-074,

' Prototype Decisions for Advanced Reactor Designs.'

! Consistent with the requirements of Title 10:of the Code of Federal Regulations -(CFR) Section 52.47(b)(2) -novel safety features of the advanced reactors-and CANDU 3 will.be i

required to be demonstrated through analysis, test programs, experience, or a combination of these methods. ; Feedback from the review process will be factored into recommended-revisions to the policy guidance, and-recomendations for the development of. licensing criteria and regulations will be made after the preapplication safety evaluation reports (PSER) are issued. - Additional issues may be developed-during the-preapplication review process; they will be

.- identified in subsequent Commission papers. ,

i s

. - _ , - _ , , - . , , - , . -, .- , .L _;.,_,n-.-, w, , ,- --,c.,,i.......

t The Commissioners '

l In an SRM dated May 8, 1992, the Commission requested the staff to prioritize the issues for Commission review. The

! staff recommends that the priority for review be consistent with the PSER issuance schedules and requer',s that-direction be provided in sufficient time to allow the staff to incor-

'> orate Commission decisions into the final PSERs. Since the .

PRISM design is scheduled as the first preapplication review, Commission attention is recuested on a highest priority for those items identiftet in the enclosure as applicable to the PRISM design.

Conclusions:

The staff requests approval of, or alternate guidance on, these proposed positions to be taken in the preapplication review of the advanced reactor and CANDU 3 designs.

Coordination: The Office of the General Counsel has reviewed this paper

and has no legal objection. The staff has forwarded a draft 4

of this paper to the ACRS for its review and comments.

Recommendations: That the Commission

• Approve the staff recommendations in Enclosure I for conduct of the preapplication reviews.

• Approve of the staff's conclusion that, based on the

)osition that the CANDU 3 design is an evolutionary-

. 1eavy-water design deriving from CANDU designs operating l

in Canada and elsewhere, a prototype CANDU 3 is not i required for design certification.

• Note that positions which change as preapplication-I review experience is obtained will be communicated to the Commission and that as the staff identifies new i issues it will inform the Commission.

i

• Note that the Commission is requested to provide highest i priority attention to those issues identified in the l enclosure as being applicable to the PRISM design.

t

!

• Note that due to the preliminary nature of the design i information on the advanced reactor and CANDU 3 designs, and the preliminary nature of the staff's preapplication

. _ - - . - . . =. . .- .-

B The Commissioners '

reviews, the staff does not recommend proceeding with generic rulemaking on any of the policy issues identified'in this paper. The staff will consider generic rulemaker.,, as appropriate, as the reviews-progress and the staff gains greater confidence in the final design information. -

James M. Taylor Executive Director for Operations f

Enclosures:

1. Policy Issue Analysis

2. Design Summaries

3. List - Reference Documents k

p 5

t 4

+ +

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

CATEGORY ISSUES PRISM MHTGR CANDU PIUS A. Accident- Evaluation X X X X B. Source Term X X X C. Containment Performance X X- X X D. Emergency Planning X X X

• 9 # E. Reactivity Control X F. Operator Staffing X X X X G. Residual Heat Removal X X- X H. Positive Void Reactivity -X X Category I. Control Room Design X X X X~

2 J. Safety Classification X Discussions of these issues are on the following pages, including a brief summary of the issue, current LWR regulations, preapplicant positions, discussion of staff considerations and a proposed recommendation for staff action. The staff considered the preapplicant's proposals in light of applicable Commission policy statements.

I DRAFT Enclosure 1

t

., 'e

- At-this preliminary review stage, the staff has limited the scope of the issues to those which could affect the licensability of the proposed design.

Additionally, if a similar issue had already been raised for the LWR designs and the staff's advanced reactor design recommendation was essentially the same, it was not repeated'in this paper. In those cases where the preapplicants proposed different cor.siderations 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.

l i

DRAFT l

1

I, -

}.

4

. A. ACCIDENT EVALUATION o um i

Identify appropriate event categories associated frequency ranges, and-  !

evaluation criteria for events that will be used to assess the safety of the )

] proposed designs.-

i Current Reaulations i

i- General Design Criterion (GDC) 4 requires the consideration of accidents in the design basis. Also,10 CFR 52.47 requires the consideration of conse-

quences for both severe accidents (throu h the required probabilistic risk f BA)-for designs which differ signif-

~

assessment)anddesignbasisaccidents'iizepassiveorotherinnovativemeans-icantly from evolutionary designs or ut

, - to accomplish safety functions.

l Preanolicants' Aceroach ,

I '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-1 rem whole body and:5 rem thyroid as their limits for a significant portion.

of their accident spectrum. The MHTGR accident guidelines invo level PAG dose limit for all sequences more probable thanper 5x10'pe reactor- the lower-year.--The PlVS p,uidelines invoke the PAGs for accident sequences more-probable than 10 per reactor-year. The accident sequences more probab_le than 10', PRISM. The per reactor-ycar. guidelines PRISM invoke _ the PAG accident evaluati,on guidelines also limit consequences from any= sequence more probable than 10' 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 regulattons as contained in lo-CFR Part 100.

The CANDU 3-preapplicant,-in-their current . safety analyse's, has excludedg 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),

unscrammed loss-of-coolant accidents (LOCAs), delayed scram events :and other events which could affect' reactivity, insertion (for example, from control system failures).. 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. '

l

- DRAFT

-+v-.4 -- we,--w-n -in i ew 4este g ry 4qv - y wy e- 4 er -eep.-u s4+., p-ws,*t.-i-P--6m-u, -r- --ei se -+ t wr ev 4w - e s.+ w y-v e s pas--- --'-9riwy r yrr --y e g-wi -4" D &PMt

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 plann_ing requirements as described in Section D of this enclosure. As discussed in Section D, the staff is not ready to make a recomendation on whether the Comission should consider a reduction in the emergency plar.ning requirements. The CANDU 3 approach which limits the scope of severe accidents examined appears to be inconsistent with the provisions of 10 CFR 52.47. The accident evaluation scheme envisioned by the staff examines challenging events to the designs to provide information for a later decision on emergency planning requirements for advanced reactors and includes consideration of the potential consequences of severe accidents. Addi-tionally, for the multi-module designs (PRISM and MHTGR), the impact of specific events on other reactor modules for the multi-module sites must be assessed.

The staff's approach is intended to be structured conservatively so that positive conclusions made on the licensability of the conceptual designs during the preapplication review will provide a reasonable basis for acceptability of the design at design certification. Several sources of uncertainty exist with the conceptual designs including limited performance and reliability data for passive safety features, lack of final design information, unverified analytical tools used to predict plant response, limited supporting technology and research, limited construction and operating experience, and incomplete quality control information on new fuel manufacturing processes. Later, during the design certification process, some -

of the conservatism could be removed based on improved understanding of the design and analytical tools through completed research.

Recommendation The staff proposes to develop a single 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 Commission policy guidance with appropriate conservatisms factored in to account for uncertainties.

• Nethodologies and evaluation assumptions will be developed for analyzing each category of events consistent with existing LWR practices.

., 4 - DRAFT

_ . , - - - .- ~ - - . . . . - . .

~

. Source term determination will be performed'as approved by the Commission

. in Section 8 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 i determine a source term and to identify a containment challenge scenario.

• External events will be chosen deterministically on a basis consistent

with that used for LWRs.

• Evaluations of multi-module reactor designs will consider whether specific events apply to some or all reactors onsite for the given scenario of

operations permitted by proposed operating practices, t

I l

DRAFT

B. SOURCE TERM issue Should mechanistic source terms be developed in order to evalutte 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 PRISH fuel to l quantify the fission product release fractions or the behavior of those l fission products migrating from the metal fuel through the sodium coolant.

MHTGR designers have proposed siting source terms for accidents based on the expected fuel integrity. The coated microsphere fuel particles in the core are predicted by the preapplicant to contain all the fission products except

for that released from the small number of failed particles resulting from in-service particle failures and added particle failures during accidents.

Insufficient data currently exists to determine whether the MHIGR fuel performance will meet these expectations.

The PlVS designer has proposed using a mechanistic LWR source term. ,

Information has been provided in the Preliminary Safety Information Document I (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 previously used for LWRs.

DRAFT e

s The CANDU 3 designer uses a source term for eacn 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 mbnorandum ($RM) related to SECV-90-016, the Commission requested the staff to submit a saper describing the status of efforts to develop an updated source term t1at 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).

4 The differences between the LWR designs and the MHTGR and PRISH 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 failu'e 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

i

l. The performance of the reactor and fuel under normal and off-norm:1 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.

4 1 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 4

design-dependent uncertainties, i The design specific source terms for each accident category would constitute 1 one component for evaluating the acceptability of the design.

1

l 4

i 4

f' e

- 8- DRAFT

. - . _ .. _,_._._,,_---m . _ - , , _ _ _ _ - _ , . - ,--- ., _

C. CONTAINMENT h12t .

Should advanced reactor designs be allowed to employ alternative approaches to traditional ' essentially leak-tight" containment structures to provide for the control of fission product release to the environment?

Current Reculations General Design Criterion (GDC) 16 requires that LWR reactor containments I provide an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment, and that containment-associated systems assure that containment design conditions important to safety are not exceeded for as long as postulated accident conditions require. GDC 38-40 set l requirements for containment heat removal, GDC 41-43 for containment j atmosphere cleanup, and GDC 50-57 for containment design, testing, inspection, ,

and integrity. Requirements for LWR containment leakage testing are I establisied in 10 CFR Part 50, Appendix J.

Preapolicants' Acoroach The MHTGR is not designed with a leak-tight containment barrier. The design relies upon high integrity fuel particles to minimize radionuclide release, and on a below-grade, safety-related concrete reactor building to provide retention and holdup of any radioactive releases. The reactor vessel and the steam generator vessel are in separate cavities within the concrete structure.

In the event of a reactor ecolant pressure boundary (RCPB) rupture, louvers in the reactor building are designed to allow the passage of gases to the environment, preventing building overpressure. The building design does not include containment isolation valves for the ventilation line from the building and has an open path to the environment via a drain line in the ,

reactor cavity cooling system (RCCS) panels. Accident dose calculations assume a constant 100 percent volume per day building leak rate, and take credit for plateout on the building walls.

PlVS, above grade, is designed with a low-leakage containment based on a pressure-sup>ression scheme that is integral with the reactor building, similar to tie ABWR and SBWR. Below grade, the concrete pool wall and floor, which is the reactor pressure boundary, and the containment are contiguous, separated only by a steel membrane.

CANDU 3 is designed with a large, dry, steel-lined, concrete containment, is without 5 percentcontainment volume per dayspray. Themaximumleakrate(usedinsafetyanalyses)The at the design pressure of approximately 30 psig.

structure is designed for a test-acceptance leak rate of 2 percent per day at the design pressure. These leak rates are significantly higher than those of a typical U.S. PWR containment.

DRAFT

1 The PRISM containment design is a high strength steel, low leakage, l pressure-retaining boundary, consisting of two components, the upper I containment dome and lower containment vessel. The upper containment is a steel dome, it differs from light-water reactor containments functionally in the following respects. The containment is specifically designed to mitigate the radioactive release consequences of severe events. The PRISM containment volume is markedly smaller than is typical of LWR containments; there is

little separation between the reactor vessel and the containment boundary; and no safety-grade containment coolers or spray systems are provided. The entire-containment structure is located below grade within the reactor building.  !

1 Discussion Each of the advanced designs and CANDU 3 maintain an accident mitigation 4

approach in which containment of fission products is a part. Two of the advanced reactor designs PRISM and MHTGR grade, providing protectio (n from external) hazards. place the Generally, reactor building below 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 com>onent of our defense-in-depth approach to regulation. However,

, the Advanced teactor Policy Statement recognizes that to encoura e 3 incorporation of enhanced safety margins (such as in fuel design in advanced t

reactor designs, the Commission would look favorably on desirabl design related features or reduced administrative requirements. New reactor designs that deviate from current practice need to be extensively reviewed to assure a level of safety at least equivalent to that of current generation LWRs is provided, and that uncertainties in the design and performance are taken into .

account.

The staff believes that new reactor designs with limited o>erational experience require a containment system that provides a su)stantial level of accident mitigation for defense-in-depth against unforeseen events, including core damage accidents. This requirement may not necessarily result in a high-pressure, low-leakage structure that meets all of the current LWR requirements for containment, but it should be an independent barrier to fission product release. The proposed criteria will need to provide an appropriate level of protection of the public and the environment considering-both the safety advantages of the advanced designs and the lack of an experience base in evaluating their performance. For evolutionary LWRs, the staff, in SECY-90-016, proposed to use a conditional containment failure

)robability (CCFP) or deterministic containment performance goal to ensure a aalance between accident prevention and consequence mitigation. During the evolutionary LWR reviews, a great deal of careful review was necessary to assure that a probabilistic CCFP would not be used in a way that could detract from a balanced approach of severe accident prevention and consequence mitigation. For advanced designs and the CANDU 3, limited experience exists

- 1,0 - DRAFT

l

)

in the analysis and evaluation of severe accidents which would lead to ,

significant difficulty and uncertainty in assessing a CCFP. For this reason, 2

the staff recomends that the deterininistic containment performance goal be 4

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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the onset of core damage. This approach is used because the

., preliminary nature of the advanced designs precludes a reliable assessment of ,

the failure probability of accident mitigation systems and, therefore, of )

containment failure probability. Further, the CCFP is grounded in a firm '

understanding of LWR safety systems and accident progression. Intrinsic

differences exist between LWR and advanced reactor technologies and their i approaches to the balance between accident prevention and mitigation. A

quantitative level of understanding of new technologies and systems comparable i to that of LWRs is not yet available. Thus, the use of a performance based criterion rather than a quant.itative one appears to be more appropriate for advanced reactor and CANDU 3 preapplication review given the current level of I knowledge of advanced reactor and CANDU 3 risk and ' ts prevention / mitigation elements.-

j Recommendation 4

The staff proposes to utilize a standard based upon containment functional performance to evaluate the acceptability of proposed designs rather than to rely exclusively on prescriptive containment design criteria. The staff l intends to approach this by comparing containment performance with the

accident evaluation criteria. -

l

• 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the onset of core damage, the specified containment challenge event results in no greater than the l limiting containment leak rate used in evaluation of the event categories.

and structural stresses are maintained within acce) table limits (i.e.,

. After t1is period, the ASME containment Level C prevent must requirementsuncontrolled or equivalent) re leases of radioactivity.

i I

6

( -

11 - DRAFT p c.., --y ,p-ev-----g , -y- r w r w ,w-. .-,- t-m r -~-.v,*w, .y. _ , . ' - . , ,w-- w , ."w w .- , --

v w -w

4 x

D. EMERGENCYPLANNING(EP)

Lung .

1

Should advanced reactors with passive design safety features be able to reduce .

) emergency planning zones and requirements?

i Current Reaulations i

Although emergency plans are not necessary for the issuance of a design certification under 10 CFR Part $2, they would be necessary for the issuance i of a combined license under Part 52 or a license issued under 10 CFR Part 50. i

, 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

~

I measures can and will be taken in the event of a radiological emergency.

l

, Currently, offsite protective actions are recommended when an accident occurs i that could lead to offsite doses in excess of the Environmental Protection

, which are 1-5 rem whole Agency's (EPA) body and 5-25 rem Protective Action Guidelines (PAG)d doses, protective actions thyroid. Atthelowerprodecte should be considered. At the higher projected doses, protective actions are warranted.

PreaDolitants' 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 PRISH design emphasizes accident prevention, long response times [36 hours) between the initiation of

, an accident and the release of any radiat90n, and containment of accidents if-they should occur.

MHTGR proposed reduced offsite emergency planning for similar reasons as those proposed for PRISM. There would be an emergency plan for.an MHTGR and the plan would include any agency that could become involved in a radiological emergency supply). The d(i.e.ifferences and reductions from a typical plan for LWRs are thatsh the MHTGR plan would have the exclusion area boundary (EAB) of 10 CFR Part 100 as the boundary of the emergency planning zone Appendix E of 10 CFR Part 50 for gas-cooled tors; reac(EPZ),

and thatas maywould there be allowed be by no rapid notification or annual drills for offsite agencies. This is based on the preapplicants' assertion that (1) the predicted dose consequences estimated at the EAB/EPZ for accidents are below the lower-level EPA sheltering PAGs and the public can be excluded from the EAB, (2) the significantly long time expected for the core to return to criticality' after being shut down by the doppler coefficient without the reactor protection system functioning (i.e., about 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />), and (3) the long time for the fuel and reactor vessel to reach maximum temperatures (i.e., about 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />)

DRAFT

during accidents. The prea >plicant asserts that the public around the would be outside the area t1at needs to be sheltered or evacuated and, plant further, there is ample time to notify and move the public during an event.

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 th'at there appears to be no credible accident sequences that would lead to serere core damage. Offsite dose for the large break LOCA is claimed to be below the lower level EPA PAGs at 500 meters distance from the containment. No specific information on emergency planning was provided by the preapplicant for review beyond the general assertion that they intend to limit offsite doses to the PAGs.

Discussion 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.

Recomendation The staff proposes that advanced reactor licensees be required to develop offsite emergency plans. Additionally, exercises, including offsite exercises, provisions for periodic should be develo)ed. emergency

-These actions are required by existing NRC regulations which include tie required establishment of an offsite emergency planning zone (EPZ). Consistent with the current regulatory approach, the staff views the inclusion of emergency preparedness by advanced reactor licensees as an added conservatism to NRC's

' defense-in-depth" philosophy. Briefly stated, this philosophy: (1) 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 l safety systems to reduce the chances that malfunctions will lead to accidents i

that release fiss!cn products from the fuel; and (3) recognizes that, in spite of these precautions, serious fuel damage accidents can happen, therefore requiring containment structures and other safety features to prevent the

~~

DRAFT

i .

l .-

l i

l release of fission products offsite. The added feature of emergency planning a 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 outilned in Section A of this enclosure will provide input to the Emergency Planning requirements for advanced reactor designs. Based in part upon these accident

! evaluations, the staff will consider w1 ether some relaxation from current requirements may be appropriate for advanced reactor offsite emergency plans.

The relaxations to be evaluated will include, but not be limited to, notification requirements, size of EPZ, and frequency of exercises. This

, evaluation will take into account the results of passive LWR emergency planning policy decisions.

i e

e DRAFT

.1- - . . - . -

E. REACTIVITY CONTROL SYSTEM Issues should the NRC accept a reactivity control system desiwn that has no control rods?

Current Reaulations General Design Criterion (GDC) 26 requires that two independent reactivity control systems be provided. One of the systems shall use control rods, preferably using a positive means for insertion. The other system shall be capable of controlling planned reactivity changes to assure fuel limits are not exceeded.

Prescolicants' Position The PIUS design does not have control rods. However, the preapplicant proposes that the design complies with the intent of General Design Criterion 26 by having two independent liquid boron reactivity control systems. The normal reactivity control system pumps boron into the primary coolant loop to control reactor power or effect a reactor shutdown; this system is only safety-grade within the bounds of the containment holation valves. The fully safety-grade reactivity control system relies on the ingress of highly borated water through the density lock from the reactor pressure vessel to scram the reactor. This ingress occurs when the equilibrium conditions across the thermal barrier of the density locks are disturbed by an imbalance between the thermal core heat generation and removal rates. Either a trip of as few as one of the four reactor coolant pumps or a reactor overpower event (with forced flow) could initiate borated water flow into the core. The reactor protection system initiates the scram function by tripping a single reactor coolant pump. Other reactivity control features of the design are in-core burnable poisons for power sha .

core size for control of xenon oscillations for slow, large, ping, and limitations in 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, psrallel 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 mecianism. ,

1 Discussion l The existing LWR regulations provide prescriptive design guidance for one reactivity control system to contain rods. Of the three advanced reactor designs, only PIUS does not have the capability to control reactivity with control rods. The PlVS design does have, however, three ways to introduce ORAFT

liquid boron into the core to control and shutdown the reactor. Two of the three rely on flow through the density lock from a common supply of borated

>ool water. The other system is the normal reactivity control system which ins a separate boron tank and is used for normal shutdown. The latter system is only safety graae within the bounds of the containment isolation valves.

Recommendation The staff concludes that a reactivity control system without control rods should not necessarily disqualify a reactor design. Adesignwithoutcontrol rods may be acceptable, but the appitcant must provide suff cient information to justify that there is an equivalent level of safety in reactor control gnd protection as compared to a traditional rodded system. This infomation must include the areas of:

a. reliability'and efficacy of scram function '

b. suppression of oscillations

c. control of power distribution

d. shutdown margin

e. operational control l

l

,16 - DRAFT

l l F. OPERATOR STAFFING AND FUNCTION t

1111t

Should advanced reactor designs be ' allowed to operate with a staffing complement that is less than that currently required by the LWR regulations.

1 i

Current Reaulations i The NRC has established the requirements for control room staffing in i 10 CFR 50.54 i control roomat(m)(2)l al times and a licensed operator or senior operator must be(iii) 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 shift crew shall include two licensed senior reactor operators (SRO), one of whom shall be designated as the shift supervisor, two licensed reactor operators (RO), and two unlicensed auxiliary operators (AO).

Preacolicants' Position i The MHTGR plant is presently four reactor modules with two modules feeding a sin le steam su) ply system. The design includes a shift-staffing level of eig t persons wio would be dedicated to plant operations; a senior licensed shi t supervisor, two licensed reactor operators in the control room, and five

, roving non-licensed operators. This results in three licensed and five non-5 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. Tne objective for the minimum number of operating staff would -include: - one SRO shift supervisor, one SRO assistant supervisor, one R0 per power block (three in the control room, and three plant R0s. This results in a minimum modules) licensed operators for nine reactor modules.

of eight 5

During normal plant operations the PIUS main control room would be manned by two R0.s and a SRO shift supervisor. The shift su required to be in the control room at all times. pervisor would not be The CANDU 3 preasplicant has not posed a specific number of licensed operators, but tie staff's expect on is that CANDU 3 will meet the current LWR staffing requirements. '

Discussion Present day LWRs would require a minimum of. one shift supervisor, one SRO. and two operators per reactor. The designers of advanced reactors have stated that the highly automated operating systems, the passive design of s.afety 17 - DRAFT

, s a

..,..,.-..w-,,~-,---.-r-.- e. - . . . , . , - _ , . . , - - - - - -,-w, --,.-, --. , - . , - - , . , . , , - - . - -

j.*

i i features, and the large heat capacity results in reactor desions that respond i to transients in a manner that demands less of the operator tfian do the

! current operating plants or evolutionary designs. The preapplicants assert

that the passive safety features and, in some cases, large coolant inventory 1 and PIUS designs may not require an operator to act or of the PRISM, intervene for severa MHTGR,l days following an accident. These designs also automate systems that start up, shut down, and control these reactors. The vendors of l these reactors have suggested that they could be operated with fewer licensed i operators and believe that this would reduce significantly the training and operating costs to licensees.

! A similar policy issue, Role of the operator in a Passive Plant Control Room, I was identified in the staff's June draft policy paper on passive

reactors. In that paper, the staff expresse 25, 1992,d concern that the man-machine interface for the passive reactors had not been sufficiently addressed and j that actual testing needed to be done on a control room prototype. The staff

! believes that position is also applicable to advanced reactors.

. l l Recommendation j t

[ The staff believes that operator staffing may be design dependent and intends-

to review the justification for a smaller crew size for the advanced reactor i i

designs by evaluating the function and task analyses for normal operation and accident management. The function and task analyses must demonstrate and 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 pro >osed number of licensed operators less one, while all other units could >e taken to a i_ cold shutdown condition from a variety of potential operating conditions

including a fire in one unit.

!

• The units can be safely shut down with eventual progression to a safe l shutdown condition under each of. the following conditions: (1) a complete l loss of computer control capability, (2) a complete station blackout, or l (3)adesignbasisseismicevent.

• The adequacy of these analyses shall be tested and demonstrated on an j - actual control room prototype.

6 i

+

DRAFT 1

. - . . . . . . . ~ - m-- o-.r,,., ...,;.,- ,.a,.- ,,,,,,w.,~r,,,.<..-et.,,we..-,%.-,,,,, u-,--,.w 3.-... v. . . - - ,y - * - m.e S w. , - , .,-po by.. ,,

G. RESIDUAL HEAT REMOVAL liin Should advanced reactor designs that rely on a single completely passive, safety-related Residual Heat Removal (RHR) system be acceptable? ,

Current Reculations General Design Criteria (GDC) 34 requires the RHR function to be accomplished using power,only safety-grade and assuming a sinslestems, fai lure within the safety system. assuming a loss of either ons 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 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 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.

Prescolicants' Position The PRISH 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 steani 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 designec 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 DRAFT l

l o>eration. and tie shutdownTheRCCSisrelieduponwhentheheattrans>ortsystem(HTS)he cooling subsystem (SCS) are inoperable. T1e HTS utilizes t steam generators and non-safety-grade feed system and condensers and is used during normal operations, startup/ shutdown and refueling. The SCS is 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 vesse .

for The PlUSheat residual design usesfrom removal a safety-grade the reactorpassive closed pool. The cooling system system consists (PCCS)ht of eig independent parallel loops located in four separate compartments that are physically separated from each other. Heat is dissi)ated through four (4 natural draft cooling towers located on the top of t1e reactor building. )One cooling tower is in each quadrant of the reactor building. The reactor pool water can be maintained at 95 *C with one loop out of service. The system is always in operation. Reactor residual heat can be removed with the condenser during startup/ shutdown and refueling conditions. If the condenser is not available, a non-safety-grade diesel-backed pump system can (.001 the pool water.

Discussion Similar issues were identified for the RHR system of the passive LWR designs.

In a draft Commission paper issued for comment on February 27, 1992, the staff identified issues relating to the ability of passive systems to reach safe shutdown, definition of a passive failure, and treatment of non-safety systems which reduce challenges to the passive systems. These issues remain open and

the staff will propose recommendations in the future for resolution, in the case of advanced reactors the safety-grade RHR systems are completely passive and are in continuous operation. Continuous performance monitoring of the passive systems is one advantage of the constant operation. The high heat capacity of PRISM and MHTGR lead to longer time periods before exceeding temperature limits. PRISM and MHTGR use the natural circulation of air to remove residual heat. PlVS uses natural circulation of water through natural draft cooling towers for its RHR system. The lack of check and squib valves, the continuous operation and use of a singic 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, i -

j DRAFT l

i j- .

l e

Recomendation l As a result of the unique design features of the PRISM, NHTGR, and PIUS j

designs, safety-related the RHRstaffsystem believes maythat reliance on In be acceptable. a single,ing carry out its futurecompletely passiv j detailed design evaluation the staff will assure that NRC regulatory

treatmentofnon-safety-relatedbackupRHRsystemsisconsistentwith j Commission decisions on passive light-water reactor design requirements. .

1 b

h i

I k

1 i .

j I

}

i 4

i 1

a k

I --

i '

i-l i

s. -

e-21 - DRAFT e

--,-nno-,,,s y m. -, m,-,r- sp w ,ws, ,w mm - e w o - -w,w,9we er a ,+N-, -~~w - ,e e,x w%a v w < ,- m g na,w-> eg new r w- -w--- ,---vm,---e,em---~ wg+-r--

8

.6 H. POSITIVE V01D REACTIVITY COEFFICIENT hat Should a design in which the overall inherent reactivity tends to increase under specific conditions or accidents be acceptable?

Current Reaulations General Design Criterion (GDC) 11 requires that the reactor core and coolant system be designed so that in the )ower operating range the net effect of prompt inherent nuclear feedback claracteristics tend to compensate for rapid increases in reactivity.

, P_t,gpolicant s' position In the PRISM design, the maximum sodium void worth, according to the preapplicant, assuming only driver fuel and internal blanket assemblies void, is nominally $5.50. If radial blanket assemblies are included, the sodium void worth is nominally $5.26 which does not include the -70 cents from gas expansion modules (GEM). Should sodium boiling begin, on a core-wide basis under failure to scram conditions with a total loss of flow without coastdown, the reactor could experience a severe power excursion and core disruption.

The aredicted temperature reactivity feedback is approximately -80 cents prior to tie onset of sodium voiding. This mitigates to some extent the positive reactivity addition. For sodium voiding to occur, multiple failures of redundant and diverse safety-grade systems would be required.

Although the overall power coefficient for a CANDU 3 reactor is claimed to be slightly negative and very close to zero, the coolant void reactivity is

)ositive throughout the fuel core lifetime. The total core void worth is i

)etween $1 and $2. The positive void coefficient is not a concern during -

normal operation, but, during a large LOCA at specific locations, void l 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 power excursion followed by core melting. The CANDU 3 design is intended to prevent an unscrammed event from occurring through the use of two separate shutdown systems each to be independent, redundant, diverse, and safety grade.

Discussion

The staff considers the existence of positive coolant void coefficients, or l 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

ii I. j i

either the PRISM or CANDU 3 designs may not necessarily lead to a large scale release of the radionuclide inventory to the atmosphere due to their

respective mitigative designs. In the CANDU 3 reactor, multiple redundant, diverse fast acting scram systems n're provided to address the positive coefficients.

! Attempts to modify the designs to reduce the effects of these positive coefficients may result in other consequences potentially as serious. For example, in the PRISH design the positive void coefficient seems to result from the design objectives of maintaining a p'assive shutdown capability and of i 4

minimizing the reactivity swing over core life. Attempts to reduce the PRISM l

! void worth might have the effect of increasing the severity of rod withdrawal

i accidents or reducing the ability to withstand an unscrammed loss of heat sink j events without core damage.

Encommaadation i' The staff concludes that a positive void coefficient should not necessarily disqualify a reactor design. The staff is proposing to require that the PRISM and CANDU 3 preapplicants analyze the consequences of events unscrammed1.0CAs,delayedscrams,andtransientsaffectingrea(suchasATWS, ctivitycontrol) that could lead to core damage as a result of the positive void coefficients.

. The staff's review of these analyses will take into account the overall risk perspective of the designs. Whether the prea>plicants will be required to t

consider changes in the designs to mitigate tie consequences of these accidents will desend on the estimated probability of the accidents as well as 1 the severity of tie consequences. +

l l

i 2

f 2

! DRAFT l

I l

,e

1. CONTROL ROOM AND REMOTE SHUTDOWN AREA DESIGN lilut Can current requirements for a seismic Category 1/ 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 Reoulithni The current LWR requirements for control room and remote shutdown area desigh are provided in 10 CFR Part 50, Appendix A, and 10 CFR Part 100. General Design Criterion (GDC) 19 requires that a centrol 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 the control room. GDC 17 requires that the electrical system for tise control room and remote shutdown equipment meet the requirements for quality and independence. These requirements are defined as Class IE in the supporting IEEE standards. GDC 2 and 10 CFR Part 100 require that structures and systems important to safety be designed to seismic Category I standards to remain functional during a safe shutdown earthquake.

preacolicants' position i

The control room for PRISH contains the instrumentation and controls for all nine reactor modules and their power conversion systems. The control room structure is nct considered safety related and, therefore, the room is not designed to seismic Category I design requirements. Additionally, no equip-ment in the control room is safety grade. A separate alternate shutdown console is located in the protected area of the reactor service building. The alternate shutdown console is within a seismic Category I structure and is equipped with the necessary Class IE controls and instrumentation to protect the core and has the required habitability control system.

The MHTGR design has, for the four modules, a non-safety-related central cora-trol room to operate the plant and a seismic Category I remote shutdown area from which to respond to accidents if necessary. Neither the equipment in the control room nor the remote shutdown area are Class IE. The remote shutdown area does not contain safety-related equipment, nor does it include a ventilation system for operator habitability, or a safety-related manual scram. This is based on the preapplicant's position that accidents do not require operator response. Tne 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

, . I 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 tl.at 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 specifie<f 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 I areas.

Each remote area contains one half of the safety-grade control equipment, e.g., the reactor trip and interlock system, control of certain isolation valves, and safety-grade monitoring systems. The manual reactor trip system is a push-button control of the main reactor coolant Sumps. Both the main control room and the emergency shutdown areas are sern ced 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 t1e 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 uf evacuation of the control room during design i basis accident conditions or hamper the control or monitoring of upset conditions as the event sequence progresses. The staff believes human performance will still play a large role in the safety of the advanced plants and CANDU 3 and that the quality of support provided by the safety-related, seismic Category I and electrical Class IE control room is appropriate.

The staff also belit:yes 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 I

l 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 s'aff's t 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 l

4 l

, 6 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 guidance for these issues, once she

! guidance is finalized.

Reconnendation

The staff recommends that until passive LWR policy for design requirements of i control rooms and remote shutdown facilities is finalized, the staff will i apply current LWR regulations and guidance to the review of advanced reactor i designs. This will ensure that plant controls and the operators will be

adequately protected so that safe shutdown can be assured in accident situations.

1 I

I L

l 1

'f G

S 4

DRAFT

e J. SAFETY CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS ,

i l 11Eut j What criteria should the NRC apply to the advanced reactor designs to identify i the safety-related structures, systems, and components?

Current Reaulations Title 10 of the Code of Federal Regulation Section 50.49(b)(1) and the current' A>pendix A.VI(a)(1) of 10 CFR Part 100 list the following criteria to define tle safety-related structures, systems and components:

a. those needed to maintain the integrity of the reactor coolant pressure boundary (RCPB);

b. those needed to shut down the reactor and maintain it in a safe condition; and

c. those needed to prevent or mitigate the consequences of accidents that could result in doses comparable to Part 100 guidelines.

Amendments to Parts 50 and 100 have been proposed (57 FR 47862) to update criteria used in decisions regarding reactor siting and design for future nuclear power plants, including the advanced LWR designs. Tnese proposed revisions include the temporary relocation of the dose considerations for reactor siting (i.e., the current Part 100 guidelines) from Part 100 to Part 50 until such time as more specific requirements are developed regarding accident source terms and severe accident insights.

Preapolicants' Position The advanced reactor designs rely on a limited number of safety-related

• systems to protect the core and the public. Some of these systems are entirely passive, with no moving compone u s and do not require operator action. The vendors believe that this reduction'in safety-related equipment results in simpler plant designs with lower costs. This also results in many structures, systems, and components, which are considered as safety related in LWR desi designs.gns, being classified as non-sofety-related in the advanced reactor Of the advanced reactor designs, only the HHTGR design is not using the current LWR criteria above for safety classification. For the MliTGR design, the only criterion for safety-grade classification is those structures, systems, and components needed to mitigate the dose consequences at the site boundary from accidents or events to below the guidelines in the current 10 CFR Part 100. Several major issues with safety classification were identified previously by the staff in the Draft PSER (NUREG-1338): (1)the RCPB is not entirely safety related, (2) no safety-related equipment is used to pressurize and depressurize the RCPB, (3) the coolant moisture monitor is not safety related, and (4) neither the control room or remote shutdown area DRAFT

1

! are safety related, and (5) no safety-related instrumentation providing j reactor protection or monitoring functions are available in the control room 4 or remote shutdown area.

1

! Discussion 1

The NRC LWR criteria are intended to require defense in depth: the advanced reactor designs include high quality, non-safety-related active systems to

>rovide defense-in-depth ca> abilities for reactor coolant makeup and decay 1 eat removal. These would se the first line of defense in the event of

! transients or plant upsets. The non-safety-related systems are a i the designers, not required for mittention of design basis evenIs,ccording but do to i provide alternate mitigation capability. In a recent draft SECY paper i covering the passive A.WRs, the NRC staff stated that it was still evaluating j 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.

Recomendations The staff intends to apply the LWR criteria for identification of safety-

related structures, systems, and com>onents to the MFITGR design. Requirements

! for non-safety-related systems will se consistent with the NRC position for

passive LWRs. We have noted that 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.

i 1

G W

4

s A. CANADIAN DEUTERIUM URAN!UM (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 CANDU 3 design evolved from other CANDU PHWRs, most notably the CANDU 6 design. The CANDU 3 is a generic standard design that has retained many key components (steam generators, coolant pumps, pressure tubes, fuel, on-line refueling machines, instrumentation, etc.) that have been proven in service on operating CANDU l power reactors. Currently, there are 25 CANDU reactors in operation in j 6 different countries and 19 under construction. The first CANDU reactor was i placed in service in 1968. CANDU experience to date amounts to over 175-years of effective full power operation.

On May 25, 1989, Atomic Energy of Canada, Limited, Technologies (AECLT) ,

informed the NRC of their intent to submit the CANDU 3 reactor design for l standard design certification in accordance with Part 52. AECLT of Rockville, 1 Maryland, is a wholly-owned subsidiary of Atomic Energy of Canada, Limited (AECL) (a crown corporation of Canada), and is the preapplicant for the CANDU 3 design. AECL in Canada is also pursuing standard design certification of the CANDU 3 with the NRC's Canadian counterpart, the Atomic Energy Control Board of Canada. AECLT's f.urrent slans are to submit a standard design certification application for CANDJ 3 in the 1995-1996 time frame.

Desian Descriotion The CANDU 3 is e 450 MWo 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 ratto of moderator to fuel is five to eight times that of an LWR. Thus, the CANDU core is larger than an LWR core generating the same power. This results in a lower core power density for CANDU 3. In addition, the CANDU 3 core is neutronically loosely coupled which results in xenon induced flux tilts that requires a relatively complicated computer operated spatial flux control system.

As in LWRs, CANDU 3 fuel elements consist of pressed and sintered uranium dioxide pellets enclosed in a zirconium cladding. Each CANDU 3 fuel bundic 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 pla,ced inside a calandria 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, l

.,1 - DRAFT Enclosure 2

r the CANDU design must be fueled on a continuous basis during power operation by an automatic fueling machin6. On-line fueling is the primary means of changing reactivity in the CANDU 3.-

for the CANDU 3 design, heavy water coolant flow through the core is uni-directional, thereby facilitating on-line fueling from one end of the reactor with a single fueling machine. The )rimary system operating pressure (nominally 1435 ay a pressurizer connected to one of the outlet headers. psig) is maintained The CANDU 3 light-water secondary system is similar to that of a PWR.

3 The fuel channel assemblies are enclosed in a horizontal, cylindrical vessel

, called a calandria that contains the low-temperature (140 'F), low-)ressure, heavy-water moderator. The calandria vessel, in conjunction with t1e integral end shields, supports the horizontal fuel channel assemblies and the vertical and horizontal reactivity control unit components. The CANDU 3 utilizes four reactivity control systems for reactor control and shutdown during normal operation, and two redundant and diverse safety-grade shutdown systems are used for reactor shutdown following a transient. A separate moderator heat removal system ensures that the moderator remains subcooled.

All systems in the CANDU 3 design are assigned to one of two groups - either Group 1 or Group 2. The systems of each group are capable of shutting down the reactor maintaining cooling of the fuel, and providing plant monitoring capability In the event that the other group of systems is unavailable.

, Group 1 systems are those primarily dedirsted to normal plant power pro-duction. The Group 2 systems include four special safety systems and other safety-related systems. These mM. Main plant safety in the event of a loss or partial loss of Group 1 systen, and mitigate the effects of accidents, including the design basir earthquake. Tfie Group 1 and Group 2 systems are,

, to the greatest extent possible, located in separate areas of the plant.

CANDU 3 employs two fast-acting, redundant, and diverse Group 2 shutdown i

systems, separate from the Group 1 reactor regulating system. Shutdown System No. 1 (5051 consists of 24 vertically inserted control rods. Shutdown System No.2(SDS2 consists of six horizontal nozzles through which a gadolinium nitrate sol tion is injected. Both shutdown systems inject into the low-pressure moderator, precluding a rod ejection accident. In addition to the two shutdown systems, the remaining special safety systems include containment and emergency core cooling system (ECCS).

t The CANDU 3 containment system includes a reinforced concrete containment structure with a reinforced concrete dome and an internal steel liner. The containment is designed with a test acceptance leakage rate of 2 percent per day. ECCS supplies light-water coolant to the' reactor in the event of a loss-of-coolant accident. Each of the four safety systems is required to demonstrate during operation, a dormant unavailability of less than 10'3 or about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per year, and be physically and functionally separate from the normal process systems and from one another. The CANDU 3 shutdown c o ng system is designed to remove heat fro;n the HTS d nominal operating temperature and pressure.

DRAFT l

r B. MHTGR (MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR)

Development History The Modular High Temperature Gas-cooled Reactor (MHTGR) was proposed to NRC by the U.S. Department of Energy (DOE) in 1986 in response to the Commission's 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 Research and a draft preapplication safety evaluation report (PSER) was issued by HRC in March 1989. DOE recently advised the NRC that the MHTGR design certification application schedule will be established in August 1993, when a DOE decision on the gas-cooled new production reactor funding will be made. The Energy Policy Act of 1992 requires DOE to submit a preliminary design approval application by September 30, 1996.

l Commercial gas-cooled reactors began with the graphite-moderated, carbon  ;

dioxide-cooled Magnox reactors developed in the early 1950's in the United l Kingdom and France. In the United States, gas reactor development resulted in l the 40 MWe Peach Bottom 1, which operated from 1967 to 1974, and the 330 MWe Fort St. Vrain, which operated from 1976 to 1989. There have been about 50 l gas-cooled reactors in the world totaling about 1000 reactor-years of operation. in this total, there has been about 50 reactor-years of experience with the HTGRs.

The BISO and TRISO (trade names) multi-layered microsphere fuel form is used in HTGRs. The BISO fuel form, a fuel kernel with two major layers, was used in Peach Bottom 1; and the TRISO fuel form, a fuel kernel with four major layers (including a silicon carbide layer), was used at Fort St. Vrain. The TRISO fuel form provides higher fuel integrity requirements than the BISO fuel and is the reference fuel for the MHTGR. DOE maintains agreements with Germany and France for the exchange of technical information concerning the integrity of the reference MHTGR fuel, and experiments will be conducted in France. As part of DOE's Technology Development Program for the MHTGR, post l irradiation testing of development fuel at Oak Ridge National Laboratory is being performed, and a technical information exchange agreement was established with Japan, which is building an experimental HTGR.

Major trends in recent HTGR designs, including the MHTGR, are the following:

(1) increased system pressure, (2) steel pressure vessels for the smaller HTGRs, including the MHTGR, versus the prestressed concrete reactor vessel for larger HTGR designs as Fort St. Vrain, (3) proposed greater fuel integrity, with a 6 x 10 4 fraction of failed fuel assumet for the MHTGR, and (4) lower enriched uranium fuel.

• Desian Descriotion The standard MHTGR plant is (cur reactor-steam generator modules and two steam turbine-generator sets. Each module is designed for a thermal output of l 350 MWt. Two reactor modules are coupled to a steam turbine-generator set to l

produce a total plant electrical output of 540 MWe.

! DRAFT t -

t The low-power density (5.9 watts /cc) reactor core is helium cooled and neaphite moderated, and uses ceramic coated (four major layers) microspheres inanorganicbondedcylindricalcompactasthefuel. The core design is intended to provide a large negative doppler coefficient to shutdown the reactor with heatu . The microsphere fuel design is stated to allow fuel temperatures as hi h as 2900 'F without significant fission product release. l The compacts are p aced in small vertical holes in the hexagonal graphite block fuel assemblies. The fuel assemblies are cooled through passages in the blocks. There are about 660 graphite blocks in the 66-column annular core region between the inner and outer reflector regions. The helfum is a single l 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, 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) into the thanwith coolant thataon the primary steam side generator tu (be leak or failure.about 925 psig), so water w 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-rehted RCCS is a set of panels surrounding the reactor vessel with a header connection to four inlet and exhaust ports above ground. This allows hot air to rise thus removing heat transferred from the reactor vessel while cold air is drawn from outside into the panels. The system 1) is entirely passive with no moving components, (2) is always operating, 3) automatically responds to rising temperatures through thermal radiation an natural circu-lation, and (4) has flow path redundancy to the cooling panels through a cross-connected header. In addition, there are two other non-safety-related, active heat removal systems: (1) the shutdown cooling system in the bottom of the reactor vessel, and (2) the main circulator / steam generator in the primary cooling loop. The non-safety-related systems are not relied upon for accident safety analyses.

The multiple barriers to fission product release are the coated fuel microspheres, the graphite blocks, the ASME Code reactor coolant pressure boundary (RCPB), and the containment. The containment is the reactor building below ground with containment isolation valves on the steam generator DRAFT

1 1

l 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.

e 6

+

4 h

4 4

4

+

DRAFT

j.'

. l l .

C. PIUS (PROCESSINHERENTULTIMATESAFETY)

Historical Develooment TheProcessinherentUltimateSafety(PlVS)reactorisbeingdesignedbyASEA Brown Boveri Atom (ABB-Atom). The concept evolved in the early 1980's from an extension of then ABB-Atom's low temperature district heating design. In .

October 1989 ABB requested a licensability review of the P!US des'gn in '

accordancewlth-NUREG-1226,andinMay1990,ABBsubmittedthePIUS for staff review. ABB plans to preliminary apply for design safety certification information document of the PlVS des (PSID)ign in thetime frame, 1994-1995 assuming a favorable preapplication review.

The PIUS design concept has already undergone tests related to the design principles. ABB has completed testing using the MAGNE Test Rig to simulate PlVS parameters such as diffusion and mixing across the primary loop / pool boundary with consideration for effects of turbulence, stratification, '

migration of boron, and others. Large scale tests of the PlVS design principles, such as flow and density lock operation, were done at the ATLE Test Ri These tests were used to validate the RIGEL code to calculate the design'g.s safety and transient performance. ATLE was a full height simulation of the PlVS pool. Other tests of the PIUS 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 Descriotion PIUS 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 PIUS design is completely sassive and is always in operation. The PlUS design consists of a vertical lo11ow cylinder, the reactor module, which contains the reactor core. The

• reactor 3,300 3

modulegallons)

M (870,000 is submerged in a large of highly borated concrete water. reactor The reactor modulevessel containinfs 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 floc 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 abysical flow barrier in the density locks between the primary loop and the morated 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 overcom,e DRAFT

_ _ _ _ , _ ___ _._________ - _ ~~__

i i

and the borated water flows into the core and shuts down the reactor. A natural circulation flow path is then established from the borated pool through the lower density-lock, up .through the core, and back-into the borated pool through the upper density lock for long term shutdown cooling. Unlike most reactors, PIUS does not employ mechanical control rods for regulating -

reactivity. Reactivity is controlled by the boron concentration anj temperature of the primary loop reactor water.

An active reactor protection system (RPS), with associated instrumentation and actuation systems, is also provided tn PIUS. -The RPS and the associated systems have the task of detecting departures from acceptable operating coaditions and -initiating coolant pump trip to cause density lock flow and a reactor scram.

Other aspects of the P]US design are ~similar to the passive LWRs being considered by the staff (AP-600 and the SBWR). Aithough PIUS is a PWR, its operating pressure (1,305 psi) is close to that of a.BWR. The proposed containment for the PIUS design is integral with the reactor building, similar

-to the ABWR and SBWR. Leak-rate has been defined as not- to exceed 1 volume percent per day at a design pressure of 26 psig., The acceptance leakage value is expected to be 0.5 percent at design pressure.

9 l

DRAFT

l l

D. PRISM (POWER REACTOR INNOVATIVE SMALL MODULE)

Development Historv ,

selected the Power Reactor Innovative.

The Small_ U.S. Department Module of Energy (PRISM)-design- as the (DOE) a dvanced liquid metal reactor (ALMR) design <

f to sponsor for NRC design certification. The conceptual design-for PRISM was

! developed by General Electric (GE) Company _ in conjunction with an industrial-team of commercial engineering firms. Research and development: support is

being supplied by the Argonne National Laboratory,- Energy Technology Engineering Center, Hanford Engineering-Development' Laboratory, and Oak Ridge i

i National Laboratory. In addition, a' steering group of utility representatives

! was involved in the PRISM design effort, ,

f I 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 -

l passive safety _ systems and greater safety-margins, reduced-cost through -

- modular design and construction, and possible future development of an actinide recycling capability. Although this last d urnative has not yet

!- been proposed in the current- application, DOE has duw sed studies evaluating j the use of actinides separated from spent fuel in' a advanced liquid metal j reactor:(ALMR) _ fast-flux core..

I The PRISM design has considered liquid-metal reactor (LMR) experience to date developed both nationally and internationally in terms of systems and:

'; components design, reliability data, and-safety assessments. This experience

! consists of operation of a number of facilities such as, EBR-II, Phenix, the--

l Fast Flux Test Facility ;(FFTF), the Joyo reactor in Japan,- and others.

L The PRISM Preliminary Safety Information Document (PSID) was1 submitted to the l NRC for review in November 1986, and the results of.an early_ NRC staff. review was the draft PSER (NUREG-1368)-issued in September 1989. In order tolobtain NRC approval'of its planned prototype, DOE plans to apply for preliminary L design approval in 1995. The DOE also' plans:to apply for standard design-j- certification in 2003 after a prototype demonstration. 'These plans are based- '

!. on the current DOE goals to demonstrate the commercial potential- for the-ALMR' i- by 2010, as ca11ed for in the Energy Policy Act of'1992, plant Descriotion-l l The PRISM plant design consists of three se>arate pcwer blocks each made up of three reactor modules. Each module has a tiermals output'of 471-MWt and an . ,

- electric output-of 155 MWe for a total (plant) output < of-1395 MWe,- The. PRISM

" design contains three turbines,3each supplied from a-power block. ' Options for i one or two power blocks are possible. PRISM operates-at'much hig a .

p . temperatures than current LWRs which will require a rigorous-evaluation of the -l

!- effects of_ creep and creep rupture on reactor vessel and: systems. The PRISM 1= design also relies on a highly automated and complex control' system utilizing 1

- digital ' processing. .  !

I.

r 1

- DRAFT 1

- - . - - - - - - - - - - - - - - - _ r-_ _ _--_w+ __ ~ .e e e- -ar--_-+-,-y-gem,..we 9 w y-w w .+ v f r i v s+y*g M g- gd y- '-a try wi g W m py y n i-Q ,- g-b ty- T'eg g' ag yd "y p y1 y F-ty N y b 'tF '9N$ 9F N"'"

p.:

i: ,

L The reactor module consists of the containment system, the reactor vessel, the core, and the reactor's-internal components. The reactor. vessel encloses and-

. supports the core, the primary sodium coolant' system, the intermediate coolant-system heat exchangers:(IHXs), and other internal components, iThe ~ vessel-is i

located just inside the containment vessel, which is located below grade-in - *

l. the reactor. silo. The reactor vessel is-penetrated only in the closuse head.

l The head is supported by the floor structure,- and the. floor structure is-

! supported by seismic isolator bearings to reduce horizontal movement'during

[' seismic events. The upper head of the reactor vessel is the closure head.

i The closure head also~ supports the-intermediate heat exchangers (IHX)'and the l electromagnetic (EM) pumps.

The main' components of the Nuclear Steam Supply System (NSSS) in' PRISMIare the I reactor module, primary sodium loop,' EM pumps, IHX, intermediate sodium loop l' and steam generators (SG). The primary sodium loop. is contained completely within the reactor vessel, which is hermetically sealed to prevent leakage of:

j the primary coolant. The EM pumps provide the primary sodium circulation.

! Synchronous machines provide flow coastdown capability to-the-EM pumps. Flow i coastdown is very important for preventing sodium boiling during a loss of EM:

j pump-power without reactor scram. Reactor-generated heat in the primary loop

is transferred through the IHX to the intermediate heat transfer' system 1

(IHTS). IHTS sodium is circulated by a centrifugal pump.. The IHTS operatas- z i at'a higher pressure than the primary loop so that, in case of a tube rupture in the IHX,'the sodium would not flow out of:the reactor vessel. A pressure of approximately-15 psig is used to assure:a minimum 10 psi positive pressure-

differential across the IHX from the IHTS to 'the-PHTS is maintained.- A .-

i- sodium-water reaction protection system mitigates _the effects of reactions j between IHTS 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.ista heterogeneous i arrangement of driver fuel and blankets.

l The PRISM core design is such that the net-power reactivity-feedback is-i negative in all- ranges of-operation, in all- transients, and in all: accidents

not involving voiding.- - For certain very~ low probability accident scenarios i involving sodium boiling, a positive void coefficient dominates and a net positive feedback can occur. In--all other situations without extensive i

voiding, an increase in temperatures produces: negative feedbacks from Doppler i and thermal expansion of the core and reltted structures that dominates the' t positive moderator density coefficient. 1The net negative temperature 1 L coefficient-is so large =that analyses predict all non-boiling transients and i' 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. -

s DRAFT f- l n ,

_ . ._ . , _ - . _ . . _ , . _ . .- ,m _ _ m - ., . . , _ _ . _ - _- . . . _ . . _ , _ _ , . . , _ _ ~ . . . _ _

i .

/

1

! 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 -

i inserted using _(1) the plant control system (PCS) for normal insertion.-

! the safety-grade reactor protection system (RPS) for rapid-insertion, and

gravity drop-into the core. If both the normal and safety-grade systems 1, 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 expansion modules (GEMS) consisting of l

tubes, closed at the top and open at the bottom, and filled with helium. If

the pumps are running, .the static pressure is high, causing the sodium level to rise to a high point in the GEM. However, with the pumps off, the static F pressure and sodium level drop, which increases neutron leakage. The reac-i' tivity. change provided by the GEMS between these two states is about j -70 cents.

! Normal shutdown cooling is achieved with the non-safety-grade condenser. If j the condenser-becomes unavailable, the: safety-grade reactor vesse'i' auxiliary

cooling system (RVACS) is used for RHR. The RVACS provides natural .

circulation air cooling of the containment vessel. The design-basis RVACS ,

i event assumes that the normal and auxiliary heat removal systems,. as well as -

i the Intermediate Heat Transport System.(IHTS) sodium, are lost immediately-l following reactor and primary EM pump trips. The'preapplicants' analysis has

shown that the RVACS heat removal rate .is sufficient to maintain fuel

! temperatures within acceptable limits, and temperatures of'the-internal I structures within the reactor vessel under American Society of Mechanical i Engineers (ASME) Level C conditions. The PRISM design also contains the nod '

! safety-grade auxiliary cooling system (ACS) to ~ assist the RVACS. The ACS uses.

natural circulation within-the steam generator- (SG) to remove heat indirectly l 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-

[ favorable 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 i coolant system boundaries, and .(3)-operation far below the sodium boiling .

i temperature, thus' obtaining the operational and analytical- simplicity :

associated with a single phase coolant.

i -

l l l 4 .

H l

i o

. l

DRAFT I i

. - - .- , -#,..;-.-,s .m,_. ..r ,,...,__-wdwn, myp'n e g,~n. r wke,-

I: j 4

i

• SECY-86-368, "NRC Activities Related to the Commission's Policy on '

the Regulation of Advanced Nuclear Power Plants," December 10, 1986 l

• SECY-89-350, " Canadian CANDU 3 Design Certification," November 21, 1989 3

• SECY-90-055, ' PIUS Design Review,"-February 20, 1990

• NUREG-1338, ' Draft Preapplication SER for the MHTGR"

• NUREG-1368, " Draft Preapplication SER for PRISM" 4

l

• NUREG/CR-5261, " Safety Evaluation of MHTGR Licensing Basis Accident Scenarios" l

• NUREG/CR-5364, " Summary of Advanced LMR Evaluations-PRISM and SAFR"

• . NUREG/CR-5514, 'Modeling and Performance of the MHTGR Reactor Cavity l-

Cooling System"-

l

• NUREG/CR-5647, " Fission Product Plateout in the MHTGR Primary System"

• NUREG/CR-5815 " Evaluation of-1990 PRISM Design Revisions" i

l l

t i

l; I j

~

l l

{ . DRAFT 1 i

Enclosure 3 i-

_ -.. ,-. _ _ .,. -. .~ . . . _ . . . , . . _ _ . . . , .-

, W2TX-P ?,T% ,s .

L ' M ~ h y ?- W :: - .

^

> n mew -

.' i . Enclosure 2 .

s j c# * " %

POLICY ISSUE November 23, 1992 (lnfOrrnation) SECY-92-393 Eg.t: The Comissioners from: James M. Taylor Executive Director for Operations Sub.iect: UPDATED PLANS AND SCHEDULES FOR THE PREAPPLICATION REVIEWS OF THE ADVANCED REACTOR (MHTGR, PRISM, AND PIUS) AND CANDU 3 DESIGHS Puroose: To inform the Comission of the staff's current plans and schedules for conducting preapplication reviews of the advanced reactor (MHTGR, PRISM, and PIUS) and CANDU 3 designs.

Backaround: In SECY-91-161, " Schedules for the Advanced Reactor Reviews and Regulatory Guidance Revisions," the staff informed the Comission of the following estimates for completion of the preapplication reviews:

• PRISM November 1992

• MHTGR December 1992

• CANDU 3 June 1993

• PIUS July 1993 The staff based these dates on broad planning assumptions including the prea cation schedules,availability pplicants' design of the Office certification of Nuclear appli-Reactor Regulation (NRR) resources to conduct the reviews, timely receipt of information to support the reviews, and a scope of review consistent with the previously issued draft preapplication safety evaluation reports (PSERs) for the PRISM and MHTGR designs.

Contacts: NOTE:

Thomas H. Cox, NRR TO BE MADE PUBLICLY AVAILABLE 504-1109 IN 10 WORKING DAYS FROM THE DATE OF THIS PAPER Brian W. Sheron, RES 492-3500 -

/

y i.

f*

The Comissioners )

J l Subsequent to issuing the estimates in SECY-91-161, a number of factors have necessitated a revision to the schedules for ,

PRISM,- MHTGR, CANDU 3, and PIUS preapplication reviews. 1 Some preapplicants have extended their design' certification- i application schedules or modified their proposed designs, i j Most staff technical review resources have been redirected 1 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

preapplicants requesting confirmation of their plans for design certification and, in some cases, a schedule for submitting additional preapplication review infonnation.

All responses were received by May 1992.

During the April 21, 1992, briefing on advanced reactor reviews, the staff advised the Comission that due to other i higher priority work, a relatively small number of staff 4

would be-dedicated to conduct the prea) plication reviews.

The staff proposed to concentrate on tiose key policy issues requiring Comission guidance and revise the preapplication

, review content and schedule to more effectively use the reduced NRR technical review resources.

In June and July 1992, the. staff held public meetings with l

each-prea)plicant to discuss the relevant scheduling infor-

mation, t1e desired scope of preapplication review,- and

schedules for additional submittals. The staff also j 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 agreed on a smaller, more focused scope for conducting the preapplication reviews. Previous planning assumed that the L scope of all preapplication reviews would be similar to the scope of the PRISM and MHTGR preapplication reviews documented as NUREG-1368, " Draft Preapplication Safety Evaluation for Power Reactor Inherently Safe Module Liquid

- Metal Reactor," and NUREG-1338, " Draft Preapplication Safety 1

Evaluation for the Modular High-Temperature Gas-Cooled 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 summary

i a r.e 1 1

fe c .

i

- The comissioners L . i 4 l

[ of this information and tte staff's rationale for sequencing the reviews. Based on-this-rationale and consideration of i available resources, the staff has identified the following-  ;

! revised estimates for completing the preapp11 cation reviews: l

• PRISM December 1993 4

• CANDU 3 December 1994 PIUS . April 1995 -3

• MHTGR December 1995 '

The. staff expects to. follow.the same review process for approval of the _PSERs as is being _used for the safety

evaluatici resorts on the evolutionary light-water reactor

! -designs and tin Electric Power-Research-Institute Utilities.

l Requirements Documentsa.Appro'ximately 6' months before

completing the review, the-staff will submit a draft-final

! PSER to the Comission. - With Comission consent, ~ the staff -

F will: forward the draft final PSER to the preapplicant, the l Advisory Comittee on Reactor Safeguards (ACRS) .and;the NRC-i Public-Document Room.- After considering-input during public j meetings with- ACRS and the preapplicant,. a final PSER will:

be forwarded to the Commission for approval..

l l The staff intends to conduct most of the preapplication .

! reviews with staff from the-NRR Associate Directorate for l l Advanced Reactors and License Renewal (ADAR), national L laboratory technical assistance, and support from the~ office o '

of Nuclear Regulatory Research-(RES).- Due to limited-staff-resources each design PSER will:bef developed in a sequential:

order. NRR technical: staff within the Associate Directorate i for Technical Assessment (ADT) will,::in_ general,_ not' participate';in the preapplication review. ADT technical staff resources are currently required for higher priorityf L operating reactor' technical = reviews and4 light-water reactor '

-(LWR) design certification. 'However, once;the draft final ,i PSER has been developed,!ADT management will review the-l report for its policy implications. The staff considers-

.this approach appropriate since the preapp11 cation review ,

considers the conceptual design, and final technic.1 l

! decisions on safety will-not be made until the_ design R

[ certification review when the ADT technical staff wil1The

! involved.-

l3 The: staff believes that the changes to-the preapplication-review scope-and schedule,'and:the approach for conducting:

NRR technical review,

provide the most effective use of NRC 1 resources. The'proposert schedules will: allow ~the staff to- o provide a timely response to the preapplicants-in important; d areas regarding their design certification application- {

plans'.9;By= emphasizing the key policyTissues for the-

~

7

- The Commissioners 1

~i-advanced reactor designs, the staff will address-the

. preapplicants' most significant questions about NRC's-licensing requirements. Resolution of these issues in the preapplication reviews is expected to allow the preapplicants to reduce the current uncertainty regarding design and design certification schedules.

The staff will continue its assessment of the schedular and resource implications of the reviews for these advanced designs. NRC preapplication review schedules may be altered to support-the recent Energy Policy Act of 1992 goals. The status of the preappitcants' 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.

Haig: There has been Congressional interest in this matter and the Chairman previously advised Senator Johnston and Congresswoman Lloyd of the schedules outlined in SECY-91-161. Therefore, the staff plans to submit this paper to the appropriate subcommittees, the Office of Management and Budget, and the Department of Energy.

s' mesWI or xecutive irector for Operations  !

Enclosure:

Summary of Input for Schedule Revision DISTRIBUTION:

Conunissioners OGC OCAA OIG OCA

, OPA ,

OPP EDO

ACRS ASLBP SECY '

i-j .c ,

a. ,

i

}y EElid l LIn March 1989, the NRC issued NUREG-1368,_" Draft Preap lication Safety = l '

i- Evaluation Report for Power Reactor Inherently safe Module Liquid Metal.

j Reactor," in which-it sumarited the results of its review of the Preliminary i Safety Information Document _(PSID) submitted in 1986. In March 1990, the U.S.

l- submitted Appendix G.to the PSID, ' Responses to

Department Issues in Draft of SER,"

Energyin (DOE)hich w it proposed-several significant changes to the i PRISM design. These-changes included increasing the power, adding an ultimate l shutdown system and containment dome, redesigning the reactor to add

[ expansion modules (CEMs), and changing to a single-wall-tube helical coil- gas 1 steam generator. .Brookhaven National-Laboratory (BNL) reviewed the revised i design and published its findings .in NUREG/CR-5815, ' Evaluations of 1990 PRISM i

I' Design Revisions." BNL'is also reviewing the performance of the GEMS and-the '

j consequences of a hypothetical core disruptive accident.- The Office of-staff is continuing to review the preap-Nuclear Reactor plicant's submittals and is writ Regulation (NRR)ing the final preapplication safety evaluation  :

l- report (PSER). The Office of Nuclear Regulatory Research (RES) performed the o i 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 L validation. -to investigate behavior of the new metal fuel,- to- assess reac-

!. tivity feedback, and to prepare for source term determination, <

i .

In a letter of March-12,1992,. DOE submitted the following schedule

i

• Preliminary Design- Approval Application - CY 1995 l

-

• Prototype Final Safety-Assessment Report 1997 i e Standard Design-Certification-Application - 2003 l

-(After prototype testing)

I This schedule appears consistent with the Energy' Policy Act of 1992 which

established DOE goals for the advanced liquid-metal reactor program to submit

- a preliminary design approval application to.the NRC by September 30, 1996, *

and to make a decision on prototype demonstration by September 30i>1998.

7 In June 1992, the National Research Council of the National Academy.of -

l Sciences published a report, " Nuclear Power: Technical-and Institutional-Options;for the Future,' in which-it discussed prerequisites needed.to l

l- preserve-the U.S. nuclear power option _and recomended that the: Federal l government: support key reactor designs. The Council recommended that the l PRISM design be the only preapplication design to receive government funding because of its unique ability as a breeder reactor.

I At a public meeting on July-1,11992, the NRC staff notified DOE that it would.

E need to delay issuing the PSER beyond the originally scheduled date of

[- November 1992. DOE noted that it had submitted all requested information to

[ the NRC:to support the preapplication review and requested that-the NRC L complete the PSER'as soon as possible to support DOE-in planning for design-

! - certification. DOE recently notified _the NRC of problems found with the W

-.r -, P- mw,-w,m i i--),.- w- g v a--w*, y-a-e-w v--

)

~s' 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 alans to conduct the PRISM preapplication review first to capitalize on the wor ( already completed and currently in >rogress. DOE has provided all necessary submittals to support the review and las been responsive to staff questions during the review. DOE's plans for preliminary design approval application in CY 1995 are supported by the National Research Council's recom-mendation of PRISM as highest priority for DOE support of the four designs in preapplication review. The staff intends to treat the PRISM fuel problem as an open issue in the PSER.

ILHlGB In March-1989, the NRC issued NUREG-1338, " Draft Preapplication Safety Evaluation Report for the Modular High-Temperature Gas-Cooled Reactor," in which it summarized its review of the PSID submitted by DOE in 1986 and 10 subsequent amendments. Responding to the draft PSER, DOE /GA submitted Amendments 11, 12, and 13 to the NRC. These recent amendments provide additional information for NRC's review of the originally proposed design.

NRR is reviewing the submittals on the fuel design and fission product transport analyses at the DOE laboratories. RES performed the early part of NRC's review and continues to support NRR's work with projects to provide formal documentation for reference in the PSER, to evaluate DOE's containment design alternatives, to investigate moisture ingress events, to assess data base adequacies, and to prepare for source term determination.

In its May 18, 1992, response to the NRC letter of February 18, 1992, DOE stated that it would not establish the schedule for the MHTGR design cert-ification until August 1993 when it expected to select the technology for the 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 identified in NUREG-1338. DOE asserted that its submittals include all the information needed by the NRC to complete its review for the final PSER. 00E believes that the final PSER is needed in 1993 for the nuclear power industry to understand that the MHTGR is a viable power reactor concept.

The National Research Council report recommended that the commercial MHTGR program be given a low priority for DOE funding because its U.S. market potential was judged to be low. However, the Energy Policy Act of 1992 established DOE goals for the MHTGR program to submit a preliminary design approval application to the NRC by September 30, 1996, and to make a decision on prototype demonstration by September 30, 1998. These new goals may result in a DOE schedule that would require earlier preapplication review of the MHTGR design and resequencing of other design reviews.

.,2 -

I

7.

l

'/ DOE has discussed >1ans to revise the MHTGR design to increase the power of the modules from tie current 350 MWt to 450 MWt at the preliminary design approval stage of design. certification. It stated that this power increase will not affect the key policy issues for the design. DOE recently informed NRR of problems in testing the reference MHTGR fuel. The preliminary failure rate for the latest test of the MHTGR fuel design is significantly higher than that needed to meet the MHTGR design criteria. DOE expects to complete the post-irradiation examinations in May 1993 at the earliest.

The staff plans to conduct the MHTGR review as the final preapplication review

. in the series of four projects, because of the uncertainties in the DOE schedule and design to be proposed for design certification application. When i DOE submits MHTGR design certification schedules the staff will reconsider the i preapplication review plans. DOE is most interested in obtaining feedback on 4 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 f 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 j standard design certification. AECLT, a wholly owned U.S. subsidiary of l Atomic Energy of Canada, Limited, (AECL) in Canada, has supported the CANDU 3 l

preapplication review by submitting a Technical Description, Conceptual Safety Report, Conceptual Probabilistic Safety Assessment, and several technology

transfer reports describing the CANDU design.

In a letter of March 18, 1992, AECLT informed the NRC that it could support a

standard design certification application in 1995 or 1996 if the NRC completed

, its preapplication review of the CANDU 3 by June 1993. On June 29, 1992, l AECLT gave the staff a schedule of submittals to support the preapplication l review. AECL has completed much of the final design work for the CANDU 3 l reactor and is negotiating to start construction in a Canadian' province which l could serve as a prototype for the CANDU 3 design certification in the U.S.

! In September 1992, AECL acknowledged that it would re-evaluate its design i

certification plans in the U.S. if Canadian construction plans did not materi-alize.

r

-The National Research Council report identified the CANDU 3 design-as a mature design that could be licensed this century. The report noted that the l

licensing process could be lengthy because of the difference in regulatory requirements between the U.S. and Canada. The Council did not find sufficient

advantages with the design to justify DOE support for design. certification.

I The staff has started some preapplication review on the CANDU 3 design. NRR is conducting two projects at DOE laboratories: a study of the CANDU 3 positive void reactivity coefficient and a review of the operation of the on-line refueling machine. RES has completed a systems study to identify- ,

j candidate event sequences for required safety analysis, and it has projects to '

assess data base adequacies, to perform preliminary transient calculations l-

. l

F iJ 9

t* .

l ** _using Canadian codes- to identify code needs for future independent' analyses, l- to initiate severe accident analyses with NRC codes, and to prepare for source term determination. RES.will also provide in-house analytical capabilities l- for itself and NRR for. the CANDU 3 design.

l' To better understand the CANDU'3' containment.perfomance and radiological-i releases, NRR is reviewing the consequences of- a large break loss-of-coolant

accident (LOCA)'with a failure to shut down. NRR is performing 4his work to-

{ support the Comission's decision on a key policy issue: the acceptability of L a design with a dominant positive void coefficient. The preapplicant'has not.

performed this analysis for CANDU 3 .and has supplied little directly relevant

, information on the event and its consequences.

1 AECLT is having problems getting proprietary information released from Canada i to the U.S. This has delayed the staff in obtaining Canadian codes thus i interrupting RES's work to use these codes for preliminary calculations. -Code-l- 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 delay the review schedule. In a letter dated September 23, 1992, the-staff informed AECLT that'an inability to transfer proprietary material to the U.S. '

i may affect the proposed'CANDU 3 preapplication review schedule.- AECLT is now

pursuing transfer of proprietary material-directly from AECL to the NRC.

! The staff plans to conduct the CANDU 3 review as the second preapplication l review because the-design and experimental-data base are already sufficiently-

! developed to su) port the review. The June 1994 PSER issuance assumes prompt

resolution of tie present problems releasing proprietary information required l for the review from Canada to the United States.

i El.Wi l In October 1989, Asea Brown Bovert (ABB) Atom requested that the NRC perform a i licensability review of its Process Inherent Ultimate Safety-(PIUS): plant-L design. ABB Combustion Engineering (ABB/CE) of Windsor,-Connecticut,-is the-

! direct representative of ABB Atom-in the U.S., and is the official.preappli- ~

l cant of record.

i-

! In May 1990, ABB/CE submitted a five volume Preliminary Safety Information

' Document (PSID) to su> port its request for a preapplication' review.. NRR has started a project witi BNL to support core neutronics modeling. RES.has _.

I completed a systems. study to-identify candidats event sequences for-required

. safety analysis, and it has projects .to-assess data-base adequacies, to

perfom preliminary transient calculations using the _ existing-TRAC code, to ,

identify code needs for future independent analyses, to initiate severe-

! ' accident analyses with NRC codes, and to prepare for source term determi-i-  : nation.- - RES will also provide.in-house analytical capabilities:for itself and' L NRR for the PIUS design, i

! In a letter.of April- 22; .1992, ABB stated that it would submit a design:

certificat.oni application in 1994 or 1995 if (1)'the NRC issues a preap- i 1- plication safety evaluation report (PSER) by April 1994 that does'not require  ;

l e

[ -

l

\

J the commercial environ-significant mer.t design at that changes time is to the favorable PIUS to that design,ABB decision. and (2)is negotiating with the Italian state utility to. support testing of the PIUS design and will givo the NRC details of its overall test plan when the basic negotiations are complete.

During an August 6,1992, meeting, ABB informed the staff of a proposed change to the PIUS design. The design change involves adding four " scram valves" and associated piping. The feed lines to these valves take' suction from the borated reactor pool water, and the valves discharge to the suction of each of the .four reactor coolant pumps. Activating the valves is expected to result in a rapid and uniform insertion of boron by a means redundant and diverse from the passive scram process. The passive scram through the density locks will still be the ultimate shutdown process. ABB plans to submit the design '

change in a November 1992 supplement to the PSID. ABB also plans to submit a PRA supplement in early 1993 for the PIUS design.

The National Research Council report concluded that the PIUS design would not likely be ready for commercial operation within the next 20 years and had a low priority for DOE support. The lack of operation and regulatory experience is expected to delay acceptance by utilities of this advanced LWR design.

The staff plans to conduct the PIUS review as the third preap)11 cation review because the design is presently at the conceptual stage and tie experimental data base for the design is still being developed. ABB is most interested in obtaining feedback-on the implementation of the key policy issues for the PIUS design. Continued emphasis by the staff in obtaining Commission guidance for resolution of these issues will provide ABB feedback in advance of the final PSER. Conducting the PIUS review third will allow ABB time to develop the design more fully and respond to staff questions without impacting the preap-plication review schedule.

l f

l

)

. , , , . , - -- . . . - . . - - :D