Text

Summary of ACRS 354th Meeting on 891005-07 in Bethesda,Md Re Definition of Adequate Protection & Generic Issue 135
	ML20059N649
Person / Time
Issue date:	10/11/1990
From:	; Advisory Committee on Reactor Safeguards
To:	; Advisory Committee on Reactor Safeguards
References
	REF-GTECI-135, REF-GTECI-NI, TASK-135, TASK-OR ACRS-2672, NUDOCS 9010170079
	Download: ML20059N649 (44)
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TABLE OF CONTENTS M

d MINUTE $ OF THE 354TH ACRS MEETING OCTOBER 5-6, 1989 Chairman'sReport(0 pen)...........................................

1 II.

DefinitionofAdequateProtection(0 pen)...........................

1

!!I. Generic Issue 135, " Steam Generator and Steam Line Overfill Issues" (0pen).............................................................

3 IV.

Meeting with Director, NRC Office of Nuclear Regulatory Research (0 pen).............................................................

4 V.

Ma intena nce of Nucl ea r Power Plants (0 pen).........................

10 VI.

Generic issue B-66, " Diesel Generator Reliability," and Associated Regul a tory Gu ide 1. 9, Rev. 3 (0 pen )............ '....................

12 VII. Generic Issue 87, " Failure of HPCI Steam Line Without Isolation" (0pen).............................................................

16 Vill. Standardized Plant Design (C ANDU-3) (0 pen).........................

17 IX.

Executive Sessions (0 pen / Closed)...................................

19 A.

Subcommittee Report s (Closed)..................................

19 t

1.

Nominating Committee / Election of ACRS Officers (Closed)....

19 B.

Reports, Letters,andMemoranda(0 pen).........................

19

[

1.

Proposed Revised Policy Statement on the Maintenance of l

Nuclear Power Plants (Report to Chairman Carr dated i

L October 12,1989).............................,.............

19 o

2.

ACRS Comments on the Safety Goal Policy and Its Relation-ship to the Concept of Adequate Protection.................

20 3.

Proposed Resolution of Generic Issue 135, " Steam Generator and Steam Line Overfill Issues" (letter to James M. Taylor, Acting Executive Director for Operations, dated

?

October 11,1989).........................................

20 C.

Other Conclusions (0 pen)......................................

20 1.

Future Agenda items.......................................

20 2.

Proposed Ru le on Hot Particles............................

21 ll' 3.

Decision Not to' Review Proposed Restart of Calvert Cliffs P1 ant.....................................

21 4

A C R S By l a w s..............................................

21 5.

ACRS Decision to Review Yankee Rowe Aplication for a License [xtension a s a Separa te i tem......................

21 l, fg A- -'

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- APPENDICES 4

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lnconvenience.

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2:45 p.m-535 p.m.t Standard Plant Ihave determined in accordance with

$30 a.m.-8 d5 c.m.: Comments by Design CANDU-3 (Open)--Briefing and Subsection 10(d) public l aw 92-463 that ACRS Chairman (Open) -%e ACRS discussion regarding proposed design of it is necessary to close portions of this Chairman will report on items of current this standardized nuclear plant.

meeting e noted above to discuss interest.

3:15p.m 6.30p.m.tMaintenance of safeguards and security information at ads o.m.-20.45 a.m. Defin/ tion of Nuclear Power Plants (Open)-Continue nucleat plants (5 U.S.C. 552b(c)(3)) and Adequote Protection (Open)-%e discussion of proposed ACRS report information the release of which would Committee will discuss and comment on regardina this matter.

represent a clearly unwarranted a proposed pos! tion paper regarding the Saturday, October 1,1989.

invsslon of personal privacy (5 U.S.C.

8.30.o.m.-J240 Nohnt Pre A CRS Reports (Open)-%parction of..' 552b(c)(6)).

definition of "adequete protection?..

e Committee Further tulormation regarding topics JJ.M o.m-22:00 Noon:Standardised Pressurized Wofer Reoctors (Open)--

wiH discuss proposed ACRS reports to be discussed, whether the meeting

%e Committee will be briefed regarding regarding items considered during this has been canected or rescheduled, the the status of the review of the standard meeting. '

Chairmen's ruling on requests for the pWR designe proposed by the JSop.m4Wp.m.t Appointment of opportunity to present oral statements Westinghouse Corporation (RESAR Sp/

ACRS Afembers (Open/ Closed)-%e and the time allotted can be obtained by

90. W Ap-400) and the Combustion Committee will discuss qualifications of a prepaid telephone call to the ACRS Engineering Company (CESSAR. System candidates proposed for consideration becutive Director, Mr. Raymoud I.

80 + ).

as ACRS members.

Fraley (telephone 301/492-6049)'

2:00p.m-2:30p.m. Meeting with portions of this session will be clo' ed between 8.16 a.m. and 5.00 p.m.

s Director. NRC Office e/ Nuclear as necessary to discuss information the Regulatory Research (Open)--The..

release of which would represent a Deled: Beptember it, tone.

Committee will discuss items of mutual clearly unwarranted invasion of

. John C. Hoyle, Advisory Commlure Monocement Officer.

. intettst.

personal privacy.

2:30p.m-3:30p.m.: Generic issue D-2:00p.m.-2:30p.mr ACRS (TR Doc. as- :sso riled 9-t1-00: 848 eml SS DieselReliability(Open)--The Subcommittee Activity (Open)-%e euneoconsreesese Committee will review and report on the Committee will hear and discuss reports proposed NRC staff resolution of this of ACRS subcommittee activities, Jed5p.m-4 d$p.nr Acc/ dent, including development of containment Advlsory Committee on Reactor generic issue.

e design criteria for future reactors and Galeguards Joint Subcommittees on Afonqpement (Open)~%e Committee proposed changes in the ACRS Bylews.

Containment Systems and Structural 5

f will revlew and report on the proposed 2:30.m-3:00p.m.:(Open)-De Engineering;lAseting '.....,,r' NRC generic letter regard'ng Co ttee willcomplete discussion of 4

. IIe ACRS Subcomniltteeson I

considerationof accidentmanagement

Items considered during this meeting.

Containment Systems and Structural activitiesin theIndividualplant Procedures for the conduct of and haminations (IPEs).

participation in ACRS meetings were Engin'eering wit hold a joint meeting on October 17.1989. at the flyett Regency 845p.m 430p.m. Mo!ntenance of published in the Federal Register on at O'llere International Airport 9300 Nuclear Plants (Open}--%e Committee October g7.1966 (53 FR 43487). In West Bryn Mawr Avenue. Rosemont. !!

willdiscuss a proposed ACRS report accordance with these procedures, oral resarding this metter, or written statements may be presented

%e entire meeting willbe open to Friday. October A 1989.

by members of the public, recordings public attendance. '

8:30 a.m.-20:00 a.m.--Access willbe permitted only during those We agende for the sub}ect meeting Authortcotion of Nuclear Power Plants ' portions of the meeting when a shallbe as follows:

(Open/ Closed)-De Committee will transcript is beingt kept, and questions Ibesday, October Jr. 2989-430 a.m.

review and report on the proposed NRC may be asked only by members of the

untilthe conclusion of business.

tule regarding access authorization at Committee, its consultants, and Staff.

%e Subcommittees will continue to nuclear power plants.

Persons deelring to make oral discuss containment design criteria for Portions of this session will be closed statements should notify the ACRS future plants with invited speakers from as necessary to discuss safeguards and Executive Director as far in advance as industry and nationallaboratories.

lants.

practicable so that appropriate

. Oral etstements may be presented by security Information for nuclear p/ssuearrangements can be made to allow the members of the public with the JRJ3 a.m-JJ:35 a.m-Genet /c

74 allure ofRPC/ Steam Line necessary time during the meeting for ; concurrence of the Subcommittee -

Without teolation (Open}--De such statementacUse of still, motion Chairman: written statements will be Committee will review and discuss the picture and televlolon cameras during accepted and made available to the.

proposed NRC staff resolution of this this meeting may be limited to selected Committee. Recordings will be permitted generic matter and matters relating to portions of the meeting as determined only during those portions of the the performance of other types of valves by the Chairman. Information regarding meeting when a transcript is being kept, in nuclear power plants.

the time to be set aside for this purpose and questions may be asked only by 21:15 a.m-22:00 Noon--Future ACRS may be obtained by a prepaid telephone members of the Subcommittees, their Actir/ tics (Open}-ne Committee will call to the ACRS becutive Director, Mr.

consultants, and staff, persons desiring discuss anticipated ACRS subcommittee Reymond F. Proley, prior to the meeting, to make oral statements should notify activities and items proposed for in view of the possibility that the the ACRS staff member named belo'w as,

consideration by the full Committee..

schedule for ACRS meetings may be far in advance as la practicable so that

!$0p.m.-2:30p.m.t Generic issue 135, adjusted by the Chairman as necessary appropriate arrangements can be made.

Sicom Generefor ondSteam Line to facilitate the conduct of the meeting.

During the initial portion of the Overfillissues (Open)-%e Committee persons planning to attend should check rneeting, the Subcommittees, along with I

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[pont1 #g UNITED STATES 09 i

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NUCLEAR REGULATORY COMMtSSION c

i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WA9HINGTON, D. C. 20665 "o. e September 28, 1989 SCHEDULE AND OUTLINE FOR DISCUSSION 354TH ACRS MEETING OCTOBER 5-7, 1989 BETHESDA, MARYLAND Thursday October 5.-1989. Room P-110, 7920 Norfolk Avenue Bethesda, Md.

1) 8:30 8:45 A.M.

Chairman's Remarks (0 pen)

Opening remarks (FJR/GRQ) (FJR/RFF) 1.1) 1.2)

Items of current interest 8:45 10:45 A.M.

DefinitionofAdequateProtection(0 pen) 2.1) Report by ACR5 subcommittee chainnan TAB 2------

(DAW /MDH) 2.2) Meeting with NRC staff representatives 10:45 11:00 A.M.

BREAK

3) 11:00 12:30 P.N.

Generic Issue-135. Steam Generator and Steam Line Overf111 Issues (0 pen)

TAB 3------

3.1) Comments by ACRS subconnittee chainnan (PGS/EGI) 3.2) Meeting with representatives of the NRC staff and its contractor 12:30 1:30 P.M.

LUNCH

/) 1:30 4

3:00 P.M.

Meetinq with Director NRC Office of Nuclear -

Regula",ory Research (0 pen)

TAB 4------

4.1) Discuss items of mutual interest, including: (CPS /SD)

Impact of FY 1990 budget reductions on the NRC research program Nuclear Safety Research Review Committee's views on the research budget reductions Status of activities to obtain greaterdiversityofresearch providers

Status of activities' to bring.

outside expertise on board in RES Contributions to the NRC Safety Research Program from the imple-mentation of the recommendations of the National Research Council

~-

l 354th ACRS Meeting Agenda 12:15 1:15 P.M.

LUNCH r-11)'1:15 4:00 P.M.

Standardized Plant Design (CANDU-3) (0)en)

/ (3:00 - h 15 - BREAK) 11.1) Comments by ACR5 subcommittee clair-man (DAW /MME)

TAB 11------

11.2) Briefing (videotape) and discussion with representatives of NRC staff and Atomic Energy of Candada, Ltd.

12) 4:00 4:45 P.M.

Maintenance of Nuclear Power Plants (0 pen) 12.1)

Continue discussion of the ACRSreporttoNRC(CN/HA) proposed

13) 4:45 -

5:30 P.M.

PreparationofACRSReports(0 pen) 13.1) Discuss proposed ACRS reports to NRC regarding:

?

13.1-1) Definition of Adequate Pro-tection(DAW /HDH) t 13.1-2) GI-135, Steam Generator and SteamlineOverfill(PGS/EGI) 13.1-3) GI-87, Failure of HPCI Steam Line Without Isolation (CN/EGI) 13.1-4) GI B-56, Diesel Reliability and Reg. Guide 1.9, Rev. 3 (CJW/MME) 13.1-5) Maintenance of Nuclear Plants (CM/HA)

14) 5:30 -

6:30 P.M.

Appointment of ACRS Members (0 pen / Closed) 14.1) Discuss qua11fications of candidates proposed for consideration as nominees forappointmenttotheACRS(CN/MFL)

(Note:

Portions of this session will be closed as appropriate to discuss information the release of which would represent a clearly unwarranted invasion of personal privacy. )

Saturday, October 7,1989 Room P-110, 7920 Norfolk Avenue, Bethesda, Md.

I 15)8:30 12:00 Noon Preparation of ACRS Reports (0 pen)

Discuss proposed reports to NRC noted above l

12:00 1:00 P.M.

LUNCH

f

.l 354th ACRS Heeting Agenda !

3:00 3:15 P.M.

BREAK i

5) 3:15 4:45 P.M.

Mainte.ance of Nuclear Power Plants (0 pen) 5.1) Goments by ACR5 subcomittee chaiman

~~~~

5.2) r g/ discussion with NRC staff regarding revisions in the proposed NRC policy statement 5.3) Discuss proposed ACRS report to NRC

6) 4:45 6:00 P.M.

FutureACRSActivities(0 pen)

T AB.......... 6.1 ) Anticipated subcomittee activities (RPS/RFF)

T AB----------

6. 2 )

Items proposed for ACRS consideration (FJR/Ntf)RPS 6.3)

Items to be discussed during December 1-2 Regulatory Policies and Practices subcomittee meeting (HWL/GRQ) i Friday. October 6. 1989; Room P-110. 7920 Norfolk Avenue. Bethesda Md.

[ 8:30 9:30 A.M.

Generic Issue B-56. Diesel Reliability I

~

and Associated Regulatory Guide 1.9 TAB 7------

Revision 3 (open) 7.1) Coments by ACRS subcomittee chaiman (CJW/MME) 7.2) Meeting with NRC staff representatives 7.3) Meeting with NUMARC representatives 10:30 A.M.

GenericIssue87.FailureofHPCISteamLine 9:30 Without Isolation (0 pen) 8.1) Comnents b ACRS subcommittee chaiman TAB 8------

Il@g(CM/EGI) y qpW (repre(egativM 10:30 10:45 A.M.

BREAK 9)10:45-11:30 A.M.

DefinitionofAdequateProtection(0 pen) 9.1)

Discussion of proposed ACRS report to the NRC on the definition of adequate protection (DAW /MDH) 10)11:30 12:15 P.M.

ACRS Subcomittee Activities (0 pen) 10.1) Re) ort of Planning and Procedures su')comittee regarding p(see R. F.

roposed changes in ACRS Bylaws Fraley memorandum dated September 9, 1989)

a.,

354th ACRS Meeting Agenda 16)2:15 3:30 P.M.

Miscellaneous (0 pen / Closed) 16.1) Complete discussion of items con-sidered during this meeting e

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MINUTES OF THE 354TH ACRS MEETING i

OCTOBER 5-6, 1989 l

The 354th meeting of the Advisory Committee on Reactor Safeguards was held at 7920 Norfolk Avenue, Bethesda, Md., on October 5-6, 1989. The purpose of this meeting was to conduct the discussions and to perform the actions described in the attached agenda.

The meeting was chaired by Dr. Remick.

i All of the discussions were held in open session except for a short session during which the Committee discussed the qualifications of candidates proposed for consideration as ACRS members.

-A transcript of selected portions of the meeting was kept and is available in the NRC Public Document Room.

[Copiesofthetranscriptareavailablefor purchase from the Heritage Reporting Corporation, 1220 L St., N.W.,

Washington, D.C.

20005.]

Chairman'sReport(0 pen)

[ Note: Mr. R. F. Fraley was the Designated Federal Official for this portion ofthemeeting.]

Dr. Remick began the meeting with a brief summary of the planned agenda and the procedures under which the meeting discussions were to be conducted. He summarized the circumstances which had led the NRC to assess fines on a number of utilities during the last month. He noted that recent actions by the State of Nevada were likely to lead to significant delays in the DOE program to construct an HLW repository at the Yucca Mountain site. Dr. Remick stated that an NRC workshop on life extension for operating nuclear plants was to be held in Reston, Va. on November 13-14, 1989, and that members interested in attending should contact the NRC staff, i

'II. DefinitionofAdequateProtection(0 pen)

[ Note: Mr. Dean Houston was the Designated Federal Official for this portion of.themeeting.]

Mr. Ward, Chairman of the Safety Philosophy, Technology, and Criteria Subcom-mittee, discussed the past ACRS actions in regard to the development and review of a proposed implementation plan for the Safety Goal Policy. He indicated that there were four principal areas of disagreement between the Committee and the latest NRC staff-proposed plan. These are:

(1) the defini-l tion of a large release, (2) the absence of a containment performance guide-H line, (3) the absence of a caveat addressing the inability of PRAs to assess human performance, and (4) the definition of adequate protection. Mr. Ward noted that, of these items, the staff has chosen not to incorporate or address items 2, 3, and 4 in the proposed plan. Mr. Ward indicated that the ACRS had discussed these disagreements with the Commission on May 3, 1989 and that the L

Commission had requested further ACRS effort only in regard to the definition of " adequate protection."

It is the Committee's opinion that the safety goal could be used indirectly as a surrogate to define adequate protection.

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o 354TH ACRS MEETING HINUTES 2

Mr. Ward further indicated that the Comittee has been silent on where the backfit rule fits into the application of safety goals and the concept of adequate protection.

He suggested that the backfit rule would be used to deal with what could be termed as the " threshold problem," that is, approaching or slightly exceeding some quantitative limit expressed in the safety goal. He acknowledged that Dr. W. Houston (NRC-RES) had prepared a draft document to compare both the NRC staff and ACRS positions on adequate protection but recommended that the Committee not coment on it in its present form.

Dr. Houston acknowledged the differences between the Comittee and the NRC staff in regard to the proposed implementation plan for the safety goal. He also expressed concerns on what he believes is a serious communications problem between the staff and the ACRS -- some or most of which may be a question of semantics.

He indicated that, from the staff's viewpoint, the determination of adequate protection has been a judgmental finding and that the staff is not recommending a quantification or definition for the term

adequate protection." He indicated that the words " adequate protection" appear in the original Atomic Energy Act but nowhere in the regulations except in the backfit rule (10CFR 50.109). The application of this rule and a need i

for a specific definition of adequate protection was challenged in the courts and the courts had declined to force the Comission to define the term. He stated that compliance with the Comission's regulations is taken as presump-tive evidence of adequate protection, but this is not considered by the NRC staff to be a definition of adequate protection.

Dr..Remick indicated that the staff had come fairly close to describing the ACRS position on' adequate protection in regard to safety goals in their document on the implementation plan (SECY-89-102).

He read statements from the ACRS report dated February 16, 1989 which said the ACRS believes that the i

safety goal should play.an important but indirect role in defining adequate protection.

Dr. Houston acknowledged that the staff had misread the ACRS l

report. Dr. Lewis stated that the ACRS emphasized that the safety goals be used to judge the body of regulations, not an individual plant.

t Dr. Remick also questioned why cost-benefit aspects were discussed in the 3

proposed implementation plan when they were not part of the Safety Goal Policy Statement.

Dr. Houston read some statements from the Policy Statement about safety trade-offs which the staff had interpreted as having an association with cost-benefit analysis.

Dr. Kerr indicated he believed that, in their l.

present form, the Safety Goal Policy and backfit rule are not consistent.

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Dr. Siess asked if the TMI-2 requirements constituted a decision by the L

Commission to raise the level of adequate protection.

Dr. Houston indicated that they were simply backfits not associated with the concept of adequate protection.

Dr. Lewis indicated that indeed these regulation changes were designed to raise the level of protection because the public believed the existing level of protection was inadequate.

Dr. Houston discussed briefly(the other three areas of disagreement between the NRC staff and the ACRS:

1) the definition of a large release, (2) the l

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354TH ACRS HEETING MINUTES 3

absence of a guideline for containment performance, and (3) a caveat address-ing the assessment of human performance in PRAs.

The Committee discussed its February 16, 1989 report on this matter and expressed an opinion that the staff had not paid enough attention to the advice given in the ACRS report. The Comittee agreed that another report to the Commission should reemphasize the positions given in the February 1989 report and indicate that the Committee is awaiting some further action by the staff to resolve the current disagreements.

The Committee decided to write a report on this matter. This report is discussed in Section IX.

III. Generic Issue 135. " Steam Generator and Steam Line Overfill Issues" (0 pen)

[ Note: Mr. E. Igne was the Designated Federal Official for this portion of themeeting.)

1 The_ Committee heard presentations by A. Notafrancesco, NRC-RES, and-Mr. R.

Neve, Scientech, a contractor, on the NRC staff's proposed resolution to Generic Issue 135.

This issue was initiated in 1986 to integrate various steam generator programs and the related safety issues.

Some of these activi-ties were already assigned as generic issues, some were part of other issues, and the remainder were ongoing staff activities dealing with steam-generator l

integrity or steam-line integrity.

Overfill, in the context of Generic Issue I

135, is defined as water entering the main steam line which can be caused by excessive feedwater flow due to control system failure or a steam generator tube rupture (SGTR) event. Control system failures were analyzed in USI A-47,

" Safety implications of Control Systems." The overfill due to SGTR is the major concern investigated under Generic Issue 135. The results of the l

Generic Issue 135 effort basically support the previous prioritization and the j

relatively low public risk associated with SGTR and steam line overfill events i

as reported in NUREG-0844.

As shown in this report, SGTR events are not L

dominant contributors to risk. The NUREG-1150 analysis produced comparable L

L risk results for Surry and Sequoyah for SGTR events.

This has led the staff l

to conclude that new requirements in dealing with the prevention of SGTRs l

could not be justified.

The industry has generally adopted the staff recom-1 mendation as documented in the resolution of Generic Letter 85-02, which the L

ACRS has reviewed and favorably commented on.

~ 1n response to Comittee questions, the staff stated:

the study was performed essentially for Westinghouse plants, and j

that a scenario where water slug could cause a pressure boundary rupture and release steam into the auxiliary building was not analyzed.

The Committee decided to write a report on this matter.

This report is discussed in Section IX.

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i 354TH ACRS NEETING MINUTES 4

~lV.

Meeting with the Director of the Office of Nuclear Regulatory Research (RE5) (0 pen)

[ Note:

portion of the meeting.] y was the Designated Federal Official for this Mr. Sam Duraiswam s

The Committee met with Dr. E. Beckjord, Director of RES, and discussed several l

items of mutual interest noted below.

Impact of the FY 1990 Budget Reduction Proposed by the Congress on the NRC Safety Research Program Dr. Beckjord stated that the President's budget request submitted to the Congress during January 1989 included a total budget of $475 million for the overall agency for FY 1990. Of this, $108 million was earmarked for the NRC Safety Research Program Support, including $5 million for support of the activities related to the High Level Waste Disposal Program.

The Appropriations Bill recently passed by the Congress included a reduction of $30 million to the overall agency budget. To accommodate this reduction, the EDO has proposed, and the Commission has approved, a reduction of $20 i

million to the NRC Safety Research Program Support budget, thus bringing the

'r total research budget down to 88 million. Dr. Beckjord stated that the current research budget includes $3.3 million for research and technical assistance to develop license renewal regulations.

l Dr. Beckjord stated that the proposed reduction of $20 million will be accom-plished as shown below.

l-Integrity of Reactor Components l-l The funding for this program will be reduced by $5.4 million, from $33.1 I

million to $27.7 million.

Specific research elements affected under this l

program are as follows:

Reactor Vessel and Piping Integrity - The funding for this element will I

l be reduced by about 52 million, from $15.6 million to $13.6 million.

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-Some of-the impacts of this reduction are.

Two of the pressurized thermal shock tests (PTSE-3 and PTSE-4) will be delayed for at least a year.

l Research on cyclic crack growth rate for ten of the vessel and piping steels being considered in this program will be deferred.

Most of the work on initial flaw distribution in pressure vessels will be eliminated.

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'354TH ACRS HEETING MINUTES 5

j Aging'of Reactor Components - This will-be reduced by about $0.9 million, from 38.4 million to 37.5 million. This reduction would:

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Reduce by 20% aging degradation assessments for electrical / mechani-j cal components and systems.

Engineering Standards Development - This element will be reduced by about j

30.4 million, f rom 31.0 million to $0.6 million. This reduction would 1

reduce' contractor assistance to the NRC in engineering standards develop-ment and transfer to ASME Code Sections 111 and XI.

Seismic'and Structural Research - This element will be reduced by $2.1 million, from-58.1 million to 56.0 million. The impact of.this reduction would:-

L" Delay for one year. studies on earthquake causal mechanisms and prehistoric earthquake occurrences.

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Eliminate research to obtain information for use in the review of 1

changes in the ASME piping design criteria.

Delays in the cooperative effort with Japan on seismic response of structures on soil sites.

\\y Dr. Siess asked on what basis the NRC. decides which research elements should-I be cut, Dr. Beckjord stated that they. have first prioritized all the research 4

elements, and those~ elements that received low priority ranking were selected as candidates for reduction.

Mr., Carroll asked whether the scope of the aging research program has been

. reduced as a result of the buiget reduction for that program. Mr. Shao stated

.that-the~ scope of-the aging research-program has not been reduced.

Because of-the budget reduction, they have postponed the testing of certain components.

Preventing Damage to Reactor Cores

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This program will be reduced by $6,0 million, from $22.0 million to $16.0 million. Specific'research elements affected under this program are as.

follows:-

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" Plant Performance - The funding for this research element will be reduced by 53.3 million, from $5.2 million to $2.9 million. This reduction would:

i Terminate experiments to confirm OTSG scaling assumptions for certain transients.

Terminate technical analysis for suggested tests on accident manage-ment in ROSA IV.

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354TH ACRS HEETING MINUTES 6

y Reactor Applications - This will-be reduced by'$2.1-million, from $5.3 million.to 33.2 m1111on. This reduction would:

Eliminate half of the baseline activities of the thermal hydraulics m

Technical Support Center.

T-Terminate work on thermal hydraulic analysis of CANDU reactors from i

the point of view of NRC licensing criteria.

Human Factors - This will be reduced by $0.2 million, from $4.7 million-to 34.5 million.

This reduction would:

Delay research on the role of Shift Technical Advisor, Delay research to integrate human and hardware reliability into-p L

assessment of advanced reactor designs.

Reliability Assessment - This will be reduced by about $0.3~ million,'from $2.8 million to 52.5 million. This reduction would delay development of an objec-I tive basis for regulatory requirements to monitor performance.

With regard to the RES' proposal to terminate the research related to the B&W OTSGs, Dr..Catton-stated that some time ago the staff felt that this. work is j

very important to understanding the behavior of B&W OTSGs. He asked on what

- basis the sta.ff now has. decided to terminate this work.. Dr. Beckjord stated:

that it is desirable to continue this work; however, since it'is not expected to contribute to the -immediate needs of the agency, coupled with the stringent budget constraints, they have decided to terminate this research.

Dr. Catton stated he believes.that at least some low-level effort should be

.j continued to maintain capability inithis area.

Dr. Sheron stated that.they i

plan to do work, as-needed.. related'to B&W OTSGs at the University of Maryland facility.

He believes that such work will help maintain necessary capability, a

in this area.

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'i Mr.: Ward stated that'the B&W plants incorporate.certain unique features, such as the OTSG. The thermal-hydraulic. behavior of these plants during transient.s and beyond-design-basis events is not well understood as compared to other PWRs. Even though the. experimental data base on this issue is very limited, 4

he does not understand why the staff plans to terminate the work related to.

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o B&W OTSG.

Dr. Beckjord stated that he agrees that they 'do not have a thorough-knowledge of the behavior of the B&W plants. However, in view of the budget constraints, coupled with the fact that this work is not expected to con -

tribute to the immediate regulatory needs of the agency, they do not believe that it is worth continuing this work.

- Dr. Catton stated that the TRAC and RELAP codes do not have adequate capabil-ity to model properly the B&W OTSG.

He believes that some level of effort j.

should be continued to alleviate this modeling deficiency of these codes. Dr.

Sheron stated that he hasn't heard any complaints about the inability of TRAC

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i 354TH ACRS MEETING MINUTES 7

and RELAP codes in analyzing the behavior of plants during transients.

They

' plan'to continue a low level of. code development work, and under that effort they might be able to take care of any modeling concerns. He plans to discuss this modeling problem with the Thermal-Hydraulic Phenomena Subcommittee in the future.

Reactor Containment Performance The funding for this program will. be reduced by $6.6 million, from $28.9 million to $22.3 million. The impact of this reduction on specific research elements under this program are as follows:

Core Melt and RCS failure - The funding for this research will be reduced by about 51.0 million,-from $7.7 million to $6.7 million.

This reduction would:

Cancel-planned confirmation of predicted natural circulation-induced' surge line failure for station blackout.

Slow work on fission product behavior within the RCS and within the containment.

Reactor Containment Safety - This will be reduced by $2.3 million, from

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511.0 million to 58.7 million.

This reduction would-Slow further development of tools for analyzing core / concrete interaction.

' Curtail the core / concrete interaction experiments.

Cancel plans for construction of experimental, facility to study i

high-temperature hydrogen combustion and detonation in containments.

Reactor Containment Structural Integrity.This element will be reduced by:50.4 million, from 52.2 million to 51.8 million. This reduction would:-

e Delay the extention of results of test to prestressed concrete

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containments.

Defer activities on structural response to hydrogen-related load-

ings, f

Reactor Accident Risk Analysis - This research will be reduced by $2.9 million, from 57.9 million to $5.0 million.

This reduction would:

Delay development and use of advanced PRA methods.

Reduce the number of plant risk studies and level of contractor support for performance of these studies.

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.' Resolving Safety Issues and Develo)ing Regulations - This program will-be reduced by 31.1 million, from 310. 3 million to 315.7 million._ This reduction would:

Delay the issuance of the Regulatory Guide on control room habi-tability.

Reduce the scope of review of risk importance and effectiveness _of-NRC regulations.

Limit value impact analysis for backfitting regulatory requirements identified in the Containment Performance Improvement Program for containments other than BWR Mark Is.

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Gramm-Rudman-Hollings Deficit Reduction Law Dr. Beckjord stated that another reduction, totalling about $25 million, to the overall agency budget may be imposed on the NRC to comply with the Gramm Rudman-Hollings Deficit Reduction Law.

If1 that happens, the research budget will be reduced by about $15-$20 million, if the research budget is reduced further, they will not be able to perform the necessary research to support the agency's mission.

In response to a question from Dr. Siess, Dr. Beckjord stated that, based on their discussions with the OMB, they do not expect major restoration of funds to the research budget in FY 1991.

Mr. Ward stated that it seems that the Commission's strategy is to reduce the research budget significantly to accommodate the budget reduction proposed by the'0MB or the Congress.

It makes him wonder whether the Commission has already. concluded that research is not important or whether it.is drifting toward making such a' conclusion.

Dr. Siess stated that since the RES has a larger program-support budget than the other NRC Offices, it normally absorbs the major share of the budget reduction.

Mr. Ward stated that it may be administrative 1y easier for the Commission to cut the research budget; but, he does not believe it is the right. thing to do.

Dr. Siess stated that it might be a subject for discussion with the Commission.-

In response to a. question from Dr. Siess, Dr. Beckjord stated that the re-search user offices have supported the research initially proposed at a $120 million-level. The Congressional Committees dealing with the NRC budget also seem to believe that the NRC research is important.

Nuclear Safety Research Review Committee's Views on the Proposed Budget Reduction Dr. Beckjord stated that the Nuclear Safety Research Review Committee (HSRRC) supports the budget level of $108 million for the FY 1990 NRC research

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354TH ACRS-MEETING MINUTES 9

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program. He believes that the $108 million level is justif1_' on the. basis of-the need for aggressive research in the following areas:

-Aging Accident Management-Human Factors.

Waste Disposal University Research Dr. Siess asked whether the NSRRC has commented on the appropriateness-of the research elements chosen as candidates for reduction.

Dr. Beckjord stated that he has not discussed this matter with the NSRRC.

. Dr. Siess requested copies of'the correspondence between the NSRRC and RES.-

Dr. Beckjord agreed.to provide copies.

Diversity in Research Providers Dr. Beckjord discussed briefly the levels of funding expected to be spent in FY 1990 for:research in national laboratories, universities, not-for-profit organizations,-ete:

National laboratories

$ 64.1 million Accident management.

3.2 million University grants 1.1 million Not-for-profit organizations 0.2 million Industry 10.8 million Contributions to the'NRC Safety Research From the Implementation of the

. Recommendations of=-the National Research Council

-Dr. Beckjord stated'that the implementation of the-National Research Council's recommendations-has-resulted in the:

Development of a research philosophy that has been useful to RES in defining the research that needs to be done.

Establishment of research review committees represented by_RES and research user offices.

Establishment of NSRRC that has been very helpful in providing comments and recommendations on the ongoing and proposed research programs.

Establishment of competitive contracting.

With regard to bringing outside research expertise on board, Dr. Beckjord stated that he still has difficulty in getting experts on board because of the salary cap, housing problems, etc.

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354TH-ACR$ HEETING MINUTES 10 i

V.

MaintenanceofNuclearPowerplants(0 pen)

[ Note: Hr. H. Alderman was the Designated Federal Official for this portion ofthe' meeting.]

Mr. Michelson, Chairman, Maintenance Practices and Procedures Subcommittee, noted in his opening remarks that a draft letter addressing the proposed

. maintenance policy statement had been previously. discussed by the Committee but not issued. He noted that the NRC staff had made some changes to the.

maintenance policy statement, and asked the staff to discuss the changes, j

l f'r. Tom King, HRC-RES, discussed the changes made to the maintenance policy statement since the-September ACRS meeting.

He noted the changes were the

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l result of discussions with the ACRS and the staff's meeting with-CRGR.-

He noted that the Office of Enforcement was preparing a set of enforcement-criteria to be used during_the 18-month period that a maintenance rule.is held in abeyance.-

Mr. King discussed the key provisions of the proposed maintenance policy statement and related activities:

'Rulemaking will' be held in abeyance for 18' months of the effective date of publication.of the' maintenance policy statement.

Industry's performance of maintenance activities will be monitored.

Maintenance team inspections will continue.

A. proposed Regulatory Guide has been published for.public comment..

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Existing requirements'related to maintenance will continue to be en-forced.

The staff is continuing its work on the development of a maintenance

standard, i

The staff is continuing its work on the' development of maintenance performance indicators.

At the end of the 18-month period, the staff will determine if there is any need 'for any additional regulatory action.

Dr. Kerr asked where the criteria for enforcement of maintenance regulations would be found. Mr. J. Luehman, Office of Enforcement, replied that most of

the criteria are in Appendix B of 10CFR50.

He noted there are others in the plant Final Safety Analysis Report and in the plant Technical Specifications.

Mr. King noted that the staff was goirg to recommend to the Commission that if a particular violation has as its root cause a maintenance problem, the staff w

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354TH ACRS MEETING MINUTES-11 can consider an escalation factor in the amount of the civil penalty. Mr.

Luehman added that the proposed change would deal with escalating the civil penalty per violations that have been assessed at a certain severity level.

Dr. Siess pointed out an apparent inconsistency between the policy statement-and the proposed enforcement criteria.

He noted that the policy statement states that an escalated penalty can be used if the violation could-have been prevented if an adequate maintenance program had been implemented. The proposed enforcement criteria just states a maintenance activity.

Dr. Remick suggested that the staff note somewhere in the policy statement that they.Will continue to work on the maintenance rule. Mr. ving replied that.the staff would _look.at the policy statement and see if son.3 words to address Dr.'Remick's concern could be added.

Mr. King discussed expected licensee actions:

_ Licensees are expected to continue improvement in the maintenance area w

<and'in implementation of maintenance programs.

Licensees are encouraged to document their commitments for. improvement.

NUMARC and INP0 are encouraged to exercise leadership in affecting improvement in maintenance.

Industry is encouraged to participate -in the development and voluntary adoption of a maintenance standard.

. Industry is encouraged in the-development and use of_ maintenance perfor-mance. indicators.

-Industry is encouraged in the participation in and use of the NPRDS-System.

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Industry is' urged to ensure _that their maintenance programs encompass all

systems, structures and components who failure could significantly impact i

safety or security.

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-Mr. King noted the changes that have been made to the policy statement since M*

the September ACRS' meeting:

7, Words have been added to acknowledge that the industry maintenance programs have-improved and to note that there are some licensees with 4

acceptable programs.

'Words have been added discussing escalated enforcement.

The discussion on the use of NPRDS has changed the emphasis from enforced

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use to encouraged use. The staff intends to achieve more timely report-E ing and the use of the data, not the addition of more requirements, q

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354TH ACRS MEETING MINUTES 12-The area of scope in the policy statement has been revised to use the scope as defined in the regulatory guide.

Dr. Kerr asked as to how many of the plants that have had maintenance team inspections were considered acceptable by the NRC staff. Mr. Gody, NRR, replied that 33 sites have been inspected. Two sites were found to be not acceptable as f ar as implementation of the maintenance program plan and one site was found to have a poor plan.

'VI.

Generic Issue B-56. " Diesel Generator Reliability," and Associated Regulatory Guide 1.9, key. 3 (0 pen)

[ Note: Dr. M. El-Zeftawy was the Designated Federal Official for this portion.

of the meeting.]

Mr. Wylie, Chairman, AC/DC Power Systems Reliability Subcommittee, briefed the full Committee regarding the subject issue.

He indicated-that Generic Issue (GI) B-56 is a_ safety issue that is closely related to the station blackout rule.' This identified the need for ensuring that the reliability of the diesel generators be at'a reliability level of 0.95 or better. The NRC staff

-has proposed that. resolution of GI B-56 be accomplished through the issuance of Revision 3 to the existing Regulatory Guide 1.9, " Selection, Design, Qualification, Testing.and Reliability of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants."

c Mr. Wylie indicated that the staff's proposed revision to Regulatory Guide l'.9

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is intended to-integrate all the requirements that are currently contained in several regulatory guides and generic letters into a single regulatory guide.

This~ regulatory guide will define the principal elements of a diesel generator reliability program which, f or the most part, is consistent with present industry practices, in addition, it will better define testing requirements, with.the possibility of elimination of cold-fast starts, and it will reduce the previous requirements for accelerated testing, in the proposed revision, the staff has assigned " alert" levels and remedial actions to be taken if a deterioration of the reliability is encountered for a diesel generator.

The staff concludes that the issuance of Regulatory Guide 1.9, Rev. 3 and the implementation of the diesel generator reliability program will obviate;the need.to. consider diesel generator aging explicitly in the licensing. renewal process.

In addition, the issuance of Regulatory Guide 1.9, Rev. 3 will not introduce any new regulatory requirements beyond those currently required lfor compliance with the station blackout rule.

Regulatory Guide 1.9, Rev. 3 was submitted for public comment and the staff has incorporated many of those comments in the present draft.. The staff has worked closely with the industry and NUMARC during the development of this regulatory guide.

The AC/DC Power Systems Reliability Subcommittee held a meeting on October 2, 1989.to discuss this issue. There are some differences between the staff and NUMARC regarding some of the items contained in the proposed revised regula-tory guide.

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354TH ACRS HEETING MINUTES 13 Mr. Wylie indicated that generally, as a result of the AC/DC Power Systems Reliability; Subcommittee meeting, the Subcomittee members consider the approach being taken by the staff in developing the subject regulatory guide to be appropriate.

Mr. A. Serkiz, NRC-RES, indicated that Regulatory Guide 1.9, Rev. 3 was.for previously discussed with the CRGR on September 14, 1988, and was issued, comment in-November 1988. The comment period was closed in March 1989 and there were 15 responses, the last being received in July 1989. Comments were-received from NUMARC, EPRI, and nine utilities, ASME, IEEE,-IMO Delaval, and one individual.

The staff is continuing its discussions.with NUMARC!s B-56' working group to address comments received and to obtain a-better insight into practices currently employed to achieve the levels of reliability prevalent.

As a result, Regulatory Guide 1.9, Rev. 3 has been considerably restructured to use INP0's U.S. Industry Plant Performance Indicator Program (PPIP) defini-tions and records structure, guidance set forth in IEEE and ASME standards, and revised Appendix D to the NUMARC-8700 document " Guidelines and Technical

' Bases for NUHARC Initiatives. Addressing Station Blackout at Light Water Reactors" (8/29/89 draft). Table 1 of the Regulatory Guide provides a cross-reference between Regulatory Guide 1-9, Rev. 3 and HUMARC's revised Appendix D.

Mr. Serkiz indicated that Regulatory Guide 1.9, Rev. 3 outlines a diesel' generator reliability monitoring and actions program.

For demand failure combinations for all diesel generators with a target reliability of 0.95, 3 failures in the lasi 20 demands or 8 failures in the last 100 demands is considered a " mild" action state 7 A " strong" action state is considered to be 4 failures-in the last 50 demands and 8 failures in'the last 100 demands.

For

-target reliability of 0.975, 3 faiTides in the last 20 demands or 4 failures in the last 50 demands or 5 f ailures in the last 100 demands is70nsidered a

" mild" action stete.

For 4 failures in the last 50 demands and 5 failures in the last 100 demands it is considered a " strong" action state. -The " mild" action state would require to review failures in the last 20, 50 -and 100 demands to determine.if there are patterns in the failure modes or causes.

The " strong" action state would require the following:

Notify the NRC that the " strong" action state had been implemented.

Ascertain the nature of the reliability problem.

Assessment actions should include one or more of: a) root cause analysis, b) analysis for patterns in failure modes, c) assessment of other plants' failure information, d) exploratory surveillance, e) exploration condi-tion monitoring, f) reliability diagnostic analysis, and g) design /

operational changes.

Document and implement corrective actions.

Revise reliability program as appropriate.

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354TH ACRS itEETING MINUTES

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' Demonstrate effectiveness of actions taken.

A " problem" diesel generator is defined in Regulatory Guide 1.9, Rev.- 3 as an f

individual diesel generator w er Nncing 3 or more failures in the last 20

' demands.. Should this case occur, a " mild" action level would be declared and the actions defined above would be endertaken.

If the " problem" diesel experiences an additional failure, such that there have been 4 failures in the last 25 demands, then a " strong" actioi level would be declared, j

l Following completion of corrective actions, restored performance of the problem diesel should be demonstrated by conducting seven consecutive failure-free starts and load-run tests. The monthly surveillance schedule should not i

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be resumed until seven consecutive failure-free start and load-run demard tests have been completed.

If, following completion of the seven consecutive failure-free tests, the same problem diesel experiences another failure such' that there have been 5 failures in the last 25 demands, this problem diesel would be. declared inoperable and an overhaul would be required.

.s Mr. Serkiz indicated that, except in those cases in which an ' applicant. pro-l l

poses an acceptable alternative method for complying with the Commission's regulations, Regulatory Guide 1.9, Rev. 3 will be used for plants for which

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the construction permit is issued after the issue date of the subject regula-1 tory guide, ~ plants for which the operating license application ~is docketed six 1

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months or more after the issue date of the subject regulatory guide, and for l-plants for which the licensee voluntarily commits to the provision-of the.

- subject regulatory guide.

1 l

Mr. A. ' Marian, NU!iARC's representative, indicated that NUMARC has been working closely with the'NRC-staff to establish a consensus on documents-to support the resolution of Generic Issue B-56.

Such documents are indeed complementary.

}

in nature. The purpose of NUMARC's effort is two-fold.

First is to develop-a

[-

diesel generator reliability program that offers guidance to utilities to.

maintain the-diesel generator target reliabilities that were established in' 1

the station blackout rule (i.e., 0.95 or 0.975). Second, is to offer resolu-

tion of Generic Issue B-56 and to bring industry performance up-to a. level of j

0.95. Mr. Marian noted that the industry performance is currently in-the order'of 0.98 in. terms of diesel generator reliability. The 0.98 value is based upon a methodology and approach that was establishe'd by EPRI and was published in NSAC.108, which accumulated data on diesel generator performance h

.from 1984 through 1986. NUMARC-8700, Appendix D,. focuses primarily onL a' L

reliability-based' program that does not include the " problem" diesel generator testing. This concept'still presents a point of difference between NUMARC and' t

the NRC staff. ~NUMARC feels that the concept of a problem diesel generator i

'doesLnot necessarily support the reliability focus from the standpoint of the station blackout rule and it is inappropriate to include-it in Regulatory

. Guide 1.9, Rev. 3.

Mr. Marian also noted that the staff is recommending u

corrective action testing that embodies demonstration of restored reliability L

of the " problem" diesel generator via 7 consecutive failure-free tests con-I ducted at a frequency of 2 to 7 days.

NUMARC's view is that such testing l

should be considered only when a " strong" action state is entered based on the L

i l-u z,

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4 e

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-354TH ACRS MEETING MINUTES 15' i

overall nuclear unit reliability. Mr. Marian commented that the current performance data indicates that industry self-initiated activities have improved diesel generator reliability.

y Mr. Marian commented that station blackout is the largest contributor to core melt. Currently, NUMARC believes that the role of the diesel generator for the design basis accident scenario as structured in Regulatory Guide 1.9, Rev.

3 constitutes a lower contributor. Consequently, NUMARC and the industry had undertaken an initiative to reduce the number of cold fast starts to 18-month 1

intervals, i

Dr. Shewmon questioned the basis for requiring the fast start testing of the 1

diesel generators, which is recommended by the NRC staff to De conducted every six months.

He indicated that the original basis was the assumption of a-guillotine breat in a large pipe. The NRC believes a guillotine break to be a very low probability' event, now essentially. incredible. Dr. Shewmon recom-i mended that the staff rethink its position and basis.for requiring the fast t

start testing. Dr. Kerr stated that he shared the same concern.

Dr. Kerr commented thet the staff's request for a-different recordkeeping system will increase the burden on the licensees, especially if the staff does not have any particular reason to assume thet the existing recordkeeping i

system needs improvement.

Mr.LCarroll expressed some concern regarding the interpretation of all the different documents that deal with the diesel generator reliability such as the IEEE, Standards, ASME Standards, NUMARC-8700, and Regulatory Guide 1.9, Rev. 3 from the viewpoint of the operating personnel and the resident-inspec-y tors.

Mr. 'Wylie commented that the NRC staff stated that the issuance of Regulatory i

Guide 1.9,'Rev. 3 will obviate the need to consider diesel generator aging f

explicitly in the license renewal process.

However, the proposed regulatory guideLhas neither documented nor supported such a' statement, and it ie not clear 1that the basis for making such a decision has been established.

l E

Mr.LWylie believed that the staff should expand more on its definition of 4

diesel generator failures and the relationship with the requirements. Mr.

. Carroll shared the same belief.

LMr. F. Rosa, NRC-NRR, said that he has some disagreement with the NRC-RES staff regarding Sections 2.3.3, 3.5, and 3.6 of Regulatory Guide 1.9,'Rev. 3.

-He believes that Section 2.3.3, which deals with the corrective action test-ing, should be modified such that following the occurrence of a degraded situation the surveillance testing interval for the problem diesel generator

-should be reduced to no more than 7 days, but not less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This 0,

test frequency should be maintained until seven consecutive failure-free start and load-run tests have been performed to demonstrate the effectiveness of-corrective actions taken. At that time, monthly surveillance testing can be resumed.

However, if subsequent to the 7 failure-free tests, one'or more Y

~~

,J y

{IA.,(

em r,

(jk

(

c354TH ACRS HEETING HINUTES 16 j

Ly;,l Ok

additional failures occur such that there are 3 or more failures in the last i

20.. tests, the testing interval should again,be reduced and maintained until 7 j

'J %

consecutive failure-free tests have been performed. The diesel generator

l t '

undergoing corrective action testing should.be considered " operable" unless V'

other license requirements necessitate' declaring such a diesel generator

" inoperable."

, 'y ' J o

W Mr. Rosa also believes that Section~3.5 (which deals with a " problem" diesel a

ilc

' generator) should be modified such that if eny individual diesel generator y,

experiences 3-or more failures in the last 20. demands, then a mild action is declared.

If, during the corrective action testing,-the same diesel generator experiences additional failures such that the number of failures in the last 20 demands is 5 or more, then such a diesel generator should be declared o,

inoperable and a major overhaul should be considered.

Following the overhaul,:

i R'

a series of 14 consecutive failure-free start and load-run tests should be uw' conducted, p

For Section 3.6 (recovery from a strong action state), Mr. Rosa believes that this section should be modified such that, following completion of corrective uJV l programmatic actions from a strong action state, restoration of diesel genera-Mi tor reliability levels for any problem diesel should be demonstrated by M

conducting 7 consecutive failure-free starts and load-run tests. All starts p'

and load-runs performed during the corrective action testing should be includ-i ed in the-nuclear unit diesel generator reliability data base so long as the-s diesel generator is declared operable, j

The Committee decided to continue its discussion and review of this matter at M<

a later date.

VII. Generic Issue 87, " Failure of HPCI Steam Line Without Isolation" (0 pen) g

[ Note: Mr. E. Igne was the Designated Federal Official for this-portion of-r

~

l the meeting.]'

m t

Mr. Michelson, Chairman of the Mechanical Components Subcommittee, presented a V

report to the Committee. Mr. Michelson stated that recent test results for motor-operated valves considered representative of the type used for BWR RWCU system isolation'left in question the ability of similar valves, including those for the high-pressure coolant injection system, to close under certain-differential pressure and flow conditions. Generic Letter 89-10, which the Committee has reviewed and reported on, addresses such valve operability i

questions and therefore also addresses the subject concern of Generic Issue

87. Hente3 the Committee agrees that the proposed resolution of-Generic Issue 87 is acceptable and is implemented by the issuance of Generic Letter 89-10 for those cases where the existing plant design basis requires the isolation valves'to be operable under conditions resulting from a large downstream pipe break.

If there are other generating units for which the existing design basis does not include isolation value operability under the conditions resulting from a large downstream pipe break, then some other resolution needs to be proposed by the NRC staff.

Nyl

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1 W --

GV 354TH ACRS. MEETING MINUTES 17 gg 5

The Committee decided to continue its discussion and review of this matter at' a later date.

i 7

VIII.

StandardizedPlantDesign(CANDU-3)(0 pen)

,y 1

[ Note: Dr. M. El-Zeftawy was the Designated Federal Official for this portion of the meeting.]

.Mr. Ward, Chairman, Advanced Reactor Designs Subcomittee, said that the NRC o

staff has recently received a request from the Atomic Energy of Canada _ Limited (AECL) to review the CANDU-3 design for a possible design certification in the w

United States. Consequently, it would be beneficial for the ACRS to be briefed at the present time on the CANDU-3 design.

Such a briefing would o

enable the ACRS to perform a technical review of such a design in the' future.

Mr. Persinko, NRC-NRR, presented a brief-background of information on this subject.

He stated that on May 25, 1989, the NRC received a letter from AECL.

The letter indicated AECL's intent to seek a standard-design certification under 10 CFR Part 52.

In response to AECL's letter, the NRC, in a letter of July 6, 1989, requested AECL to submit a licensing review basis and submittal schedules upon which AECL intends to submit the application and proposed acceptance criteria.

This is in accordance with procedures which have been 4

used for similar type documents that have been suamitted on the ABWR and other-i standardized plants.

i

Currently, the staff is preparing an information paper on the subject to the Commission. The staff is also in the process of deteloping another Commission j

paper on the review priorities of. the_ evolutionary sind advanced reactors.

In that paper, CANDU-3 will be-considered-in the prioritization.

.Mr. G.-Kugler, Deputy to the Vice-President, AECL Technologies, stated that the AECL has been looking at the U.S. market:for several years with the intention of selling one CANDU plant in-the U.S. this next. decade.

Mr. L. Rib, AECL representative, presented a video tape describing the histori-cal, background of AECL and the CANDU-3 design. The AECL was established as Federal Crown Corporation-in April 1952.to develop and promote peaceful

?y applications of atomic energy. The major AECL facilities are Chalk River Nuclear Laboratory (approximately 2200 staff), Whiteshell Nuclear Research establishment (approximately 1000 staff), and Sheridan Park (approximately 100 staff). The Chalk River Nuclear. Laboratory contains an NRU reactor, a freon j

heat transfer loop, and metallurgical hot cells. The Whiteshell Nuclear Research facility contains a WR-1' reactor, containment test facilities, a high-temperature fuel channel test facility, and an underground storage test facility.

Sheridan Park is the site of the CANDU engineering offices and the

,~

Sheridan. Park' engineering laboratory.

'AECL considers the CANDU-3 to be an advanced but proven nuclear power generat-

^

ing system. While significant advances in design over earlier CANDU systems are incorporated in CANDU-3 design, the key components (such as steam

^

~

pp~3 t

, %m,., '

s U

g, NW $

354TH ACRS MEETING MINUTES 18 generators, coolant pumps, pressure tubes, fueling machines, etc.) are identi-a cal to those already proven in service on operating CANDU reactors. Thus the reactor physics, fuel, fuel channel, and thermal hydraulics aspects of CANDU-3 h

can be considered to be well proven.

g, The CANDU-3 is the smallest version of the CANDU pressurized heavy water reactors developed in Canada. With a net electrical output of 450 MWe, the CANDU-3 complements the established mid-size CANDU-600 t

The CANDU-3 design uses natural uranium fuel and and heavy water coolant and moderator. The coolant and moderator are separated from each other by regu-larly and relatively loosely spaced fuel channels, the coolant being hot and 1"

pressurized, and the moderator cool and at essentially atmospheric oressure.

'w The CANDU-3 core has-very little excess reactivity and, as a consequence, all gm reactivity coefficients are very small.

The CANDU-3-core is larger than an LWR core generating the same power. This leads to a generally lower overall core power density, but also,to potential-xenon' induced flux tilts which are 4

controlled by an automatic computer-operated spatial flux control system.

Due 4

to the use of natural uranium fuel, there-are no criticality accidents possi--

g'

-ble in transportation or storage of CANDU fuel..

[

Mr. Kugler indicated that, for the CANDU-3 design, a LOCA is accompanied by a-positive void reactivity effect. This effect is, however, small. To counter-L act it, two fully capable, fast-acting, diverse, and independent shutdown h

to demonstrate an unavailability not greater than 10"gstable during operation systems are provided. Each shutdown system must be t unavailability on a

demand.

L The CANDU-3 fuel is contained in short (20":) fuel bundles containing approxi--

L mately 50 pounds of natural uranium dioxide' fuel' pellets.. The fuel design is relatively simple compared to LWR fuel.

Defective fuel bundles can be located r

and removed on-power leading to a relatively._ clean reactor coolant system.

,o l

h Mr. Kugler stated that since heavy water is expensive (approximately U.S.

W'

- $120/ pound), the CANDU design has always focused on leak-tight systems and P

upgrading of heavy-water in case of any ' leakages. Most of the tritium in~-

l CANDU reactors appears in the moderator, and adequate design features and

?

procedures are provided to ensure acceptable ~ levels of exposure to plant p.

personnel, ml The CANDU seismic design is based on two earthquake levels:

a design basis 1

1 t

earthquake (DBE) and a lower level of earthquake called the site design

~

earthquake. The site design earthquake (SDE) is assumed to occur in the f

long-term period after a LOCA.

The philosophy of protection against common mode events is centered on the W

two-group separation concept. Structures and systems of the plant are divided W

in two groups (Group 1 and Group 2), and are functionally and spatially separated to the extent feasible.

All process systems belong to Group 1 and

/

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'W>~

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3 w Q x4 y y.,

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. 354TH ACRS MEETING MINUTES.

19 all safety systems belong to Group 2.

Group 2 is designed to the highest -

quality standards.

Mr.- Kugler stated that the spectrum of postulated' initiating events for safety -

analysis consideration will be. studied by means of a-deterministic-accidentL

. analysis that-is similar to the LWRs.. A basic difference is, however, that a

while in the U.S. each initiating event is analyzed in combination with the.

worst s' ingle failure, in Canada the accident analysis also assumes the com-

'bination of the initiating event with the total unavailability of any one of

]

4 the special safety systems.such as reactor shutdown systems or emergency core -

' cooling system.

The~ design-of the. electric power system in.CANDU-3-is also based on.the l:

two-group concept. The Group 1-electric system supplies power to process loads, whereas the Group 2 electric system supplies power to ~ safety loads.

l I~

l

'i U

All CANDU reactors have used highly reliable computers for the control of process systems, such as reactor regulating system, boiler level and pressure-control, pressurizer level, and pressure control. Digital control is also 4

L Lused for the actuation of safety systems, including the two reactor shutdown systems.

3 The Committee members commended the AECL's representatives on their effort and d

g

-for furnishing the. video-tape presentation to the ACRS.

IX.

Executive Sessions (0 pen / Closed) j LA.

Subcomittee Reports.(Closed)

~

m

-1.

Nominating-Comittee/ Election of ACRS Officers (Closed) 1

,.o,<

Mrs.-Mabel Lee was the Design'ted Federal Official for this p

l[ Note:

a portionofthemeeting.]'

i 4

See Official Use Only-Supplement.

]

.B.

Reports. Letters, and Memoranda (0 pen)

I 1.

Proposed Revised policy Statement on the Maintenance of Nuclear Power Plants (Report to Chairman Carr, dated October 12, 1989)

'The Comittee.recomended that the Commission not rush to judgment p'

on-this matter. The Committee expressed its opinion'that it would d

h, be best to spend effort in determining just how serious the problem.

of inadequate. maintenance may be and then what, if any, actions need to be taken. The report contained comments by Dr. Kerr expressing gl' concern that the Comission actions on maintenance may be moving lL, toward an inadequately. defined goal and made suggestions as to how the goals might-be better defined.

p n

.y 5

l 4

3:

y ji 354TH ACRS MEETING' MINUTES.

20 j

k 2.

ACRS Comments on the -Safety Goal Policy and Its Relationship to the h

H Concept of Adequate Protection (Report to Chairman Carr, dated L

October 11, 1989) 1 l

The Comittee stated that in general its pcsition on the Safety Goal l:

Policy and its relationship to the concept of adequate protection I

remains as stated in previous ACRS reports. The Committee stated a

1 that it believes that the-.backfit rule, as just'one part of the

~

general body of regulations, should be regarded as subordinate to y

the Safety Goal Policy. The Comittee stated that it has not yet

~ '

developed a position as to whether the backfit rule is entirely' consistent with this Policy.

j 3.

Proposed Resolution of Generic Issue 135, " Steam Generator and Steam J

^

Line Overfill-Issues" (Letter to James M. Taylor Acting Executive a

p Director for Operations, dated October 11,-1989),

y The Committee' stated that the staff's proposed resolution of this Generic Issues is-incomplete and made comments as to matters that need to be addressed.

q C.

Otherconclusions(0 pen) u

?

1.

~ Future Agenda items a.

The Comittee did not complete its discussion of the NRC

l staff's' proposed resolution of Generic Issue B-56, " Diesel-d Generator Reliability," and agreed to schedule time for addi-

.l tional discussion at the November 16-18, 1989 ACRS meeting.

L The Comittee-asked for written statements of the positions of

.the NRCistaff and NUMARC regarding this matter.

(Mr.

u El-Zeftawy'has the follow-up action on this matter.)

y b.

The Committee did not complete its discussion of the NRC

'l staff's proposed resolution of Generic Issue 87, " Failure of HPCI Steam Line. Without Isolation," and agreed to schedule time' for additional discussion during the November 16-18, 1989 ACRS meeting. (Mr. Igne has the follow-up. action on this matter.)

J c.

The Comittee was briefed by E. Beckjord, RES,.regarding.the.

impact-of the budget reductions imposed by the Congress on the NRC Safety Research Program. The Comittee discussed the budget reductions-that had been imposed on'the NRC.over recent.

years and the manner'in which these reductions had been dis-tributed among the various NRC offices. The Comittee noted i

that the RES has been absorbing major portions of the budget reductions each and every year, and expressed concern about the continually dwindling research program budget.

It decided to discuss with the Comissfon during a future meeting with the Commissioners as to how the Commission sets priorities in

i 74 --w i

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5

.3 354TH ACRS MEETING MINUTES-21 i

allocating resources to various NRC offices. (Mr. Duraiswanty has the follow-up action on this matter.)

Dr. Siess requested that copies of all the correspondence-

'between the Nuclear. Safety Research Review Committee and RES be provided to the ACRS.

E. Beckjord agreed to do so.

(Note:

Copies were obtained from RES and sent to all members by Mr.

Duraiswamy on October 12,1989.).

A d.

The Comittee decided to schedule a discussion during the-3' Easton meeting on the impact that the NRC staff's current regulatory practices might.have on the management of licensees-plants (e.g.,assumptionofmanagement-role)andtheappropri-m ateness of this impact.

(Mr. Quittschreiber has the follow up-a action on this matter.)-

2.

Proposed Rule on Hot Particles 1

^

The Committee agreed that it did not need a briefing on the NRC j

-staff's recent actions on the proposed rule:on hot particles during the November meeting prior to a planned staff briefing of-the-1 Comissioners scheduled for late November-1989.. Mr. Carroll, Chairman of the Occupational and Environmental! Protection-Systems t

Subcomittee, will review the staff's proposed resolution package

'a when it is available and take the appropriate action.- (Mr. Igne has the follow-up action on this matter.)J '

3.

Decision Not to Review Proposed Restart of Calvert Cliffs Pla'nt The Comittee decided not to review the4 proposed restart of the

'Calvert Cliffs plant.

t 4.-

ACRS Bylaws

~

i s

-The Comittee approved proposed changes in:the ACRS Bylaws. dealing with additional coments by members in ACRS' reports, and the proce-y dures under which the-ACRS would review plant restarts.

(See: letter from R. F. Fraley to ACRS Members and ACRS' Staff, dated October 19, 1989.)

]

5.-

ACRS Decision to Review Yankee Rowe Application for a License Extension as'a separate Item-1 The first of the future applications for license extensions is expected to be for the. Yankee Rowe plant.

NRC staff review activi-ties on this plant are also expected.to be carried out in parallel with the development of the staff's generic guidance / requirements for licensees' seeking license extensions. The Comittee agreed to review the Yankee Rowe application for a license extension as a separate item.

[ Note: This-review is the responsibility of the m

.s

'].

lt I y

n 4

354TH'ACRS MEETING MINUTES 22 ACRS' Subcommittee:on Westinghouse and Combustion Engineering Reactor Plants.]~ (Mr. El-Zef tawy has the follow-up action on this matter.)-

D..

FutureActivities(0 pen) 1..

Future' Agenda l

The. Committee agreed to the tentative future agenda shown in Appen-dix II.

2.

Future Subcommittee Activities

]

m.

=

A schedule of future subcommittee meetings was distributed to members (see. Appendix III).

The 354th ACRS Neeting was adjourned at 5:20 p.m. Friday, October 6, 1989.

5 l

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l f

1 TABLE OF CONTENTS APPENDICES MINUTES OF THE 354TH ACRS NEETING.

OCTOBER 5-6, 1989 i

I.

Attendees 11.

Future Agenda 111.. Future Subcommittee Activities IV.

Subject:

Resolution (Proposod) of GI-135, " Steam Generator and Steam Line Overfill Issues," dated August 14, 1989 with Attachments:

DRAFT (undated) memo to J. Taylor from E. Bockjord (INTERNAL COMMITTEF,USE ONLY) and DRAFT NUREG/CR-4893, Technical Findings Report for Generic Insuo 135:

Steam Generator and Steam Lino Overfill Innues (INTERNAL COMMITTEF, USE ONLY).

{.

,o o

..oo 354Til ACRS IV-2 4

MEETING WITl! DIRECTOR, OFFICE OF NUCLEAR REGULATORY RESEARCH, NRC Slides used by speaker du' ring the presentation Table of Contents r

Schedule Table 1,~~RES FY 1990 Budget Chronology (INTERNAL COMMITTEE l

USE)

Table 2, RES Dudget Information for FYs 1989-1991 (INTERNAL COMMITTEE USE)

Identification of License Renewal Activities (INTERNAL COMMITTEE USE)

Advanco copy of memorandum to Chairman Carr from J.

Taylor,

Subject:

FY 1990 APPROPRIATIONS REDUCTION (INTEIUIAL COMMITTEE USE)

NUREG-1325, Disposition of Rocnmmendations of the National Research Council in the Report Revitalizing Nuclear Safety Research", dated Juno 1988.

5 MAINTENANCE OF NUCLEAR POWER PIANTS Slides used b'y speaker during the presentation Table of Contents Schedule Status Report DRAFT

" Maintenance of Nuclera Power Plants; Revised Policy Statement" (IN.TERNAL COMMITTEE USE).

Memorandum to R.

Fraley, ACRS from B. Morris,' Division of Regulatory Applications,

'RES,

Subject:

Revised Policy Statement on Maintenance, dated Sept.

28,

1989, with Attachment -INTERNAL COMMITTEE USE ONLY.

6.1 LIST OF FUTURE SUBCOMMITTEE A'CTIVITIES

~

t 7

GENERIC ISSUE D-56, " DIESEL GENERATOR RELIABILITY" Table of Contents Schedule Statu's Report with Attachments:

I.

DRAFT Proposed REGULATORY. GUIDE 1.9 (TASK RS 802-5)

(IPTERNAL COMMITTEE USE).'

II.

DRAFT NUMARC 87-00, Guidelines and Technical Basis for Numarc Initiatives, Rev.

1, Appendix D, EDG Reliability Program (INTERNAL COMMITTEE USE).

8 GENERIC ISSUE 87, FAILURE OF HPCI STEAM LINE WITHOUT ISOLATION Tabic of Contents Status Report Status Report Memorandum to R. Fraley from R. Houston,

Subject:

Resolution of Generic Issue (GI-87),

"HPCI Steam Line Break Without Isolation,"

dated July 17, 1989 with attached DRAFT MEMORANDUM,

Subject:

" Resolution of GI-87,..............."

(INTERNAL COMMITTEE USE)

I W

. y.! o o*

h 354th ACRS Meeting Minutes IV-3 HANDOUTS Tak 2

ADEQUATE PROTECTION AS IT RELATES TO SAFETY GOALS; ACRS AND STAFF POSITIONS DRAFT SECY on above subject - INTERNAL COMMITTEE USE ONLY.

BRIEFING ON STATUS OF MAINTENANCE INITIATIVEb AND 5.1 SRM

SCHEDULE, 9:00 a.m.,

friday, september 15,

1989, Commissioners' Conference

room, One White Flint

North, Rockville, Maryland (Open to public attendance.)

6 Memorandum to ACRS Members from G.

Quittschreiber, ACRS,

Subject:

Regulatory Policies and Practices Subcommittee Meeting, December 1-2, 1989, Easton, Md.

11 STANDARDIZED PLANT DESIGN (CANDU-3)

Memorandum dated Sept. 13, 1989 to Thomas H.

Cox, NRR from Drew Persinko,

NRR,

Subject:

Summary of July 27, 1989 Presentation by AECL on CANDU-3.

11 Document:

"CANDU 300 Technical Outline," Document PPS 01010-003 published by Atomic Energy of Canada Limited, CANDU Operations, Sheridan Park Research Community, Mississauga, Ontario LSK 2B2.

10.1 Memorandum for ACRS Members from R.

Fraley,

Subject:

ACRS Bylaws - Proposed Revisions, dated October 3, 1989 - INTERNAL COMMITTEE USE ONLY.

a I

ML20059N649

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