Text

Summary of 900712-13 363rd ACRS Meeting.Encls Include List of Attendees & Future Agenda
	ML20058C581
Person / Time
Issue date:	07/12/1990
From:	; Advisory Committee on Reactor Safeguards
To:	; Advisory Committee on Reactor Safeguards
References
	ACRS-2713, NUDOCS 9011020009
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TABLE OF CONTENTS MINUTES OF THE 363RD ACRS MEETING JULY 12-13, 1990 I.

Chairman's Report (Open).................................

1 Systematic Assessment of Licensee Performance (Open).....

3

.II.

III.

EPRI Requirements Document for Advanced LWRs (Open)......

5 IV.

Requirements for Essentially Complete Design (Open)...... 12 V.

Emergency Operating Procedures and PRAs for Shutdown Modes of Reactor Operation (Open)........................

15 VI.

Nuclear Power Plant Operating Experience (Open)..........

20 L

VII.

Fire Damper Reliability (Open)...........................

23 Executive Scssions (Open)................................_24 VIII.

A.

Subcommittee Reports (Open)..........................

24 1.

Thermal-Hydraulic Phenomena.....................

24 2.

ACRS Annual Report to the U.S. Congress.........

25 B.

Other Matters (Open).................................

26 Anticipated Foreign Meetings.........................

26 C.

Summary / List of Follow-Up Matters....................

27 F

D.

Future Activities (Open).............................

30 1.

Future Agenda....................................

30 2.

Future Subcommittee Activities...................

30 y0p

/

9011020009 900712 PDR ACRS 2713 PDC 0

e 11 APPENDICES MINUTES OF THE 363RD ACRS MEETING JULY 12-13, 1990 I.

Attendees II.

Future Agenda III.

Future ACRS Subcommittee Meetings IV.

Contact:

Brenda M. Brush, Executive tJ)ysses spacecraft, including its planned AstronomicalSciences; Notice of Secretary of the Task Force, National launch in October 1990.

Intent to Propero en Environmental Science Foundation, room 546.

Comments on the draft EIS were Impact Aesesement Telephone Number: 202-337-5012:TDD:

previously solicited from Federal, State evasesARY: %e National Radio 357 4 807, tnd local agencies and members of the Astronomy Observatory (NRAO). a Purpose ofMeeting:To hear from two public through notices published in the National Astronomy Center operated by final witnesses and to develop Federal Register; NASA notice on Associated Universities,Inc., under prehminary findings and February 22,1990 (55 FR 6326):

contract with the Nauonal Science recommendations for Foundation action

- Environmental Protection Agency notice Foundation (NSF),is proposing to to catalyze removal of barriers tr

on February 23,1990 (55 FR 6443).

construct a fully steerable.100 meter.

participation in science and engineering Copies of the draft and final statement class radio telescope in Green Bank, carects for persons with disabilities.

have been furnished to the CouncU on Pocahontas County, West Virginia. The Afinutes: May be obtained from the

' Environmental Quality; the Green Bank site for the telescope.

Executive Secretary at the above Environmental Protection Agency; the owned by the NSF,is located in the address.

De artments of Air Force, Commerce, National Radio Quiet Zone which was Agenda Tuesday, July 17: 9 a.m.:

De ense, Energy,llealth and lluman established in 1958 specifically to Presentation by Commissioner Nell C.

Services, and Transportation: the protect the electromagnetic environment Carney of the Department of Education; Nadonal Academy of Sciences: the of the radio telescopes located and to be 10:15 a.m.: presentation by Dr. llarry Office of Management and Budget; to built on the site.

Lang ithe Rochester Institute of appropriate State and local agencies; The construction of the Green Bank Technology; 11:15 a.m.]uly 17 through 3 and to numerous private organizations.

Telescope entalls site preperstion inr a p.m. July 18: members will develop, in Copies of the final statement and final wheel and track, elevation over azimuth, * [188 ",I,QP,','If,'"g"*ghd I

Safety Analysis Report may be configuration on a foundation of a cast.

examined by contacting the Freedom of in. place concrete ring of approximstely b Task Fm's fint W Information Act Office at any of the 50 meters (165 feet) diameter, and the Accommodation 1f you ptan to attend following locations.

assembly and erection of a solid-surfsce ge meeting and require any kind of a) National Aeronautics and Space Mector of a projected diameter of accommod please notify the p

g,

,FY'

~

A ministretion, Washington, DC 20546 approximately 100 meters (330 feet).

(202 453-2939),

Operation, to atart in 199a, will be M. Rebocca Winkler, c

nu os an i enuca to ee commluceMonogementofrim.

(b) NASA, Ames Research Center Moffett Field, CA 94035 (415-694-4190).

n Ba si e (c) NASA, Goddard Space Flight The NSF willprepare an Center, Greenbelt MD 20771 (301-286-

. Environmentallmpact Assessment prior 6255) to the beginning of construction. All NUCLEAR REGULATORY (d) Jet Propulalon Laboratory NASA interested Federal, state, and local

' COMMISSION Resident Office 4800 Oak Grove Drive, agencies and private organizations are /

Pasadena, CA 91109 [816-354-5179).

invited to submit, by July 30.1090, Advisory Committee on Reactor

-(e) NASA, Johnson Space Centerc comments and/or requests for further Safeguards;Revloed Meeting Agenda llouston, TX 77058 (713 483-3071).

Information on the proposed (f) NASA, Kennedy Space Center, construction.

In accordance with the purposes of Kennedy Space Center, FL 32899 (407-ADonESe As40 P0erf 0F C0erfAct sections 29 and 182b. of the Atomic '

~

National Science Foundation,1800 G _

Fmgy Act (42 U.S.C. 2039,2232b), the

'867-2201).

(g) NASA, Langley Research Center' Street, NW., Washington. DC 20550.

Advisory Committee on Reactor Hampton, VA 23665 (804-864-01:0*,

A ttne Dr. julian Shedlovsky, (202/3$7-Safeguards will hold a meeting on July -

(h) NASA. I4wis Research Cente".

9752).

12-14,1990, in Room P-110,7920 Norfolk 21000 Brookpark Road, Cleveland. Oil Deted: June 22,1eo0, Avenue, Bethesda, Maryland. Notice of this meeting was published in the 44135 (216-43&-2902).

Devid A.a-s" Federal Register on May 22,1990 (55'FR '

(1) NASA Marshall Space Flight Aseissant Dimtor.Mothematicolond 21126).This revision incorporates j

Center, Huntsville, AL 35812 (205 -544 Phpica/ Sciences,Nationa/ Science 0031).

Foundation.

additional sessions on Thursday and

Friday, I

(j) NASA, Stennis Space Center.

[FR Doc. 90-15o58 Filed 6-27-00; 8:45 emj Thursday /uly12, JS9a Room P-Jf0, l

' Stennis Space Center MS 39529 (801-salam coom rees.ew 7920 Norfolk Avenue, Bethesda, Md.

66V2164).

B:30 a.m.-8:45 a.m.: Chairman's Additionally, interested porties may cbtain copies of the final EIS from the Meeting Remarks (Open}--The ACRS Chairman will briefly report regarding items of National Technical Information Service Name: Task Force on Persons with current interest.

by calling 703-4874650 and requesting Disabilities.

8:45 a.m.-9:45 a.m. Systematic the document by its title.

Place National Science Foundation, Assessment of Licensee Performance l'

Dated: hne 22 m 1800 G Street, NW., Washington, DC (Open)-Representatives of the NRC 20550.

Uloward Robino, Jr.,

staff and industry will brief the Date July 17 and 18,1990.

Committee and discuss proposed Associate AdrninistratorforManagement, Time / room July 17: 9 a.m.-5 p.m.,

changes in the SALp rocess based on a p

(FR Doc. 90-tso?1 Filed 6-D-oo; 8.45 eml room 540. July 18: 8:30 a.m.-3 p.m room survey of the regulatory impact on plant salasscootrswe w M O.

opetutions, l

'?'520 Federr1 xegistzr / Vol. 55, No.125 / Thursday, lune 2a,1990 / Notices

' 10.m a.m.-12:00 Noon and 7.w p.m.-

brief the Committee on the status of the adjusted by the Chairman as necessary 2.00 p.m. EPR/ Requirements for ongoing work on fire damper reliability.

to facilitate the conduct of the meetingi Advanced Light-waterReactors

!!:30 a.m.-12:15p.m.: Future ACRS persons planning to attend should check (Open)-%e Committee will review and Activities (Open)--The members will with the ACRS Executive Director if

- report on the staffs Safety Evaluation discuss anticipated ACRS subcor nittee such rescheduling would result in major Report regarding Chapters 1-5 of the activities and items proposed fc.

Inconvenience.

EpRI Requirements Document for consideration by the full Committee.

I have determined in accordance with Advanced LWRs. Representatives of the I:15p.n-2:45 p.m.: Preparation of Subsection 10(d) Public Law 92-463 that NRC staff and ERpt will participate as ACRS Reports (Open)-%e Committee it is necessary to close portions of this appropriate, will discuss proposed reports to NRC meetlng noted above to discuss internal 2:15 p.m.-3:15 p.ma Requirements for regarding items considered during this personnel practices of the agency (5 An Essentially Complete Design meeting.

U.S.C. 552b(c)(2)), information the (Open)-Representatives of the NRC 2:45 pm.-J:45 p.m.: A CRS release of which would represent an staff will brief the Committee and Subcommittee Activities (Open)-%e unwarranted invasion of personal discuss the status of the development of Committee will discuss procedures for privacy (5 US.C,552b(c)(6)), and requirements for an essentially complete conduct of ACRS subcommittee and proprietary Information applicable to -

design for evolutionary light. water subgroup meetings.

the matter being discussed (5 U.S.C.

' reactors.

4Mp.m.-5:30 pn.:Genericissue B-552b(c)(4)).

3:15p.m.-4:45 a.m.: Emergency 54 'WmlCenatorReRab##y" Further information regarding topics Operating Pmcedures and Probabilistic (Open)--ne Committee will review and to be discussed. whether the meeting Risk Assessmentfor Shutdown Modes report on the NRC staffs proposed has been cancelled or rescheduled the ofReactor Operation (Open)--

rmluuon of Gericl,ssue B-56," Diesel Chairman's ruling on requests for the

-l Representatives of the NRC staff will Gerator Reliability. Representatives opportunity to resent oral statements of the staff and the NUMARC will

- brief the Committee regarding tl e status and the time al otted can b,' obtained by of emergency operating procedures and, pa,rt a

rop a prepaid telephone call to the.*CRS

,,,f Executive Director, Mr. Raymond 1.

PRAs for shutdown modes of reactor ACRS Reports (Open)-ne Committee o

persuon, will continue discussion of proposed Fraley (telephone 301/492-8049),

4:45 pm.-5:30pm.:ACRS ACRS reports to the NRC, as between 7:45 a.m. and 4:30 p.m.

Subcommittee Activities (Open)-The appropriate.

Deted: June 22,1990..

Committee will hear and discuss reports Saturday, fuly 14,1990 Room P-Jfo, John C. Hoyle, regarding the status of subcommlttce 7920 Norfolk Avenue, Bethesda, Md.

Advisory Committee Management Officer.

activity in designated areas o 8:30 a.m.-11:30 a.m.:Pieparation of

[FR Doc.15o28 Ftled 6-27-em et45 sm) responsibility including thermal

ACRS Reports (Open)-%e Committee hydraulic phenomena and the scope and will complete preparation of ACRS

.i

.(

nature of the ACRS annual report on the reports to the NRC.

i NRC research program.

11:30 a.m.-12 30p.m.: Miscellaneous (Docket Nos. 50-628,50-629, and 50-630]

5:30p.m.-6:25pm. NRCPersonnel (Open)-%e Committee will complete Policies andPractices (Closed)-%e the discussion of items considered Artrona Pubile Servloe Co, et al.,

Committee will discuss the status of during this meeting and related matters.

Faculty Operating Uconee Nos. NPF-proposed NRC personnel action.

procedures for the conduct of and 41, NPF-61, and NPF-74 Palo Verde nis session will be closed to discuss - participation in ACRS meetings were Nucteer Generating Station Receipt of l

Internal personnel practices of the published in the Federal Register on Petition for Director's Decielon Under t

agency and information, the release of September 27,1989 (54 FR 39594). In 10 CFR 2.206 j

.i-which would represent an unwarranted accordance with these procedures, oral

-[-

invasion of personal privacy,' '

or written statements may be presented Notice is hereby given that a petition

. Friday, July JJ.1 spa Room P.-110.

- by members of the public, recordings pursuant to i 2.206 of Title to of the a

. 7s30No.-folk Avenue, Bethesda, Md.

will be permitted only during those

- Code of Federal Regulations (CFR) of I

8:30 a.m.-1RJ0 am. NuclearPower portions of the meeting when a May 22,1990 has been filed with the l

1 Plant Operatig F.rperience (Open/

transcript is being kept, and questions Commission by Mrs.1.inde E. Mitchell i

Closed)--Repren.itatives of the NRC

- may be asked only by members of the (Petitioner). Petitioner states that she is staff will brief the Committee and Committee, its consultants, and staff.

employed by the Arizona Public Service j

j.

discuss recent operating events and Persons desiring to make oral Company (licensee) as an associate incidents including the discovery of flaw statements should notify the ACRS electrical engineer at the Palo Verde indications and cracks in reactor.

Executive Director as far in advance as Nuclear Generating Station (Palo pressure vessel heads and in a primary practicable so that appropriate Verde) Petitioner alleges that serious j

. system pressurizer, malfunctiona of arrangements can be made to allow the violations exist at Palo Verde in the -

molded case circuit breakers, failure of necessary time during the meeting for systems for emergency lighting and fire L

operators to pass requalification exams, such statements. Use of still, motion protection and thatlicensee personnel a proposed change in the frequency of picture and television cameras during acted improperly to ** water down" NRC steam turbine stop valve testing in '

this meeting may be limited to selected inspection findings, suppress other i

Westinghouse nuclear plants, and portions of the meeting as determined serious violations, and discredit an NRC '

miscellaneous items as appropriate.

by the Chairman. Information regarding inspector, in addition. Petitioner alleges Portions of this session will be closed - the time to be set aside for this purpose that NRC Region V agreed to " water as necessary to discuss Proprietary may be obtained by a prepaid telephone down" inspection report findings and

)

Information applicable to these events, call to the ACRS Executive Director. Mr.

retaliated agalnst the NRC inspector in j

l Ja45 a.m.af J:30 a.m.: Fire Damper Rayrnond F. Fraley, prior to the meeting.

question. Petitioner claims that these j

Reliability (Open)--Representatives of -

in view of the possibility that the actions will chill efforts by NRC the NRC staff and of the industry will schedule for ACRS meetings may be inspectors and employees of NRC-

-j l

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'o UNITED STATES P

NUCLEAR REGULATORY CC>MMISSION l

{

r,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS W ASHINGTON, D. C. 206SS o

%,,, g #,4 Jr n 29, 1990 (rev.)

SCHEDULE AND OUTLINE FOR DISCUSSION 363RD ACRS MEETING JULY 12-14, 1990 Thursday. July 12, 1990, Room. P-110. 7920 Norfolk Avenue. Bethesda, Md.

8:45 A.M.

Chairman's Remarks (Open) 1)

8:30 1.1)

Opening Remarks (CM/GRQ) 1.2)

Items of current interest (CM/RFF) 9:45 A.M.

Systematic Assessment of Licensee 2) 8:45 Performance (Open)

TAB 2-----

2.1)

Report by ACRS Subcommittee chairman regarding proposed changes to the SALP program (HWL/GRQ) 2.2)

Meeting with members of the NRC staff and representatives of the nuclear industry as appropriate 10:00 A.M.

BREAK 9:45 l

12:00 Noon EPAT Reauirements Document'for Advanced LWRs 3) 10:00 (open) 3.1)

Report by ACRS Subcommittee Chairman i

TAB 3-----

regarding NRC Staff SER on Chapters f

1-5 of the EPRI Requirements Document for Advanced LWRs (CJW/MME) 3.2)

Meeting with members of NRC staff and EPRI representatives as appropriate 1:00 P.M.

LUNCH

'12:00-2:00 P.M.

Continuation of Discussions on EPRI 3) 1:00 Reauirements Document for Advanced LWRs (Open) 3.3)

Continue-the meeting with NRC staff and EPRI representatives 2:15 P.M.

BREAK 2:00 3:15 P.M.

Reauirements for Essentially Complete l

4)

.2:15 Desian (Open)

TAB 4-----

4 1)

Report by ACRS Subcommittee Chairman regarding status of the development

)

L

=,

l:

2 of requirements for an essentially complete design (CJW/MME) 4.2)

Briefing by members of the NRC staff and representatives of the nuclear industry as appropriate 4:45 P.M.

Emeroency Oooratina Procedures and PRAs for 5) 3:15 Shutdown Modes of Reactor Ooeration (Open) 5.1)

Report by ACRS Subcommittee Chairman regarding status of EOPs and PRAs T AB 5----.

for shutdown modes of reactor opera-tions (JCC/PAB) 5.2)

Briefing by members of the PRC staff and representatives of the nuclear industry as appropriate 5:30 P.M.

Activities of ACRS Subcommittees (Open) 6)

4:45 T AB--------- 6.1)

Report of June 14, 1990 Thermal-Hydraulics Phenomena subcommittee meeting regarding the status of selected research program activities (IC/PAB) 6.2)

Report of activities regarding nature / scope of ACRS Annual Report to the U.S. Congress (IC/SD) 6:15 P.M.

FRC Personnel Action (Closed) 7)

5:30 7.1)

Discuss results of NRC personnel' action and related Committee action (CM/RTF)

(Note:

This session will be closed to discuss internal personnel practices of the agency and information the release of which would repretient an unwarranted invasion.of personal privacy.)

Friday. July 13. 1990. Room. P-110. 7920 Norfolk Avenue. Bethesda, Md2 10:15 A.M.

Nuclear Power Plant Operatina Exoerience 8)-

8:30 8.1)

Hear and discuss reports of power Plant operating experience and TAB 8-----

events noted below: (JCC/PAB) 8.1-1)

Update on the status'of investigation of recent incidents of reactor vessel head cracking (Fitzpatrick and Quad Cities plants) 8.1-2)

Cracking of pressurizer vessel cladding (Haddam Neck plant)

O 8.1-3)

Malfunction of molded case GE circuit breakers 8.1-4)

Failure of operators to pass requalification tests (Brunswick plant) 8.1-5)

NRC review of the Westing-house Owners' Group study justifying support of re-duction of turbine stop valve test frequency 10:15

- 10:30 A.M.

BREAK v;

10:30

- 11:30 A.M.

Fire Damoer Reliability (Open).

9.1)

Report of Mechanical Components Subcommtitee meeting on June 6, 1990 regarding fire damper re-liability (CM/EGI) 9.2)

Briefing by members of the NRC staff and representatives of the nuclear industry as appropriate

10) 11:30

--12:15 P.M.

Future ACRS Activities (Open)

T AB-------- 10.1 Anticipated subcommittee Activi-ties - (RPS/RFF)

T AB-------- 10. 2 )

Items proposed for full Com-mittee consideration (CM/RPS) 1:15 P.M.

LUNCH 12:15 2:30 P.M.

Prenaration of ACRS Rooorts (Open) 11) 1:15 11.1)

EPRI ALWR Requirements Document (CJW/MME) 11.2)

SALP Program (HWL/GRQ)

(Tentative) 12.3)

Requirements for Essentially Complete Design (tentative)

(CJW/MME)

-2:45 P.M.

BREAK 2:30 3:45 P.M.

ACRS Subcommittee Activities (Open) 12) 2:45 12.1)

Discuss proposed memo to OGC regarding conduct of ACRS subcom-mittee and subgroup meetings per FACA (CM/RFF/RPS) 6:00 P.M.

Prenaration of ACRS Recorts (Open) 13) 3:45 13.1)

Continue preparation of reports on items noted above-l

4 i

Saturday, July 14, 1990. Room P-110, 7920 Norfolk Avenue. Bethesda, Md.

14) 8:30 11:30 A.M.

Preparation of ACRS Reoorts (Open) 14.1)

Preparation of ACRS reports as appropriate

15) 11:30 12:30 P.M.

Miscellaneous (Open) 15.1)

Complete discussica of items con-sidered during this meeting and related matters

)

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NY MINUTES OF THE 363RD ACRS MEETING JULY 12-13, 1990 The 363rd meeting of the Advisory Committee on Reactor Safeguards (ACRS) was held at Room P-110, 7920 Norfolk Avenue, Bethesda, Md.,

on July 12-13, 1990.

The purpose of this meeting was to discuss and take appropriate actions on the items listed in the attached agenda.

The entire meeting was open to public attendance with the exception of a portion that dealt with personnel practices and procedures.

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

(Copies of the transcript are available for purchase from Ann Riley & Associates, Ltd., 1612 K Street,-N.W., Washington, D.C.

20006.)

I.

Chairman's Reoort (Open)

(NOTE:

Mr.

R.

F.

Fraley was the Designated Federal Official for

.this portion of the meeting.)

Mr. Michelson, the full Committee Chairman, convened the meeting at 8:30 a.m. with a brief summary of the planned meeting schedule and the provisions under which the discussions were.to be held.

He stated that the Committee had received neither written comments nor requests for time to make oral statements from members of the public.

Items-oLCurrent Interest Mr.

Michelson stated that the following items are of current interest:

o The Commission has recently acted on SECY-90-016,

" Evolutionary Light Water Reactor (LWR) Certification Issues and.Their Relationship to Current Regulatory Requirements."

The Staff Requirements Memorandum (SRM), dated June 26, 1990, related to this matter states that the Commission approved some of the recommendations and disapproved some other 1-recommendations proposed by the NRC staf f in SECY-90-016.

For l

example, 'the Commission did not endorse the core damage frequency value of 10-8/ year

instead, l~

it supports the value of 10~' proposed by the staff; l

/ year for core damage frequency unless the staff can demonstrate that designs with this L

frequency do not meet the NRC safety-goals.

Comments by the l

Commissioners reflect ACRS recommendat' ions included in its

~ report of April 26, 1990 in several areas.

l l

l

~

0'.

h 363rd ACRS Meeting Minutes 2

o The SRM dated June 15, 1990 states that the Commission has approved, with several comments, the staf f's proposed plan for implementation of the NRC Safety Goal Policy. Some preeminent comments include:

The' staff should examine further the guideline for the general plant performance which specifies that the overall mean frequency of a large release of radioactive materials to the environment from a reactor accident 6

should be less than 1 in 10 per year of reactor operation.

The Commission believes that the basic concept of a plant performance objective that focuses on accidental releases from the plant and eliminates site characteristics, as suggested by the

ACRS, is appropriate.

The staff should evaluate and advise the Commission whether such an objective can be developed and how it would be useful.

The staff should routinely consider safety goals in developing. and reviewing regulations and regulatory practices. To achieve this objective, the (staff] should establish a formal mechanism including documentation for ensuring that future regulatory initiatives are evaluated for conformity with the safety goal.

o The Inspector General's report on the conduct of the former NRC-Executive Director for Operations (EDO)

Mr.

Stello regarding the- "Fortuna affair" has been referred to the Department of Justice for possible criminal action.

o The General Accounting Office (GAO), at the request of the Congress, has begun a study of the NRC relationship with-the Institute of Nuclear Power Operations (INPO).

o The recent briefing to the Commission on the status of the operating plants indicates the following:

No new plants identified as " problem plants" Surry Units 1 and 2 have been removed from the list of problem plants.

Therefore, no further NRC special attention is needed beyond the current level of Regional Inspection.

Four operating nuclear units (Calvert Cliffs 1 and 2 and Nine Mile Point l' and 2) remain in the list of problem plants.

363rd ACRS Meeting Minutes 3

Three units (Browns Ferry Units 1, 2, and 3) will remain shutdown until improvements have been made.

o Availability of equipment in a shutdown condition has been identified as an additional Emerging Technical Issue.

o Of the 27 topics identified by the Systematic Evaluation Program (SEP) review of eleven old operating plants, only the following four remain to be resolved:

Turbine missiles I

Containment design and inspection Reactor safety system and engineered safety system electrical isolation Pipe break effects on systems and components.

II.

Systematic Assessment of Licensee Performance (Open)

(NOTE:

Mr. M.

D.

Houston was the Designated Federal Official for

-this portion of the meeting.)

Dr.

Lewis, Chairman of the Regulatory Policies and Practices Subcommittee, provided opening remarks indicating that, as directed by the. Commission, the staff is reevaluating the Systematic Assessment of Licensee Performance (SALP) program.

He stated that the staff presentation would be a status report on its effort and that no Committee action was expected at this time.

Feview of Current SALP Process.- Mr. W. Russell, NRR Mr. Russell, Office of Nuclear Reactor Regulation (NRR), discussed the objectives of the SALP program, the process employed in' conducting a SALP review, and the membership and functioning of the SALP Board.

He stated that the four principal objectives of the SALP program are to:

o Improve licensee performance o

Focus management attention on performance o

Support allocation of NRC_ resources o

Improve the NRC regulatory program.

He indicated that the SALP report was one of the major documents used in the senior management review and assessment.

Mr. Russell discussed the SALP process and provided insights from his involvement in SALP reviews during his time as the Regional Administrator in Region I.

He described some situations where he

363rd ACRS Meeting Minutes 4

personally could not accept the ratings put forth by the SALP Board in its draft report, and indicated how his opinion was derived based on more information than that available to the Board.

In such cases, he would send the draft report back to the Board members for their reconsideration.

At the conclusion of the SALP Board's effort, the Regional Administrator reviewed and issued an initial SALP report to the licensee.

This initial report was then discussed with the licensee at a public meeting.

Dr. Wilkins asked if licensees tend to be constrained in their discussion of this matter at public meetings.

Mr. Russell stated that it was about " fifty-fifty" -- some were very reserved while others openly challenged the ratings.

All licensees provide written responses to the SALP report.

Mr. Russell noted also that the staff provides an opportunity for a mid-cycle SALP review.

The usual SALP cycle is 18 months, but may be shorter for those plants that require more attention.

The mid-cycle review is expected to enhance communication on issues related to performance.

Mr. Carroll asked about the checks and balances built into the

process, mostly in respect to fairness by the Regional Administrator. Mr. Russell discussed the role of senior management and the EDO and indicated that there are sufficient safeguards in the process to overcome individual biases.

PJooosed Chances to SALP-Procram - Mr.

G. Grant, NRR Mr. Grant, NRR, discussed the major proposed changes to the SALP program.

These changes were in the following areas:

o Performance category definitions o

Revised evaluation criteria o

Revised SALP report format o

coordination prior to deviations from the SALP manual chapter o

Suspension of SALP evaluations o

Flexibility in extending the assessment period.

The two key proposed changes involved:

o The. addition -of another performance category U

for unacceptable performance o

A revision of the evaluation criteria from those based on

" responsiveness to the NRC" to those based on a " level of performance."

The proposed definition of Category U includes the following statement:

" Generally, use of a Category U rating will' be confined

363rd ACRS Meeting Minutes 5

-to plants that are or have been in a shutdown conditior., or, if operating, under a performance-based NRC order."

Br.

Lewis questioned the clarity of this definition and discussed various situations where it could conceivably be misinterpreted.

In respect to SALP ratings, Dr. Shewmon asked if there was an attempt by the staff to rate the plants on a mean basis as acceptable, and if the staff had looked at averages of the SALP results.

Mr. Russell stated that there were no quota allocations for the scoring and that averaging of the results for whatever comparison is considered inappropriate.

Dr. Siess asked if there was a correlation between the SALP rating and the time devoted to inspections.

Mr. Russell stated that there might be some indirect correlation between the two but that it would be difficult to draw conclusions since every plant has a given core inspection prog-am.

Mr. Grant concluded his presentation with a brief discussion of the revisions to the SALP report

format, coordination prior to j

deviations from the SALP manual chapters, suspension of SALP evaluations, and flexibility in extending the assessment period.

Dr. Lewis expressed his appreciation for the staff presentation and indicated'that the Committee would most likely discuss this matter again in the near future.

III.

EPRI Recuirements Document for Advanced LWRs (Open)

(NOTE:

Dr. M.

El-Zeftawy was the Designated Federal Official for this portion of'the meeting.)

Mr.

Wylie,-

Chairman o*

the Improved Light Water Reactors-Subcommittee, noted that t.ne purpose of this session is to discuss the staff's draft Safety Evaluation Reports (SERs) for. Chapters 1 through.5 of. the Electric Power Research Institute's (EPRI's)

Advanced Light Water Reactor (ALWR) Requirements ~ Document.

Mr.

Wylie indicated that EPRI, in conjunction with the utility-sponsored ALWR' Steering Committee, prepared a

compendium of technical requirements applicable to ALWR design.

This document represents a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond.

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363rd ACRS Meeting Minutes 6

ERR Presentation - Mr. T. Kenyon, NRR Mr. Kenyon, Project Manager, NRR, stated that the NRR staff has prepared draft safety evaluation reports (DSERs) to discuss its review of the EPRI-ALWR Requirements Document on a chapter-by-chapter basis.

In September 1987, the staff issued the f. cst DSER which addressed the Requirements Document Executive Summary and Chapter 1, "Overall Requirements," of the ALWR program.

Chapter 2,

" Power Generation Systems," was evaluated in the second DSER that was issued in February 1988.

The third DSER, issued in May 1988, covered Chapter 3,

" Reactor Coolant System and Reactor Non-Safety Auxiliary Systems."

The fourth DSER, issued in June 1988, covered Chapter 4,

" Reactor Systems."

The fifth DSER was issued in February 1990.

Key issues evaluated in the fifth DSER include:

o EPRI's ALWR public safety goal o

Severe accident prevention and mitigation o

Severe accident containment performance criteria o

Hydrogen generation and control o

Source term issues o

Fire protection o

High/ low interface design o

ATWS issues o

operation of residual heat removal (RHR) system with reduced reactor coolant system inventory o

-Station blackout o

Core-concrete interaction o

High pressure core-melt ejection o

Equipment survivability o

Inservice testing of pumps and valves o

Resolution of certain generic safety issues.

Mr. Kenyon stated that the Standard Review Plan (SRP) was used as guidance, but the level of detail did not permit a completeness review.

The staff has assumed that all carrent regulatory requirements would be met by a design that complied with the EPRI-ALWR Requirements Document, except where deviations are identified in the document and if the staff identified a

potential incompatability between EPRI-proposed design requirements and current regulatory requirements.

As a-result of the NRC review, a number of items discussed in the DSERs regarding Chapters 1

through 5

remain outstanding.

Currently, there are about 160 open items.

These issues fall into one of four categories:

o Issues that require satisfactory resolution before the staff can complete its review of that particular chapter of the Requirements Document

s 363rd ACRS Meeting Minutes 7

o Issues for which staff review of other related chapters of the Requirements Document has not yet been completed o

confirmatory issues for which the staff will ensure follow-up of commitments in the Requirements Document o

Issues that require satisfactory resolution in support of a vendor / utility-specific application.

Mr. Kenyon stated that EPRI is modifying its Chapter 1 in a rollup document to identify areas of compliance with the Commission's regulatory requirements.

The rc11up document is expected by the end of the' summer of 1990.

The staff is attempting to complete DSERs on Chapters 6, 7,

8, 9,

12, and 13 by the end of the summer 1990.

Schedules for DSERs on Chapters 10, 11, and Appendix A to Chapter 1 will be determined after EPRI has provided additional information-requested by the staff.

The staff is currently requesting two (or one if the ACRS desires) more interim letters from the ACRS on this subject.

Mr. Kenyon noted that the interim ACRS letters are needed to allow incorporation into a rollup-document and to assist the staff in the preparation of a final SER.

The final SER is expected to be completed by late spring of 1991.

ALWR Utility Steerina Committee Presentation - Mr. E. Kintner, ALWR Utility Steering Committee Mr. Kintner, Chairman of the ALWR Utility Steering Committee, briefed the Committee regarding the EPRI-ALWR Requirements Document.

He stated that there is a growing momentum towards the EPRI ALWR program. ~The Nuclear Power Oversight Committee (NPOC) has created a-Subcommittee to coordinate the ALWR program activities.

A six-year and $200+ million program has been initiated - by the Department of Energy (DOE) to fund detailed design development and certification of General Electric (GE) and Westinghouse passive plant designs.

There is growing domestic and international support for EPRI-coordinated design reviews of passive plant designs. EPRI is paying major. attention to ' the NRC efforts for defining appropriata process and priorities for review of standard plant designs.

The goals of the ALWR program for future nuclear power generation are:

o.

Real improvements in safety for both public acceptance and investor confidence o

Stabilized. regulatory basis o

Standardized designs that meet utility requirements

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363rd ACRS Meeting Minutes 8

o Design development and certification schedules that meet urgent utility baseload capacity requirements o

Potential for reduced capital ccsts.

Mr. Kintner indicated that EPRI is concentrating on simplification through the plant emphasizing significant additional engineering margins (e.g.,

thermal margins and negative reactivity requirements).

The EPRI-ALWR Requirements Document is intended to create a sound technical foundation for the next generation of LWRs.

It is & mechanism to resolve issues and provide basis for dialogue and technical honesty in the decision for future plants.

The philosophy of the EPRI-ALWR Requirements Document differs from the past and current approach to safety.

In the past, the approach to safety has been one which is fundamentally to protect the public against accidents once they occur and, therefore, has concerns rated very heavily on accidents after they have occurred. The EPRI program focuses on avoiding accident initiators that present any kind of a threat to the public, and concentrates on designing a more reliable and safer nuclear plant rather than on reducing the consequences of accidents.

EPRI PRESENTATION Overview - Mr. J. Trotter, EPRI Mr.

Trotter,- EPRI, presented an overview of the EPRI-ALWR Requirements Document.

He stated that this document consists of three volumes.

Volume I, Executive Summary, is a management-level synopsis of the Requirements Document that includes:

o Design objectives and philosophy o

overall physical configuration and features of a future nuclear plant design o

Steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant.

Volume II addresses the evolutionary (approximately 1350 MWe) plant requirements that include overall performance and design requirements

-(Chapter 1) and requirements for systems and structures (Chapters-2 through 13).

Volume III addresses the passive (approximately 300-600 MWe) plant ALWR requirements.

The EPRI Requirements Document applies-to the entire nuclear plant and incorporates resolution of generic safety issues and of optimization issues.

The document reflects the industry consensus on principal safety, performance, and design

363rd ACRS Meeting Minutes 9

issues.

Mr. Trotter stated that there has been modest change from the process defined in NUREG-1197, " Advanced Light Water Reactor Program," to update the process for EPRI Requirements Document treatment of issues (the cutoff date was originally July 1,

1986, but was changed to January 1, 3990).

The rollup will proceed with all chapters and is expected by the end of August 1990.

Chapter 5, Source Term and Levere M cJfsent Issues - Mr. D. Leaver, EPRI Mr. Leaver, EPRI, discussed Chapter 5,

Source Term and Severe Accident Issues.

He stated that the EPRI criterion for hydrogen detonation is that containment mixtures of 13 percent hydrogen or less are sufficient to avoid detonation.

Design reflood and depressurization rates bound in-vessel hydrogen generation during recovery or attempted recovery, and thus 75 percent clad oxidation is a conservative bound for in-vessel events.

The reactor coolant system (RCS) depressurization and cavity flooding combine to bound ex-vessel hydrogen generation within the margin afforded by the unrecovered cases.

Rapid debris quenching limits oxidation for credible core debris initial discharge fractions.

Mr. - Leaver noted that using a conditional containment failure probability (CCFP) criterion for the containment performance

'l criteria would be an unnecessary and counterproductive regulatory requirement for ALWRs. A ragged containment is required regardless of calculated core damage frequency.

EPRI believes that a

containment vent for severe accident protection is an unnecessary, undesirable, and potentially unworkable design feature.

The ALWR Requirements Document offers extensive accident prevention features to meet regulatory and investment protection objectives.

Some of the extensive EPRI-ALWR accident prevention features are:

o A significant reduction in transient initiation frequency o

Improved reliability and diversity of on-site AC sources (e.g.,

' third emergency diesel generator for third safety division for BWRs) o Improved decay heat system removal reliability L

o-Higher pressure residual heat removal o

Improved depressurization capability.

l Some of the EPRI-ALWR Requirements Document features for improved mitigation capability rely on:

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363rd ACRS Meeting Minutes 10 o

Preventing direct containment heating Cavity configuration to capture, contain, and cool core debris o

o Cavity flooding capability via direct path from proximate water source.

The EPRI ALWR requirements will meet the NRC safety goal with 3

margin, via existing requirements.

The EPRI public safety criterion is more stringent.

A dose of 25 rem is a low dose, causing no observable health effects.

Mr. Leaver stated that the ALWR will be a fundamentally better plant through the EPRI-ALWR Requirements Document, and there is a strong utility consensus to standardize future plants around the EPRI requirements.

Some of the potential issues are source term and the technical basis for emergency planning zone reduction.

Dr. Catton expressed concern regarding the use of computer codes such as MAAP in the ALWR program, and indicated that the documentation for these codes has been very poor and is not readily available.

Dr. Catton expressed concern that in the NRC staff's review, it does not seem that there is an approved method or study to determine how to handle the hydrogen stratification in containment for the new designs.

Dr. Shewmon stated that additional information is needed regarding the embrittlement in reactor vessel steel and aging problems.

Dr.

Siess requested additional information regarding-EPRI's proposals for dealing with the issues associated with the operating basis and safe shutdown earthquakes issue for the ALWR and asked to be kept informed about any future meetings between the staff and EPRI representatives regarding this matter.

Mr.

Michelson questioned EPRI's compliance.with NUREG-1197 regarding the process, scope, and cutoff date.

Mr. Michelson said that it is not clear if the fire or internal flood events are considered as severe accident issues in the EPRI-ALWR Requirements Document.

Mr.

Michelson noted that additional information is needed to investigate further the environmental qualifications of the advanced control complex for future designs.

Overview of Certain Issues in Chapter 5 - Mr. X. P. Abadie, EPRI Mr.- Abadie,

EPRI, presented an overview of certain issues in Chapter 5 of the EPRI-ALWR Requirements Document.

He stated that w

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4 363rd ACRS Meeting Minutes 11 Revision 1 of Volume II is still in the review process and will be submitted at the end of August 1990.

Systems Overview - PWR The RHR consists of two divisions (each with motor-driven pump and heat exchanger) and circulates water from RCS through heat exchanger to RCS.

The emergency feedwater (EFW) system consists of two divisions, each with two pumps (1 motor-and 1 steam-driven).

The EFW supplies water to steam generators from dedicated supply tanks following loss of main and startup feed pumps.

The safety injection (SI) system consists of two divisions, each with two high-head motor-driven pumps, and delivers water from in-containment refueling water storage tank (IRWST) and accumulators to RCS.

It applies direct vessel injection that eliminates branch lines / valves and reduces required flow.

i The automatic depressurization system (ADS) consists of two divisions, each with two valves in series, vents RCS fluid to IRWST.

The containment spray system-(CSS) consists of two divisions, each with motor-driven pump and heat exchanger.

It supplies water from IRWST to containment spray headers, or back to IRWST for long-term post-LOCA cooling.

The containment is a large dry type with sufficient volume to meet 75%/13% hydrogen limits.

The design pressure is based on LDB events.

Systems Overview - BWR The ADS would include safety / relief valves and permit RCS flooding by DHR pumps.

It is initiated by' low RCS water level and high drywell pressure.

The reactor core isolation coolant (F C) system has one steam turbine-driven high pressure pump, with water supply from dedicated condensate reservoir or suppression pool.

The standby liquid control (SLC) system has two high pressure pumps and two parallel, electrically operated injection valves.

For containment spray, it includes spray of both wetwell and drywell regions and water supply from suppression pool.

The containment is a vapor suppression type.

363rd ACRS Meeting Minutes 12 The hydrogen control system requires ignitor system (preferred by utilities) with inerting system that has been demonstrated in l

previous designs and GE ABWR.

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Ryggarv_of the BWR/DSER Issue 1 - Mr. L.

Fidrych, EPRI l

Mr. Fidrych, EPRI, summarized some of the BWR/DSER issues.

He stated that, to satisfy regulatory requirements (such as the ATWS l

rule) and to invoko performance requirements (such as adequacy of boron mixing), specific requirements b+'e been added to Chapter 5 of the EPRI-ALWR Requirements Document

deal with:

o Automatic standby liquid control o

Effective distribution of boron injection

)

o Safety classification of containmer; spray l

l i

o Main Steam Isolation Valve leak rate j

BWR suppression pool fission product scrubbing.

o Additional requirements were included in Chapters 2, 3,

and 13 to clarify seismic and quality group classification of main steam line to support use of main steam line and condenser for hold-up and plate-out instead of leakage control system.

Mr. Michelson expressed concern regarding the legal standing (if l

any) of the EPRI-ALWR Requirements Document and the etaff's j

commitments regarding this issus.

Mr. Michelson asked if the staff has truly studied the leak-before-break issue for the evolutionary designs.

The Committee decided to continue its discussions of this matter during the August 9-11, 1990 ACRS meeting.

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IV.

Recuirements for Essentially Comolete Desian (Open) l l

[ NOTE:

Dr. M. El-Zeftawy was the Designated Federal Official for this portion of the meeting.)

Mr.

Wylie, Chairman of the Improved Light Water Reactors l

Subcommittee, briefed the Committee regarding the completeness of the design issue for future plants.

He stated that the final l

rule's provision on the scope of completeness of design (Section L

52.47) reflects a policy that certain designs, especially designs

(

that are evolutions of the LWR plants now in operation, should not i

be certified unless they include all details of the plant design that are essential for safe operation of the plant except the site-I specific elements.

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O 363rd ACRS Meeting Minutes 13 NRR PRESENTATION - Mr. M. Virgilio, NRR Mr.

Virgilio, NRR, briefed the Committee regarding the NRC interpretation of the level of detail requind for design certification under 10 CFR Part 52.

He stated that the level of avenil associated with a design certification is analyzed in terms of three factors:

o The scope of an application for design certification o

The material to be developed by the applicant and made available for audit o

The information certified by rulemaking.

The staff has examined four levels of detail, the corresponding degree of standardization achieved, compliance with Part 52, and the safety and economic benefits derived from each.

The degree of standardization resulting from these levels of detail and the certification process will:

Provide identical

physical, functional, and Level 1

performance characteristics of all structures, systems, and components affecting safety except for site-specific characteristics.

Level 2 - Provide physically similar and identical functional and performance characteristics of all structures, systems, and components affecting safety except for site-specific characteristics.

Provide identical functional and performance Level 3

characteristics of all systems, structures, and components except for site-specific characteristics.

Provide at least a

product-line type of Level 4

standardization.

Mr. Virgilio stated that Level 1 is probably not commercially

feasible, because the level of detail required would make it difficult to ensure continued availability of components with all the certified attributes over the life of certification.

Level 2 provides the maximum degree of standardization while avoiding Level I concerns.

Level 3 characterizes the industry proposal.

Level 4 would not constitute an acceptable application under 10 CFR Part 52, because it is not sufficient to allow the staff to reach its final conclusion on all safety issues in a one-step process.

The staff's licensing review of an application for design certification for all levels will deviate from the traditional

1 363rd ACRS Meeting Minutes 14 r';.. ice, with the addition of inspection, tests, analyses, and

?ptance criteria (ITAAC).

The staff believes that ITAAC will i

side reasonable assurance that a plant which references the sign is built and will operate in accordance with the design certification.

The staff used the heating, ventilating, and air conditioning (HVAC) system as an example to show how much detail would be expected for each of the four levels.

The staff estimates that $800 million to $1.09 billion are required to develop a complete design.

This estimate refers to Architect Engineering and utility engineering costs (design and design implementation) to the point of fuel load, not including site and QA/QC engineering.

Mr. Virgilio noted that the staff is seeking advice from the ACRS and guidance from the Commission regarding the level of detail to be required in an application for design certification and subsequent rule certifying the design under 10 CFR Part 52.

The staff has just completed its SECY paper (SECY-90-241) to the Commission on this subject.

Mr. Michelson expressed concern regarding which decament the NRC will be reviewing and certifying for the future designs; for example, is'it the Standard Safety Anelysis Report (SSAR) only or does it include the references also?

Mr. Michelson asked why the NRC staff does not have a comparison with the contents of the EPRI-ALWR Requirements Document in dealing with the completeness of the design issue, and how the EPRI-ALWR Requirements Document folds into the certification process.

Mr. Ward questioned the criteria and level of PRA that are proposed to be performed for the future designs.

DR)Le Power Company Presentation - Mr. Rehn, Duke Power & Light Company Mr. Rehn, Duke Power & Light Company, outlined the recent work that was performed by the industry to implement the design certification process under 10 CFR 52, for example:

o Design information will be analogous to that in a Final Safety Analysis Report, minus as-built construction, as-procured (nameplate) details that have been submitted sometimes, and site-specific details.

o A detailed delineation of ITAAC, with appropriate cross-references to the SSAR/FSAR, will be provided.

o Results of the PRA will also be included in the application.

a 363rd ACRS Meeting Minutes 15 Mr. Rehn stated that the practical workability or flexibility will be considered in the following:

o As-built deviations o

Startup, operating, and maintenance problems o

Obsolescence o

Equipment improvements.

Some of the ITAAC objectives are to retain as much as possible the existing NRC regulatory inspection and review processes, and to incorporate the NRC " Sign-as-You-Go" approval process pioneered by NRC Region II and Georgia Power and Light in the Vogtle Readiness Review.

Validation attributes will typically be used to verify physical plant assumptions or inputs used in preapproved analyses that demonstrate conformance with an acceptance criterion.

Mr. Michelson stated that it is not clear how to incorporate the PRA as part of the application for certification and what the regulatory mechanism would be if the PRA numbers have to be changed after certification.

Mr. Wylie questioned the ca.aracterizations of system interf aces and the process to be used in the certification.

Dr. Catton reiterated his previous concern that in the staff's proposal regarding the four levels of detail, it does not appear that there is an approved method or study to determine how to handle the hydrogen stratification in the containment for the new designs and how to certify it.

NUMARC_ Presentation - Mr. Rowden, NUMARC Lawyers Committee Mr.

Rowden, Chairman of the Nuclear Management and Resources Council (NUMARC) Idawyers Committee, described the "two-tiered" approach proposed by industry for design certification.

He stated that the two-tier structure is simply a means for formatting and documenting in the design certification (DC) rule, the certified and noncertified parts of the design, and specifying the change mechanisms governing each in accordance with 10 CFR Part 52 requirements.

The first tier would contain:

o A description of the certified design based on SSAR section 1.2, with details comparable to that in current SERs o

The full array of inspections, tests, analyses and acceptance criteria that 10 CFR Part 52 requires.

363rd ACRS Meeting Minutes 16 i

The second tier wouldt o

Reference the entire SSAR design description.

By referencing the SSAR in the DC rule, the NRC would documcnt the features and commitments that were the basis for NRC approval (beyond those certified in the first tier) and document the

" matters... resolved in connection with the issuance...of a j

design certification," (per Section $2.63 (a) (4).

The second tier would be associated with the rule certifying the design (but Dg1 be part of the certification itself) and would include a change process like the current 10 CFR 50.59 that would allow changes without prior NRC review as long as no unreviewed safety question is presented.

Mr. Michelson questioned the treatment of open items after the final design approval (FDA) is granted.

Mr. Michelson noted that there is a difference that needs to be l

clarified between the EPRI numbers (60 70%)

and the NUMARC numbers (~ 33%) regarding the level of design effort completion at the certification submittals.

The Committee deferred its action on this matter pending additional review / discussion by the Improved Light Water Reactors Subcommittee on August 8, 1990.

The full Committee will continue its discussion of this subject during its August 9-11, 1990 meeting.

V.

Emeraency Operatina Procedures and PRAs for Shutdown Modes of Reactor Operation (Open)

(NOTE:

Mr.

P.

Boehnert was the Designated Federal Official for this portion of the meeting.)

Mr. Carroll, Chairman, Plant Operations Subcommittee, noted that the ACRS has, over the paFt year, become concerned with the issues of the use PRAs and emergency procedure guidelines (EPGs) to quantify and mitigate, respectively, plant risk during shutdown operations.

During this session, the staff will brief the Committee regarding its efforts to deal with this issue.

HRR Presentation - Mr. W. Ritssell, NRR Mr. Russell, NRR, noted that the staff has initiated a program entitled " Shutdown Risk."

He noted the issues involved, the basis for concern, significant technical issues, and the staff's action plan to deal with this matter.

In response to a question from Mr. Carroll, Mr. Russell said the program's focus will be on events that occur while fuel

e 363rd ACRS Meeting Minutes 17 is in-vessel.

Mr. Russell said that NRR is interested in ACRS feedback on their program approach.

Mr. Russell stated that numerous events involving loss of DHR and/or loss of AC power, loss of inventory, and inadvertent criticality precursors have convinced the staf f that additional study of shutdown risk is necessary.

In addition, U.S. and foreign PRA results show that the risk during shutdown is a significant contributar (20-30%) to total core damage frequency.

Mid-loop L

operatien is deemed a dominant shutdown-risk contributor.

Foreign operating experience has shown that errors committed during shutdown can be evidenced during startup as a potential accident procursor.

While the focus of the staff program will be on operations during Modes 3-5/6, they will examine also the risk potential for startup/ shutdown events.

Mr.

Russell described the current regulatory requirements controlling shutdown risk.

NRR has concluded that a comprehensive re-evaluation of shutdown risk is needed, because the above requirements were developed on a piecemeal basis.

Mr. Russell stated that the Shutdown-Risk Program is a broad-scope program, covering all aspects of shutdown risk.

The duration of 1.5 years, and NRC will coordinate this the program will be effort with the industry.

Program review by the ACRS/ Committee to Review Generic Requirements (CRGR) / Commission will be sought, as appropriate.

Mr. Russell said continued plant operation during the conduct of the program is acceptable, on a judgmental basis, because of the reduced decay heat source term, and stored energy seen during shutdown.

He said that larger plant syctems margins are also available.

The major elements of the staff's action plan include:

o Evaluation of operating experience, current PRAs, foreign activity, as well as plant instrumentation, proceduros, and

training, o

Investigation of outage

planning, management, and configuration control.

o Performance of'PRA case studies (Grand Gulf & Surry).

o Analysis of risk management schemes versus technical specification controls, o

Revision of regulatory requirements as necessary.

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363rd ACRS Mceting Minutes 18 In response to questions from Mr. Michelson, Dr. Sheron, Office of Nuclear Regulatory Research (RES), agreed to investigate the risk potential from the dropping of a heavy load on the core.

Dr. Wilkins urged NRC to cw; sider the implications of the dynamics associated with the reduction of decay heat versus risk potential.

NRR staff said they would do so.

Mr. Carroll asked if the industry Owners Groups (OG) are expanding their EPGs to cover shutdown operations.

Mr. Russell stated that the NRC is not aware of any such OG activity.

Mr. Minnick urged prompt interaction with the industry on this program.

Mr.

Russell said they plan to meet with industry representatives next month.

Mr. Carroll asked about the status of the NRR response to the ACRS concern regarding the ability of licensees to be able to close containment on an efficacious basis if RHR is lost during mid-loop operations. Mr. Russell said that during September 1989 NRR issued a temporary instruction to the Regional Offices that should result in obtaining the information desired.

However, the responses are not due to NRR until two years after issuance of the temporary instruction.

Mr. Russell stated also that the issue of containment integrity will be addressed as part of the Shutdown-Risk Program.

BEA_Erjapantation - Dr.

B.

Sheron Dr. Sheron, RES, discussed the RES supporting studies related to shutdown risk.

He stated that RES has been concerned for some time about the risk posed during shutdown operations; PRA results also indicate that this risk is significant.

Referring to the PRA results shown, Dr.

Lewis said that -the uncertainties associated with the core damage frequencies (CDFs) cited render these values meaningless.

He stated that the decision to take action on this issue should not hinge on these PRA estimates of CDP.

The scope of the study will include:

o Evaluation of all operating modes, other than full power o

Analysis of internal events -- including internal fires and l

floods and refueling events (up to tne point the fuel leaves

(

the reactor) o Human factors analyses.

t

363rd ACRS Mocting Minutes 19 Dr. Lewis questioned the usefulness of the human factors studies, again given the uncertainties involved.

Dr. Sheron stated that they may have to finesse this point by use of EPGs/EOPs to reduce the uncertainties associated with potential operator actions.

Dr.

Sheron stated that RES will perform Level I PRA studies, utilizing the HUREG-1150 results from the Grand Gulf and Surry plants.

Sandia and Brookhaven Laboratories will be the respective contractors. Containment evaluations and consequence calculations will be performed as necessary.

A Senior Consulting Group (SCG) will be used to oversee this effort.

Dr, Lewis requested a list of the SCG members.

Dr.

Sheron stated that the intent of the study is to obtain insights regarding potential accident initiators and scenarios as soon as possible.

To this end, a two-phase approach is being used.

Phase I will consist of a coarse screening analysis to identify i

potentially significant initiators.

Phase II will include a detailed analysis of the critical operating modes and accident scenarios.

Completion of the Phase I activities is targeted for early-to-late 1991, with the Phase II results due about one year later.

Dr. Lewis expressed doubts regarding the ability of the SCG, by itself, to be able to ensure a quality PRA product is provided, given that the intended user of the PRA is not experienced in its employ.

Dr. Sheron indicated that the PRA effort will no doubt evolve as time passes.

He suggested that RES meet with the cognizant ACRS subcommittee assigned to this issue within the next 2-3 months to discuss the details of the PRA effort.

MVEAEtc Comment _a - Mr.

W.

Rasin, NUMARC Mr. Rasin, NUMARC, provided comments on the NRC _ Shutdown-Risk Program.

He stated that NUMARC feels it is a mistake to apply PRA to the resolution of this issue, the reasons being that it will take too long, and that the plants are in the process of doing their Individual Plant Examinations (IPEs).

Given the deadline on the IPE effort (3 years), and the uncertainties associated with the human error analyses, use of PRA is impractical.

In response to a question from Mr. Carroll, Mr. Rasin stated that the appropriate industry action at this point would be to evaluate the effectiveness of the licensees' implementation of the NRC's current regulatory requirements (Generic Letters, Orders, etc.).

Mr. Carroll said the ACRS will follow closely the progress of the Shutdown-Risk Program.

Arrangements will be made to hold a meeting

363rd ACRS Meeting Minutes 20 of the ACRS Plant Operations Subcommittee with RES to discuss the PRA studies.

VI.

Nuclear Power Plant Operatina Experience (Open)

(NOTE:

Mr.

P.

Boehnert was the Designated Federal Official for this portion of the meeting.)

Mr.

Carroll, Chairman of the Plant Operations Subcommittee, introduced the discussion of this

matter, stating that the following nuclear power plant operating events would be presented by NRR representativest Stat 9s of Vessel Head Crackino Incidents and Pressuriser Claddina Cracking - Mr.

B.

Elliot, NRR

]

Mr. Elliot, NRR, noted the following points regarding the subject topics:

o The cracking seen on the reactor vessel head of the Quad Cities plant was caused by intergranular stress corrosion cracking (IGSCC).

Because of a mismatch between the flange and head closure portions of the head, a weld overlay was made.

Grinding of the stainless steel overlay used during this operation induced residual stress.

This, combined with 1

the very high oxygen content of the water in the vessel head (14-17 ppm vs a normal value of 5ppk), resulted in the IGSCC.

In response to a question from Mr. Carroll, Mr. Elliot said the cracks penetrating ir.to the base metal (maximum penetration is - 0.250 inchesi may slowly propagate somewhat, but the licensee will conduct periodic inspections to track this situat).on, o

Initial exanination of the vessel head of the Fitzpatrick plant showed indications of both surf ace and subsurface flaws.

The subsurface flhw was initially thought to extend 50%

through-wall.

Subsequent investigation has shown that the subsurface is a metallic inclusion or segregate resulting from the fabrication process.

In response to a question from Dr. Shewmon, Mr. Elliot said

.the inclusion is probably composed of manganese sulfite.

1 Fracture analysis of this inclusion shows the head is acceptable for use.

The inclusion is such that its orientation and location result in very low stress intensity l

factors; i.e.,

there is little chance of crack propagation.

[

363rd ACRS Meeting Minutes 21 Mr. Michelson suggested that NRR examine the fabrication records for the vessel head to see if a volumetric inspection had been performed.

Mr. Elliot indicated he would do so.

o During a

visual examination of the internals of the pressurizer of the Haddam Neck plant, cracks were discovered in the stainless steel wall cladding.

The cracks are located in two general areas of the lower internals: from the second

[-

support plate to the heater elements, and around the retaining L

basket of the diffuser plate.

Some indications around the mid-wall area were determined to be metallic inclusions.

The deepest penetration into the base metal was - 0.035 inches 2.5-inch-long crack.

NRR believes that the most from a probable cause of the cracking is thermal fatigue induced by relatively cold water from the RCS.

The situation is L

aggravated by the phenomenon of thermal stratification of the water in the surge line, allowing a slug of quite cold water to enter the pressurizer.

A fracture mechanics evaluation, conducted in accordance with Section XI of the ASME Code, indicates that the cracks do not need to be removed.

In response to concerns expressed by Drs. Wilkins and Shewmon, NRR staff said they will require the licensee to inspect periodically the pressurizer over its remaining life to ensure its integrity.

Messrs.

Carroll, Minnick, and Dr.

Shewmon asked if NRR requires periodic inspection of the BWR vessel internals. Hr.

Elliot stated that visual inspection is performed.

As a result of further discussion, NRR was requested to provide a presentation to the ACRS on this issue when they have formulated a position or the scope and extent of inspection y

required here vis-a-vis Life Extension.

r=

Egilute of Operators to Pass Recualification Examination at Brunswick Units 1 and 1 - Mr.

D. Lang, hRR Mr. Lang, NRR, stated that 22 of 47 operators and 7 of 8 crews failed either the requalification examination or first operating evaluation.

The cause was determined to be inadequate training.

Details noted include:

o on a requalification exam, 9 of 12 senior reactor operators (SROs) and 5 of 8 reactor operators (ROs) failed; 3 of 4 crews also failed.

An evaluation of the remaining operators were conducted shortly thereafter to determine if the other crews were up to snuf f and to gauge the extent of the problem.

Eight of 27 operators and 4 of 4 crews failed this test.

E M

363rd ACRS Meeting Minutes 22 o

The licensee (Carolina Power & Light - CP6L) shut down both units on May 20, 1990, and NRC issued a Confirmation of Action letter on May 21, 1990.

o Additional operating evaluations were conducted on June 9-10, 1990, and a sufficient number of operators / crews passed to support a 4-shift rotation.

The licensee was given permission (by the cognizant Regional Administrator) to restart the units on June 10, 1990.

The CP&L submitted a summary root cause analysis on June 30, 1990; this report is undergoing staff review.

In response to committee questions, the following points were noted:

o The method and content of the examinations had changed between the last two examinations; however, the licensee was well aware of the changes madis.

About 10 percent of all licensees' training programs have been r

found deficient.

For a program to be considered acceptable, 75 percent of examinees must pass.

o The deficient training program at Brunswick had been accredited by INPO.

The Committee requested copies of the licensee's summary root cause report noted above, and the last SALP report that addresses evaluation of the licensee's training program.

GE Molded Case Circuit Breakett - Mr. S. Alexander, NRR Mr. Alexander stated that the Oyster Creek licensee purchased 170 GE molded case circuit breakers (MCCBs).

Pre-installation testing r$ owed that 5 of 7 breakers were defective (failed to trip on demand);

specifically, the Phase C thermal overcurrent trip function was disabled by a misorientation of calibration screw spring clips.

The breakers had been orderSd for use on safety-grado equipment.

The MCCBs were manuf actured at a plant in Puerto Rico; under-voltage relays (UVRs) were installed at a GE plant in Knoxville.

The UVRs were tested, but the MCCBs were not retested after the UVRs were installed.

GE subsequently dedicated these MCCBs, but relied on the testing conducted at the Puerto Rico plant to do so.

Root causes of this problem cited by Mr. Alexander were defective manufacturing procedures, training, and QC.

In addition, the Knoxville plant lacked f acilities/ procedures for post-installation retest of the MCCBs normal functions.

363rd ACRS Meeting Minutes 23 GE has taken corrective actions to:

Institute post-installation testing at Knoxville o

o Correct / prevent QC deficiencies o

Correct interference problem with the breaker o

Revise the dedication program o

Evaluate this incident for a 10 CFR Part 21 report.

The NRC staff has issued an Information Notice (90-13) to alert the licensees about the problems associated with the MCCDs.

Inspections of the vendor's facilities have been conducted, and more are planned.

As a result of Committee questioning, NRR committed to inform the Committee regarding the intended safety-grade use of the MCCBs at the Oyster Creek plant.

Owing to lack of time, the briefing on the results of the NRC staff's review of the Westinghouse Owners Group study justifying support of reduction of turbine stop valve testing frequency has been deferred to the August 9-11, 1990 ACRS meeting.

(NOTE:

This item has been subsequently deferred to the September 6-8, 1990 ACRS meeting.)

VII.

Fire Damoer Reliability (Open) - Mr.

C. McCracken, NRR (NOTE:

Mr.

E.

Igne was the Designated Federal Official for this portion of the meeting.)

Mr. McCracken, NRR, discussed briefly the inoperability of fire dampers in nuclear power plant air ventilation systems.

Before his presentation, two curtain-type fire dampers of different sizes typically used in nuclear power plants were shown and their actuating mechanisms demonstrated to the Committee.

Mr. McCracken stated that tests performed to dete a.ne the ability of the curtain-type fire dampers to fully close under operational air flow conditions revealed that these dampers failed to close under such conditions.

The design, fabrication, and testing of fire dampers have been done in accordance with the provisions of the UL Standard 555, " Fire Dampers and Ceiling Dampers."

This Standard assumes that ventilation systems are shut down automatically when fire occurs, and hence does not require testing of the operability of fire dampers under operational air flow conditions.

However, several nuclear plants maintain air flow in the system while fire dampers isolate local fire areas.

The staff issued Information Notice 89-52, " Potential Fire Damper Operations Problems," during June 1989, alerting all licensees

a e

363rd ACRS Meeting Minutes 24 about the failure of curtain-type fire dampets to fully close under ventilation system operational air flow conditions and suggesting that:

These dampers be tested under " worst-case" air flow conditions Ventilation system be shut down administrative 1y during a fire situation.

In response to a question by Mr. Michelson, Mr. McCracken stated that the dampers are fire-rated for 3

hours and are also seismically qualified.

In response to Mr. Carroll's question as to what an essentially complete design with respect to licensing of standard plant implies, Mr. McCracken stated it means that you have basically everything but the nameplate; it will cost the applicants a lot more money to reach that level of design before they are certified, but he feels comfortable with this definition.

VIII.

Executive Sessions (Open)

A.

Subcommittee Reparig (Open) 1.

Thermal Hydraulic Phenomena

[ NOTE:

Mr.

P.

Boehnert was the Designated Federal Official for this portion of the meeting.]

Dr. Catton, Chairman of the Thermal Hydraulic Phenomenon Subcommittee, reported on the matters discussed during the June 14, 1990 meeting.

The Subcommittee discussed the status of the 2D/3D research

program, and calculational tools for accident management.

Key points noted by Dr. Catton include:

o The history of the 2D/3D program was noted.

This cooperative program (U.S.,

FRG, and Japan) ran a number of experiments on large-break LOCA issues in three test facilities.

The total cost was about

$500 million. Mr. P. Damerell, MPR Assoc., has dono a good job of summarizing and intarpreting the test data for RES.

However, program funding ends in FY-1991; data may be lost, lacking careful analyses.

This data deserves to be fully preserved.

The Subcommittee plans a meeting in the fall of 1990 to discuss the issue of archival of this data.

o A presentation was given on the Engineering Analyzer developed by Brookhaven National Laboratory (BNL)

363rd ACRS Meeting Minutes 25 1

by Dr. Wulf f.

The BNL uses this Analyzer to perform realistic transient analyses for BWR plants at about i

ten-times faster than real-time.

The code used to support the Analyzer (HIPA), is of high quality.

l Most recently, the Analyzer has been used in support of the BWR stability issue, and has provided results in a timely and comprehensive manner.

The larger thermal hydraulic codes (e.g., TRAC) have not been able to do as well.

o Dr. Fabic, AEOD, discussed the Engineering Plant Analyzer (EPA) he has developed for NRC use.

The code used (RETRACT) was originally doveloped by Dr.

Fabic and a colleague when both were employed by the

]

Singer-Link Co.

Unfortunately, too little information was presented at too rapid a pace.

Dr.

Fabic has promised to provide detailed documentation on RETRACT in the near future.

Dr. Catton said he will reserve judgment on the code's capability until j

he reviews the above material.

o RES discussed its program on development of accident i

management capability.

This program emphasizes dealing with accidents that go beyond initial core damage, and makin' maximum use of existing plant equipment.

Rogarding the analysis requirements needed to devise accident management strategies, RES noted that:

no codes are required to identify accident management strategies, but codes will be required to evaluate selected strategies.

RES also believes that accident management actions can be devised that are independent of the phenomenological uncertainties that exist in the codes to be used.

Regarding Dr. Fabic's presentation, Dr. Wilkins noted that Dr. Fabic extended an invitation to the ACRS members to observe his code "la action" on a PC located in Dr.

Fabic's office.

2.

Annual ACRS ReDort to the Conaress (Open)

(NOTE:

Mr.

S.

Duraiswamy was the - Designated Federal Official for this portion of the meeting.

Dr.

Catton, Chairman of the Safety Research Program Subcommittee, stated that.

as instructed by the Committee, he met with Mr.

Rathbun, Office of Congressional Affairs, on July 12, 1990 to discuss the possibility of meeting with selected staff members of the Congressional NRC Oversight Committees under the guidance

363rd ACRS Meeting Minutes 26 of Mr. Rathbun to obtain their views on the ACRS intent to provide a more comprehensive report to the Congress on 'he NRC Safety Research Program than that which has be6.4 provided during the past three years.

Mr.

Rathbun agreed to consider making necessary arrangements for Dr. Catton to meet with selected members of the Congressional oversight Committees.

He suggested

~

that the ACRS discuss this matter with Chairman Carr or the Commission prior to proceeding further.

D.

Qther Matters (Open) o Anticipated Forelan Meetinas The Committee is presently considering participation in two foreign meetings during 1991:

Meetina with Japanerp Reoresentatives - (Hitachi,

Toshiba, Tokyo

Electric, MITTI, and Japanese Advisory Committee)

The Committee proposed to meet with the Japanese Representatives in Japan during the Spring of 1991.

The Committee agreed to limit the scope of this meeting to the discussion of issues related to ABWRs and Westinghouse SP-90 standard plant design.

Second Ouadricartite Meetina of Advisory Committag.g (Germany, France, Japan, and possibly USSR)

In response to the invitation by the Federal Republic of Germany, the Committee proposed to participate in the Second Quadripartite Meeting of Advisory Committees which is tentatively scheduled to be held during September 1991.

The location of this meeting has not yet been chosen, but is likely to be in France.

The scope of this meeting would be to discuss reactor safety mattet, of a generic nature.

The Committee informed Chairman Carr about this matter through the Summary Report of the 363rd ACRS meeting dated July 31, 1990.

In the event a choice must be made whether to participate in the meeting with the Japanese representatives or.in the Second Quadripartite Meeting of Advisory Committees, the Committee decided to give priority to the Second nuadripartite Meeting of Advisory Committees.

-.i----.--.-.------.--

363rd ACRS Meeting Minutes 27 C.

Summarv/ List of Follow-Un Matters o

Dr.

Catton informed the Committee that Mr. Rathbun, Director of the Office of Congressional

Affairs, suggested that the ACRS discuss with Chairman Carr its intent to expand the scope and content of the annual ACRS report to the Congress on the NRC Safety Research Program prior to discussing this matter with selected Congressional staf f members.

(NOTE: Dr. Catton met with Chairman Carr on July 31, 1990 to discuss this matter.)

(Mr. Fraley and Mr. Duraiswamy have the follow-up action on this matter.)

o The Committee suggested that the AC/DC Power Systems Reliability Subcommittee discuss the proposed resolution of Generic Isuue B-56, " Diesel Generator Reliability,"

and the status of the resolution of the differences between NUMARC and the NRC staff on this matter prior to referring this item to the full Committee for consideration.

(NOTE:

A meeting of the AC/DC Power Systems Reliability Subcommittee was held on August 8, 1990 to discuss this matter.)

(Dr. El-Zeftawy has the follow-up action on this matter.)

o The Committee suggested that the Improved LWRs Subcommittee discuss SECY-90-241,

" Level of Detail Required for Design Certification Under Part 52," 4.sted July 11, 1990, and develop information for use by the full Committee in its review of this matter. The members suggested that the ACRS staff invite NUMARC to participate in this meeting.

(NOTE:

A meeting of the Improved LWRs Subcommittee was held on August 8,1990 to discuss this matter.)

(Dr. El-Zcftawy has the follow-up action on this matter.)

The Committee assigned the following issues, that may be o

appropriate for consideration in the ELWR certification process, to cognizant Subcommittees and-suggested that these Subcommittees discuss these issues and recommend a

course-of action for consideration by the full Committee:

Systems Interactions (Safety Philosophy, Technology, and Criteric - Mr. Ward)

Quality Assurance Requirements for

Design, Construction, and Operation of ELWRs (Quality and Quality Assurance - Dr. Siess) 1

363rd ACRS Meeting Minutes 28 (Mr. Houston and Mr. Igne have the follow-up action on this matter.)

o Mr. Michelson suggested that the ACRS staff try to find out the status of translation (from Japanese to English) of the MITTI report on ABWRs.

(Dr. Savio has the follow-up action on this matter.)

o Mr. Michelson, the ACRS Chairman, delegated the authority for " coordination of the selection and appointment of consultants" to the ACRS Vice Chairman Mr. Wylie.

(Mr.

Fraley and Mrs. Lee have the follow-up action on this matter.)

o The Committee agreed to invite Mr. Fitzgerald, OGC, to the August 9-11, 1990 ACRS meeting to discuss the OGC's interpretation of the Federal Advisory Committee Act requirements with regard to conducting ACRS Subcommittee /Subgrcup meetings.

(NOTE:

This item has been subsequently deferred to the September 6-8, 1990 ACRS meeting to accommodate the availability of Mr.

Fitzgerald.)

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

o Mr. Ward, Chairman of the Human Factors Subcommittee, stated that he plans to review the public comments and the staff's resolution of these comments related to 10 CFR Part 55, " Fitness for Duty Requirements for Licensed Operators," prior to making a decision whether to' hold a Subcommittee meeting to discuss this-matter or refer it to - the full Committee for consideration.

(Mr.

Alderman has the follow-up action on this matter.)

o

-The Committee agreed to continue its discussion of Chapters 1-5 of the EPRI-ALWR Requirements Document-and the associated staf.f's Safety Evaluation Reports, and a proposed report to the Commission during the August 9-11, 1990 ACRS meeting.

(Dr. El-Zeftawy has the follow-up action on this matter.)

l o

Dr. Siess suggested that Cognizant Subcommittee Chairmen review the appropriateness of the priority rankings l

proposed by the staff for various Generic Issues and i

provide comments in writing to Sam Duraiswamy by August 20, 1990.

(Mr. Duraiswamy has the responsibility to coordinate this task.)

i l'

l

l e

363rd ACRS Meeting Minutes 29 o

Dr.

Lewis, Chairman of the Regulatory policies and Practices Subcommittee, recommended that the Committee discuss matters related to Systematic Assessment of Licensee Performance in the near future.

(Mr.

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

o During the discussion of the Emergency operating Procedures and PRAs for Shutdown Modes of Reactor Operation, the following requests were made.

(Mr.

Boehnert has the follow-up action on this matter):

Dr. Lewis requested a list of the members of the Senior Consulting Group established by RES to provide advice to them in performing PRAs to evaluate shutdown risk. Mr. Cunningham, RES, agreed to provide this list.

Mr. Michelson asked whether RES plans to evaluate the risk from dropping a heavy load on the core in the Shutdown-Risk Study.

Dr. Sheron, RES, agreed to investigate this matter in the Shutdown-Risk Study.

Mr. Carroll requested that the ACRS be provided with a copy of the NRR Temporary Instruction Manual issued for use by the personnel at the NRC Regional Offices in obtaining information on the specifics of the containment closure requirements while a plant is in a shutdown mode.

(Copies of this manual have been received by Mr. Boehnert and will be distributod to all members.)

o Dr. Catton stated that as suggested by the Committee, the Thermal Hydraulic Phenomena Subcommittee reviewed the Engineering Plant Analyzer developed by AEOD during its June 14, 1990 meeting and decided that no full Committee action is necessary on this item. The Subcommittee plans to discuss this matter further in the future.

(Mr.

Boehnert has the follow-up action on this matter.).

o During the discussion of Reactor Operating Experience, the following requests and commitments were made.

(Mr.

Boehnert has the follow-up action on this matter):

The Committee requested a briefing from the NRR staff on their proposed plans to address inspection of BWR reactor vessel internals vis-a-vis the concern for the potential loss of structural integrity as a result of cracking.

This briefing is to be scheduled when NRR has formulated its plans l

i

b-363rd ACRS Meeting Minutes 30 to address this issue as part of its life extension program.

Mr. Elliot, NRR, agreed to provide this briefing.

Dr. Catton requested that the Committee be provided with copies of the Brunswick plant licensee's report that includes a root cause analysis of the failure of a large number of reactor operators to pass the requalification examination.

Mr.

Becker, NRR, agreed to provide copies of this report.

Mr. Michelson requested a copy of the last SALP report for the Brunswick plant, in particular the portion of the report that addresses the review of the operator training program.

Mr. Becker, NRR agreed to provide a copy of this report.

Mr. Alexander, NRR, committed to provide information i

to the Committee un the intended safety-related use l

of the molded case GE circuit breakers at the Oyster Creek plant.

D.

Future Activities (Open) 1.

Future Acenda The Committee agreed on a tentative schedule for the next Committee meeting (Appendix II).

2.

Future Subcommittee Activities A list of future subcommittee meetings was distributed to the members (Appendix III).

The meeting was adjourned at 4:50 p.m. July 13, 1990.

i

i t

g P

.g'

.e i

f I

APPENDICES MINUTES OF THE 363RD ACRS MEETING JULY 12-13, 1990 F

i I.

Attendees II.

Future Agenda III.

Future ACRS Subcommittee Meetings IV.

Subject:

Coherence in the Regulatory Process.

Presentation materials provided during the weeting.

e 3-EPRI REQUIREMENTS DOCUMENT FOR ADVANCED LWRs e

Schedule Status Report with Attachments:

e Attachment I:

DRAFT Safety Evaluation Report on Chapter 5

.of EPRI-ALWR Requirements Document (INTERNAL COMMITTEE USE ONLY).

Attachment II:

ACRS Report to Chairman. Carr

+,

from F. Remick dated September 12, 1989,

Subject:

Electric Power Research Institute Advanced Light Water Reactor Requirements Document.

Attachment III:

Staff Requirements Memorandum from S.

Chilk, Secretary,;for J.

Taylor-and C,

-Michelson,.

dated December 15, 1989, Re:

SECY-89-334, Recommended PriorG tes for Review of' Standi Plant Des'<Jns.

i Presentation-materials providt.". during the meeting.

e 4

.'4 BRIEFING ON REOUIREMENTS FOR ESSENTIALLY COMPLETE DESIGN e

Schedule Status Report with Attachments:

e.

- Attachment I:

10 CFR Part 52, Early Site Permits;.

Standard Design Certifications; and Combined. Licenses for Neclear Power

. Reactors, dated April 28, 1989

- Attachment II:

ACRS Report dated August.. 12,-1986, Re:

ACRS Comments'on Proposed NRC Standardization Policy Statement o

L

\\.

c I

~

1 te oye; b

i' T363rd:ACRS.M r, ting MinutOc Appendix IV-2

- Attachment'III:

ACRS Report dated October 15, 1986, Re:

ACRS Comments on Draft NUREG-1225, " Implementation of NRC Policy on Nuclear Power Plant d

Standardization" Attachment IV:

ACRS Report dated June 7, 1988, Re.

NRC Proposed Rule on Early Site

Permits, Standard Design Certification, and Combined Licenses for Nuclear Power Plants

- Attachment V:

ACRS Report dated January 19, 1989, Re:

Draft Final Rule on Standardization and Licensing Reform, 10 CFR Part 52

- Attachment VI:

ACMS Report dated February 15, 1989,

~

Re:

Final Rule on Standardization and Licensing Reforms, 10 CFR Part 52

- Attachment VII:

ACRS Report dated November 24, 1989, Re:

Module 1 of the Draft Safety Evaluation Report for the Advanced Boiling Water Reactcr Design

- Attachment VIII:. Remarks by Commissioner J. Curtiss, j

at-the'1990 Nuclear Power Assembly, j

'3' Washington, D.C., May 22, 1990 i

. Attachment IX:

SRM M900427 from S. Chilk, Secretary, o

for J. Taylor, Execucive Director for Operations,.Re:xxx

.]

.l 5:

EMERGENCY OPERATING PROCEDURES AND PRAs F_QR SHUTDOWN MODES OF l

BEACTOR OPERATION e

Schedule Status Report with Attachments:

l-o

-ACRS Letter on the resolution of Generic Issue 99, j

L L

" Improved Reliability of RHR Capability in PWRs,"

-dated February 16, 1989

)

P. Boehnert's 6/20/90 memorandum reperting on 6/8/90' l

)

Commission meeting:

Briefing on rrport of NRC IIT e

L investigation of the Vogtle loss-c.f-power event'of l

March-'20, 1990 P.

Boehnert's 6/8/50 memorandum that ' provides a a

(

summary of the NRC IIT report (NUREG-1410) on the

~

ul 3/20/90 Vogtle. incident l

6.1 REPORT OF THERMAL / HYDRAULIC (T/H)

PHENOMENA SUBCOMMITTEE.

MEETING' j

Status: Report with attached WORKING COPY-of-Minutes of o

,m' i the.. June 14, 1990 meeting of the Subcommittee on T/H Phenomena (INTERNAL COMMITTEE USE ONLY).

c.-

h

~..

ge i

363rd ACRS' Matting Minutos IV-3 8

BRIEFING ON RECENT EVENTS AT OPERATING REACTORS e

Schedule Status Report with Attachments:

o Excerpt from proposed Minutes of the May 1990 ACRS Meeting detailing the NRR presentation on vessel head cracking at the Fitzpatrick and Quad Cities plants Excerpt from P. Boehnert's 5/23/90 memo reporting on NRR weekly briefing of events at Haddam Neck and Brunswick plants Excerpt from P.

Boehner,'s 5/18/90 memorandun reporting on NRR weekly briefing of problems with GE molded case circuit breakers NRR Safety Evaluation approving use of H methodology to support relaxation of turbine valve test frequency e'

Presentation materials provided dr. ring the meeting.

i.

L 10.1 LIST OF' SCHEDULED ACRS 'COMMITTEls AliL SUBCOMMITTEE MEETIN_QE L

10.2 (Handout) FUTURE ACRS ACTIVITIES - 364TH ACRS MEETING -

AUGUST 9-11, 1990 HANDOUTS 1.2 EVOLUTIONARY LIGHT WATER REACTOR (LWR) CERTIFICATION ISSUES' AND'THEIR RELATIONSHIP TO CURRENT REGULATORY-REOUIREMENTS e

SECY-90-016,-dated Jan. 12, 1990, same abject.

J e

ACRS Report dated April -- 26, 1990 to Chairman Carr, O

Subject:

Evolutionary Light Water Reactor Certification Issues 'and Their Relationship to Current Regulatory Requirements.

o!

Memorandum dated April 27, 1990 from J. Taylor, EDO for Commissioners,

Subject:

Staff

Response

to ACRS l~

Conclusions Regarding Evolutionary Light Water Reactor Certification Issues with enclosure A, Evolutic7 cry Light Water Reactor Certification Issues and Their "Slationship to Current.. Regulatory Requirements - [stafs. s evaluation of ACRS recommendations).

Staff Requirements Memorandum dated June 26, 1990 for J.

h e

Taylor, EDOlfrom-S. Chilk, SECY,

Subject:

SECY-90 Evolutionary Light Water Reactor ~(LWR)

Certification Issues and Their Relationships to Current Regulatory-Requirements.

9.1" FIRE DAMPER REJ,IABILITY o

Schedule-Status Report _

Summary / Minutes of the Mechanical components Subcommittee e

Meeting on June'6, 1990.

Presentation materials provided during the meeting.

3

J

.,. 4

. ~.

. 363rd ACRS.Mssting Minut0s Appandix IV-4 HANDOUTS Outstanding Issues - ALWR Utility Requirements Document, pp.

1-7.

NRC INFORMATION NOTICE NO. 90.43 - Mechanical Interference with Thermal Trip Function in GE Molded-Case Circuit Breakers Staff Requirements Memorandum dated June 15, 1990, for J.

Taylor from S.

Chilk, SECY,

Subject:

SECY-89-102, Implementation of the Safety Goals.

SECY-90-241, ; Level of Detail Required for Design Certification

. Under Part 52, 7/11/90 Memorandum to ACRS Chairman nnd Members dated 10 July 3990 from Ray y

Fraley,

Subject:

Arrival of Summer Interns.

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)

1J

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