ML19003A384

SECY-95-050: Change in Plans for Design Certification Review of the Candu 3U Reactor

ML19003A384
Person / Time
Issue date:	03/03/1995
From:	Taylor J NRC/SECY
To:
References
NUDOCS 9503130330, SECY-95-050
Download: ML19003A384 (42)

Text

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F.ELEASED TO THE PDR :

~ . ~'l/9.r iniflj ~

, POLICY ISSUE March 3, 1995 (Notation Vote) SECY-95-050

[QB: ThP C?mmissioners EB2ft: James H. Taylor Executive Director for Operations

SUBJECT:

CHANGE 114 Pl.ANS FOR DESlGN CERTIFICATION REVIEW OF THE CANDU 3U REACTOR PU!!POSf.:

TQ inform the Co11r.1ission of a change in the plans of the Atomic Energy of t 111urla limited Techn~logles, Inc., (AECLT) to submit a revised safety analysis re~vrt (SAR) for final design approval (FDA) and design certification under 10 CFR Part 52 for the CANDU JU design, and to request approval for modification of the previously proposed review plan.

CATEGORY:

This paper covers a signi~1cant schedule and resource change in staff activities.

f.IACICG!\Ql.lilQ:

The CANDU JU Is a 450 H\le, heavy water-moaerated and cooled, pressure tube reactor developed by Atomic Energy of Canada, Ltd. (AECL). AECL developed the design from previous CANDU reactors, most notably the CANDU 6, a 600 t!We NOT!:: TO BI: HADE PUBLICLY AVAU.ABLE WHEN THE FINAL SRH IS MADE AVAILABLE CONTACT:

G. Harcus, PDAR 415-1111 Cl. S* oletti, POAR 415-1104

• The Commissioners design. There are 25 CANDU reactors In operation and 19 under construction around the world. CANDU reactors have operated for over 175 effective full power years.

In December of 1988, a U.S. company, AECLT, the U.S. representative of AECL, was created as the preappllcant for the CANDU 3U licensing effort in this country. In a letter to the U.S. Nuclear Regulatory Commission (NRC) on Hay 25, 1989, AECLT Informed the NRC of its Intent to seek design certlflca-t Ion of the CANDU 3 under the provisions of 10 CFR Part 52.

On ~pril 8, 1993, the staff submitted SECV-93-092, "Issues Pertaining to the Advanced Reactor (PRISM, HHTGR, and PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory Requirements.* The staff described five policy Issues associated with the CANDU 3 design: accident evaluation, source term, containment performance, positive void reactivity, and design of the control room and remote shutdown area. On March 24, 1994, the staff issued SECY-94-079, "Schedule and Resource Estimates for CANDU 3 Design Certification Review.* In this paper, the staff estimated the resources and confirmatory research required to review the CANDU 3 for design certification. The staff estimated 105 full-time equivalents (FTE) and S2.2 million for the Office of Nuclear Reactor Regulation (NRR) to complete a 54-month review schedule starting In fiscal year (FV) 96, and 23 FTE and $18 million In confirmatory research.

On September 30, 1994, AECLT submitted an application for FDA and design certification under 10 CFR Part 52 for the CANDU 3U design. The staff completed an acceptance revlP.w of the application and sent AECLT a letter on Its findings on December 15, 1994 (Attachment 1). The staff informed AECLT that a docket number had been assigned to the application to facilitate public access to the correspondence and rovlew Information, but the staff did not intend to develop a review schedule until the updated safety a11 .. lysls report (SAR) and s'hedules for all outstanding Information had been submitted. The staff assigned docket number STN 52-005 and requested a response In 30 days to Include a schedule for rnl:,..ltt1ng the missing Items.

On January 19, 1995, AECLr sent the Conrnlsslon a letter In which It stated that the response would be delayed until no later than February 8, 1995, pending a review of work planning and scheduling.

On January 30, 1995, the staff met with AECLT to discuss the CANDU status (see viewgraphs, Attachment 2) AECLT again expressed concern regarding the proposed fees for the CANDU review. AECLT proposed changing the tinting of some of the major NRC mtle~tones for the review effort to start In l.pril 1997.

They al~o proposed a different allocation of staff resources for the review.

Under their proposal, the FTf expenditures would be low In the beginning and at the end, but would grow more rapidly and to a hlg~er peak than staff had proposed In SECY-94-079. In response, the staff described how reviews are conducted and how staff time is scheduled. The staff also stated that the fee issues would be decided by the Conrnls)lon and that a paper was being prep~red for the Conrnission on this subject (SECY-g5-035, "Reassessment of Fee 8ill~ng

The Conmlssloners Practices and Fee Policy for Office of Nuclear Regulatory Research (RES)

Activities Associated with Design Certification (DC) Applications").

On February 2, 1995, AECLT responded to the staff's December 15, 1994, letter, by submitting a schedule for updating the SAR and submitting other required Information (Attachment 3), AECLT proposed to update the SAR In stages with the final update to be submitted In January 1997 and proposed for NRC to start the design certification review In April 1997. In July 1998, AECLT would submit other* required Information such as the Level Ill probabilistic safety assessment (PSA) 4 nspectlon, test, analysis, and acceptance criteria (ITAAC);

design acceptanc~ criteria; technical specifications; test programs; and severe accident mitigation design alternative (SAMOA). AECLT proposed for the staff to issue the final CANDU safety evaluation report In October 1999, 30 months after the staff starts the review. AECLT also proposed that the NRC only do limited work on certain generic Issues and policy issues over the next 2-plus years. During this period, AECLT Indicated that It will limit funding for NRC review costs to S3 million over that time period. AECLT stated that 1* rould not go forward with the review If the NRC costs were to approach SS011.1lion.

The proposed schedule to complete the SAR and proposal to review only generic licensing Issues would substantially delay the staff from the schedule and resource estimates In SECY-94-079. Further, the proposed plan for reviewing selected Issues and constraining the level of effort departs from normal practices and could result In sl9nlflcant resource an~ schedule Inefficiencies.

In response to the AECLT request, NRC staff plans the following actions:

Hold discussions with AECLT on the Issues listed In Enclosure 2 to their letter of February 2. 1995, to better understand the type and scope of rev1ew they propose and to determine which areas should be considered highest priority for review by NRC staff, using appropriate contractual assistance.

2. Assign NRR staff to work in the area or areas of first priority.

3 Inform AECLT wl ?n we ha*1e spent 90 percent of the a*;ailable resources.

This notification will state that w~ will cease work as soon as the resource I imit is reached, unless A[CLT wist.es to provide additional support

4. r.ease all NRC activities in areas not dir~ctly related to the requested rev1ew areas. including NRR contract dct1vities and deslsn-speclflc Office of Nuclear Regulatory Research (RES) activities, lncluolng the development and use of Independent audit codes.

The C0Dt11lssloners 5. If AECLT requests detailed schedules and cost estimates for the tasks undertaken, we would charge the costs of developing the estimates to the applicant. Such costs are likely to be significant bec3use Information may be needed from many technical staff,

6. Begin desir~ certification review of the SAR after AECLT completes and submits alf necessary revisions. According to their letter, the revisions and submlsslo~s will be completed by January 1997, which will delay NRC from beginning a full-scale review by 15 months from October 1995 as stated In SECY-94-079. Therefore, delays In all planned review activities, s~pportln~ research activities and potential staff resource shortages would significantly delay completing the FDA. However, the staff cinnot develop a detailed review schedule until July 1998, when all of the required information, including the completed PSA, ITAAC, technical specifications, tests programs, and SAMOA, has been submitted.

7. Send AECLT revised cost and schedule estimates when the generic licensing Issue review has been completvd. Completing certain generic reviews before the design certification review may result In some limited savings In specific areas of the design certification process. However, the Inefficiencies of conducting a resource limited review effort prior to the design certification review, which may Include starting and stopping work as resources expire, will likely result In some net increase In the cost estima!e.

A. Continue to hold periodic ~enior management meetings with the Atomic Energy Control Board of Canada. However, defer any planned expenditures of resources, such as the planned exchange of personnel until the start of the design certification review.

RESOURCES; The FY 1996-1997 NRC budget contains re!ources In accordance with the review plan proposed In SECY-94-079. Re~ourcei to conduct the design certification review for the CAl'DU 3U design In respunse to AECLT's revlsPd proposal will be addressed during the FY 1996-2000 Internal Program/Budget Review process.

RECOl!MEHQATIOH:

Approve the ~taff's ,1an for reviewing CAHDU JU Issues, this plan Includes:

deferring the design certification review, doing a limited review of selected Issues Identified by AECLT within Its resource constraints, cessation of contract activities and RES efforts unrelated to specific AECLT requests. and give AECLT revised estlmatrs for the design certification review when the staff Is ready to begin this effort.

The Co11111lssloners -s-On Harch 1. 1995, the staff was told by AECLT that there are Indications that the Board if Directors of AECL has placed the CANDU design certification on hold. A~C~T Is presently seeking clarification of the Implications of the Board action. AECLT also Indicated that they may be visiting Individual Co11111lssloner~ sometime next week, when they have further Information. Should AECLT wish to pursue continuation of NRC's review, staff reco11111ends that the proposed plan outlined In this SECY be adopted. Attachment 4 1s a proposed letter the staff pla~; to send AECLT If they Indicate they wish to proceed with limited review.

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Attachments:

C ~;cut1vei0trector for Operat Ions I. Staff Acceptance Review Letter

2. AECLT Viewgraphs from Jan. 30, 1995 Heeling

3. Letter from AECLT (Feb. 2, 1995)

4. Draft Letter to AEC'

• Commissioners' comments or consent should be provided directly to t~~ Office of the Secretary by COB Friday, !larch 17, 1955.

Commission Staff Office comments, if any, should bP submitted to the Commissioners NLT Friday, March 10, 1995, with an infor-mation copy tn the Office of the Secretary. If the paper is of such a ndtur~ that it requires add1t1onal review u~d comment, tht.* t.:omm1si,uonf!rs and the Secretariat should be apprised of when C'Ommcnts may be expected.

OlSTRlllUTlON:

Comm1ss1oncrs OGC

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, Cl, C. -

December 15, 1994 Mr. A. D. Hink, President AECL Technologies, Inc.

9210 Corporate Boulevard Suite *10 Rockville, Maryland 20850

Dear lifr. Hlnlt:

SUB.I~!:'!', RESULTS OF THE ACCEPTANCE REVJEV FOR A~CL TECHNOLOGIES' APPLICATfON FOR FINAL DESIGN APPROVAL AHO DESIGN CERTIFICATION FOR THE CAHDU 3U DESIGN In a letter dated Septl!llber 30, 1994, AECL Technologies (AECLT) submitted Its application for final design approval (FDA) and standard d~slgn certification (DC) under Part 52 of Title 10 rf the Code of Federal Reak *lions (10 CFR) for the CAHDU 3U design. The contents of the appltcatton were provided In the form of a CAHDU 3U Safety Analysis Report (SAR) consistent with the fonaat In Regulatory Gulde 1.70 and the Standard Review Plan. The application letter acknowledged that certain lnfor11atlon required by 10 CFR 52.47 WAS mtsstng fr1111 \he application. The missing Information was Identified as the inspections, tests, ~nalyses, and acceptan~~ criteria (ITAAC), technical specifications, severe accident mitigative design alternatives (SAPl>A), and the f*1lure 1110des and effects analyses (FKEA). It should also be noted that the required Level II and Level Ill probabilistic risk 1~sess11ents (PRAs) and deslg~ acceptance crlterl~ (OAC), If any, were not Included as part of the CA!lllJ 3U application.

In accordance with 10 CFR 2.101, the staff perfor11ed an acceptance review to deter11lne If the CAHDU 3U was sufficiently c011plete to enable the staff to carry out the design certification review. The staff has deter11lned that a stgniftcant amount of tnfon11atlon ts either missing, or In a for11 which would cause the staff to expend a great deal more resources to com,lete the DC review than previously anticipated. The staff has previously indicated tt could proceed with the review In the absence of certain infol'lllatton: JTAAC, technical spuciflcatlons, and SAMOA. This position was based on t11r staff's experlenc~ '.hat the areas In question would not require a detailed review early In the review process. However, a schedule for sublllittal of these and other *tsstng tte*s ts needed. Subllllttal of these items will influence the scheJule b~* which the CAHDU 3U review is carried out. Early submittal of all Information Is necessary to assure that staff can develop ano maintain an effective r~vte* schedule. In that regard, the staff underst1nds that tt ts your Intention to submit a PRA, completed through level III, '.n about a year.

Furthermore, It should be noted that tt Is the staff's tnten~ to keep the nUllber of DAC to a atnl111t1111.

Attachment t

A. D. Hink -?-

HRC also requires a clear Identification In the SAR that CAHDU JU 111eets the applicable U.S. codes and standards and the HRC's General Design Criteria (GDC). While the SAR Indicates that aany of the Canadian (CSA) codes and standards cited are equivalent ~o the existing U.S. codes and standards, the equivalence Is not explained In sufficient detail to demonstrate th~t the CSA r.odes and standards do Indeed meet our requ1ret1ents. Where Canadian standards are necessary because there are no U.S. standards or acceptance criteria, an equivalent level of safety analysis to the GDC should be provided. provides further details related to lnfonaatlon needed by the stiff to continue the revle-.1 cf tl:e CAHDU JU design. Please be aware that the enclo~ure does not represent a cU1prehens1ve list of deficiencies; It Is 11mlted to those found during the 1lmlted acceptance 1 *view. Other issues will be Identified to AECLT through requests for additional Information during the review.

The staff has assigned Docket number STH-52-005 to the CAHDU JU application to facilitate public access to correspondence and review Information. A copy of the Federal Register notice ls enclosed (Enclosure 2) for your information.

AECLT shoulci reference this docktt number when submitting the requisite J8 updated copies of the CAHDU JU SAR pursuant to 10 CFR 50.4 for the st~rt of the DC review. The staff does not plan to develop a detailed review schedule until the updated SAR and schedules for all outstanding information have been submitted.

In response to your application letter, the staff has considered the resources needed to issue an FDA for the CAHDU JU design. The staff has examined the review proce~s and the full-time equivalent (FTE) staff required for the completion of each of the evolutionary plant DC ~evlews. While there should be so'lll! potential resource savings from the staff experience In c'nductlng DC revle.*\, i~ is expected that this will be offset by the potential difficulties lnhPrent In reviewing a non-light water design. The acceptance review confirms that there ar2 a number of significant Issues which potentially cculd require enhanced NRC review efforts. Therefore, the staff still considers the resource and schedule estimates made in the Harch 24, 1994, Co11111ission paper (SECY-94-07~. *~chedule and Resource Estimates for CANDU J Design Certification Review") to be appropriate at this time. This estimate included 105 *TE and $2.2 *1llion for the Office of Nuc'!ar Reactor Regulation (NRR),

and 2J FTE and $18 *llllcn for the Office of Nuclear Regulatory Research (RES). Two additional factors ~ill influence the ultimate schedule and the fees actually billed to AECLT for the DC review: how closely the application cocnplles with the Information re~ulred by the Standard Review Plan; and the results of the ongoing agency reevaluation of the NRC fee structure related to research needed to support licensing of advanced reactor designs. Until both of these Issues are resolved, the staff cannot provide a more complete estimate of the costs and schedule to complete the CANDU JU review.

A. D. Hink . 3-In this regard, AECLT should provide ~lthin 30 days of the date of this letter a schedule for sublllittal of the updated CANDU 3U ~AR and the other missing infol'lllation ~uch as the ITAAC, l~:~nical spe:ifications, and SAMOA. We understand that unless we hear differently from you, the staff plans to continue It! limited work on some key issues such as void reactivity and shutdo1t11 system reliability. If you have any question regarding this letter pleast contact the NRC project manager, Dino C. Scaletti at (301)504-1104.

Sincerely, t?~~c:JiJ~~

~s*H~ Crutch e d, for Advanced Reactor and l.1cense Re1;ewal Office of Nuclear Reactor Regulation Docket No.52-005

Enclosures:

I. Request for Additional Information

2. federal ~s~ Notice cc w/enclu)ures:

See next page

ENCLOSURE 1 REQUEST FOR ADOITIONAL IHFORKATION 210.1 Classification of Structures, Sy .eas, :nd Cl)llponents (SSCs)

a. SAR Section ~.2 and other SAR Sections contain references to (1) a series of Canadian Standards (CAH(CSA N-285 through H-290) that provide requirements for safety c ~sstfir~t1ons, design and fabricatior, quality assura~cl, and seis*ic qualification, and (2) a series of Safety Oestgn Guides that apparently provide dP.sign guideline~ for Euch subjects as sets*tc analyses, ~ode class~ffcatton, and ptpe rupturft protectfon. SAR Subsection 3.2.7.2 states that where ther<* ts an 1ndivtdual 1*efere11ce in the CAH/CSA Standards to ASHE Code Section III, Division J or 2, ft ts the intent of the CAHOU 3U design certification process to Incorporate all Articles ~n iiSME Subsr~tfons fiB, NC, llD, NF, CB, CC, and Divisions 1 and 2 Appendices 1n their entirety. Huwever, for some SSCs classified a1 CSA Class 2, 3, IC, 2C, or JC, SAR Table 3.2-1 identifies the principal construction codes and standards as both ASHE Section III and one of the CAN/CSA Standards. Because the certified de$ign rule will be a part of the NRC regulations, all references should be aade to U.S. codes and standard~. The SAR s~ould clearly discuss the acceptability of th' de!lgn In those areas where the design deviates frOlll the criteria In the ASHE Code or U.S. industry standards.

b. SAR Table 3.2-1 lists the ouality assurance for all SSCs as one of the Z299 series Quality Assurance Standards. SAR Section 3.2.4.3 contll~s only a brief description of these standards. With respect to the inforaatlon in the SAR, It Is not cle;r to the staff which of these standards, if any, contain a c0111ltll"nt to 10 CFR 50, A~pen dlx 8. The SAR should either state that Z299 meets &11 the require-ments of Appendix B, or replace Z299 with a COllllltment to Append!~ B for all SSCt classified ~s DBE (Seis*lc Category 1).

210.2 Basis for the Alternative Safety Assessment Atomic Energy of Canada Limited Technologies (AECLT) Indicates that a :.uk-Before-Break (LBB) appr*oach may be us~ to demonstrate that catastrophic pipe failure has a very low probability of occurre~ce.

If the LBB approach ts to be used In the CANDU 3U design, the details of LBB methodology and acceptance criteria should be subllltted in the SAR for staff review.

210.3 Computer Progruis It Is not acceptable to provide only 3 list of computer prograi:s.

Additional tnforaatton for each progr1111 tn accorda1,ce with the guidelines of SRP 1.9.1.11.2 must be provided In the SAR to demonstrate that the program has been verified for Its applicability and validity.

210.4 EX)1rtll!lntal Stress Analysts SAJ Section 3.9.1.3 states that no expert11ents are required to qualify an~ 11ech1ntcal systEtlS or coeponents; however, shake tests may be requt~d on SOllll fuelling 111chtne c011Ponents whtch wilt be specified in accordance with Section 3.10. It should be noted that expert11entat stress analysis ts not an equipt1ent qualification test. Therefo~e, AECLT should clearly state i~ expert11ental stress analysis is r.ot used tn the CAHDU 3U design. If tt is to be used, sufficient infol"llation must be presented in the SAR c01mitting that the requiregents of AJ>rendlx II to ASME Code, Section Ill, Division 1 will be 11et.

210.5 Seisaic process Qualification Testlnq of Safety-Related Mechanical Syste11s Per the guidelines of SRP 3.9.2, SAR Subsection 3.9.2.2 should include tnfonaatton on the seismic analysis aethodotogy and approach for all Category I systet11s, C011Ponents, equipment and their supports.

210.6 Dynamic Response Analysts of Reactor Structure ~nd Fuel Channel Assemblies Under Operational Flow Transient And Steady-State Conditions

a. AECLT Indicates that the dynamic rcsponsP. analysts ts not planned at this tlllli! because the operational flo-t transient and steady-state conditions do not produce large loads to justify such analysts.

Because the CAHDU 3U design Is the first of a design revle~i by the staff, documentation or references containing the results of req~tred analyses and tests (either from a prototype of CAHDU 3U or an~ other slAllar CANDU plants) must be provided for staff review.

In Sections 3.9.2 and 3.9.3 of the draft SAR, the following ddltlonal information as described in SRP Section 3.9.2 should be

-ovided for verifying the design adequacy of the reactor vessel and iternals:

!tailed infonnatton on the reactor vessel and Internal c0111ponent sign, Including their configurations, major functions *~d design

• ameters, material requirements, definitions of operational and

lLad condition design loads such as LOCA and SSE, design eptance criteria such as stress ind deflection limits, and

~ods and procedures for conducting *:;*; and analyses to ensure ir structural Integrity and functional oper*blltty.

etalled description of the preoperatlonal vibration asse~:i':ll!nt

9ram for verifying the design adequacy against flow-Induced

-rations, lncludin~ designation of the prototype reactor, resu1ts pre-testing vibration prediction analjsls a~d acceptance criteria, preoperatlon1l flow testing program and pl,nned instru-mentation for vibration monitoring, and the post-testing visual Inspection program.

~10.7 Ccrrela 1*~s of Reactor Structure Asslllbly Vibration iests with the Analytical Result' AECLT lll'Jlcates that the CAHIXI 3U Jesigr, Is based on proven technology and that the key features of the CAHDU 3U design are essentially Identical tQ those of operatlr.g plants. Since the CAHDU 3U design is the first of 1 desl~n reviewed by the staff, a discussion 1111st be pro-vided which d~scrlbes the meth:ld~ use I (either r~r the CAHDU ~u design or for any previous sl*llar operating plants) to correlate the Lest results with those frOll dyn111lc analyses.

210.8 C0111Ponent Supports When snubt-lrs are utilized a~ supports for snfety-r6late~ systeins and

.. CO:ll?on*nts, SAR Subsection 3.9.3.4 should Incorporate relevant provisions as specified In the SRP 3.9.3 for establishing acceptable snubber operability assurance.

210.9 Control Rod Drive Syste* (CRDS)

a. At.** 1.T lnd'catps that the SRP acceptance crlt!rla are not ;:ppl lcable to the CA.'iD:J JU desl9:i ar.1 that the Canadian Standard CSA-N285.0 is used In th2 design of C!UJS. At this tl111e, the staff has not confirmed that the SRP at::eptance criteria are not applicable to the CANDU 3U design nor doo~ the staff have a clear guidance position on hOY. to revl!!W the acceptability of applying Canadian Codes and Standards to the CA!~ 3U design. HO\oJ9ver, a sl111ple reference to Canadian ~tandards Is not accep.able. The SAR should discuss ""1y the SRP criteria are not considered applicable and clearly Identify all criteria In these standards that are applicable to the CAHDU JU design.

b. In Secl!on 3.9.~ of tt.e draft SAR. provlt!r. a det~tled description of th~ testlr.g t~d analysis conducted for ensuring control ro~

hosertlon and safe shutdovn of the reactor under faulted plant conditions such as sets11lc and LOCA ev.Jnts.

210.10 Calandrla Vessel Internals (CVI) and F~elllng Hachlr.e (FH)

AECLT lndlc~tes that th! SRP criteria are not app1l~able to the CAHDU 3U design and th1t design alternatives lncludln!J tl1e use of Ca'ladlan Safety Design Guides and Standards are applle~ to thll desl£n of CVl A! dis-cussed In 3.g,4, the SAR should discuss wily the SRP cr~terla are not considered applicable ~nd rlearly describe the design :lternatlves and l~entlfy all criteria In those standards and d~slgn guides that are appllcaLle to the CAHOU 3U design.

210.11 In-Service Test.Ing of Pugps and 1fahes The lnfonnatlon presented In this section should be expanded to ad:iress lmplementat1on of staff positions specified In s~CY-90-016 and llRC GenPrlc letter 89-10. Althougt- a detalled IST program .iill be developed by the COL appllcaht, an lST plan of sufficient infor11atlon must be submitted to demonstrate that all safety-related PIC!IPS and valves Including ~afoty/relief valves of the CA.~DU 3U design can Ilg adequately tested at the required frequenr.y. Justlficatlo"s for testing at cold shutdown or refuellir.g outage 1111st be provided for those pumps and valves that cannot bo testP~ quarterly.

210.12 Evaluation of Safety Issues The following issues should be addressed:

Issue A-1 *11aterha1111er*

Issue 70 "PORV and Block Valve Reliability*

Issue 7'> *unanalyzed Reactor Vessel Thermal Stress During Natural Convection tooldown*

Issue 87 "Failure of High-Pressure Coolant Injection Steamllne Without Isolation*

Issue II I' ! "Perf~rmanc~ Te~tir.g of PWR Safety ~nd Relief Valves" 210.13 Seismic Qualification of Seismic Category I Jnstrumentation and f.lectrlcal Equipment

a. The title ot SAR s~ctlon 3.10 and its contents should be expanded to Include not only seismic but also dynamic qualification of both electrical equlpme~t and mechanical equlproent.

b. Information and co=nltments must. :,t pro~lded In the SAR de1110nstratlng that relevant quallf lcatlon criteria specified In SRP 3.10, Rev. 2 have been met.

220.l Sections 3.3.1.2 and 3.3.2.3 of the CANDU JU Safety Analysis Report (SAR) ~late that the method for deterv.*nlng pressures generated on str11ctures due to des 1gn wind Is spec. if led In accordance wl th the requirements of the National Bulljlng CodP. of Canada Although Table 3 3-3 of the SAR state~ that the load de.ennl* .* tion criteria of SRP 3.3.l and 3.3.2 (I.e., ANSI ASB.2 and ASCE Paper 3269) are complied with, the extent and manner of compliance Is not clear to the staff.

AECLT should ~*~~rly ~emonstrate how COlll!>llance with the U.S. codes for wind load determination has been achleve11 and do* *1111ent the results 111 the SAR Th~ SAR should clearly dlscu~s the acceµtabl1 *ty of *he d~slgn In those areas 1-'here the design deviates from the criteria In the ASHE Cooe or U.S. Industry st.aodards.

22~ 2 Section 3.5.3 discusses the concrete missile barr*cr design crlt<*oa but does not address the design ~f steel missile-resistant ~Jrrlers. AECLT should addre~~ compliance with SRP 3.5.3 guidelines for the design of steel missile-resistant barriers.

Z20.3 The CAii.JU 3U st~ndard design defines two levelY of earthquake: doslgn basis earthquake (08£' and site destgn earthquake (SOE). T~e design load comblnatl~ns specified In Section 3.8 Imply t~at t~e DBE Is equiva-lent t~ the ~afe shutdown earthquaKe 'SSE) and SOE Is equi**alent to the operating basis earthquake (OBE). The d1111>tng ratios spectfted tn Tabla 3.7.1-2 coeply wtth RGI.GI for the OBE but the SOE da1111ing ratios do not comply wtth the OBE damping ratios of RG I.6I. AECLT should provide Justtftcatton for ustng d111Ptng ratios that are not tn COll;>ltance wtth RG 1.61.

220.4 Table 3.7.I-l of the CAHDU 3U SAR provides the design basts earthquake ground response spectra tn the horizontal dtrectton for different damping values. In thts table, the spectftc frequency values of the lower, tnterlledtate and upper frequency ranges listed are not defined.

AECLT should clearly define these frequency ranges to enable the staff lo delel"lltno the COllP11ance of CAHDU 3U design response spectra wtlh the RG I.60 response spectra.

220.5 Table 3.7.5.1-1 of the SAR states that the CAHDU 3U design ground response spectra envelope ts the RG I.60 spectra tn the horizontal direction but the vertical design response spectra ts less t~an the RG 1.60 vertical response spectra. AECLT should provide Justiftcatton for using vertical design response s~eclra that are not tn compliance wtth RG I.60.

220.6 Table 3.7.5.1-2 of the SAR states that the seismic system analysis does 11ot c0111ply with SRP 3.7.2 guideline for considering 5% additional eccentricity to account for accidental torsion. AECLT should Justify how accidental torsion effects wl~l be considered In the CAHDU 3U standard design In order to comply with the guidelines of SRP 3.7.2.

220.7 Section 3.7.2 of the CANDU 3U SAR does not provide the details of the seismic soil-structure Interaction (SSI) analysis for compliance with the guidelines of SRP 3.7.2. This lnfonnatlon should be provided.

220.8 Section 3.7.3 of the Cl'\HDU 3U SAR docs not provide the crtterta for the analysis and design of cable trays, conduits, HVAC and their supports.

This Information shoul~ be provided.

220.9 CAHDU 3U standard design e11Ploys modular construction for various components of the containment Internal structures. Although seismic analysis and deslg~ procedures for various systems and subsystems are presented In Sections 3.7 and 3.8, these generally address conventional safety related structures which may not be totally apollcable lo structures COlllPrlsed of the lllOdular units used In the CAHOU 3U design.

Sections 3.7 and 3.8 should discuss seismic behavior and ri~~lgn analysis methods for the CAHDU 3U modular construction.

230.l In Chapter 2, each section and subsection needs a COL applicant action Item statement so the COL applicant will know what has to be done to assure that the site fits within the design assumption envelop. All references should be to U.S. codes and standards, HRC regulations, HRC regulatory guides and the HRC Standard Review Plan. The SAR should clearly discuss the acceptability of the design In those arta~ where the design deviates frOll the criteria In the ASHE Code or U.S. Industry standards. In particular, the SAR should discuss how an equivalent design .. rgln Is *alntalned.

230.2 The SAR should provide one s111111ary table that lists all the as~llllll!d site paruieters for which the CAHDU 3U plant Is doslgned rather than having it spread over various parts of the SAR.

230.3 The SAR should provide Instructions for the COL applicant (or early site permit) on the type of lnfor111tlon needed and types of studies to be perfor11ed to show that a site meets the geological, seismological and geotechnlcal engineering requlrlllf!nts.

230.4 The staff view Is that sites with the potential for tectonic faulting at or near the ground surface are not acceptable for nuclear power plants.

The SAR ~ho~ld state this.

230.5 Provide a COlljllete description of the seismic Instrumentation characteristics (I.e., solid state components, digital recording, bandwidth, dyn1111lc range, ability to promptly determine responses spectra and cumulative absolute velocity, etc.)

230.6 Provide a dlscusslo~ of the need for the COL applicant to have a program plan to perform pre-earthquake planning and post earthquake actions and an outline of such a program.

240.l State the ground water level and the external flood level for which tne plant Is designed.

252.1 Turbine Missiles This section of the SAR should Include a figure showing the +25i degree low-trajectory turbine missile ejection zone. Further, the SAR should co11111lt to meeting RG 1.115, "Protecl1on Against Low-Trajectory Turbine Missile,* Revision I which specifies that the prob;blllty of unacceptable damage from turbine missiles be less than 10* per reactor-year. Consistent with the staff's position taken for recently licensed plants, the probability of turbine missile generation should be no 5

greater than 10* per reactor-year for unfavorably oriented turbines ind 10* 4 for favorably oriented turbines.

a. Paragraph 4.5.2.3 of the SAR indicates that Zirconium-Niobium pressure tubes meeting the requlre111ents of Canadian Standard CAH/CSA-N285.6.1 will be used In the CA.~DU 3U reactor. The SAR should Identify all criteria In these standards and guides that are different or deviate from those In the ASHE Code or Standards endorsed by NRC. The SAR should clearly discuss the acceptability of the design In those areas where the design deviates from the criteria In the ASHE Code or U.S. Industry ~tandards.

b. Paragraph 4.S.2.3 of the SAR also stat~s that Canadian Standards CAN/CSA-N28S.6.8 and CAN/CSA-N2BS 2 wi . be used for the selection of llOdlfled Typg 403 stainless steel aaterial and the use of a rolled Joint design between t~e pressure tubes ~nd the Type 403 aaterial. AECLT 111.1st id~ntlfy all criteria In these s**ndards and guides that are different or deviate fr1>11 those In the ASHE Code or Standards endor~ed b!' f.ilC. T:1e SAR should clearly discuss the acceptability of the dtislg:. In those areas where the design deviates fr1111 the criteria in the ASHE Code or U.S. industry standards.

c. Paragraph 4.5.4.1 of the SAR state~ th~* lattice tubes and castings shall be fully radiographed over the aaxillUll feasible vol1111e.

Section III of the ASHE Boller and Pressure Vessel Code (ASMt Code) requires that nucle*r C011POnents requiring radiography 1111st be ex .. lned over the entire vol1111e. The SAR should explain why CAHOU 3U c011ponents cannot be examined over the entire vol1111e as required by the ASHE Code and discuss the acceptability of such an approach.

252.2 Heat Transport Syste11 and Connected Systems

a. Figure 5.2-1 of the SAR shows the heat transport system pressure boundary flow diagram. Using this dlagr111, the SAR should Identify the materials of construction for the major pressure retaining comp1nents, e.g., DzO storage tank - Aurtenltlc Stainless steel Type 304L; 020 feed pllllps - Austenltic Stainless type 304L casing, 304L Impeller, type 410 shaft; shutdown/bleed cooler - shell carbon steel, channel carbon steel, tubes carbon steel; bleed valves - body carbon steel, disk carbon steel, stem type 410 stainless steel; piplrg carbon steel.

b. Figure 5.6-1 of the SAR shows the 110derator syst~ flow diagram.

Using this dlagr111, the SAR should Identify the aiaterlals of construction for the major pressure retaining cD11ponents of the moderator system.

c. The SAR should Identify the Impeller material for the heat transport and shutdown cooling pumps.

d. The SAR should Identify all structural materials that will be exposed to high neutron fluence.

252 3 Engineered Safety Feature!

This section must Include 'subsection o.2.7, Fracture Prevention of Containment Pressure Boundary and coawlt that the containment liner, containment penetrations, equipment and personnel hatches will meet the ASHE Code,Section III fracture toughness requlrer.ients.

252.4 Steam and Power Conversion Syste~

Subsection 10.3.6 should be expanded to discuss those measures that have been taken In the CANDU 3U design to address the concerns of

• roslon/corroslon caused by single-phase or t110-phase erosion/corrosion ph.::~o.. non as doc1111ented In Generic Letter 89-08, "Eroslon/Corroslon-Indu*. "Cl Pipe Wall Thinning.*

262.l The deta'led test progr111 Individual Test Descriptions were not Included In the CAi'OU 3U SAR; however, the applicant acknowledged the requirement to provide that fnform1tlon to the staff, 263.1 Th11 AECLT rel hbll I ty usunnce progru (RAP) wu developed using the staff's fnterl* position for a ~P. as stated In SECY-93-087. AECLT w~ll need to revise the D-RAP to reflect the NRC's final position on RAP to Include a descrlp*tton of the essenthl ele11ents of D-RAP, dehlls on how the applicable regulation for 0-RAP fs satisfied, and the ITAAC to verify lmpleeentatlon of D-Rl.P prior to fuel load.

290 .1 The *a In contro1 area (MCA) HVAC syste11 ts a non-safety syste11 vh1 ch does not confol'll to the single failure criterion for providing safety-related filtration and cooling functions. Therefore, the HCA does not meet the requtre*ents of GOC lg of 10 CFR Part 50, Appendix A.

420.l Section 1.8 CONFORMANCE TO NRC REGULATORY GUIDES The recoar.iendatlons of Regulatory Gulde 1.75 with regard to separation and Independence of electric circuits ar~ discussed only In SAR Section 8. AECLT should also discuss how the design meets the Intent of Regulatory Gulde 1.75 for the l&C systems discussed In SAR Section 7.

The design should show that low-energy signal cables are routed separall!ly from power cables, and t!lat safety-related redundant l&C circuits and coraponents are separated and Isolated from non-safety-related circuits.

In AECLT's ldenttflcat*on of the lnstr1111entatton and control systems Important to safety and the acceptance criteria for these systems, many Canadian standards art refere~ced. However, there Is no comparison provided between the criteria In the Canadian standards to that In comparable U.S. standards.

420.3 AECLT stated In Section 7.1.2.5.2 that a Failure Hode and Effects Analysis (FHEA) ts not provided at this time since the detailed schematic and loop diagrams are not available. RG 1.70 calls for an FHEA for protection systems and components. The staff considers an FHEA for the CANDU 3~ Special Safety Systems to be essentl* for Its review.

420.4 Section 7.2 SPECIAL SAFETY SYSTEHS AECLT stated In Section 7.2.1.2.8 that the technology proposed for the trip computers has been used In the Darlington Nuclear Generating Station. However, there ts no documentation describing the hardware and software design, the verification and validation processes, con-figuration management and other aspects of the digital system design.

The Inspections, tests, analyses, and acceptance criteria (ITAAC) for these systems have not been tncludP.d. The staff Intends to address th1s lack of design detail tn t *anner similar to that for the GE ABWR and ABB-CE S7ste11 80+ designs by certtflcatlon of a design develop11ent process as described In SECY-92-053.

420.5 In Section 7.7 (Control Systeas not Required for Safety), there ls no discussion of the lnstrU11entation and controls for the on-power fueling machine and Its Interface with safety-related syste*s. This ls a unique feature In the CAHDU plant design which 111y pose potential system Interaction concerns that should be addressed by AECLT.

435.1 Section 8.1.4.3.4.1 of the SAR only states that the design of the CAHDU 3 ts In COllPllance with the requlr11111nts of BTP-ICSB 4 (Rev. 2, 1981 July). The staff requires details of how the CAHDU 3 design complies with thts BTP In order to conduct Its review.

435.2 Section 8.1.4.3.4.4 of the SAR only states that the design of the CANDU 3 ls In COllPllance with the requirements of BTP-ICSB 18 (Rev. 2.

1981 July). The staff requires details of how the CAHDU 3 design c011plles with this 8TP In order to conduct Its review.

435.3 Section 8.1.4.3.4.5 of the SAR only states that the design of the CAHDU 3 ls In compliance with the requirements of BTP-ICSB (Rev. 2, 1981 July). The staff requires details of how the CAHDU 3 design complies with this BTP In order to conduct Its review.

435.4 Section 8.1.4.3.4.6 of the SAR states that the design of the CAHDU 3 electric protection system complies with the requirements of DTP-PS~ 1, and the details of voltage s2tpolnt and time delay of the tripping shall be analyzed In the detailed design stage. The staff requires additional lnfonnatlon on how the CAHDU 3 design ccmplles with this DTP. AECLT should provide details of the design and how It complies with each position of the BTP.

435.5 With regard to THI Item 11.E.3.l, Emergency Power Supply f~r Pressurlz~r He1ters, Section 8.1.4.3.5.1 states that there ls alternate cocnpllance for this Issue as the necessity for pressurl,er he1ters for continuing natural circulation can be challenged and dismissed by ~pproprlate an*lysls. The staff requires the details of this analysis In order to properly perform Its review of this Issue.

435.6 There Is no detailed discussion of how operational ~xperlence was incorporated Into the CAHDU 3 design. Specifically, In the electrical power area, a discussion of how the design Incorporated the experience Identified In Generic Letter 88-15, Electrical rower Systems - Inade-quate Control Over Design Process, Is required.

435.7 The staff requires additional lnfonratlon on the CAHOU 3 lighting design described In SAR Section 9.5.3 In order to conduct Its review. The following areas of Information sho~ld be addressed:

a) Illumination ranges for the normal, standby, and essential/emergency lighting ;ystems are required.

b) Additional design details are required on the different lighting systems.

440.l C1111puter Codes used for Transient and Accident Analysts:

AECLT needs to provide detailed descrtptir.ns of all the COllPUter codes used tn the core design discussed In SAR Chapter 4 and the transient and accident analyses docU11111nted tn Chapter 15 of the SAR. The doc111111ntatton of each c011puter code should Include a discussion of the purpose of the code, description of the calculattonal models, and vertftcatton of the c011puter code against test data or applicable operating data. The code doc1111entatton should dl!llOnstrate Its acceptance to the staff by showing tts validity of the governing mass, energy, and 11011C1nt1111 equations. AECLT ts also be requested to show the correct use of l!llPlrtcal correlations and ste11t-heavy water lnterfaclal relationships, accuracy of the n1111ertcal solution scheme Including llOdel*ng techniques, and adequacy of benchmark comparisons with existing data.

For the computer models used In the loss-of-coolant-accident (LOCA) analysis, the applicant 11111st use, (as required for LWRs In Item (a)(l)(I) of 10 CFR 50.*6), an evaluation model that Includes sufficient supporting Justification tft show that the analytical techniques realistically describe the behavior of the reactor syste~

during a LOCA. C011Parlsons to appllc2ble experimental data must be made and uncertainties In the analysis 11111thod and Input 111Ust be Identified and a~sessed so that uncertainty In the calculated results can be estimated. Alternatively, an evaluation model used for the LOCA analysis may be developed using an ~~~~each consistent with the requlreinents for LWRs described In Appendix K to 10 CFR 50. The discussion should Include the references of test reports used to validate the models simulating key phen0111na that could occur d1,rlng transients and accidents.

471.1 Operational considerations - Section 12.l should Include a section on how operating experience from current generation plants has been Incorporated Into the CAHDU JU design. This section should provide ex1.111ples of design changes and operational Improvements oased on lessons learned from past plant designs.

471.2 Contained radiation sources - Section 12.2 should Include a description of contained radioactive sources (such as waste tanks. heat exchangers, filters, holdup tanks, resin tanks, etc.) In the plant. Information provided for each source should Include source location within the plant, component material and gec.metry, and radionuclide contents with associated source strengths.

471.3 Plant layout drawings - Section 12.3 should contain detailed plant layout drawings which show: plant radiation zones, shielding w~ll thicknesses personnel and equipment decontamination areas, health physics facilities, controlled access areas, contamination control areas, chemical and analysis labs, post-accident sa~pling stations, and tht counting rooa. These plant layout drawings should be provided for each plant elevation and should also Include the location of plant equl11111nt and coaponents as well as no~l and post-accident personnel traffic patterns.

471.~ Radiation protection design features - In accordance with Chapter 12 of the SRP, Section 12.3 of the CAHDU 3U SAR contains

• good description of facility desl~n features Incorporated to ensure that personnel exposures are 111lnt*lned ALARA. However, the SRP also states that Section 12.3 of the SAR shoulrl describe hOll the plant design considers such *aJor exposure accU11Ulatlng functions as ~lntenance, refueling, In-service Inspections, dec11111lsslontng, etc. Section 12.3 of the SAR should include this lnfo1111tlon, as well as 1) a discussion of source ter11 control, 2) A description of how robotics have been Incorporated Into the plant design to *lnl*ize personnel doses, and 3) a description of any accessible plant areas having a potential for dose rates greater than one Sv/hr.

471.5 Dose 1ssess111nt - Section 12.4 should contain a dose assessment performed In ac~ordance with the guidelines of Regulatory Guide 8.19,

• occupation Radiation Dose Assessment In Light-Water Reactor Power Plants Design Stage Han-Rem Estimates*.

471.6 OBA dose consequence COlllJ>Uter codes - In Sections 15.l.3, et at., codes used by AECLT In the CAHDU consequence analysts are referenced. These codes need to be described, and the basts for their validity needs to be presented.

471./ Calculatlonal 111ethodologles and asswnpttons - The dose calculation methodologies and input assumptions used to analyze the radloloytcat consequences of postulated accidents need .o be provided. The information provided in this area needs to be adequate to allow NRC to perform Independent calculations of the dose consequences of DBAs.

471.8 Accident source term - The source term used In the radiological consequence analyses should be specified. NRC has done extensive work on the accident source tel"lll for light water reactors (LWRs); HRC has little experience with heavy water reactors. Therefore, the basts for the CANOU accident source tenn needs to be described.

620.l The staff's revlev of HW111n Factors Engineering (HFE) topics can be performed In three levels: prograJ1111attc, Implementation plan, and completed (see attached infonnatlon). AECLT should Identify the level for each topic area. At present It appears that the HFE program plan and the Operating Experience Review may be COlllJ>leted. Host of the other topics can be treated at an implementation plan level of detail. V&V appears to be at a prO'jrannattc level of detail. Please discuss.

620.2 While no specific doclJlllent reference ts provided, the SAR discussion seems to Indicate thit a Human Factors Engineering Program Plan (HFEPP) exists which may pro1lde some of the additional detail required for the review. AECLT should provide the HFEPP.

620.3 Great emphasis ts placed on the use of the predecessor plant's analyses and operating experience as a blsis for the CANDU 3 design. AECLT should provide a discussion of the relevant HFE analyses perfol'lll!d for that design which provide starting points for the new design 620.4 The spectftc doc11111ntatton ""11th ts already available for review to support the HFE discussions and conceptual design in the SAR and that will be developed as part of the detailed design should be clearly identified.

620.5 The SAR contains atsstng figures, as well as cross-references, vhtch when checked, provide little additional 1nforaat1on (e.g.,

Ftgure 18.3.4-3 atsstng; Section 18.1.6.1.2, *c1111plete ltst of docuaents,* ts *isstng: and there are numerous references to Section 18.1 ""11th do not really expand upon the toptc under dtscusston). These aspects of the SAR should be llOdtfted.

620.6 Several review topics were either atniaally addressed or not addressed at all. AECLT should provide detailed tnfor111tton on these topics.

Most notably, these include:

Task analysts (*tntmally addressed)

HRA-HFE tntegratton (appears to be omitted)

H1n1111U111 Inventory (appears to be omitted)

COO/ITAAC/DAC (appears to be omitted)

GENERAL GUIDANCE HEE Review leyels; The HFE PRH (NUREG-0711) cin be used to revlev ippllcant submlttals at three levels: Progr...atlc Revleti, l*pleaent*tfon Plan Revff!ttl, and Co*plete £1e*ent Revleti. At a Progr1,.1tlc Review level, the SAR does no~ Include detailed methodology and, therefore, detailed evaluations using the HFE PRH acceptance criteria are beyond the scope of the staff review for design certification.

At a progr111111atlc level review, the HFE PRH criteria are used to determine whether the progriJll provides a top-level Identification of the substance of each criterion which, after design certification, will be developed by the applicant Into a detailed Implementation plan. The value of the prograanatlc level review Is that It provides assurance that the hapleaientatlon plan will address all HFE PRM criteria. Applicant c01111lt111ent to the development of such a detalltd l1111le111entatlon plan should be described In ITAAC/DAC. lhe staff will review this plan during post certification revlev activities. ITAAC/DAC are also needed for completing the Implementation plan and providing the results to the ~taff for review.

To perform an lapleaentation Plan Review, the applicant's submlttals should describe the applicant's proposed metho~ology In sufficient detail for the staff to determine whether the methodology will le*d to products that meet the Hf[ PRH acceptance criteria for the clement. An Implementation plan review provides the applicant the opportunity to obtain staff review and concurrence on the applicant's full approach prior to design certification. The actual completion of the plan will then likely take place after design certification.

Such a review Is desirable from the staff's perspective since It provld the opportunity to resolve methodological Issues and provide input early In the analysis or design process when staff concerns ca~ more L slly be addressed than when the effort Is completed. While some Implementation plans can be reviewed on their own merits, the staff may request a sample analysis which demonstrates the application of the i;iethodology and Its results. ITAAC/DAC are needed for completing the Implementation plan and providing the results to the staff for review.

A Coaplete Eleaent Review can only be performed when the finished products (e.g., main control room (HCR) design) are available for the staff to evaluate. This means that the applicant h*s submitted an analysis results report(s) and design team review report(s). An analysis resul*s report provides thr results of the applicant's efforts on an HfE PRH element with respect to the review criteria. A reviewer will utilize the report as the main source of Information for assessing compliance with the review criteria.

An applicant's design team review report provides the Independent evaluation of the activities addressed for the element by the design team. On resolution of staff conc1rns regarding the analysis or Its results, the review topic can then be closed prior to design certification.

Piping [)eslgn; In accordance with 10 CFR SZ.47(a)(Z), an application for a standard design should contain sufficient design lnfon11tlon to en3~1e the HRC to make final safety determinations. For piping design, acceptable approaches consist of either (1) having docllJll!nted and available for audit a complete design of all safety-related piping systees or (Z) providing C0111Prehenslve descriptions of design acceptanre criteria (DAC) for piping In the SAR and sa~le representative piping analyses. The piping DAC should contain Information In the foll~wlng areas:

• applicable codes and standards

• llll!thods to be used for crapletlng the piping design analysis

• modeling techniques

• pipe stress analyses criteria

• pipe support design criteria

• criteria for postulating high-energy line breaks

• Leak-before-break (LBB) approach applicable to CAHDU-3 (analyses are required for all candidate LBB piping)

The SAR should clarify the approach and upgrade Secttuns 3.7.2, 3.9.2 and 3.9 3 of the draft SAR as necessary to address the above areas of concern.

ENCLOSURE 2 Unites! States Nyclear gp.;~j,t;;v '21J1tsston Atotlc Energy of Canada. Ll1ltesl Technologies Recelot of Apollcatlon for Qeslgn Certlflc*tion Notice Is hereby given that the Nuclear Re9ulatory COlllllsslon (the Ca..lsslon) h~s received an application frOll At01lc Energy of Canada, Ll1lted Technologies dated Septetlber JO, 1994, filed pursuant to Section 103 of the Atocalc Energy Act and 10 CFR Part 52, for the standard design certification of the CANDU JU Pressurized Heavy Water Reactor Plant.

A notice relating to the rulemaklng pursuant to 10 CFR 52.51 for design certification, Including provisions for participation of the public and other parties, will be published In the future.

The CAHOU 3U Is a 450 Hiie pressurized heavy water reactor design. A unique feature that distinguishes the CANDU JU from the current generation of ll9ht water reactors designed In the United States Is the use of natural uranlwn fuel contained In a pressurized heavy water coolant system and a separate heavy water moderator system. The CANDU 3U application Includes the entire power generation comp;~~. except those elements and features considered site specific.

The staff has determined that the application does not contain all Information required by 10 CFR 52.47. A docket number Is being assigned to the application to facilitate public access to correspondence and review lnfor1atlon. Although no formal review S* "~'ule will be established until an updated Safety Analysis Report and *.* 1edules for all Information required has been received, the NRC staff will continue limited review of

• 2*

the applicltion. This is consistent with the letter of Septellber 30, 1994,

• *** th*t no 111Jor activity will be initiated by the NRC beyond the acceptance review ****

• A copy of the application ts available for public inspection at the C011111tsslon's Public OocUlll!nt RoDll, the Gel114n Building, 2120 L Street, N.V., Vashtngton, o.c. Previous corresp~ndence on thts application ts ftled under Project nUllber 679. The new docket established for thts appltcatton ts STN*52*005.

Dated at Rockville, Haryland thts 15th day of December1994.

~~ told, s and for Advanced Reacto ''' "''

License Office of Nuclear Reactor Regulation Renewal

CAHOU ProJe!:t Ho. 679 Docket Ho.52-005 cc: louts H. Rib, Licensing Consultant AECL Technologies 9210 Corporate Boulevard, Suite 410 Rockville, Kiryland 20850 A. H. Hortada Aly, Senior Project Officer Advanced Projects licensing Group St~dles and Hodtflcatton Division AtDlllC Energy Control Board P.O. Box 1046, Statton B 270 Albert Street Ottava, Ontnrto, *anada KIP 5S9 Manager, Safety l Licensing - CAHDIJ-3 AECL CAHDU, Western Region 441A-2nd Avenue Hort~

Saskatoon, Saskatchevan Canada S7K 2Cl L. Hanning Muntzlng Newman l Holtzlnger, P.C.

1615 l Street, N.W., Suite 1000 Washington, D.C. 20036 Steve Goldberg, Budget Examiner Of ftce of Hantgement and Budget 725 17th Street, NW.

Washington, D.C. 20503 Director, C.\HDU JU Safety l licensing AECL Technologies, Inc.

9210 Corporate Boulevard, Suite 410 Rockville, Maryland 20850 A. D. Hink, President AECL Technologies 9210 Corporate Boulevard, Suite 4\0 Rockville, Maryland 20850

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NRC/AECLT Meeting J&nusry 30, 1995 AGENDA:

1. Introduction

- AECLT: Purposu of Meeting/Background

- NRC: Current Status of NRC Cost/Schedule Review

2. AECLT: Compa**ison of FDA Review Costs

3. AECLT: Evaluation of fDA Review Costs

4. AECLT: Proposed FDA Review Resources and Schedule

5. AECLT: The Next Step

'. ~

~ AECL Technclogles Inc.

I 'I 49M COST COMPARISON EVOLUTIONARY DESIGN - FDA 13M I I 11 M l

ABWR SYSTEM SO+ CANOIJ NRC/AECLT llEETINO January 3D, 1HS

NRCfAECLT Meeting January 311, 1995 AECLT POSITION ON RESEARCH:

• Standard Review Plan Review of Safety Margin and R&D Support of CANDU Design is Adequate and Sufficient

• No Additional Research is Required, CANDU Technology is Mature and Based on Operating Experience and AECL Research programs

.... ~

k_ AECL Technologies Inc.

COSTS WITHOUT RESEARCH EVOLUTIONARY DESIGN - FDA 25M 13M 11 M I -

ABWR SYSTEM8o+ CANDU NRC/AECLT MEETtNO J1111111ry :so, 1191

~ AECL Techr:t>logles Inc.

NRC INITIAL ESTIMATE SECY-94-079 {NRR ONLY)

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FFORT 20 20 20 TOTALS:

( TE) 15 - FlES: 105 PROGRAM 15 15 SUPPORT: S2.2M 10 10 TOTAL:S25M 5-I I I ..I ...I .._

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NRCIAECLT MEETING January 30, 1995

NRC/AECLT Meeting January 30, 1995 COST ESTIMATE EVALUATION - PROPOSED EFFICIENCIES:

• Start with Program for Resolution of Generic/Policy Issues

• Conduct Technical Issues Seminars During Detailed Review (SNUPPS approach, tends to minimize number of RAls)

• Apply "Load Follow" versus "Base Loaded" Resources

• AECLT Improve SRP Comparisons in SAR

• Reach Agreement to Pursue Regulatory Efficiency Generally and to Control Budget and Schedule S~cifically to Achieve This Objective

k AECL Technologies Inc.

AECLT PROPOSED PROGRAl't 11111 111& 1117 1/111 1111 1:00 1l'01

---~__._.......__.__L l Ll _ _L_f _L l_J I I I I I_ I _I Ll EVENTS BAS'°D ONNRC I I A>L.

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-- ......I AECLT I cu ACTIONS MCft ...... FINAL oco E Ch.11 ll'3UltC*

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MEl!TINOS I 11&11 ITAACIDAC TilCH IJHCS SDmNARSI TI!ST A&SmA~CH SHC tlAMDA ITAAC MV.

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- 23

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~ I NRCIAECLT MEETING Januaiy 30, 1995

• -\ !'."

NRC/AECLT Meeting January 30, 1995 THE NEXT STEP:

• Develop Agreement/MOU Regarding Control of Cost and Schedule 11 Conduct Periodic (Weekly/Biweekly/Monthly)

Project Status Meetings

- Monitor Status

- Antic~pate Problems

- Facilitate Responses to Questions

NRC/AECLT Meeting January 30, 1995 CANDU GENERIC/POLICY ISSUES:

CODES AND STANDARDS ACCIDENT ANALYSIS SOURCE TERM CONTAINMENT PERFORMANCE POSITrJE VOID REACTIVITY coN-rROL ROOM - SECONDARY CONTROL AREA SEISMIC DESIGN QUALITY GROUP BOUNDARIES FIRE PROTECTION ON.POWER FUELLING HUMAN FACTORS ENGINEERING

-f~A BACKUP SLIDE

~* ~~~~~~~~~~~~~~~~~~~--.-~~~~~--="'-~~-

AECL Technologies Inc.

~ 210 Corpor111 Boulev1rd Suit* '10 RoctviUe, M1t'j\and 20850 USA 1-llOO*USA-AECl 13011417~7 FIX 1301) '17-0748 Telex .OJ.<<2 February 2, 1995 Docket No. STN-52-005 File No. 09000401 Control No. 950202001 Document Control Desk U.S. Nuclear Regulatory Commission Mail Station P1 -137 Washington, DC 20555

Subject:

RnponH to NRC Requnt for lnfonnatlon Revardlng CANDU 3U Application for Final Dealgn Approval and Design Certification Refs: 1. NRC letter to AECl.T (0. M. Crutchfield to A. D. Hink) dated December 15, 1994: Results of the Acceptance Review for AECL Technologies' Appilcatlon for Final Design Approval and Standard Design Certification for the CANDU 3U Design

2. NRC Meeting Summary, dated January 12, 1995: Summary of Meeting Held with AECL and AECLT in Ottawa, Ontario Gentlemen:

This letter responds to the results of the ~u*c staff accepamce review of the subject application as documented In Reference 1.

AECL Technologies Inc. (AECLn has developed a schedule for submittal of additional informailon as requested in Reference 1. The schedule for updating the Safety Analysis Report (SAR) and submitting other information (ITMC, Technical Specifications, etc.) is provided in Enclosure 1. The schedule set forth in the enclosure :s based on the review milestones developed by the NRC staff in SECY-93-097; howevc.r, the milestones have been relocated to coincide with an AECLT*

proposed s.:hedule for the review. This schedule was presented to members of the NRC staff at a January 3Ci, 1995 meeting to discuss cost and schedule for the CANDU review. The schedule proposes a period of generic licensing issue review followed by the specific plant design review. This timing is based on the current AECL CANDU plans and schedules for developing design Information for the full CANDU plant product line. AECLT is planning to meet with the NRC staff to identify 9502080187 950202 Attachient CANDU m the USA JJ)CO1 PDR ADOCK 0~20000~

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Document Control Desk Control No. 950202001 Page 2 and address several technical and safety issues that are generic to the review of CANDU vlants using NRC regulatory requirements and guidance. Therefore, the period prior to initiating the specific design review would be used productively to resolve goneric issues such as these identified in SECY-93-092 and In Referenco 2.

These Issues involve all major aspects of CANDU technology; therefore, their evaluation and resolution would result in a more efficient review of the application.

Reference 1 contained specific questions in the lorm of a request for additional information. Many of these questions will be answered, and result In SAR updates, during the planned generic issue review effort. AECLT Intends to provide responses to the remaining questions In an SAR update submitted no later than Jsnuary 1997.

AECLT proposes to initiate discussion:i with the NRC staff In the near future to identify and schedule reviews or generic CANDU Issues. A list of candidate issues 1s provided In Enclosure 2.

At the January 30 meeting, AECLT again stated that (1) the cost for an FDA review should be more in line with that of other avolutionary plants (with due consideration of unique dLsign differences) and (2) that AECLT could not proceed under the previous estimate of approximately $50 million if that amount were proposed to AECLT for NRC cost recovery. AECLT is looking forward to further discussions after the decislori of the NRC Commil's1on regarding confirmatory research costs and NRR's reasse&sment of resources for the CANOU review are available.

In the final paragraph of Reference 1, ongoing program support tasks are mentioned. It is requested that NRC staff provide AECLT with task descriptlons and an estlmate of the anticipated fees that will be associated with these tasks over the next year

• to March 31, 1996.

Should you have any '1Jestlons regarding this letter, please contact the undersigned.

Very truly yours, ,"'

!ii II f~

M H. Fletcher Director, Safety and Licensing Enclosures 1 Proposed Program Milestones

2. CANDU Generic/Policy Issues cc: D Scaletti NRR

Document Control Desk Enclosure 1 Control Nil. 950202001

" Proposed Program Mlloatcmea The following milestones are based on the program for CANDU re\ * '" discussed with the NRC staff at a meeting on January 30 1995. The proposea program involves an initial tv10-year review of generic/policy issues follo.ved by a three*year detailed review of thn CANDU dr:sign. The sarient points affecting U1e program schodule are* (1) recognition of CANDU as a well-developed reavtor plant design; (2) the necessity of pursuing innovative and efficient review methods to limit costs; (3) the need for AECLT to provide timely and complete infonnation to the NRC, (4) maintaining a level of effort within AECLT budget during the generic review phase, i.e., $1M (US) the first year building up to $2M in the sect>nd year; and (5) elimination of costs for research that duplicates existing CANDU research. The milestones are also shown graphically on Figure 1.

• Present

• 4197 Review of CANDU generic licensing issues

• Prestmt

• 1/97 AECLT updates to SAR

• 4/97 NRC staff initiate specific design review

• 5197 (approx.) AECLT initiate review seminars in selected design topics for NRC staff ll 'mica! reviewers

• 1/98 AECLT submit r ;ponse to NRC RAls

• 1198 AECLT submit Level I PSA

• 11198 NRC issues Draft Safety EvalL.ution Report

• 7/98 AECLT submit Level II & Ill PSA, ITAAC & DAC, Technical Specifications, Initial Te::.t Program Test Abstracts, and Severe A:cident Mitigation De3ign Alternatives

• AECLT respond to Draft Safety Evaluation Report and submit Final SAR, Final Technlcal Specifications, and Final ITAAC & DAC

• 10/99 NRC issues Final Safety Evaluation Report for Review

• 4/00 AECLT submit Design Conti ol DocuMent

\

• 9100 FDA issued

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~ AECL Technologies Inc.

AECLT PROPOSED PROGRAM 1116 1111 1117 1118 1111 1/IHI 1/01 I I I I I I I I I I I I I I I I I I I I I I I I '

I EVENTS

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S.r.R SAR I llUI ITAl'C suau Ch.11 UPDATI! I ITAACIDAC C*E".

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SOURCE TaRM SUllNARSI SAMOA TI!CH. REVtEW J---OTHVI U.llTIHOll OENl!RIC ISSUES REVlllWl!IAR UPDATl!S FIGURE 1

Document Control Desk Enclosure 2 Control No. 950202001 CANOU GENERIC/POLICY ISSUES 1 Codes a11d Standards (st-uctures, pressure-retaining systams, testing/inspeclJon, electrical)

2. Accldent Analysis* (even: selection and evaluation, categorization of events, acceptance criteria, assumptions, analytical tools, severe accidents, external events)

' 3. So1.orce Term* (mechanistic evaluation: fuel performance, transport, event-specific source terms)

4. Containment Performance* (offsite dose limits, ASME Level C lfm:>: or equivalent)

5. Positive Vold Rea~ (probability, consequences)

6. Control Room/Secondary Control Area* (CANDU philosophy, NRC GDC-19) 7 Seismic Design (assumptions for SOE, seismic margins)

8. Quality Group Boundaries (rc;:::tor COl>lant pressure boundary, coriminment penetrations, Class 6 systeir:s, differences between U.S. and CANDU definitions)

9. Fire Protection (CANDlJ methodology and NRC Standard Review Plan)

10. On-Power Fuelling (safety Issues and operating experience)

11. Human Factors Engineering (CANDU human factors e11gineering plan and

" criteria)

• Issue identified in SECY-93-092

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UP41TED STATES NUCLEAR REGULATORY COMMISSION

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"" WASHINGTON, O. C. 20!565 Hr. H. D. Hink, President AECL Technologies, Inc.

9210 Corporate Boulevard Suite 410 Rockville, Maryland 20850

SUBJECT:

RESPONSE TO YOUR LETTER OF FEBRUARY 2, 1995

Dear Hr. Hink:

Enclosed Is a copy of SECY-95-XXX, "Change In Plans for Design Certification Review of the CANl.:J 3U Reactor.* In t11e Corrmisslon paper. the staff details plans for the CANDU 3U review in response to the February 2, 1995, letter from Hr. Fletcher of your staff.

In our letter of December 15, 1994 we responded to your request for limited continuing program support. The staff plans to continue a low level of effort In accordance with your needs on key issues such as void reactivity and shutdown system reliability. We will track our resource expenditures closely, inform you when we are close to your budget limit, and stop work when the limit Is reached, unless you Indicate that further support Is available. We request that you give us your budget limits for our fiscal year (October l through September 30) to &ssist us In our planning. We have also advised the Office of Nuclear Regulatory Research of your letter and stated that they should cease all design-specific research in support of the CANDU application unt 11 further notl ce.

The staff has three contracts nearing completion; these contracts address fuels, core physics, and reactivity analysis. The contractors are preparing the draft reports for these projects, which should ue complete within the next few weeks. The Frbruary and March costs to finish the three draft reports should be less than $20,000. The staff Issued one contract for FY95 for

$128,562 to assess the CANDU ~hut down system reliability.

Certain areas of your proposed review schedule will be difficult for us to accorrmodate as discussed In the Co1M11sslon paper The AECLT schedule calls for the staff to Issue the final CANOU safety evdluatlon report (SER) In October 1999, allowing only 30 months for t~e review. Past staff experience with evolutionary Jnd passive design review schedules, and the limited resources available will not support your request for a 30-month schedule.

The staff cannot prepare a detailed review schedule until July 1998, when it has received all of the required Information such as the completed probabilistic safety assessment, Inspection, test, analysts, and acceptance criteria; technical specifications; tests programs; and severe accident mitigation design alternative.

Attachment 4

Mr. A. D. Hink If you have any question regarding this letter please contact the NRC project manager, Dino C. Scalettl at 301-415-1104.

Sincerely, Dennis M. Crutchfield, Assotlate Director for Advanced Reactors and license Renewal Office of Nuclear Reactor Regulation Dock~t No.52-005 cc w/enclosure:

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