ML20154M653

Summary of ACRS 306th Meeting on 851010-12 in Washington,Dc Re TVA Organizational Changes,Source Term for Nuclear Power Plant Accidents,Emergency Planning,Advanced Reactors & Annual ACRS Rept on NRC Safety Research Program

ML20154M653
Person / Time
Issue date:	10/10/1985
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2365, NUDOCS 8603140389
Download: ML20154M653 (213)

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TABLE OF CONTENTS '

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MINUTES OF THE . _ d U wd a!,,;lgdl 306TH ACRS MEETING --

OCTOBER 10-12, 1985 WASHINGTON, D.C.

I. Chairman's Report (0 pen) ................................. 1 II. TVA Organizational Changes (0 pen) ........................ 2 III. Davis-Besse Nuclear Power Plant (0 pen) ................... 6 IV. Source Term for Nuclear Power Plant Accidents (0 pen) .... 12 V. Emergency Pl anning (0 pen) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 VI. Meeting with NRC Commissioners (0 pen) ................... 25 VII. General Electric Standard Safety Analysis Report (GESSAR II) (0 pen) ...................................... 28 VIII. Advanced Reactors (0 pen) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 IX. Annual ACRS Report on the NRC Safety Research Program (0 pen) .......................................... 30 X. Executive Sessions (0 pen) ............................... 31 A. Subcomi ttee Ass ignments . . . . . . . . . . . . . . . . . . . . . . . . . . . 31

1. TVA Organizational Changes .................... 31

2. State of Nuclear Power Plant Safety ........... 31

3. Davis-Besse Nuclear Plant ..................... 31

4. Report of the Procedures and Administration Subcommittee Meeting en July 30, 1985 ......... 32

8. Reports, Letters and Memoranda ..................... 33

1. Additional Comments on the EPA Standards for a High Level Radioactive Waste Repository ..... 33

2. Advanced Reactor Policy Statement ............. 34

3. Impacts of Natural Phenomena on Off-Site Emergency Response ............................ 34

4. k nsideration of Earthquakes in Off-Site Emergency Planning ................... 34 E !CNA Z3 CEICEML Cortinos ry ~ -

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I TABLE OF CONTENTS (Cont.)

MINUTES OF THE 306TH ACRS MEETING

5. ACRS Action on Proposed Regulatory Guides ...... 34 C. Generic Issues ...................................... 34

1. Pressurized Thermal Shock ...................... 34

2. Natural Ability Selection of Reactor Operators . 35

3. Seismic Margins ................................ 35 4 Source Term .................................... 35 D. Future Schedule ..................................... 35

1. Future Agenda .................................. 35

2. Future Subcommittee Activities ................. 35 E. D. L. Basdekas Letter to Congress ................. 35 F. ACRS Meeting Dates for CY-1986 ...................... 36 G. Indian Point Special Proceeding . . . . . . . . . . . . . . . . . . . . . 36 H. INPO Visits to Nuclear Power Plants ................. 36 11

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TABLE OF CONTENTS APPENDICES TO MINUTES OF THE 306TH ACRS MEETING OCTOBER 10-12, 1985 Appendix I - Attendees.......................................... A-1 Appendix II - Future Agenda..................................... A-7 Appendix III - Schedule of ACRS Subcomittee Meetings........... A-9 Appendix IV - Tennessce Valley Authority Presentation to ACRS.............................................. A-41 Appendix V - NRC Activities Regarding TVA Management............ A-50 Appendix VI - Toledo Edison Presentation to ACRS................ A-51 Appendix VII - Staff Presentation on Davis-Besse................ A-92 Appendix VIII - Staff Introduction and Intended Application of NUREG-0956....................... A-103 Appendix IX - NRC Presentation - Status of Implementation of Severe Accident Policy......................... A-104 Appendix X - Principal Features of New Source Term Analytical Procedure............................... A-124 Appendi x XI - Source Tenn Science. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-133 Appendix XII - Emergency Planning and Severe Low Frequency Natural Phenomena..................... A-139 Appendix XIII - Severe Weather and External Events (Natural) in Current Emergency Plans........... A-145 Appendix XIV - Probabilistic Estimates of Exceeding Seismic Design Leve1s........................... A-154 Appendix XV - SECY 84-320...................................... A-159 Appendix XVI - Referenceability Terms.......................... A-168 iii

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.r Fed;ral R: gist:r / Vel. 50. Nr. tes / Friday, September 27.1965/ N:tices~ 3g189 be re! eased to other government encono sounca catseonse:

sgencies, which often receive proposals Bethesda, Maryland, from 8:15 a.m. to O. from the same PrincipalInvestigator/ Information is taken from materials submitted by the appucants.

5:00 p.m. Copies of comments received NC "' #* may be examined at the NRC Public w .c w onstonma systuu c'THe azaurran Act: enoes camase saevissous w7 ed "

sernsymo. Accesamo, neramma. an,o ~

sesposmo or naconos we Tha systsas:

None. Regulatory guides are available for Dated: September 23.1985. Inspecuen at the Commission's Public stoaaos: .

Document Room.1717 H Street NW.,

Herman G. Fleming.

Information is stored in an automated NSFPrivacy Act Officer. Washington DC. Copies of active guides data base on disk and magnetic tapes. may be purchased at the current

'(FR Doc. 85-2379 Filed s.as.as; E45 am) Covernment Printing Office price. A metamvastury: a u sso caos m w subscription service for future des in Information can be retrieved by name specific divisions is avallable ugh usht or Social Security Number of app!! cant. the Government Printing OfBce.

NUCt. EAR REGULATORY ' Information on the subscription service lon sareouanos: COMMISSION and current prices may be obtained by

• Authorised Users: Employees who Regulatory Guides;leeuence and writing to the Superintendent of maintain records in this system are Avsflability Documents. U.S. Covernment Printing

, instructed to grant access to other Office. Post OfBee Dox 37082 d .

employees on a need.to.know basis as The Nuclear Regulatory Commission WasMngton, E 20013-7082.

1 specifically authorized by the Privacy has issued revisions to three guides in (5 USC. 552(all j Act Officer. lts Regulatory Guide Series. This series Dated at Sdver Sprins. Maryland this 23rd

= Physica/Sc has been developed to describe and day of September tasa.

employs secur/egucedst Building ity guards. Building is make available to the public methods

- acceptable to the NRC staff of For the Nuclear Regulatory t'aa"=61on.

locked dunng non. business hours when Robert 3. Minogue.

, guard is net on duty. Room tn which implementing specific parts of the d pirector,o '

, records are kept is locked during non. Commission's regulations and,in some it, search. ffic,,fy ci,arJtgulaeory

? business hours. cases, to delineate techniques used by the staffit evaluating specific problems (TR Doc. es.231M Filed e-as.as; E45 am)

Precedura/Sofeguards: Access to ausse caos ruse.ew cc: puter files is controlled by the use of or postulated accidents and to provide "

l passwords. Access to source data files guidance to applicants concerning ~"

is strictly contro!!ed by program staff. certain of the information needed by th Advtoory Comndttee on Reactor S..*;fA

. staffin its review of a RETINTioN AMo otsposAu permits and licenses. pplications for Safeguards, Meeting Agende

"%iA' Computer files are cumu!ative and Regulatory Guide 1.84. Revision 23 In accordance with the purpose of

" Design and Fabrication Code Case section 29 and,182b. of the Atomic

, maintained indefinitely.

Acceptsbility. ASME Section ID.

Energy Act (42 U.S.C. 2039,2232b). the systm mA=4 canal Aho AccMSs: Division 1." and Regulatory Guide 1.85. Advisory Committee on Reactor (Computer files). Chief. Revision 23. Materials Code Case Safeguards will hold a muting on A& inistrathe Systems Branch. Office Acceptabilit . ASME Section DI. .

! Division 1." at those code cases that October 10-12.1965. In Room 1046.1717 I ofInformation Systems. National H Street. NW. Washington DC.Netice are generally acceptable to the NRC of this meeting was published in the

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Science Foundation, Washington. D.C.

20550. ataff for implementation in the licensing Federal Register on August 21.1965.

I -

of light-water. cooled nuclear power The agenda for the subject meeting l momcamm enocawnss: plants. Revision 4 to Regulatory Guide will be as follows:

. 1.147. " Inservice Inspection Code Case

To determine if a record exists write Acceptability. ASME Section XI. nursdaI' October 14,1385

) to: NSF Privacy Act Officer. National -

' l Science Foundation. Washing *on. D.C.' Division 1."IJsts those code cases that &#AR-&d5ARtR*PortofACAS 20550. are generally acceptable to the NRC Choirman (Open)--The ACRS Chairman i

i staff for implementation in the laservice will report briefly regarding items of The National Science Foundation inspection of light. water. cooled nuclear reserves the right to require sufficient current interest to the Committee.

power plants. These three guides are identification to positively identify the periodically revised to update the Ad5AR-fac0AXrAdvanced individual making the request. Complete ustings of acceptable code casesReactors and to (Open)-Re members wi!!

discuss proposed ACRS comments to procedures are found at 45 CFR Part 613.

include the results of public comment the NRC regarding the proposed -

naconos access enocaounes:

• and additional staff review. Commission policy statement on '

1 Comments and suggestions in See "NoGflcation Procedures" above. advanceci reactors. Members of the NRC connection with (1) items for inclusion Staff will Requester should reasonsbly specify the in guides currently being developed or participate, as appropriate.

record contents being sought. JaVOAR-1&JOARr Topicefor (21 improvements in all published guides Meeting with NRC Commissionerir cowissTwo naconos emoceounes: are encouraged at any time. Written (Open)-The members will discuss l comments may be submitted to the -

i See " Notification Procedures" above. ACRS recommendations to the NRC Requester should reasonably identify Rules and Procedures Branch. DRR. .regarding EPA standards for high level i

the record and specify the information ADM. U.S. Nuclear Regulatory waste disposal Commission. Washington, DC 20555.

I to be contested and state the correcu.e Comments may also be delivered to , 22J0AR-JJ MARrMeeting with ac% se :;;hra .d the reasons (cr the NRC Commissioners (Open)-ne Room 4000. Maryland National Bank correction, with supporting justification. members will meet with the NRC '

Building. 7735 Old Georgetown Road.

Commissioners to discuss ACRS .

3S195~ Federal Registee / Vcl. So, N .188 / Prid:y, September 27, 1985 / N:tices =

c;mments on the EPA standards for high proposid ACRS comments regarding ACRS me: tings m y be cdjusted by tb

' tvel waste disposal. application of the pRA for these units to Chairman as necessary to facilitate the 21:30 A.M-1245 P.M TVA this nuclear station. conduct of the meeting. persens

.)rganiutionalChanges (Open}- 5::0 PM-6-30 PN.: Nuclear Pdwer planning to attend should r+"k w'th the Representatives of the NRC Staff will Plant Sofery-Reloted issues (Opes)- ACRS Executive Director if su'ch brief the hittee regard.ng proposed The members will discuss a pmposed rescheduling would result in major ch:cges in the TVA organization to ACRS report on the state of nuclear inconvenience.

correct deficiencies in the constructica power safety. Members of the NRC Staff t have determined in accordance with -

cud operation of TVA plants, will participate, as appropriata. subsection 10(d) Pub. L 92-463 that it is 1:45 PJf.-J:45PNa Doris.Besse necessary to close portions of this Saturday, October 12, toes NuclearPowerPlant (Open)-%e meeting as noted above to discuna mimbers will hear and discuss a report 8 30 AN.-fr30PNa Prepomt/on of Proprietary Information [5 U.S.C.

by representatives of the NRC Staff ACRS Reports (Open/C!osed)-The 552b(c)(4)]. and detailed secunty regarding the results of the start.up test members of the Committee will discuss information [5 U.S.C. 552b(c)ta)l.

program for Unit 1 of this nuclear proposed reports to the NRC regarding Further information regarding topics station. Representatives of the licensee items considered during this meeting. In to be discussed. whether the meeting .

will participate, as appropriate. addition, proposed ACRS comments has been cancelled or rescheduled, the 3:45 PR.-5:45 PR.:Cenern/ Electric regarding the use of natural abdity Chairman's ruling on ra;aests for the StaydctdSofety Acolysis Report testmg of nuclear power plant operators opportunity to present oral statements (Gu AR fil P; en/ Closed}-%a will also be considered. and the time allotted can be obtained by Committe e - J discuss the proposed J:J0 PN.-2:30 PNa Quantitatm. a aid telephone call to the ACRS ACRS report regarding the FDA for thie Sofety Cools (Open)-The members will Exe utive Director.Mr.Raymond F.

type nuclear statica. hear and discuss the repwt ofits Portions of this session will be closed subcommittee on proposed quantitative

• Fraley (telephone 202/634-3265)'EDT.

between 8.15 A.M. and 5.00 P.M as needed to discuss Propnetary safety goals for nuclear power plants.

i 130 PX-,22 PJfa ACRS Procedures Deted: September D 1986.

Information applicable to this plaos and detailed secunty arrangements for this and Prec: ices (Open)-The Committee John C. Hoyle, type of fac:lity, will hear and discuss the report of its Advisory Comimtree Management OTTeer 5.45 PN.-6:15 PNa Regulotory subcommittee on prcposed changes in [FR Doc. SS-DM6 Fileh- 6 45, a 45 am)

Activities (Open}-The mernbers will ACRS procedures and practices based acoa,7 % .

j on meetmgs ofits subcommittee and the hear and discuss the reports ofits subcommittee on proposed changes to report of its Panel on ACRS

- NRC Regulatory Cuides,includmg Effectweness. Advisory Committee on Reactor eteorological measurement programs: JW PN.-440 PN.: ACRS Annual Safeguards; ComMned Site Evaluation iteria for power. Instrumentation, and Repert to t3e US. Congress on the and Extreme Extemal Pttenomena;

• control portions of safety systems and Preposed NRCSofety Research Meeting instrurnent setpoints for safety.celated Program and Budger (Open)-%e members will discuss the proposed The ACRS Subcommittees on Site

- systems, Evaluation and Extreme External se:m and content of its annual reports Friday. October 11.1963 to the U.S. Congresa on proposed NRC Phenomena will hold a combined safety research programs and budgets. meeting on October 9.1985. Room 104&

- RJO AN.-!W AR.: Source Terms - 1717 H Street. NW. Washington. DC.

forNuclect PowerPlant Accidents Procedures for the conduct of and participationin ACRS meetings were %e entire meeting will be open to

[Open)-The members will hear the report of its subcommsttee and discuss pubhshed in the Federal Registae os public attendance.

October 3.1984 (49 FR 193). In The agenda for the subject meeting proposed NUREC oG58. Reassessment of Source Tum. Members of the NRC accordance with these procedures. oral shall be as follows:

or wntten statements may be presented Wednesday. October 9.1935-3:30 Staff will participate as appropriate,

!!.30 AM-1:30 PM: Emer;ercy by members of the public. recordings o.m. unni the conclusion of business.

Planning (Open)--The Commattee will will be permitted only durtcg those The Subcommittees will:(1) Evaluate.

- hear and discuss the :eport of its portions of the meeting when a from a probabilistic approach the subconmuttee on consideration of transcript is being kept, and questions relative Importance of various natural extreme e.xternal phenomena in may be asked only by members of the phenomena. and priontize them in terms emergency planning. Representatives of Committee.its consultants and Staff. of their potentialimpact on offsite the NRC Staff and invited experts will Persons desiring to make oral emergency planning. considering the statements should notify the ACRS likehhood that such phenomena might patiopate, as apprepnate. cause an accident that would require the 2J0 PM-*J0 PNa CeneralElectric Executive Director as far in advance as Standard Sa' sty Analysis Report practicable so that appropnate implementation of offsite emergency (CESSAR D)(Open/C!osed)-The arrangements can be made to allow the plans. and (2) review the proposed final members wdl continue discussion of the necessary time durmg the meeting for amendments to 10 CFR Part Part $0.

Committee's report to the NRC regard.ng such statements. Use of sull, motaon Appendix E. Consideration of cn FDA foe this type of nuclear plant. picture and television cameras during Ear hquakes in Emergency Planning.

Portions of this session wdl be closed this meetics may be limited to selected SECY-85-283, dated August 21,1985.

cs nem==y to discusa Proprietary portions of the meeting as determined and develep an ACRS consensus on this Information appheable to this matter by the Chairman. Information regarding issue for the Commission.

nd details of secunty arrangements int the time to be set aside for this purpose Oral statements may be presented by g

's type of plant .

may be obtained by a prepaid telephone members of the public with the 4:30 PR.-5.30 PR. Zodian Point call to the ACRS Executive Director. R. concurrence us tne sum.mmmu..

( duclear Generstug S4acoa f/mts 2 and F. Fraley prior es the meeung. In view of Chairmam written statements will be 3 (Open)-The mesobers wdl diamma the possibility that the schedule foe . accepted and made available to the ,

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n UNITED STATES NUCLEAR REGULATORY COMMISSION B- i/ . ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g ,,, g

& WASHINGTON, D. C. 20555 Revised: October 2, 1985 SCHEDULE AND OUTLINE FOR DISCUSSION 306TH ACRS MEETING OCTOBER 10-12, 1985 WASHINGTON, D. C.

Thursday, October 10, 1985, Room 1046, 1717 H Street, NW, Washington, D.C.

1) 8:30 A.M. - 8:45 A.M. Report of ACRS Chainnan (0 pen) 1.1) Opening Statement (DAW) 1.2) Itemsofcurrentinterest(DAW /RFF)

2) 8:45 A.M. - 10:00 A.M. Advanced Reactors (0 pen) 2.1) Report of ACRS Subcomittee regarding proposed NRC Policy Statement on Advanced Reactors (JCM/MWC/MME)

3) 10:00 A.M. - 10:30 A.M. Items for Meeting with NRC Comissioners (0 pen) 3.1) Discuss proposed ACRS coments regarding EPA Standards for Hi Level Waste Disposal (DWM/OSM)gh

4) 10:30 A.M. - 11:30 A.M. Meeting with NRC Comissioners (0 pen) 4.1) Discuss topics noted above

5) 11:30 A.M. - 12:45 P.M. TVA Organizational Changes (0 pen) 5.1) Briefing by representatives of the NRC Staff regarding proposed changes in the TVA nuclear organization (AJC) 12:45 P.M. - 1:45 P.M. LUNCH

6) 1:45 P.M. - 3:45 P.M. Davis-Besse Nuclear Power Plant (0 pen) 6.1) Report o,f ACRS Subcomittee regarding restart of Unit 1 following loss of main and auxiliary feedwater (FJR/HA)'

6.2) Meeting with representatives of the NRC Staff and the Licensee, as appropriate

l , s, 306th ACRS Meeting Agenda 7) 3:45 P.M. - 5:45 P.M. General Electric Standard Safety Analysis Report (GE55AR II) (0 pen / Closed) 7.1) Discuss proposed ACRS report regarding the request for an FDA for this type of nuclear steam supply system (00/RKM)

(Note: Portions of this session will be closed as required to discuss Proprietary Information for this facility and detailed security arrangements for this type plant.)

8) 5:45 P.M. - 6:15 P.M. Regulatory Activities (0 pen) 8.1) Report of ACR FSubcommittee (CPS /SD) regarding proposed changes in NRC Regulatory Guides regarding:

8.1-1) Meteorological Measurement Programs for Nuclear Power Plants (R.G. 1.23. Rev. 1) 8.1-2) Criteria for Power, Instru-mentation, and Control Portions of Safety Systems (Task No. IC 609-5) 8.1-3) Instrument Setpoints for Safety-Related Systems (R.G. 1.105, Rev. 2)

9) 6:15 P.M. - 6:45 P.M. FutureActivities(0 pen) 9.1) Discuss anticipated ACRS Subcomittee activities (MWL) 9.2) Proposed activities for ACRS consideration (RFF)

O e

gs 306th ACRS Meeting Agenda Friday, October 11, 1985, Room 1046, 1717 H Street, NW, Washington, D.C.

10) 8:30 A.M. - 11:30 A.M. Source Tenn for Nuclear Power Plant

' Accidents (NUREG-0956. Reassessment of Source Term) (0 pen) 10.1) Report of ACRS Subcommittee regarding NUREG-0956 (WK/MDH) 10.2) Meeting with representatives of the NRC Staff and the nuclear industry, as appropriate

11) 11:30 A.M. - 1:30 P.M. Emergency Plan anin (0 pen) 11.1) Report of ACRS Subcommittee on consideration of extreme environ-mental phenomena in emergency planning (DWM/OSM) 11.2) Meeting with representatives of the NRC Staff, the nuclear in-dustry, and invited experts, as appropriate 1:30 P.M. - 2:30 P.M. LUNCH

12) 2:30 P.M. - 4:30 P.M. General Electric Standard Safety Analysis Report (GE55AR II) (0 pen / Closed) 12.1) Discuss proposed ACRS report to the NRC regarding the request for an FDA for this type of nuclear steamsupplysystem(D0/RKM)

(Note: Portions of this session will be closed as necessary to discuss Proprietary Material applicable to this type facility and detailed security arrangements for this type of nuclear steam supply system.)

13) 4:30 P.M. - 5:30 P.M. State of Nuclear Power Plant Safety (0 pen) 13.1) Report of ACR5 Subcommittee regard-ing the state of nuclear power plant safety (WK/AJC)

14) 5:30 P.M. - 6:30 P.M. Probabilistic Risk Assessment (0 pen) 14.1) Discuss application of the PRA for the Indian Point Nuclear Generating Station, Units 2 and 3 to the Indian Point facility (00/RPS) 3

306th ACRS Meeting Agenda Saturday, October 12,1985, Room 1046,1717 H Street, NW, Washington, D.C.

15) 8:30 A.M. - 11:30 P.M. Preparation of ACRS Reports (0 pen)

, 15.1) Discuss proposed ACRS reports regarding:

15.1-1) Policy Statement on Ad-vanced Reactors (JCM/P9fE) 15.1-2) NUREG-0956, Reassessment of Source Tem (WK/MDH) 15.1-3) Emergency Planning (DWM/OSM) 15.1-4) Davis-Besse Nuclear Station (tentative) (FJR/HA)

16) 11:30 P.M. - 12:30 P.M. Annual ACRS Report on the NRC Safety Research Program (0 pen) 16.1) Discuss proposed scope and fomat for the ACRS report to the U.S.

Congress on the proposed FY 1987-88 RSR program and budget for the NRC (CPS /SD) 12:30 P.M. - 1:30 P.M. LUNCH

17) 1:30 P.M. - 2:30 P.M. ACRS Procedures and Practices (0 pen) 17.1) Report of ACRS Subcommittee on the July 30, 1985 meeting of the ACRS Subcomittee on Procedures and Administration and its Septem-ber 27,1985 meeting on the Report of the Panel on ACRS Effectiveness (DAW /RFF)

18) 2:30 P.M. - 3:30 P.M. Preparation of ACRS Reports (0 pen) 18.1) Discuss proposed ACRS reports regarding:

18.1-1) State of Nuclear Power Plant Safety (WK/AJC) 18.1-2) Application of PRA to the Indian Point Station (D0/RPS) 18.1-3) Natural Ability Selection of Reactor Operators (GAR /JOS)

19) 3:30 P.M - 4:00 P.M Ouantitative Safety Goals (0 pen) 19.1) Report of ACRS Subcomittee on items considered during its meet-ingofOctober9,1985(D0/RPS)

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PROPOSED MINUTES OF THE E I7- . ,

2 he O R -1 .

WASHINGTON, D.C.

The 306th me.eting of the Advisory Comittee on Reactor Safeguards, held at 1717 H Street, N.W., Washington, D.C. was convened by Chairman D. A.

Ward at 8:30 a.m., Thursday, October 10, 1985

[ Note: For a list of attendees, see Appendix I. P. G. Shewmon was not present on Saturday, October 12.]

Chairman D. A. Ward noted the existence of the published agenda for this meeting, and identified the items to be discussed. He noted that the meeting was being held in conformance with the Federal Advisory Committee Act and the Government in the Sunshine Act, Public Laws92-463 and 94-409, respectively. He also noted that a transcript of some of the public portions of the meeting was being taken, and would be available in the NRC's Public Document Room at 1717 H Street, N.W.,

Washington, D.C.

[ Note: Copies of the transcript taken at this meeting are also available for purchase from Ace-Federal Reporters. Inc., 444 North Capitol Street, Washington, D.C. 20001.]

I. Chairman's Report (0 pen)

[R. F. Fraley was the Designated Federal Official for this portion of the meeting.]

Chainnan Ward noted that the Supreme Court refused to prevent restart of the Three Mile Island, Unit 1 Power Plant and reported that the Plant is currently at 15 percent power, and is scheduled to shortly go to 100 percent power if all goes well during the intermediate stages. He noted that the staff has check off points and is following the start-up procedure very carefully.

D. A. Ward indicated that the NRC Office of Research has agreed to keep the ACRS informed by giving advanced notice of technical meetings involving its project managers and NRC contractors so that the ACRS can send observers if desired.

D. A. Ward mentioned that D. W. Moeller has participated in a second INPO team evaluation. Members were reminded by the ACRS Executive Director of the standing invitation by INP0 to have members accompany INP0 teams ~w hen they perform operational evaluations of plant operations. Members should infonn the ACRS Executive Director of their interest and availability.

D. A. Ward also mentioned that Members should inform A. Newsom, ACRS Assistant Executive Director for Administration, regarding their plans for foreign travel in view of the limitations on ACRS travel funds.

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 a

, e, II. TVA Orcanizational Changes (0 pen)

[ Note: A. J. Cappucci was the Designated Federal Official for this portionofthemeeting.]

J. W. Huffham, TVA Manager of Nuclear Licensing, contrasted the differences between their organization, effective early in 1984, and that which is in effect today, as of July 1985 (see Appendix IV). He explained that TVA has defined problems with certain plant programs and has redefined the organizational relationships. He pointed out that while TVA has reorganized several times over the past years, this is the first time for reorganizing relationships.

TVA has also been able to recruit and hire experienced and talented managers from outside of TVA.

J. W. Huffham explained that prior to July 1985, TVA was two massive organizations --

one dedicated to engineering and construction, and the other to operations. There was no central focus that led all of these activities to a common goal.

Therefore, in January 1984, H. Parris was vested as the Manager of the Power and Engineering Organization. H. Parris had approximately 20,000 people to manage under this organization.

Accountability was very difficult and an attempt was made to pull nuclear responsibilities together under one person, to define functions not readily identifiable. C. Michelson suggested that TVA's problems have not been generated in that year and a half between the two organizational charts. The problems were generated considerably earlier. J. W. Huffham agreed that TVA problems have been evolutionary. He acknowledged that prior to this early 1984 organization, there was an Office of Engineering and Construction and a separate Office of Power, each of which reported to the General Manager. There were generally bad comunications between the engineering and construction organization and the organization concerned with power production. The July 1985 reorganization reduced H. Parris's span of control from 26,000 to about 14,000 individuals. This reduced his overall responsibilities.

J. W. Huffham explained the chain of comand at each plant site. He indicated that there was a site director who was supported by the plant manager, design services manager, modifications manager, and site services manager. G. A. Reed noted that TVA has a training facility at the Sequoyah site. He asked where this falls in the organization. J. W. Huffham explained that its function is still under the Division of Nuclear Services and, since it functions for other plants, it is not part of Sequoyah's organization. C.

Michelson indicated that it was his impression that TVA had decentralized, to some extent, its engineering services. J . W.

Huffham agreed that the contingency of design services within the site is under the design services manager who reports directly to the site director. The site director is responsible for all design projects on that site.

.', / MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 J. W. Hr7fham discussed the objectives of the TVA nuclear program.

He indicated that the July 9 organization put H. Parris in charge of all of the resources thought necessary to achieve four long-term objectives:

. Consolidate nuclear resources under a tightly f7cused umbrella

. Acquire, retain and train management talent to effectively manage TVA's nuclear activities

. Establish priorities to limit TVA's activities to those that can be accomplished in an excellent manner

. Develop a team of experienced and qualified key personnel to provide leadership and direction to TVA's nuclear program J. W. Huffham noted that three managers were removed from the Brown's Ferry site organization (Plant Manager, Assistant Manager for Maintenance, and Operations Supervisor). G. 'l. Reed suggested that some sort of evaluation and selection process was applied to these individuals. J. W. Huffham explained that it was an evaluation of past management performance history ione internally.

It involved evaluating a manager's relationsnips with his employees. G. A. Reed asked if TVA is utilizing #ormalized tests to evaluate management potential. No formal answer was given.

J. W. Huffham discussed the current TVA organization in which H.

Parris has now approximately six people reporting to him. The site directors report through a Manager of Nuclear Operations directly to H. G. Parris. J. C. Ebersole asked the extent to which TVA is now decentralized into projects and the extent of central control as exercised from Knoxville, Tennessee. He was particularly interested in the extent to which decisions are made at the site director level. J. W. Huffham indicated that while there is a core engineering function that remains in the Office of Engineering in Knoxville, site specific design functions have been moved to the site to provide direct support to the site and direct involvement in the activities of the site. J. C. Ebersole supposed that since TVA is so decentralized, the basic critical technical decisions appear to be made by the site director at each project and not in Knoxville. R. Parker, TVA Assistant Director for Quality Assurance, agreed. He explained that to reinforce that philosophy the site director has been designated as the owner / operator of that site and actually establishes a contractual relationship with the TVA Office of Engineering for those services. The engineering function is a service function to the site director. This is not the case for sites under construction where the engineering function has not been decentralized.

C. Michelson pointed out that when one looks at organizational structures at TVA from 1980 to 1983, one finds that TVA has come full circle on such functions as engineering and construction. The current chart looks like one from several years ago. The functions -

,', , MINUTES OF THE 306TH ACRS MEETING Octob:;r 10-12, 1985 were combined into one and now they are broken down again. He questioned whether the problems generated in the period 1980 to 1983 or 1984 and the people who were responsible and in charge at that time have been a part of this reorganization. Most of the people shown are old time TVA people. J. W. Huffham disputed this, pointing out that there was a group of ranagers who had been moved to Chattanooga who sit at a round table under the direction of H.

Parris. It is not as separated as it used to be. C. Michelson indicated that he was a believer in the closeness of the working level to the managers. The working groups appeared to be still about 100 miles from the managers.

J. W. Huffham indicated that TVA has placed a single manager in charge of its nuclear program and has established a management team to provide overall direction to the nuclriar program. All responsibilities of operations have been consolidated under a single manager of nuclear operations and TVA has consolidated responsibilities for engineering and construction under a single manager. The Manager of Quality Assurance has been elevated in the organizational structure, as has the Manager of Licensing. There is also now a corporate entity established to set policy and guide the total nuclear program toward a goal.

F. J. Remick noted that, before this reorganization, TVA was planning to remove the simulator from the training center and put simulators at the individual sites. He asked if this type of '

decentralization is still planned. J. W. Huffham indicated that it

~

was and that evastruction was underway at Brown's Ferry. C. J.

Wylie asked if the sites are autonomous in contracting, engineering and construction services. J. W. Huffham indicated that at the

, sites the site director has complete control over operations, maintenance and engineering and construction as well as the cuality of engineering and construction. He did note that there will be a small group at the central office to define requirements and consistency between sites. D. A. Ward said that he was not sure that the presentation had been entirely successful at identifying what seemed to be a serious problem. He noted that TVA was regarded as a flagship of the nation's nuclear program. It is readily apparent from TVA's own self-identification, and presurably what the NRC Staff will relate to the Comittee, that it is TVA's internal ability to organize and manage the operations of a nuclear plant that appears to be the problem. He reminded the Comittee that it needs to decide how it wishes to involve itself in the resolution of these problems.

H. Thompson, NRC, indicated that the Staff and the Comission have been concerned about TVA management. Their decline in performance over the past five or six years has been identified for the Comission in SECY-85-231 (see Appendix V). The SECY document highlighted concerns the Staff had with management structure, aspects of accountability and communication difficulties that are apparent. There were some clear atiferences in performance between the various TVA sites. He mentioned concerns expressed by TVA employees regarding harassment and other activities that were going

',, l MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 on at the operating plants. Through the efforts of a super SALP Board, established by the EDO, it was felt appropriate in a regulatory sense to issue a 50.54(f) letter listing the major deficiencies identified as part of the SALP process, as well as some management concerns which were discussed earlier by J. W.

Huffham. He suggested that the key for TVA is to establish an approach that has accountability, that has people communicating and working together to resolve problems, and that has people working together in an environment that will address not only licensing concerns but the operational concerns that TVA has. He suggested that an effective organization will have to better develop working relationships at the first line supervisor level. One of the problems identified was the loss of some key managers from the TVA organization and the need for a fresh look at their organizational and management structure. p. G. Shewmon wondered whether TVA's payscale was a part of their problem. H. Thompson agreed that the ability to have a very high salary would be an incentive to attract a lot of people. He noted that Davis-Besse has had much greater success at recruiting than TVA because of TVA salary caps.

J. C. Ebersole noted the limitations of the TVA Board which is made up of poli +' cal appointees. He wondered how involved they were in anything at a reactive mode. H. Thompson thought that the TVA Board was involved despite the fairly limited nuclear background of the individuals. J. W. Huffham indicated that the TVA Board had been very supportive of the establishment of the current management team.

H. Thompson indicated that the NRC Staff intends to review any I

response from TVA to the 50.54(f) letter. The Staff expects to have at least sixty days before any start-up action and does not anticipate start-up of any facility or the licensing of any facility before the Staff has had the opportunity to review it.

One thing the Staff wishes to see is that TVA has the confidence to select the right people and can implement the program. D. A. Ward asked if the Staff has any criteria for judging whether the changes in organization and management approach are adequate. H. Thompson indicated that this is a difficult area for the Staff to develop hard and fast criteria. The Staff does look at operating experience, qualifications of individuals with operating experience who have been put in management positions. The Staff is also looking for good communication skills and good listening skills, all the way down to the level of first line supervisor. C.

Michelson asked questions regarding the soundness of TVA's organizational philosophy which has developed over a period of years and may have been a contributor to this problem. He asked what is being done to change the organizational philosophy. H.

Thompson certainly agreed that that was an issue but it will have to await TVA's response to the 50.54(f) letter.

D.A. Ward asked the Committee for its thoughts on how the ACRS ought to be involved in this issue. G. A. Reed thought that it would not be appropriate for the Comittee to be involved in the day to day details. F. J. Remick thought it might be a good idea

,' ,' MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 to schedule an information briefing by TVA when they have responded to the 50.54(f) show cause letter. J. C. Ebersole indicated that because of the extreme decentralization of TVA's operations into projects, he thought the Committee ought to concentrate on interacting at the working levels with the project organizations.

W. Kerr thought that if the ACRS were to make a meaningful contribution to this issue it ought to try to establish by itself or encourage some other organization to establish criteria by which the performance of a nuclear plant is judged. Some sort of objective criteria must be developed before they can be applied to TVA or any other organization. H. W. Lewis contended that the Committee is not in a position to judge the manaoement efforts at any particular plant at this time. He recognized the fact that the NRC does not have any objective criteria at this time for judging management competence at a plant. He thought that in spite of the difficulties, the ACRS ought to come to grips with the issue since it is an important safety problem.

III. Davis-Besse Nuclear Power Plant (0 pen)

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

F. J. Remick explained that the purpose of this session was to review the status of the start-up plan of the Davis-Besse Plant following the loss of main and auxiliary feedwater on June 9, 1985.

The meeting was requested by the NRC Staff. This session is not a review of the NRC incident investigation report nor is it a discussion of Comissioner Asselstine's request regarding this issue. F. J. Remick then briefly reviewed the June 9 event. He noted that Toledo Edison, the Licensee, is making a number of administrative and procedural changes as well as scme hardware fixes.

J. Williams, Toledo Edison Senior Vice President for Nuclear, described various management changes to the nuclear mission organization (see Appendix VI). He spoke of restructuring the engineering function so that the nuclear plants would be supported by an in-house engineering organization to the greatest extent possible with as little as possible reliance on outside consultant assistance. He indicated Toledo Edison's objective to recruit to fill key positions to shore up the organization for the restart effort. Toledo Edison upper management approved an increase to the Davis-Besse station personnel complement from 690 to 930 and raised salary levels. He recognized that the Davis-Besse Plant had a weak maintenance organization without sufficient talent in the leadership of that group. One of the programmatic changes made to the organization for start-up was to augment the talent throughout that organization and elevate the head of that organization to the level of Assistant Plant Manager of Maintenance. The Operations Superintendent was also elevated to the level of Assistant Plant Manager of Operations.

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

-b [ MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 )

4 C. .

i J. Williams discussed aspects of performance enhancement and a SALP improvement program at Davis-Besse. .He briefly mentioned preparations of detailed position descriptions for the new organization, development of a czpenent system data base with system descriptions to provide the operators' the design basis of systems, and a program of management training.- He indicated that

. Appendix R compliance is being accelerated.. The objective will be

, to have most issues addressed in 1988 rather than by the end of

{- 1989 as agreed upon with the Staff. G. A. Reed indicated that he was surprised at the staffing level of about 900 individuals .for tu single unit nuclear plant. J. Williams explained.- that the

. actual plant organization will be slightly over 400, the remainder being outside consultants. G. A. Reed asked if Davis-Besse is

. using any form of validated aptitude testing or selection processes in order to judge the talent it is acquiring. J. Williams i

indicated that he was bringing in high quality individuals and engineers that he had confidence in helping to select those individuals. G. A. Reed asked if Davis-Besse is going to use EEI tests. J. Williams indicated negatively. D. Okrent asked where in the new organization the responsibility lies for review, evaluation and decision making concerning the adequacy of non-safety and safety systems. J. Williams indicated that .this responsibility

, lies in the Facilities Engineering Group with heavy participation of operations people. D. Okrent asked where in the organization

, there would be knowledge of severe accident phenomenology. J.

Williams indicated that the responsibile*y would be shared by the Nuclear Plants Systems Group and the Nuclear Facility Engineering Group. D. Okrent expressed concern that even after staffing up, the Davis-Besse staff will not have sufficient talent with sufficiently deep knowledge of the behavior of the plant beyond the FSAR to deal with all operating situations.

J. Wood, Toledo Edison, discussed equipment concerns specific to

( the June 9 event. He described the resolution of equipment concerns. He explained how the equipment was subjected to trouble-shooting and investigation methodology where the findings

led directly to corrective actions and generic implications.

Generic implications such as maintecance, training, operator interaction were folded into a final resolution. He discussed issues associated with the main feed pump turbine, auxiliary feedwater valves AF599 and AF608, the power-operated relief valve (PORV), and nuclear instrumentation neutron source range detectors.

For each of these he explained a concern, findings, corrective actions, and generic implications. He explained that motor operators on the auxiliary feedwater valves were not properly adjusted, an issue which had generic implications for all motor operator valves in the plant.

G. A. Reed pointed out that the PORY in the plant which failed to close properly after the third time it opened during the incident was an internal pilot-operated valve which operates in a hydrogen borated environment (top of the pressurizer). It is subject to boren crystal scum. He suggested that this might be an improper application of thd PORV. G. A. Reed asked what Davis-Besse plans

[ MINUTES OF THE 306TH ACRS MEETING ' October 10-12, 1985

., .r, to do with respect to the maloperation of this particular type of valve. J. Wood explained that the operator has been provided with the indication in the control . room that he needs to use his equipment, such as the block valve, in order to mitigate a failure to close. He explained that a preventive maintenance program will

'be - developed for the source' range, intermediate range and power range detectors to address the problem of inadequate grounding of shields at the preamplifier due to paint and lack of star washers.

H. W. Lewis asked if Davis-Besse is comfortable that the preventive maintenance program will not do more harm.than good since a lot of -

electronic failures have developed from too much maintenance. J.

Wood indicated that Davis-Besse expects to develop a program that will balance the equipment expected life time _against the maintenance intervals to mitigate any negative aspects to the preventive maintenance effort.

Sushil Jain, Toledo Edison, indicated that most of the systems involved in the June 9 event related to the removal of decay heat from the reactor core, maintenance of start-up systems, and the steam and feedwater used for initiation of the auxiliary feedwater system. To attack these problems a task force was forried with the special objective to remove -the frequency of demand for emergency decay heat removal and also reduce the automatic responses that are required for initiation of auxiliary feedwater. The task force had additional objectives of improving 'the reliability of this system by removing the potential for common mode failures and also evaluating other diverse and redundant means available which could be installed to improve the overall reliability. The overall goal was to provide systems at Davis-Besse which would bring about an improvement conenensurate with the NRC's specified Standard Review Plan. J. C. Ebersole asked if Davis-Besse had a calculated estimated reliability for the feedwater system prior to the June 9 event. S. Jain indicated that there was no PRA model for the main feedwater system but a detailed model for the oh-a-per challenge basjs. Before June 9 that calculated number was on the order of 10~ per demand for both trains. Toledo Edison's recommendations regarding improvements to the system including the installation of the electric pump were based on minimizing the challenges to the system as well as maximizing system reliability of the system when challenged. J. C. Ebersole asked if the SFRCS will still dominate whether the feedwater systems will run with the addition of the new electric pump. Will it be subject to the same interception of flow as the two turbine-driven pumps? The electric auxiliary feed pump does not have an interface with the S8RCS.

S. Jain indicated that Toledo Edison will improve perfomance by improving the SFRCS power supply which contributed to spurious actuation of the system. The SFRCS logic will be modified so that isolation of main feedwater and main steam line will not automatically occur because of isolation of those two systems. A major program for the motor-operated valves has been undertaken to improve the performance of the system once it is challenged.

Several changes are being made regarding whether to leave the valves open or repower the valves to minimize the number of valves

, .' MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 ,

that have to reposition in the auxiliary feedwater system. This will eliminate the possibility of spurious mispositioning. The cause of the overspeed trip of the June 9 event, believed to be condensation on the steam lines (steam hitting the cold steam lines all the way to the turbine), will be corrected by putting the steam emission right next tc the turbine. This will minimize the condensation on the line and therefore minimize the overspeed trips. Operator error from misactuating the SFRCS will be corrected by revisions to the SFRCS panel (relocations of the manual actuations).

S. Jain mentioned removal and resizing of pump suction strainers.

D. Okrent asked if there is any down side risk to that modification. S. Jain indicated that the pump clearances have been evaluated to see what kind of debris traverses the strainers and the overall reliability of the pump has been observed with this change.

S. Jain explained that Davis-Besse is investigating longer term decay heat removal reliability improvements such as improvements in the Davis-Besse feed and bleed capabilities.

D. Okrent mentioned a paper given at a recent conference in Brussels, Belgium by Carl Fleming. Olli Moshi and R. Kenneth Gallagher entitled "The Systematic Procedure for the Incorporation of Common Cause Events and the Risk and Reliability Models." He asked if the NRC Staff was familiar with that paper. He explained that conclusions reached in that paper are that, for a typical three train auxiliary feedwater system, as found in several U.S.

nuclear plants, a realistic failure frequency ger challenge with all support systems avail per demand. The NRC Staff calculates 10~gble, is about He or better. 1 xhighly 10_ recommended that both Tolede Edison and the Staff review this paper and advise the Ccmmittee at some future time whether they agree or why they disagree with the conclusions drawn.

J. Lingenfelter, Toledo Edison, explained that a system review and test program is being performed to address the two major issues raised by the June 9 event. Specifically these are the concern over the effect of inadequate maintenance for those systems which were directly involved with the June 9 event and concern that the test program at Davis-Besse was not necessarily adequate in all cases to prove all functions of safety systems. The objectives of the program are to identify significant recurring maintenance and operations problems, identi fy testing required to assure that systems will perform their specified functions, and to conduct a test program to assure that the systems are fully functional. A total of 31 systems were divided into five groups: (1) primary systems; (2) electrical systems; (3) instrument control; (4) support systems; and (5) secondary systems.

J. Lingenfelter explainea that the system review program is based on an examination of historical documentation which includes licensee event reports, NPRDS data, maintenance work orders.

( h MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 facility change requests, Davis-Besse internal modification mechanisms and transient assessment analyses. The review also includes human engineering discrepancies. A second portion of the program involves evaluation of all the past testing that has been performed, including surveillance testing, periodic testing, as well as preventive maintenance testing to determine whether the program adequately tests all the functions of the systems. Based upon that review, groups will prepare outlines for new and revised tests that will be required prior to start-up.

F. J. Remick presented a question by J. C. Mark regarding plant access and security. K. Meyer, Toledo Edison, indicated that action is being taken to improve the access capability for the operators. He noted that Davis-Besse, however, is not significantly altering the normal access requirements or the security requirements that are in place.

J. Stolz, NRC, indicated that certain restart evaluation items derived from Staff concerns, as expressed in the 50.54 (f) letter that was given to the Licensee on August 14, 1985. Toledo Edison responded to the 50.54 (f) on September 12 with a proposed course of action regarding the NRC schedule. The Staff expects to have an SER prepared by early December and is committed to brief the Comission prior to the restart. Concerns expressed in the 50.54 (f) letter can be divided into three categories. The Staff requested that the Licensee continue (troubleshooting) an investigation of the events surrounding the incident on June 9, report on the specific plant findings, and address the programatic and management issues which contributed to the event and the performance of the Plant (see Appendix VII). A. Deagazio, NRC, indicated that the completion of the event investigation would involve an identification nf equipment malfunctions and operator errors, determination of tne root causes of the malfunctions and errors and their implications to the restart of the Plant, and corrective actions to assure the reliability of systems which mitigate loss of main feedwater events. W. Kerr asked what the Staff thought was the required reliability of the systems which mitigatelossofmajnfeedwater. J. Stolz indicated that the Staff has a target of 10-L. Rubenstein, NRC, explained that the Staff has a predecisional request for tge CRGR tg clearly state that all PWRs will have to meet the 10~ to 10- per demand target reliability criterion range. He thought Davis-Besse would be working toward that criterion. J. C. Ebersole suggested that the Palo Verde plant would have difficulty meeting such a reliability goal since they do not have the bleed and feed optior.. L. Rubenstein noted that the Staff does not include bleed and feed in their reliability analysis such that credit is not even given to that possibility. A.

Deagazio indicated that one of the Staff's concerns was for the Licensee to specifically explain what the capability of his system was for feed and bleed operations. Davis-Basse had previously l

reported to an ACRS Subcomittee that the PORV makeup and high pressure injection combinations could maintain cooling, if action

1

{ k, MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 i

4 f

were taken in an appropriate time frame. G. A. Reed questioned the Licensee's assertion for a system having a single unreliable PORV. *

' D. A. Ward also was concerned that the non-safety grade feed system was single failure vulnerable. From the system description it- is l 4 vulnerable to a single valve failing to open, which would cut off

all fl ow'. G. A. Reed expressed his dissatisfaction with the Staff's acceptance of a single PORV train and the. upgrading of the auxiliary feed system. He thought the Staff ought to hold out for

~

redundancy and diversity for decay heat removal, such as ' two or three PORVs appropriately located, able to blow down and allow the bleed and feed operation to proceed as an alternate means-of decay I heat removal.

A. Deagazio indicated that the Staff has asked Toledo Edison to address the inability to place the start-up feedwater , pump into service during the June 9 event. He indicated, however, that the

+

installation of the new electrically-driven start-up feedwater pump, in the long term, should address that issue. W. Kerr spoke of the requirement of the Staff to disable the start-up pump during that June 9 event because it is considered a non-safety grade piece

of equipment of questionable reliability . He asked if the Staff

is convinced that keeping that start-up pump unavailable rather 1 than having it available makes sense in terms of risk reduction.

> D. A. Ward asked if the issue has generic implications. W. Kerr i also thought the Staff ought to ask that question. He suggested l whether the Staff requirement that kept the pump from being

[ available was based upon the assumption that there was risk from i

[ breaking a high pressure line.

l R. H. Wessman, NRC Division of Licensing, discussed NRC generic i i technical actions under review and under consideration for commitment of resources as a result of the Davis-Besse event. He -

identified a number of issues, which could be considered as short-tenn generic issues and as potential long-term generic issues (see Appendix VII). The short-term generic issues identified are i currently under examination by the NRC Division of Safety

< Technology to determine cost-benefit considerations and to l determine reliability improvements that might result. These issues

! are being ranked as either high, medium or low priority. He I

explained that the first issue deals with the inability to remove decay heat because of problems with the auxiliary feedwater system. l This issue derives from three subparts that have emerged from the Davis-Besse event:

( . Difficulties with auxiliary feedwater valves where they go [

l closed and can not be reopened when they are needed i

l; . Difficulties in restoring steam turbine auxiliary feedwater

( pumps to operation should they trip  !

l

. Failures in the steam and feed rupture break mitigations systems that might yield reliability problems on an overall basis

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

! .I Y

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 Two other issues mentioned were the development of emergency

, procedures, operator training and . plant systems appronriate for determining when feed and bleed should be initiated and physical i

security system should be constrained, recognizing- that the latter -

could deny. timely access to . vital equipment. He indicated-that a dozen potential long-term generic issues have been identified, the first five of which have primarily been derived from the incident investigation effort, itself. These included the PORY and its block valve that are being considered as part of generic issue 70, and the . aspect of the PORV concerning its environmental qualification which is being considered as part of USI A-45.

R. H. Wessman indicated that I&E is preparing a bulletin dealing with potential failures of limitorque valves, similar to the 599 and 608 valves in the auxiliary feedwater system at Davis-Besse.

I&E is also considering issuance of several I&E notices dealing with classification of emergencies and timely notification of the

. NRC concerning a plant emergency. AEOD has under consideration two long-term generic studies, one dealing with the assessment of safety-related motor-operated valve failures, and failure modes

affecting valve performance under design basis conditions. They plan to conduct the study of steam turbine-driven performance as

, related to overspeed trips of the turbine. C. Michelson noted j that, for the RCIC system, in the case of boiling water reactors, General Electric has provided an overspeed trip reset which could be operated from the control room when it is deemed to be safety related. If it is not deemed to be safety related, then one had to set it locally. In the case of auxiliary feedwater (when it is

deemed to be safety related), he asked why there is not an 4 overspeed trip reset in the control room. R. H. Wessman indicated that this is one of the matters included in the Staff's evaluation t

of short-term generic . issues regarding auxiliary feedwater system

, reliability and how to deal with the turbine once it has tripped.

I IV. Source Term for Nuclear Power Plant Accidents (0 pen) 1

[ Note: G. R. Quittschreiber was the Designated Federal Official for this portion of the meeting.]

W. Kerr explained that the Subcommittee and ACRS consultants have met with the NRC Staff three times to discuss the draft report NUREG-0956 " Reassessment of the Technical Bases for Estimating Source Tenns ." He explained that after the Three Mile Island accident, observations indicated that a considerably smaller amount of iodine was released to the containment than previously anticipated based upon the results in WASH-1400. It was concluded by some that calculations of the source terms and of the consequent risks were much too high and that a more realistic examination of the chemical and physical processes that might be expected to occur during the severe accident would indicate that the risk was considerably less than calculated. As a result of this and other considerations, the NRC Staff has undertaken a rather extensive research program to investigate the pnysical processes and to attempt to formulate models which will predict the course of a MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 i

! severe accident. The Subcomittee concluded that NUREG-0956 (which incorporates the results of the research and the new computational  ;

4 models) can be characterized as a status report for a task begun but far from finished.

F M. Silberberg, NRC, defined the intended scope or intended areas cf 1

applications as the Staff now views it for the methodology for analytical procedures described in NUREG-0956. He discussed the  !

source tem reassessment study program relationships (see Appendix a

'VIII). He pointed out that NUREG-0956 embodies work that took  !

l place' from late 1982 through 1984, largely a contractor effort.

l which was then factored into NUREG-0956 by the Staff in developing its position on a current technical basis for source term  !

estimation. The major basis for NUREG-0956 was source tem 4

. analyses done by Battelle Columbus Laboratory for five reference plants. Additional initial scoping estimates were done by Sandia '

National Laboratory on uncertainty, and by Oak Ridge National Laboratory regarding validation ' support on the available data base i

as well as separate assessments of containment behavior on source term outcome. He indicated that two reports have evolved from the

, program; one is NUREG-1037, concerning containment performance and the other is NUREG-1079, on containment loads. Another important i element of the program is the American Physical Society independent j

! review which supplements the Staff's own peer review process.  ;

NUREG-0956 is one of the centerpieces of what will evolve into what is called regulatory implementation of source term infomation, and

regulatory control of severe accidents in the context of the

mandates given in the severe accident policy statement. On the ,

j regulatory side, a linkage is formed by NUREG-1150, currently under ,

. development, which is an outgrowth of a Severe Accident Risk Reduction and Rebaselining Program. NUREG-3150 will, using the

. NUREG-0956 procedures, be used to do a rebaselining of the risk for six reference plants. Five of these plants that were studied in ,

i BMI-2104 and noted in NUREG-0956 and an additional plant, the la Salle County Station, Units 1 or 2 will be studied for NUREG-1150. '

j M. Silberberg indicated that new analytical procedures in NUREG--

0956 are being used to calculate source terms for the risk i rebaseline of the six reference plants. He mentioned that there are ten areas which are potential candidates for changes in <

regulatory practices. The source term information will also be

! used regarding the Staff position on what methods should be used to 1

search for outliers in the systematic evaluation of operating plants. The analytical procedures described in NUREG-0956 will provide a basis for either audit calculations or other review procedures by which the Staff would review analyses made by the utilities in their systematic plant evaluation. This is currently identified as the IDCOR generic applicability study. He hoped that  ;

the Committee would find that the new technology described in  ;

i- NUREG-0956 is much better than the current'y used methods and that  ;

it should be used to replace current methods. The methods were

! said to have had extensive trial use and extensive review by i qualified and independent peer reviewers. He suggested that these  ;

i methods are appropriate for the ongoing rebaselining effort and I

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 adequate for proceeding into the start of a regulatory implementation phase.

M. Silberberg noted that NUREG-0956 was officially published in the Federal Register on August 7, 1985 and only a few technical coments have been received so far. The ACRS Subcomittee had asked about steam explosions and, as a result, some further discussions on the Staff's position on in-vessel steam explosions are scheduled, especially regarding the impact on the source term differences between the IDCOR and NRC models.

D. Okrent wondered how the Staff would reconcile majority and minority positions in the Steam Explosion Working Group report as the issue of steam explosions is not covered in NUREG-0956. Will it be resolved through engineering judgements or some well-backed scientific basis? J. Mitchell, NRC, indicated that there is a

, mall discussion of steam explosions and a small NRC-sponsored research program that will address the issue mentioned in Chapter 7 of NUREG-0956. She explained that the Staff position will be based on the Steam Explosion Review Grcup consensus report in which this group will address the issue from a risk point of view. D. Okrent noted that there is not an in-depth treatment of accident secuences for each of the reactor containment types. Most of NUREG-0956 seems to be a discussion related to the Surry Power Station. He asked when the Staff expects to have an in-depth treatment for each of the reactor types, J. Mitchell indicated that the calculations are now in progress in NUREG-1150 and will be available sequentially by mid-1986 for review.

D. Okrent asked if there was some frequency of occurrence for accident sequerces below which the Staff would ignore the sequence or its contribution to the source term. J. Mitchell indicated that the Staff has not explicitly considered a de minimis value but has accepted the WASH-1400 sequences. D. Okrent asked if the Staff excludes sequences whose estimated frequency is less than some number. M. Silberberg explained that the Staff thought it not appropriate to discard sequences at this point without looking further at the magnitude of the release fraction in terms of an early containment failure. D. Okrent asked if the Staff has gone through WASH-1400 to see what things have been assumed not to fail and not to occur that could, in fact, lead to early containment failure. J. Mitchell indicated that the Staff has not arrived at that point but it is part of the work plan. D. Okrent thought that that ought to be considered at an early stage in the Staff's

, thinking. R. Meyer, NRC, explained that NUREG-0956 addresses the I

method of calculating the source term only and has nothing to do with probabilities. This risk exercise that takes account of the probability of occurrence of new sequences (not recognized before) is out of the scope of NUREG-0956 but is part of NUREG-1150. D.

Okrent continued to stress that the NRC would need to use a threshold cutoff as a logical step in the development of NUREG-1150.

l L

\, ,.

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 Z. Rosztoczy, NRC, explained that the implementation program for the Severe Accident Policy involves an evaluation of six reference plants, and an evaluation of individual plants based upon the reference plant evaluation. Changes in rules and regulatory practices are based upon lessons learned from the analyses of similar accidents from the six reference plants. These rule changes will involve both source term-related changes, potential severe accident-related changes and the resolution of outstanding issues by)

IX . comparison of NRC He indicated thatcalculations withof the first phase those of IDCOR the Severe (see Appendix Accident Research Program has been completed. It has provided a much better understanding of the physical phenomena associated with severe accidents and is providing the Staff with calculational methods which can be used to analyze severe accidents. This methodology is now being applied to the reference plants. Important results from the evaluation of the reference plants are core damage frequency, radioactive material released from the containment, and potential exposure risk to the public. Factoring the IDCOR information into the NRC's data will yield basically two conclusions: One of the conclusions is whether the reference plants are sufficiently safe for severe accidents. A second conclusion will be what guidance should be given to other plants in order to assure that they are also sufficiently safe as far as severe accidents are concerned.

Parallel to the Staff's effort to develop guidelines and procedural criteria, as mentioned in the Policy Statement, IDCOR is developing the methodology such that individual plants can be reviewed once the review of the reference plants is complete. Regarding resolution of outstanding issues, some issues have been identified in connection with the IDCOR review. He noted the interest of the ACRS in the question of external events and seismic events. He indicated that the IDCOR calculations for evaluation of the reference plants are done for internal events and do not include seismic events or external events. Nevertheless, five recent PRAs which are available did handle external and seismic events. He indicated that the Staff was looking at two basic approaches regarding conclusion of the question of external events. One involves the transfer of information from the five PRAs that did treat them to the reference plants to complete the reference plants evaluation. Another possibility involves development of some simplified methodology for seismic analysis which would be used in the future in risk assessments.

J. C. Ebersole brought up the notion that the Staff is approaching the point of discounting the notion of an abrupt and complete pipe break. He asked how the Staff intended to treat a pipe burst with gross metallurgical failure. Z. Rosztoczy indicated that the basis of the Staff's overall work is to consider all events beyond the design basis that could be significant contributors to public risk.

He discussed how the Staff intended to treat the possibility of the removal of pipe restraints and more detailed attention to leakage detection. P. G. Shewmon asked that, if there were no pipe breaks larger than one-tenth of the cross sectional area of the allowed pipe break, would that change the accident sequences significantly.

". 5 MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 Z. Rosztoczy indicated that an analysis is oncoing but results are not yet available. Preliminary indications are that the large pipe break, even with the frequency presently assumed by the Staff, is a small contributor to risk. C. Michelson asked if the Staff had postulated a pipe break at other locations than where there were pipe whip restraints for the severe accident studies such as arbitrary intermediate locations for pipe restraints. Z. Rosztoczy indicated that the goal is to pick up those possibilities which are potential contributors to overall risk but he could not guarantee that every combination had been investigated.

Z. Rosztoczy indicated that the Staff intends to make a complete assessment of uncertainties as part of its application of the methodology in NUREG-1150. Since there is also a need for simplified source terms for regulatory applications, the Staff is considering reform of source terms in the following areas:

. Detailed source term calculations for individual plants

. Use of tables or some type of procedures for plant types

. A simple bounding source term applicable to most plants Z. Rosztoczy explained that detailed source term methodology will be done in such a manner that there will not only be information on the amount and timing of fission product releases from the containment, but also the calculational methodologies will predict the exposure of equipment located inside the containment. W. Kerr asked for a brief indication of how the conclusiuns of NUREG-0956 will be used to determine whether containment performance criteria are needed. Z. Rosztoczy explained that the methodology in NUREG-0956 will be applied to the six reference plants. The Staff should be able to draw some conclusions from the results as to whether the current design requirements or containments are sufficient and what benefits can be gained from including additional containment requirements. W. Kerr suggested that the key is whether the containment has sufficient strength, a possible containment criterion. He asked how the Staff is going to determine whether ccntainment performance is appropriate. Z.

Rosztoczy suggested that one containment criterion might be that the estimated frequency of early containment failure is below some probability. W. Kerr thought that was a legitimate approach but not available directly from a PRA. He suggested one approach might be to avoid the contribution of the containment but look just at total risk and also defense-in-depth. He still wondered how the Staff would determine whether one needed criteria using all the information provided in the PRA. Z. Rosztoczy indicated that the Policy Statement suggests a combination of deterministic analyses, supplemented by PRA results to evaluate the necessity for containment criteria. D. Okrent suggested that early containment failure, should it occur, might impose greater risk than previously ancicipated. He suggested that the staff may have to look beyond the calculated man-rems to think about the frequency of release of

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 radionuclides such as noble gases to determine the acceptability of some containments.

R. Meyer, NRC Office of Research, stated that the assumptions and limitations in the WASH-1400 methodology in light of the concerns after the Three Mile Island accident became the driving force for the current source term study. The source term work in NUREG-0956 was initially fission product release and transport study work (see Appendix X). Initial efforts on the source term project were an attempt to model more mechanistically the processes not adequately described in WASH-1400. The ORIGEN computer code was used to model fission product inventory in the fuel rods. Temperature, pressure, fluid flow rate, and other thermal hydraulic properties were modeled in the MARCH code. A major effort (mechanistic analyses) to calculate fission product retention in the reactor coolant system was undertaken because of the fact that no credit was given in WASH-1400 for fission product retention in the reactor coolant.

The entire source term suite of codes is described in the seven volume report, BMI-2104, where it is used to calculate source terms for some sequences. This gave a general framework that the Staff has used for their seluences.

D. Okrent asked if the Staff has a reasonably good insight into whether one can anticipate failure of the piping or steam generator due to fission product heating. R. Meyer indicated that the Battelle suite of codes does not have the capability to calculate the multi-dimensional recirculating flows needed to do that analysis. It is one major area of uncertainty that is being handled by other programs. J. C. Mark asked if the Staff knows what chemical forms will be produced if one dumps molten fuel on concrete. R. Meyer indicated that the chemical / thermodynamics that control those processes are fairly well known and modeled in the VANESA Code. The VANESA code has 125 chemical reactions set up and solved simultaneously with several kinetic steps in the calculation related to the flow rate of M.he gases that are released from the ablated concrete. A concern is the large uncertainty in the computer code that feeds the VANESA calculation and describes the basic core-concrete interaction, the ablation rates, temperatures, as well as production of gases from decomposition. The basic processes of core-concrete interaction are the subject of a program at the Sandia National Laboratories. H. Etherington asked if the MARCH code provides for alternate assumptions on the behavior of the corium and concrete. He suggested that s or,e large scale experiments are needed. R. Meyer indicated tha t the Staff is attempting to fill in this missing information with calculations using the CORCON code.

D. Okrent asked if the Staff's calculations allow for the thermal disintegration of the reactor cavity walls in existing structures above the reactor cavity. R. Meyer eyolained that the calculation for core-concrete interaction considers the concrete in the cavity which includes the horizontal, bottom, and vertical sides of the cavity but does not include any ceilings or other nearby structures. D. Okrent suggested that for Mark III containments.

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, e, MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 L

recent . staff estimates have indicated very adverse effects on the reactor ' cavity walls which might cause them to collapse, thus

, impact the overall question of how the molten core behaves. J.

Mitchell explained that for the Mark III the main question is the probability of pool bypass if the walls collapse because heat build-up causes the concrete to crumble. This opens a potential path for bypass. If the water coming out of the concrete does not alter the path for fission products through the suppression pool.

- then it does not really impact the source term. P. . Cybulskis.-

Battelle Memorial Institute, indicated that the MARCH Code will predict how much erosion there is under the concrete wall as a function of time. The code does not perform the structural calculation that would tell whether the wall would collapse. After running the MARCH calculation to see if there is significant or large erosion of the wall, one would have to do a side calculation to determine structural stability.

J. Mitchell explained the Staff's conclusions regarding source tenn science. She explained that the new analytical procedures in NUREG-0956 have been extensively reviewed and documented. One of these reviews has been a validation exercise reported in an Oak Ridge National Laboratory document that was under way at the time i- the analytical procedures were being developed. She mentioned the American Physical Society's review of the analytical procedures in their report (see Appendix XI). She stressed that the analytical

, procedure is not a production set of codes. It requires great care on the part of the analyst. One must consider all parts of the calculations. She indicated that no external events were considered in BMI-2104, not because the source terms that might i arise from external events were the same as might arise from t

internal events, but because they did not present a challenge to the computer codes different from the kinds of challenges considered in the sequences already chosen. The Staff considered high pressure and low pressure, fast sequence and slow sequence and early containment failure in containment. External events present the same kind of challenges. She noted that containment behavior for most accident seauences was the largest single factor affecting the source terms. Source terms were also found to depend strcngly on plant design and construction details. She pointed out that there is no uncertainty analysis available to date, something that is a very complex problem. An extensive effort is under way not only to derive estimates of the uncertainties associated with the source term but also to develop procedures for getting those estimates. This is being done in NUREG-1150 and is not part of NUREG-0956.

J. Mitchell indicated that the source term analytical methods in NUREG-0956 are an improvement over those derived from TID-14844 assumptions and WASH-1400 methods. She indicated that the Staff recommends that the new source term analytical methods be used recognizing that there is additional research to be done, as well as areas of major uncertainty which must be resolved. She noted that close coupling between the research effort and the regulatory effort will be required in assessing uncertainties and evaluating MINUTES l0F THE 306TH ACRS MEETINGL ' October 10-12, 1985 ,

-l technical issues associated with the new source term analytical procedure.

The Comittee' briefly discussed the nature of public; coments on i NUREG-0956. M. Silberberg indicated that there will be another peer review done in conjunction with NUREG-1150. M. W. Carbon asked if it was the Staff's belief that ' the' American Physical Society Study Group thought that more. research had - to be done -

before the Staff took regulatory action. Did that Study Group urge only that research be done before the Staff could understand all of the details? J. Mitchell explained that the NRC Staff position is that the Staff must complete the research and one of. the major.

conclusions of the American Physical Society was that - the NRC should continue its strong research effort. They were silent on

, regulatory uses. D. Okrent questioned whether the focus of the research effort will have the potential for significant impact on either regulatory action 'or steps that licensees will' take which have a feedback. He asked if the Staff has identified types of research' and kinds of information that are still unavailable and needed. J. Mitchell indicated :that the major areas of uncertainties have been singled out as having the potential for major impact on ~ source terms. R. Meyer identified Chapter 7 of NUREG-0956, a listing of uncertainty studies, which cross-correlate with comments from the American Physical Society and IDCOR as a description of research still required.

V. Emergency Planning (0 pen)

-[ Note: 0.S. Merrill was the Designated Federal Official for this portion of the meeting.]

D. W. Moeller explained that last spring the NRC Comissioners asked the ACRS to advise them of the complicating impacts to the response of off-site organizations to emergencies caused by a seismic event that occurred contemporaneously with a major accident

'at a nuclear power plant. The ACRS' coments on that subject are sumarized in its report to the Comissioners dated June 10, 1985.

Based upon the input from this Comittee as well as ~ many other organizations, the NRC Staff has formulated a suggested position in SECY 85-283. During a meating of the ACRS with the Comissioners in August 1985, the Cmittee was asked to review all types of natural phenomena th6 mt 5t affect nuclear plants, as well as the accompanying emerwr - r ponse. This same subject was discussed with the Comisswrori 3 the NRC Staff on September 9,1985. To follow up on this responw, a joint meeting of the Subcomittees on Site Evaluation and Extreme External Phenomena was held on October 9, 1985 for the purpose of developing data in three areas:

Range of probabilities for occurrence of various natural phenomena Potential contribution of each of these phenomena to accidents that might lead to severe core damage I ' '

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985

., < i Potential impact of these phenomena on emergency response The natural phenomena included in the Subcomittee discussions were seismic events (earthquakes), hurricanes and high winds, heavy rains and floods, and tornados. There was also some discussion of blizzards, ice storms, fog, tsunamis, brush fires, and even

. volcanic eruptions.

D. W. Moeller indicated that the Subcomittee qualitatively evaluated and ranked the various phenomena and formulated an opinion regarding the relative severity. He stated that the full Comittee should be able to prepare some written remarks after listening to presentations by the NRC Staff and a representative from FEMA.

M. Jamgochian, NRC, discussed earthquakes and their potential impact on emergency planning. He explained that prior to December 1981 the Staff requested that licensees, especially those in California, make a limited assessment of earthquakes primarily up to the SSE. This was dcne on a case-by-case basis. When the Staff requested that San Oncfre conduct such a study for consideration of evacuation time estimates, the Licensing ' Board decided that it would be appropriate to consider earthquakes beyond the SSE and their cceplicating effects on emergency planning. At this point, the Comission stated that no consideration need be given to the ccmplicating effects of earthquakes in emergency planning. The San Onofre decision was reaffirmed in August 1984 in the case of Diablo Canyon. The Comission, at that time, suggested that the matter ought to be explored in a broader sense by a rulemaking procedure (See Appendix XII). The proposed rule was published in the Federal Register in December 1984 It stated basically that neither evacuation time analyses nor emergency plans need consider the complicating effects of earthquakes. In July 1985, the ACRS met with the Comission and stated that it saw no technical reason for the exclusion of earthquakes from the natural phenomena considered in off-site planning. On September 9, 1985, the Staff made a presentation to the Comission proposing that limited consideration of severe low frequency natural phencmena bt considered in emergency planning. This was contrary to the proposed rule published in December 1984, which had three alternatives presented to the public for their consideration and coment:

Promulgate the proposed rule "Neither emergency plans or evacuation time estimates need consider the complicating effects of earthquakes" Leave the issue open for adjudication on a case-by-case basis Promulgate a final rule which limits the consideration of the complicating effects of severe low frequency natural phenomena on emer!)ency planning M. Jamgochian briefly discussed the extensive public cements received on the proposed rule. He indicated that, after careful

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,, MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 l

1 evaluation of the comments, a proposed final rule was submitted to the Comission. It stated that emergency plans should assure that the following capabilities exist relative to the complicating impacts of severe low frequency natural phenomena characteristic of the site:

The capability should exist to return people to the site in case of a natural phenomenon and an accident The ability should exist to reestablish two-way communications with off-site authorities Some preplanning should exist with off-site authorities regarding road networks, disrupted comunications, and alternate routes of travel with consideration of earthquakes D. Okrent noted that the Commission indicates that there should be plans for emergency evacuation or sheltering. He asked the rationale for the Commission's position. E. Jordon, I&E, explained that the Commission's position is based on an attempt to provide an improved level of protection for the public by having a plan that could respond principally to fast-breaking accident scenarios. The very long-term scenarios do not necessitate much preplanning but an additional measure of capability would be provided for the public for that type of accident. D. Okrent asked if the Comission believes that earthquakes are not fast-breaking scenarios or cannot cause fast-breaking scenarios within the nuclear power plant. E.

Jordon indicated that the intent is to provide the same protection regardless of the external event.

J. C. Ebersole asked if, in the event of an earthquake or other severe disaster, there must be the capability for radio transmission from a nuclear power plant. M. Jamgochian indicated that Appendix E requires redundant means of comunication. E.

Jordon indicated that every plant does have radio ccmunications.

C. P. Siess asked if radio comunication is required. E. Jordon indicated that he was not sure; however, diverse comunication paths are required since comunication with local enforcement officials is mandatory. The radio turns out to be the most convenient method as a backup. D. Okrent expressed some concern regarding the extent that preplanning for a seismic event in excess of the SSE might be relevant. E. Jordon agreed that how much is appropriate, based upon the risk and cost of additional planning, is an issue. The Staff recommendation has evolved from " don't do any more" to "do as much as proposed in this particular rule." The Comission is not sure what the appropriate level of additional planning should be.

E. Jordon clarified the scope of the responsibilities of the NRC and the Federal Emergency Management Agency (FEMA) regarding emergency preparedness. He explained that the NRC is responsible for reviewing off-site plans for licensed plants with respect to emergency preparedness and developing regulations that apply to utilities in tems of their plans. FEMA's responsibility is to

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MINUTES OF THE 306TH ACRS MEETING October '10-12,1985 i

perform the reviews of off-site. plans and, as they develop i regulations that affect the States, provide guidance for the off-site. reviews. The NRC normally works in conjunction with FEMA in developing criteria and guidance and coordinates rule ~ changes..

In the formulation of emergency preparedness, response organizations, Federal, State, and local governments and the utility are knit together with a comunications network'(see

, Appendix XIII). It is necessary for. the utilities to be able to j assess plant conditions and make the necessary modifications to the' .

State, local, and Federal plans regarding c' aracterizing the severity of a particular problem. The capability of the emergency preparedness system is augmented by training and exercises '

involving Federal, State and local authorities as well as the i

utility. There is an annual exercise for the utility only. C. P.

Siess asked what the effect on the rule would be if local

authorities simply refuse to exhibit appropriate concern about earthquakes at. a particular site. E. Jordon explained that the j regulation applies directly to the utility and there is only indirect pressure exerted on the local authorities.

E. Jordon discussed classifications which are referenced in Regulation 5047, Appendix E, and the present NUREG-0654 which I

called for actions related to external events. D. W. Moeller asked

! if every nuclear plant has an earthquake monitoring instrument that imediately tells them whether an earthquake is equal to the OBE or i

. the SSE or larger. E. Jordon indicated that it depends on the age ,

! of the plant and the locale. Plants that have high seismic ,

2 probability have more seismic instrumentation and the newer plants, i in general, have more equipment. E. Jordon discussed the 3

differences in licensee and State and local authority actions for unusual events, alerts, site area emergencies, and general emergencies. He discussed the existing evaluation criteria related to external events. In addition to reporting on facilities and equipment available, the licensee must have provisions to obtain  !

meteorological, hydrological, and seismic data. The licensee is to

, provide evacuation routes for on-site personnel- including alternatives for inclement weather, traffic, and radiological conditions. State and local authorities are to have plans for

! dealing with potential impediments such as seasonal impassability j of roads for use as evacuation routes and contingency measures.

E. Jordon emphasized that the Staff does not believe that this rule

• is intended to or should modify the surroundings in the vicinity of a nuclear plant. The rule urges the use of existing facilities and personnel to the best advantage to implement planning and decision making. He noted that protective actions, such as those just mentioned relating to the existing evacuation criteria, are not always synonymous with evacuation. It is also important to consider sheltering, even though it may not always be the best option for a large part of the population.

The Committee briefly discussed the TERA study done for the Diablo Canyon Nuclear Plant to address the Staff's questions on emergency preparedness regarding the earthquake hazard for that plant. E.

MINUTES OF THE 306TH ACRS MEETING October 10-12,1985 Jordon indicated that the study was very detailed and more than would be expected with regard to this proposed rule. D. W. Moeller asked if the estimates of physical damage from the TERA study were j sufficient. -Was the idea of alternative evacuation routes  !

, .regarding returning on-site personnel to the site after an event -l adequately treated? E. _Jordon expressed the belief that the very

~ detailed study done for damage to bridges and overpasses for the Diablo Canyon site would be excessive for other plants. He thought that consideration of alternate evacuation routes would be a useful approach'as part of an emergency plan.

i L. Reiter, NRC, discussed three studies of probabilistic estimates of exceeding seismic design levels (See Appendix XIV). He noted that the Committee should be aware that these probabilities ought-not to be viewed as hard numbers, as one may be dealing with just sophisticated ways of handling speculation. The first of the three studies was conducted by the Lawrence Livermore National. Laboratory which hopes to calculate the chance of exceeding ground motions at all nuclear power plants east of the Rockies. The second was a series of utility-sponsored studies related to PRAs that have included calculations of seismic hazards. The last study presented

. some initial preliminary results from the EPRI program for nine sites. He referred to graphs of the annual probability of exceeding peak acceleration relative to design-related values, such

, as the OBE and the SSE for particular sites. He cautioned that the use of this one parameter description of ground motion is a compromise and peak acceleration is a poor indicator of damage.

Both P. G. Shewmon and H. W. Lewis asked how it is detemined  ;

whether a site has experienced an OBE level earthquake. P. G. '

Shewmon asked if the sites in the studies have accelerometers which

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can identify an OBE level earthquake. L. Reiter indicated that these are studies which distribute earthquakes according to particular seismic zones and then average them over the seismic zone. Some sites do have accelerometers but not all of them work all of the time. The only case of a definite verification of exceedence of the OBE was for the Humboldt Bay Plant in 1975. P.

G. Shewmon expressed concern regarding the validity of these ,

predictions as there is such a paucity of experimental data to

, confirm the predictions. M. W. Carbon cited the lack of knowledge l as to why earthquakes have occurred east of the Rocky Mountains,

, such as those at New Madrid and at Charleston. L. Reiter indicated i that much more is known about New Madrid than about Charleston.

i The seismic zone has been identified for New Madrid, as well as some history on past earthquake frequency. The exact location of - ,

the seismic zone for Charleston has not yet been identified, nor the reason why an earthquake would occur there. There is some evidence accruing that there have been some large earthquakes there in the past, but the evidence is much less than for an earthquake ,

that occurred at Boston though the size of the earthquake at Boston was much less than either of those at New Madrid or Charleston, C.

P. Siess asked what the Staff knows about the Central Stable Region. L. Reiter indicated that, besides New Madrid, there was a j discovery in Western Oklahoma of a fault which apparently ruptured

in the past 2000 years. While there has been relatively little I-

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, ,, MINUTES OF THE 306TH ACRS MEETING- October 10-12, 1985 l

seismicity there within ~ the past 100 years, this was clearly the i

~ first indication of a " capable surface fault" in this - portion' of '

the United States.

L. Reiter explained a series of probability plots for the 08E. SSE, two times the SSE, and four times the SSE. He pointed out that the EPRI results shown on the graphs are preliminary and that changes, particularly in ground motion models, may be forthcoming. A'so, magnitude five or greater earthquakes were considered and rock.

sites were assumed. He ' stressed that there was little understanding of ' the causative mechanisms for the earthquakes included in the studies. The EPRI results appear to be an order of magnitude lower, in general, than the medians for the Livermore and those of other sites studied. He pointed out that ground motion is the significant issue and explains much of the differences between the three studies. He did not endorse the results at this point but indicated that for studies using similar ground motion models there will be some coalescing with these results. J. C. Mark asked if there is a place where an earthquake smaller than magnitude 5 on the Richter Scale could generate ground motion larger than the 0BE.

L. Reiter indicated that it can occur if one has a shallow earthquake. A more interesting question is whether a small earthquake can cause damage. He explained that while it it possible for an earthquake of a magnitude less than 5 to cause extensive damage, the best course of action is to see ~ that when analyses are made, smaller earthquakes are segregated from the results which include larger earthquakes. The consequences of smaller earthquakes do not readily mix with thoca for larger earthquakes.

L. Reiter presented some insights on the behavior of transportation facilities during earthquakes. He indicated that while the studies included earthquakes from intensity 8 to 9 on down, there was indication of disruption in some cases as low as intensity 6. As a result of a 1971 San Fernando earthquake and the severe damage to certain transportation facilities, such as bridges, there was an acceleration in retrofitting of bridges in California. As a result of the TERA report, which looked at the bridges around the Diablo i Canyon Nuclear Plant, some bridges were retrofitted.

M. E. Sanders, FEMA, explained that FEMA and NRC have some differences of opinion with respect to the language in the proposed final rule. He cited the portion in the rule that states that the licensees should have the ability to assess damage to the plant, to translate this information into projections of the expected or actual offsite hazard, and to be able to communicate this 9fonnation to offsite parties. He agreed that this is a licensee resynsibility but indicated that it also involves State and local governments. He suggested that the rule include some language to indicate that State and local governments have necessary redundant communications capability to accept the information that comes from licensee so tnat they can act on it. He objected to the introduction in the rule by the Office of Policy Evaluation of the notion that consequences should be emphasized at the expense of

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 causes. He suggested that this would be an expansion of the requirements on State and local governments and would make their implementation of the additional requirements called for in the proposed rule more difficult. He indicated that FEMA has a program that encourages State and local governments to identify the hazards to which they are subject. When these determinations are made, the State and local governments will key their plans to.these specific hazards. He contended that the NRC proposes to talk only about the consequences of an emergency event, irrespective of what caused the event, by eliminating reference to severe natural phenomena characteristic to a site.

VI. Meeting With NRC Comissioners (0 pen)

[Present at the meeting were Chairman N. J. Palladino and Comissioners F. M. Bernthal, T. M. Roberts and L. W. Zech, Jr.]

Chairman Palladino indicated that the Comission is currently considering a proposed letter to the Environmental Protection Agency (EPA) on its standards for high level waste disposal and noted planned NRC rulemaking to incorporate the EPA assurance requirements into NRC's 10 CFR Part 60. The Comission is very interested in the views of the ACRS in this matter, particularly regarding comments in a July 17, 1985 letter to the Comission which expressed some concerns regarding demonstration of compliance with the proposed EPA Standards. ACRS Chairman D. A. Ward indicated that there is a strongly-held consensus by the Comittee regarding the EPA's Standards and several members would have additional comments after the opening remarks of D. W. Moeller.

D. W. Moeller reminded the Comission that after a 30 day comment period these Standards will become official. Since the Standards will apply to facilities being proposed by 00E for a high level radioactive waste repository, facilities which will ultimately be licensed by the NRC, the ACRS thought it very important that it discuss this matter with the Comission. He noted that the ACRS Waste Management Subcommittee has been following the development of the Standards for the past several years with representatives of EPA, and that EPA has consulted with a variety of outside organizations regarding input to the formulation of the Standards. EPA also asked for advice on their proposed Standards from their own Science Advisory Board which then established a High Level Radioactive Waste Disposal Subcomittee.

This Subcommittee reviewed the Standards in detail and offered advice and suggestions to the EPA on how they should be modified.

The Science Advisory Board made a number of recomendations, two of which the Comittee believes are particularly important. The first conclusion of the Science Advisory Board was that the Standards were overly restrictive or too stringent. The Board recomended that release limits incorporated within the Standards be increased by a factor of 10. The second recommendation of the Science Advisory Board was that the quantitative probabilistic conditions of the release limits be made dependent on EPA's ability "to provide convincing evidence that such a condition is practical

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 to meet and will not lead to serious impediments, legal or otherwise to the licensing of high level waste geological repository." The Board went on to say that if such evidence cannot be provided, the EPA should adopt qualitative criteria such as those previously suggested by the NRC. The Board further recomended that analysis of repository performance be required to demonstrate that there is a less than a 50 percent chance of exceeding given containment requirements. Events whose median frequency is less than 1 in 10,000 within the the 10,000 years posed for consideration of cumulative radionuclide releases to the accessible environment, need not be considered.

D. W. Moeller indicated that the ACRS joins the Science Advisory Board in these particular criticisms. The ACRS finds additional problems that were noted. Among these are the fact that the Standards are not consistent in terms of using a risk-based approach. He noted that the ACRS is of the opinion that the overly restrictive Standards could lead to the rejection of some sites <

that might otherwise be acceptable. Another problem noted by the Subcommittee was that the release limits were based upon a generic environmental model which the EPA applied. This model was used to calculate the magnitude of the release that would have to occur to produce a particular dose. The Committee questions whether this generic environmental model will actually apply to the specific sites that are chosen. The Comittee believes that the Comission has to seriously consider the possibility that the Standards now established could lead to serious impediments to the licensing of a high level waste repository. The Commission should also consider whether it would be more advantageous to challenge the Standards now and get the problems straightened out at this point or wait until later when more difficult problems occur downstream.

D. W. Moeller indicated that the Committee does not understand the stringency of the EPA high level waste repository Standards as compared to other risks considered acceptable in our everyday lives. He wondered whether the EPA was using a consistent approach in applying risks to all types of environmental problems, such as the risks associated with a toxic chemical waste disposal facility or the health affects from uranium mill tailings. The high level waste repository Standards are 3 to 6 orders of magnitude more stringent than the risks associated with related Standards being applied to other problems in EPA's jurisdiction. He indicated that, as a policy decision, he urged the Comission to look once again at these EPA Standards. While the Comission has not signed off on the Standards, the NRC Staff has approved them. When there is a choice between comparing a very restrictive Standard loosely applied versus a more realistic or less restrictive Standard that could be properly verified and enforced, the ACRS prefers the latter. W. Kerr noted that the ACRS has not really explored the Science Advisory Board recomendation that EPA develop qualitative Standards if it could not demonstrate enforceability of its current ones.

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 l

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Comissioner Zech thought that the ACRS concerns and l recommendations have been incorporated by the Staff in its  ;

submission to the Comission. D. W. Moeller noted that the ACRS  ;

still does not agree with the Staff. Comissioner Bernthal ,

expressed some confusion regarding the units in which the numbers '

were presented. He saw no comprehensive analysis from either the NRC Staff or even the ACRS to reduce these numbers to understandable quantities. In addition, he asked what EPA's rationale was for picking these numbers. D. W. Moeller suggested that the EPA endorsed figures that appeared most acceptable to environmental groups and other organizations that were asked to coment. He suggested the possibility that the EPA agreed to what was expedient instead of values that were scientifictlly evaluated.

In answer to a question by Comissioner Bernthal, D. W. Moeller indicated that ne understood that the risk evaluation group within EPA which should be involved in setting Standards for a high level repository was not involved in the development of the Standards in any way. Questions were asked as to why the NRC Staff as well as DOE agreed to the Standards. D. W. Moeller suggested that they might have thought that the Standards would be applied on a trial basis and changed if found unworkable at a later date.

D. Okrent suggested the possibility that the Standards may have been influenced by a proposal from the Natural Resources Defense Council (NRDC) that the radioactive waste when put in the ground should produce no more effect than would have been produced had natural uranium been left in the ground. In any event the relaxation of the Standards is really EPA's business according to law. It is the NRC's job to regulate according to the Standards.

However, the NRC should be satisfied that they are workable. C. P.

Seiss suggested that the NRC Staff believes that it is possible to build a high level waste repository that will meet the Standard and it recognized that it may be difficult to establish it before the hearing board. He suggested that the Staff may be naive to think that the words " reasonable assurance" in the EPA regulation will do any more good than they would in the NRC regulation.

H. Miller, Branch Chief of the NRC Repository Projects Branch, and author of SECY-84-320 (see Appendix XV) indicated that the Staff considered all these items in its negotiations with EPA and was very much concerned with the ability to implement these regulations. He indicated that the Staff was particularly concerned about the ability to come to closure in the hearing I process on a finding that the Standards are being met. He suggested that it is speculation that probicns will develop l regarding the implementation of the Standards dowti the road. He I noted that the Staff, regarding the specific sites, is right now I identifying potential disruptive events that might impact the  :'

implementation of the Standards. The Staff judges that the Standards can be implemented. He suggested thot ACRS concerns may be centered on the cost-benefit aspects of implementing a probabilistic standard. The Staff had given a fair amount of deference to EPA, given the fact that by statute they are chartered with setting these Standards. The Staff did not attempt to redo

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985

=. v.

their cost-benefit analyses. In answer to a question by Comissioner Bernthal, H. Miller noted that the Standards were published in the Federal Register September 15, 1985 with a 60 day waiting period before they become legally effective. Comissioner Bernthal wondered if the Comission would have to mount a fonnal legal challenge within the 60 day period. Chairman Palladino suggested that unless the Comission does mount a legal challenge it appears that the Standards will become effective.

Chairman Palladino indicated that the Comission ought to get ACRS .. expanded comments in writing as soon as possible. He suggested that the Comission ought to have a briefing by the NRC Staff shortly to see what options are available and what the Comission wants to do with regard to these options. He thought it might be worthwhile to get the ACRS coments before such a briefing. He also indicated intention to ask OELD to prepare a discussion regarding a legal course of action with regard to these EPA Standards.

Commissioner Bernthal asked the ACRS for specific instarces where the EPA Standards are unreasonable. D. W. Moeller indicated that the requirements are primarily in terms of the release limits which, in turn, are supposed to reflect 1,000 health effects per 10,000 years. It is this table of 10,000 years cumulative release that is a problem. Both the release amount and the probabilistic requirements were thought by the Subcomittee to present difficulties in demonstrating the degree of confidence expected that the Standards could be met. Chairman Palladino thought that the NRC briefing should definitely go into technical matters. The Comission ought to make sure after the briefing that it understands the issues and the differences between the ACRS and NRC Staff positions and can make a determination of where the Comission wants to go from there.

VII. General Electric Standard Safety Analysis Reoort (GESSAR II) (0 pen)

[ Note: R. K. Major was the Designated Federal Official for this portion of the meeting.]

The members spent considerable time discussing the report on GESSAR II by D. Okrent. They agreed on a revised scope for the report in which the FDA for GESSAR II would be addressed in one report and ACRS coments on improvements in the safety of future nuclear power plants would be addressed in another report. A table on the reference ability of GESSAR II, as well as CESSAR F and RESAR SP/90 was offered by C. Thomas, NRC (see Appendix XVI). M.

W. Carbon was asked to prepare the draft report on improvements for future nuclear power plants and H. Etherington was asked to draft the report regarding the FDA for GESSAR II. Both will be submitted for consideration during the 307th ACRS meeting.

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 VIII. Advanced Reactors (0 pen)

[ Note: M. El-Zeftawy was the Designated Federal Official for this portionofthemeeting.]

J. C. Mark indicated that a meeting of the Advanced Reactors Subcommittee was held on September 25, 1985 at which the proposed policy for regulation of advanced nuclear power plants was discussed. He indicated that several modular design concepts were mentioned in an NRR presentation. These included an HTGR concept, a 350 MWt modular design, as well as two liquid metal reactors, a Power Reactor Inherent Safety Module (PRISM), 425 MWt modular design, and a Sodium Advanced Fast Reactor (SAFR), a 900 MWt modular design. The proposed Advanced Reactor Policy Statement was issued for public coment on March 26, 1985 and ACRS comments were expected. The Subcommittee chose to comment after the public comments had been reviewed by the Staff. On August 21,1985, after review of the public comments, the NRC Staff issued a revised policy statement (SECY-85-279). The Commission is in favor of early interaction in this matter and has set the regulatory climate to facilitate a licensing review.

J. C. Mark indicated that the proposed revised policy statement defines advanced reactors as " reactor designs which are significantly different from the present generation of light water reactors". He suggested that a decision must be made on what is actually an advanced reactor. The PIUS reactor, liquid metal reactors, as well as a large HTGR would be considered advanced as would be SAFR modular design. The advanced BWR design is considered evolutionary, and not advanced, within the scope of the policy statement. He questioned whether the advanced reactor group at NRR was capable of appropriately following through on a thorough licensing review for the advanced reactor designs. The public comments suggested that the' Staff reduce the prescriptive nature of the NRC regulations. Members of the Subcommittee agreed that most criteria already are not prescriptive.

J. C. Mark indicated that the proposed policy contains a list of 11 general characteristics that the Comission believes should be attributes of a reactor design for it to be considered advanced.

Among these were the attribute that the design ought to consider the defense-in-depth philosophy by maintaining multiple barriers to radiation release. He thought that the general characteristic suggested by the Staff that designs reduce potential radiation exposures to plant personnel ought not to go forth in the policy statement as stated. He pointed out that the first draft of the policy statement talked of an " enhanced margin of safety". This statement was missing from the August revised draft. He interpreted this to mean that the Staff now will not mandate improved safety but encourage safety improvements. D. Okrent thought that this statement could be modified to make it acceptable. F. J. Remick defended the policy statement in that he felt the.11 attributes were not meant to be a definition of an advanced reactor but were "should include" design guidance. D. A.

Ward agreed that the 11 attributes make sense as guidance but not as a definition.

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 R. Hernan, NRC, explained that 14 requirements packages are to be sent to prospective vendors who expect to submit advanced reactor design!. This will be the primary interaction with the nuclear industry The Staff's main effort at present is to tie down

_ generic 1: sues and unresolved safety issues that could affect advanced designs. J. C. Mark noted that there is considerable discussion regarding general design criteria (GDC) for the advanced reactor designs. He suggested that the Staff ought to start over and develop a fresh set of GDCs or use those GDCs for light water reactors and build upon them. He thought that the suggestion to make up other GDCs specific to a particular proposed design would not be appropriate as a real GDC. M. W. Carbon noted that the Clinch River Breeder Reactor (CRBR) was to meet a revised set of GDCs that were tailored to the CRBR design. He indicated his concern regarding the question of redundancy and diversity which might be inappropriately carried over from the past. J. C. Mark suggested that it might be appropriate to send some response to the Staff which makes suggestions on how they might make decisions. H.

W. Lewis indicated that any response to the Staff should note that the level of current safety should be maintained.

IX. Annual ACRS Reoort on the NRC Safety Research Program (0 pen)

[ Note: S. Duraiswamy was a Designated Federal Official for this portion of the meeting]

C. P. Seiss mentioned that at the retreat held at Harper's Ferry on November 19, 1984, members felt that too much time was being spent on the annual ACRS report to Congress on the NRC Safety Research Program. It was thought that the Comittee could satisfy the law by drafting a short (few pages) document signed by the ACRS chairman. He suggested that the Cemittee use the report to the Ccmmission on the NRC safety research program and budget for FY 1987, dated June 11, 1985, as a guide for the upcoming ACRS report to the Congress. The Comittee could revise or update that report to meet its obligation. Regarding succeeding reports, he thought that the Comittee ought to spend less time on the money issues and more on cosmic issues. The report to Congress ought to have the scope of the report to the Comission (5 pagas) and might contain less advice to the NRC Staff. After February 1986, the Comittee needs to consider a new approach for the report to Congress. The Comittee might forget about dollar numbers and concentrate on research needs, what the NRC is doing, and priorities for research.

G. A. Reed agreed that the Comittee ought to deal with the research program on a cosmic level. W. Kerr stressed that the Comittee should try to develop a document to which the Congress would pay some attention. H. W. Lewis suggested that the ACRS ask for relief from its obligation to prepare an annual report to the Congress but adhere to its requirement to advise whenever particular issues come up. C. P. Seiss indicated that the major discussion on the upcoming report to Congress will be in the full Comittee . He indicated his plan not to schedule any Subcomittee r

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 meetings, nor any discussion on the annual report to the Congress at the 308th ACRS meeting in December.

I-X. ExecutiveSessions(0 pen)

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

A. Subcomittee Assignments

1. ,TVA Organizational Changes The Comittee was briefed by re of the Tennessee Valley Authority (TVA) presentatives and the NRC Staff regarding changes to the TVA nuclear organization in response to management problems at the corporate and nuclear site levels. Several suggestions were made by members about the nature of future ACRS involvement regarding this issue but the Comittee made no decision on a course of action for ACRS participation. (Note -

This matter will be referred to the Management Committee forfurtherconsideration.)

2. State of Nuclear Power Plant Safety The Comittee discussed the deliberations of the Subcommittee on the State of Nuclear Power Plant Safety and requested that W. Kerr draft a memorandum for circulation to all ACRS members regarding the next step in this effort. By memorandum from W. Kerr dated October l

12, 1985, members are asked to submit by the November meeting a description, of not more than one-half page in length, of what they individually consider the most important safety issue regarding operation of nuclear power plants and what they believe is an outstanding strength of the system in assuring safety. The results of this survey will be discussed at the December ACRS meeting.

3. Davis-Besse Nuclear Plant i The Comittee discussed possible courses of action to i

address the start-up plan for Davis-Besse and those changes resulting from the recent loss of auxiliary feedwater event at this plant. Note was taken of the request by the Comissioner Asselstine to respond to three questions:

i . Review of the Davis-Besse feedwater system

. Evaluation of NRC's inspection and enforcement program for Davis-Besse l

L

O 3 ,

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 l

. Implications of operating experience on the adequacy of the B&W plant design Dr. Okrent thought the ACRS definitely ought to coment on the start-up plan for Davis-Besse and the Davis-Besse design as well as the overall effort of the NRC Staff to resolve this matter. The Davis-Besse Subcommittee was assigned to review the restart of Davis-Besse and the Subcommittee on Regulatory Policies and Practices was assigned to evaluate the investigative effort by the NRC Staff (OIA) of NRC inspection and enforcement activities related to the June 9, 1985 incident. Action on a review of the adequacy of the B&W plant design and other issues will await further clarification of the ACRS role expected during the week of October 21,

4. Report of the Procedures and Administration Subcomittee Meeting on July 30, 1985 In connection with a proposal by D. A. Ward for a risk-based review of NRC rules and regulations, the members decided that additional information is needed regarding the present and anticipated NRC Staff use of risk considerations in decision making. J. MacEvoy, senior ACRS fellow, was assigned to work up background infonnation on this issue. His re presentation by the 307th (November) portmeeting.

ACRS is planned for The Comittee agreed to simplify the ACRS annual report to the U.S. Congress on the NRC proposed Safety Research Program and budget planned for February 1986. The June 1985 ACRS reput to the Comissioners will form the basis for the reduced scope and length (approximately five pages long) of the report to Congress. The ACRS will discuss longer range methods by which the scope of the report to the Congress can be reduced subsequent to preparation of the February 1986 version.

The Comittee discussed and endorsed, with one caveat, proposed revisions to the Memorandum of Understanding (MOV) with the EDO to provide opportunity for ACRS review of changes in the NRC Standard Review Plan (SRP). Note was taken of a special case where a portion of the SRP might be substituted for a regulatory guide with no corresponding indication in the SRP. Members wanted to be sure that they would have the opportunity to comment on the effect of such a chante in guidance. The ACRS Executive Director, R. F. FraLey, said that the matter would be reviewed and the MOU suitably modified to cover such an eventuality.

The Comittee decided that it should continue to review Naval facilities when asked by DOE /000 under the existing DOE /NRC MOU. The members agreed, however.

MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 that the ACRS should request that documents that the Cemittee continues to receive (approximately 12/ year) about naval reactor siting for both U.S. and foreign naval nuclear ships should be terminated.

The Comittee noted that four of the NRC Comissioners have now expressed qualified support for a National Academy for Training Reactor Operators as put forth in a pending bill before the U.S. Congress originated by Senator Moynihan. F. J. Remick indicated that INPO has recently established a National Accreditation Institute for Reactor 0)erator Training to satisfy the concerns of Senator Moynihan. Although the ACRS had previously decided not to address this issue, it was agreed that the ACRS staff (J. O. Schiffgens, staff engineer) should prepare a package of background information including the Congressional hearing at which the Comissioners testified so that this matter can be discussed during the 307th ACRS meeting.

R. F. Fraley noted the limitation being applied to INPO documents and data received from the NPRDS consistent with the MCU between the NRC and INPO. Requests to cbtain additicnal copies for distribution to interested ACRS members and/or ACRS consultants on occasion have been turned down by INPO. While these limitations seem inappropriate, members concluded that copies of the INPO documents provided to the Committee should be adequate for ACRS purposes, particularly since INPO would be reluctant to provide any copies at all if some control of the distribution were not required by NRC.

Draft minutes for the Procedures and Administra" ton Subcomittee Meeting held on September 27, 1!85, regarding the first thirteen recomendations of the Effectiveness Panel plus additional coments by Panel Chairman M. Muntzing regarding meeting " logistics" were distributed. Cemittee member coments were requested by Chairman Ward for consideration during the next meeting on this subPct (November 12, 1985 Subcomittee meeting).

B. Reports. Letters and Memoranda

1. Additional Coments on the EPA Standards for a High Level Radioactive Waste Repository The Comittee prepared a report to the Comissioners regarding the EPA Standards for a High Level Radioactive Waste Repository which were published on

, September 19, 1985 by the U.S. Environmental Protection Agency (EPA).

..'e, MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985

2. Advanced Reactor Policy Statement The Committee prepared a report to the Commissioners of discussion of the proposed Statement on Regulatory Policy for Advanced Reactors as presented in SECY-85-279 dated August 21, 1985.

3. Impacts of Natural Phenomena on Off-Site Emergency

Response

The Committee prepared a report to the Commissioners of its review and evaluation of the relative importance of various natural phenomena which initiate or occur in coincidence with accidents at nuclear power plants and have the potential for significant impacts en emergency response activities.

4. Consideration of Earthquakes in Off-Site Emergency Planning The Committee prepared a report to the Commissioners of its review of the proposed amendments to 10 CFR Part 50, Appendix E, Consideration of Earthquakes in Emergency Planning.

5. ACRS Action on Proposed Regulatory Guides The Comittee prepared a memorandum to the E00 in which it concurred in the regulatory positions of the Proposed Revision 2 to Regulatory Guide 1.105,

" Instrument Setpoints for Safety-Related Systems" and proposed Regulatory Guide, Task No. IC 609-5,

" Criteria for Power, Instrumentation, and Control Portions of Safety Systems."

In addition, the ACRS concurred in the NRC Staff's proposal to issue for public comment the proposed Revision 1 to Regulatory Guide 1.23, " Meteorological Measurement Program for Nuclear Power Plants."

Subsequent to the public comment period, the Committee expects to review the proposed final version of the Guide together with the public comments and the NRC Staff's responses to them.

C. Generic !ssues

1. Pressurized Thermal Shock The Committee briefly discussed the latest comments by D. L. Basdekas regarding the adequacy of NRC requirements to preclude thermai shock to the reactor pressure vessel. Sidney J. S. Parry, ACRS senior fellow, was asked to investigate how representative the weld sample from the H. D.

,,,e MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 Robinson Nuclear Unit 2 was of a vessel beltline weld regarding materials properties. D. L. Basdekas concerns regarding the severity of transients will be examined by ACRS consultant, I. Catton. P. A.

Boehnert, ACRS staff engineer, will prepare a package of calculations for I. Catton to include recent transient calculational work by T. Theofanous done for the NRC Staff.

2. Natural Ability Selection of Reactor Operators A proposed report by G. A. Reed was discussed briefly during this meeting. It was deferred for further discussion during the 307th (November) ACRS meeting.

3. Seismic Margins A proposed letter regarding seismic margins was discussed briefly. No specific action was taken regarding this letter due to a lack of time. It was deferred for Comittee action during the 308th (December) ACRS meeting. A brief discussion will be scheduled during the 307th ACRS meeting (November).

4 Source Tenn l Consideration of a proposed ACRS report regarding the NRC Staff reassessment and new position on accident source terms was deferred until the 307th ACRS meeting.

D. Future Schedule

1. Future Agenda The comittee agreed on tentative agenda items for the 307th ACRS meeting, November 7-9, 1985 (See Appendix !!).

2. Future Subcomittee Activities A schedule of future Subcomittee activities was distributed to members (see Appendix III).

E. D. L. Basdekas Letter to Congrass The Comittee decided not to reply to D. L. Basdekas' criticism of the ACRS in a letter to the Hon. M. K.

Udall, Chairman, Subecmittee on Energy and Environment.

Comittee on Interior and Insular Affairs, U. S. House of Representatives dated August 22, 1985. H. W. Lewis was informed that he could respond if he desired to do so expressing his personal opinion on this issue as an individual member of the public.

l . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _

,,,.. MINUTES OF THE 306TH ACRS MEETING October 10-12, 1985 F. ACRS Meeting Dates for CY-1986 ACRS meeting dates for CY-1986 have been scheduled as noted:

Meeting Date 309th January 9-11, 1986 310th February 13-15, 1986 311th March 13-15, 1986 312th April 10-12, 1986 313th May 8-10, 1986 314th June 5-7, 1986 315th July 10-12, 1986 316th August 7-9, 1986 317th September 11-13, 1986 318th October 9-11, 1986 319th November 6-8, 1986 320th December 11-13, 1986 G. Indian Point Special Proceeding The Committee declined to endorse a proposed ACRS report introduced by D. Okrent regarding the outcome of the l

Indian Point Special Proceeding based on its PRA. The use of the Indian Point PRA numerical core melt frequency estimates in the draft report was characterized as

" bottom fining." W. Kerr thought it not acceptable to base the entire judgment regarding nuclear power plant safety on core melt frequency,a quantity known to have a large uncertainty associated with it. He noted that even the NRC Staff has not yet adopted such a posture. C. P.

Siess asked D. Okrent his reason for concentrating on core melt frequency and D. Okrent cited R. 8. Minogue's concerns regarding the downplaying of accident prevention, the defense-in-depth concept, and core melt frequency which he considered a surrogate for accident prevention. D. Okrent believes that, especially for high power nuclear plants at one of the most highly populated sites in the U.S., the NRC should continue to strive for a substantial reduction in both the predicted core melt frequency and the uncertainty therein. H. Denton has written a similar letter in which he endorses working toward low core melt frequencies. It was noted that a PRA was performed for Indian Point Units 2 and 3 and that a systems interactions study has been performed for Indian Point Unit 3. These studies were used to evaluate the adequacy of the design of these plants. This is a level of effort beyond what is normally required. D. A.

Ward acknowledged D. Okrent's genuine concern to make the Indian Point plants safer in the long term but took note of the apparent lack of enthusiasm on the Comittee to E

MIhUTES OF THE 306TH ACRS MEETING October 10-12, 1985 join him in a written report. He was informed that he is now free to send a personal letter regarding this matter if he so desires. D. Okrent has sent a letter to the Comission dated October 12, 1985 regarding this matter.

ACRS members J. C. Ebersole, D. W. Moeller , G. A. Reed, and C. J. Wylie joined D. Okrent as signatories of this report.

H. INP0 Visits to Nuclear .Nwar Plant Members were reminded by the ACRS Executive Director of the standing invitation by INPO to have members accompany INP0 teams when they perfonn operational evaluations of plant operations. Members should inform the ACRS Executive Director of their interest and availability.

The 306th ACRS Meeting was adjourned at 3:20 p.m., Saturday, October 12, 1985.

O APPENDIXES f TO l MINUTES OF THE 306TH ACRS MEETING OCTOBER 10-12. 1985 l ACRs- E36 6 I

i

APPENDIX I NRC STAFF ATTENDEES 306TH ACRS MEETING Thursday, October 10, 1985 0FFICE OF NUCLEAR REACTOR REGULATION OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS T. King S. M. Coplan S. Sands P. Williams R. Hernan DIVISION OF HUMAN FACTORS SAFETY E. G. Adensam H. Thompson D. Persinko C. Thomas E. D. Throm 0FFICE OF NUCLEAR REGUATORY C. McCracken RESEARCH W. Raulson A. DeAgazio F. A. Gastanzi G.W. Knighton M. C. Ley REGION III M. Rubin F. J. Miroglia I. N. Jackiw R. Wessman O. D. Parr L. Rubenstein REGION II R. M. Bernero D. C. Caletti D. Verreli B. Hardin i

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i

INVITED ATTENDEES

305TH ACRS MEETING l 4

! Thursday, October 10, 1985 TOLEDO EDISON COMPANY BABCOCK & WILCOX  ;

2  ;

j J. Wood J. Williams J. H. Taylor  !

i R. Peters M. S. Fertel B. Dunn  !

, S. Seny R. Borsum i J. Hirsch

S. Jaw i' J. Lingerfelter SHAW PITTMAN l l T. Myers i

D. Lewis i TENNESSEE VALLEY AUTHORITY i j '

- K. Whitt J. Hughan  !

! R. Parkey

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J. Hufham  :

I C. Elke I

!' BECHTEL POWER CORPORATION I i

J. Ray

! J. Fay i

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PUBLIC ATTENDEES 305TH ACRS MEETING Thursday, October 10, 1985 J. Hannah, Associated Press G. Teal, BLCPAR D. L. Foreman, General Electric R. Villa, General Electric F. K. Alderson, Arizone Public Service Company C. B. Brinkman, Combustion Engineering L. Connor DSA K. Troxler, DLC K. Campbell, NUS R. Dali Rogers, Rockwell International J. Nurmi, Qatel A. J. Pressesky, ANS D. L. Foreman, General Electric M. Coffman, CAP Brodcast D. Feinstein, Wash. Radio Press M.A. Kos, Capitol Broadcast O B. Slepp, Capitol Broadcast S. Savage, NUS P. Hildebrandt, MPR Associates O

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f

} t NRC STAFF ATTENDEES j

l i- 306TH ACRS MEETING t

Friday. October ll,1985 i OFFICE OF NUCLEAR REACTOR REGULATION j l

L. Soffer I R. Hernan .

C. Thomas

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a D. C. Scaletti i

j M. Rubin l t

G. P. Marino I i

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INVITED ATTENDEES 307TH ACRS MEETING Friday, October 11, 1985 FEMA _

, M. E. Sanders

J. Rumbarger W. R. Cunning i

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i a

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PUBLIC ATTENDEES 307TH ACRS MEETING Friday. October 11,1985 P. Cybulskis, Battelle A. J. Pressesky, ANS X. Campbell, NUS F. R. Norm Bechtel J. Nurmi, Qatel R.Borsum, Babcock & Wilcox J. O. Berga, EPRI K. A. Troxler, DLC M. Danam, SPP P. F. Riehm, KMC A. Emrich, DuPont-DOE R. Villa, General Electric D. Foreman, General Electric S. Savage, NUS M. Blatt, Consolidated Edison (NY)

D.'.u e dC

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APPENDIX II O FUTURE AGENDA -

November ACRS Meeting .

/

Palo Verde Nuclear Generating Station--Briefing 3 hrs + letter by the NRC Staff regarding the operating expeafence and the startup test program for Unit 1 Committee to Review Generic Requirements--Briefing I hr by Chairman, CRGR, regarding activities of this NRC Committee General Electric Standard Safety Analysis Report 4 hrs (GESSAR II) -- Continue ACRS review of the FDA for this standardized NSSS Report of the ACRS Subcomittee on the State of deferred to Nuclear Power Plant Safety -- Discussion of the December most significant safety-related issues in need of resolution ,

Meeting with NRC Comissioners to discuss ACRS 1 1/2 hrs report on consideration of extreme environmental n

v phenomena in emergency planning Briefing by NRC Staff of Recent Events at Operating Nuclear Power Plants 2 hrs Status briefing by IE representatives regarding the I hr Development of Outage Inspection Program Beaver Valley Nuclear Power Station Unit 2 -- OL 3 hrs review Safety Goal Policy--ACRS r= view / comment on the 5 hrs

draft EDO paper on the two-year trial use o' the 1983 Comission Safety Goal Policy and tho' future use of quantitative safety goals Use of Aptitude Selection Testing in the Nuclear Power Industry -- ACRS comment Seismic Design Margins--H. W. Lewis letter 1/4 hr regarding overemphasis on seismic risk USI A-17. " Systems Interactions in Nuclear Power deferred Plants" -- Review /coments on NRC Staff's proposed resolution of USI A-17 V

/ -7

p (V I Reports of ACRS Subconnittees Joint Reactor Radiological Effects and Fire deferred to Protection Subcommittees regarding their review December of the increased N-16 radioactivity and need for  ;

fire protection when using hydrogen addition to BWRs to reduce IGSCC Joint Waste Management and Metal Components deferred Subcommittees regarding their review of several key Waste Management issues Subcommittee on ACRS Procedures and Activities deferred regarding the remaining recommendations of the Panel on ACRS Effectiveness Requalification of Nuclear Power Plant Operators-- deferred to Reply to Commissioner Asselstine's inquiry dated December February 21, 1985 Discussion of specific topics needing attention deferred as part of the proposed NRC long-range program i \

Alternate Proposals for Incident Investigation-- deferred Discuss alternate proposals of the ASLBP and OPE regarding investigation of nuclear accidents and j incidents Report of ACRS Subcommittee on Reliability deferred Assurance regarding reliability of valves

! \

4

APPENDIX gli

=== <0//p>/r4 O ACRS SUBCOMMITTEE MEETINGS

_ Joint Reactor Radiological Effects and Fire Protection, October 18, 1985, 1717 H 5treet, NW, Washington, DC (Merrill/Aldeman), 8:30 a.m., Room 1046.

The Subcommittees will review the increased N-16 radioactivity and fire protection problems associated with hydrogen addition to BWRs to reduce Intergranular Stress Corrosion Cracking in reactor coolant piping.

Dr. Moeller Dr. Okrent Dr. Carbon Mr. Reed Mr. Ebersole Dr. Siess Mr. Michelson Mr. Wylie Joint Waste Management and Metal Components. October 24 and 25, 1985, 1717 H 5treet, NW, Washington, DC (Merrill/Igne), 8:30 a.m., Room 1046. The Subcommittees will review: (1) High-Level Waste Program Programatic Overview and Approach -- Products, Activities and Schedules; (2) Definition of High-Level Radioactive Wastes; (3) NRC Staff's General Technical Approach to Identify Licensing Information Needs -- Overview of Perfomance Assessment Methodologies and Issues; (4) Final Waste Form Package Reliability Generic Techni.:al Position; and (5) NRC High-Level Radwaste Form and Container Materials Research and Technical Assistance Programs.

Dr. Moeller Dr. Carter

/n Dr. Carbon V) Mr. Etherington Dr. Mark Dr. Kassner Dr. Parker Dr. Steindler Dr. Shewmon Dr. Clark Beaver Valley Power Station Unit 2, Site visit October 31 and subcommittee meeting November 1,1985, Pittsburgh, PA (Aldeman), 8:30 a.m. The Subcomittee will review the application of the Duquesne Light Company for an operating license.

Mr. Wylie Dr. Kerr Mr. Ebersole Dr. Remick Regulatory Policies and Practices, November 1, 1985, 1717 H Street, NW, l Washington, DC (Cappucci), 8:30 a.m., Room 1046. The Subcomittee will l

discuss SECY-85-208 and recomendations made by Judge Cotter of the ASLBP, and OPE related to the establishment of an incident investigation organization within NRC.

Dr. Lewis Mr. Michelson l Dr. Remick i

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A9

O Reactor Operations, November 4, 1985, 1717 H Street, NW, Washington, DC (Major), 1:00 p.m. Room 1046. The Subcomittee will review recent operating experience.

Mr. Ebersole Mr. Reed Dr. Kerr Dr. Remick Mr. Michelson Mr. Ward Dr. Okrent_ Mr. Wylie Long Range Plan for NRC, November 6, 1985, 1717 H Street, NW, Washington, DC (Major), 8:30 a.m., Room 1046. The Subcomittee will continue discussion oii developing coments on a long range plan for the NRC. Topics to be discussed are primarily technical issues related to the regulation of '.

nuclear power plant safety and safety regulation over the next 5 to 10 years.

Dr. Carbon Dr. Remick (p/t)

Dr. Lewis (p/t) Mr. Wylie Dr. Moeller Joint Reliability and Probabilistic Assessment and Safety Philosophy, Tech-nology,andCriteria, Washington, November DC (Savio), 9:00 a.m. , Room 6,1985(tentative),1717H5treet,NW,(1 1046. The Subcomittees will:

(m continue the review of the two-year trial use of the Proposed Safety Goal s

Policy (2) review the NRC Staff proposed resolution for USI A-17. " System Interactions in Nuclear Power Plants," and (3) review the status of the ongoing NRC Staff work on steam generator overfill.

Dr. Okrent Dr. Mark Dr. Kerr Dr. Remick (p/t)

Dr. Lewis (p/t) Dr. Siess Management, November 6,1985, Washington, DC, (Fraley), 2:00 p.m., Room 1010/1008. The Subcomittee will discuss priorities for the December ACRS meeting agenda and the organization of Committee assignments.

Mr. Wa rd Mr. Ebersole Dr. Lewis (p/t) 307TH ACRS MEETING, November 7-9, 1985, Washington, DC, Room 1046.

Procedures and Administration, November 12, 1985 Washington, DC (Fraley),

8:30 a.m., Room 1046. The Subcommittee will continue its discussion of the recommendations of the Panel on ACRS Effectiveness and reorganization of subcomittee assignments.

Mr. Ward

, Mr. Ebersole

( Dr. Lewis Dr. Siess 4 -10

Millstone Nuclear Power Station Units 1, 2 and 3, November 18 and 19, 1985 Waterford, CN (Schiffgens),1:30 p.m. on Nov.18 and 8:30 a.m. on Nov.19.

The Subcommittee will review the Northeast Nuclear Energy Company's application for conversion of the Provisional Operating License (POL) for Millstone Unit I to a Full Tenn Operating License (FTOL).

Dr. Shewmon Mr. Ward Dr. Moeller Emergency Core Cooling Systems, November 22, 1985, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 a.m., Room 1046. The Subcommittee will continue its review of the proposed revision of 10 CFR 50.46 and Appendix K.

Mr. Ward Dr. Catton Mr. Ebersole Mr. Schrock Mr. Michelson Dr. Sullivan Mr. Reed ~ Dr. Theofanous Dr. Tien Human Factors, November 25 & 26, 1985, 1717 H Street, NW, Washington, DC (Schiffgens), 8:30 a.m., Room 1046. The Subcommittee will complete its review of current reactor operator requalification procedures and initiate review of proposed final rulemaking on 10 CFR 55 and three related Regulatory Guides.

N Mr. Ward Dr. Remick Mr. Michelson Mr. Wylie Decay Heat Removal Systems, December 2 (tent.) & 3, 1985, 1717 H Street, NW, Washington, DC (Boehnert), afternoon only on 12/2 and 8:30 a.m. on 12/3, Room 1046. On Dec. 2 the Subcomittee will discuss the issue of AFW reliability, and on Dec. 3 the Subcomittee will continue the review of the NRR resolution position for USI A-45, " Shutdown Decay Heat Removal Require-ments."

Mr. Ward Mr. Reed Mr. Ebersole Dr. Catton Mr. Etherington Mr. Davis Mr. Michelson Qualification Program for Safety-Related Equipment, December 4,1985 (tenta-tive), 1717 H Street, NW, Washington, DC (Cappucci), 8:30 a.m., Room 1046.

The Subcomittee will discuss resolution and implementation of USI A-46. ,

Mr. Wylie Mr. Reed i Mr. Ebersole Dr. Siess Mr. Michelson Mr. Ward

% Dr. Lipinski T

d _308TH ACRS Meeting, December 5-7, 1985, Washington, DC, Room 1046.

/ -

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a Emergency Core Cooling Systems, December 10 and 11, 1985 Palo Alto, CA (Boehnert), 8:30 a.m.

The Subcommittee will continue the review of the joint NRC/B&WOG/EPRI/B&W joint IST Program. A visit is planned to the EPRI-sponsored facilities supporting this Program and the Stanford Research Institute and Science Applications, Inc.

Mr. Ward Dr. Catton Mr. Ebersole Mr. Schrock Mr. Etherington Dr. Sullivan Mr. Michelson Dr. Tien Mr. Reed Dr. Theofanous Mr. Wylie Quality and Quality Assurance In Design and Construction. December 13, 1985, Washington, DC (Major), 8:30 a.m., Room 1046. The Subcomittee will discuss with the NRC Staff such programs as CAT, IDVP, TDI, and readiness review to ensure quality in nuclear plant design and construction. Further, a discussion with the Staff of their program to deal with allegations at the OL stage (e.g., Comanche Peak). Emphasis should be on comparing the resources required by the various programs and the effectiveness of the programs in assuring quality of plant design, construction and readiness for operation.

('g Dr. Remick Dr. Siess i Mr. Michelson Mr. Ward Dr. Okrent Mr. Wylie Mr. Reed Safety Research Program, (CLOSED), January 8,1986, Washington, DC, (Duraiswamy), 8:30 a.m. , Room 1046. The Subcomittee will discuss the NRC Safety Research Program and Budget for FY 1987 and the OMB initial mark, and gather information for use by the ACRS in its preparation of the annual report to the Congress on the related matter.

Dr. Siess Dr. Moeller Dr. Carbon Dr. Okrent Dr. Kerr Dr. Shewmon Dr. Mark Mr. Ward Dr. Michelson Safety Research Program, February 12,1986(tentative), Washington,DC, (Duraiswamy), 8:30 a.m., Room 1046. The Subcommittee will discuss the OMB final mark and a final draft of the ACRS report to the Congress.

Dr. Siess Dr. Moeller i Dr. Carbon Dr. Okrent Dr. Kerr Dr. Shewmon Dr. Mark Mr. Ward q Dr. Michelson

! A - / 2-l

-E-Westinghouse Water Reactors. (CLOSED), Date to be detennined (late November).

Washington, DC (Cappucci). The Subcommittee will begin the PDA review of the Westinghouse Advanced Pressurized Water Reactor (RESAR SP/90).

Mr. Ebersole Dr. Shewmon Mr. Etherington Mr. Ward Mr. Michelson Mr. Wylie Dr. Siess Mr. Davis Human Factors, Date to be determined (November), Washington, DC (Schiffgens).

The Subcommittee will explore methods for deciding what actions should be automated in nuclear power plant operation.

Dr. Ward Dr. Remick Mr. Reed Mr. Wylie Mr. Gimmy Fort St. Vrain, Date to be determined (November / December), near Longmont, CO (McKinley). The Subcommittee will tour the facility, explore technical Poblems addressed during the recent extended outage, and discuss management changes made as a result of the licensee's independent assessment of manage-ment controls.

T Dr. Siess Dr. Kerr Q Dr. Carbon Mr. Ebersole Dr. Remick Mr. Ward Reliability and Probabilistic Assessment, Date and location to be detemined (Fall, tentative) (Savio). The Subcommittee will review the probabilistic risk assessment for Millstone 3.

Dr. Okrent Mr. Michelson Dr. Kerr Dr. Siess Mr. Ebersole Mr. Ward Dr. Lewis Mr. Wylie Dr. Mark

_ South Texas Units 1 and 2, Date to be detennined (January), Washington, DC (El-Zeftawy). The Subcommittee will review Houston Lighting and Power Company's application for an operating license.

Dr. Mark Dr. Lewis Dr. Axtmann Mr. Michelson Mr. Ebersole Dr. Siess

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.O Scram Systems Reliability, Date to be detemined , Washington, DC (Boehnert).

The Subcomittee will discuss scram breaker reliability for B&W and CE plants.

The Subcommittee will also continue the review of the status of ATWS Rule implementation effort and related issues.

Dr. Kerr Mr. Davis Mr. Ebersole Dr. Lee Mr. Ward Dr. Lipinski Mr. Wylie CE Nuclear Plants, Date to be determined, Washington, DC (Boehnert). The Subcommittee will discuss the issue of rapid depressurization for CE plants without PORVs.

Mr. Wylie Dr. Lewis Mr. Ebersole Mr. Reed Dr. Kerr E

O A-M -- -

f~.s SCHEDULE OF ACRS SUBCOMMITTEE MEETING U

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS OCTOBER 18, 1985 JOINT REACTOR RADIOLOGICAL EFFECTS (MERRILL/ ALDERMAN)

AND FIRE PROTECTION Moeller, Carbon, Ebersole, Michelson, Okrent, Reed, Siess, Wylie PURPOSE:

To review the increa ed N-16 radioactivity and fire protection problems in using H 2 addition tr. BWRs to reduce IGSCC in reactor coolant piping.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Comments to Staff. Needed by no specific date.

What will be done at this meeting?

p Review of H 2 addition regarding N-16 activity and fire hazards.

What would be the consequence of postponing this meeting?

Meeting suggested by Committee members. Postponement of meeting will not have any serious consequences.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Paper by V. Benaroya on Reactor Coolant System Chemistry Control.

2. EPRI workshop report on subject.

3.

NUREG/CR-3551, " Safety Implications Associated With In-Plant Pressurized Gas Storage and Distribution Systems in Nuclear Power Plants."

(The at,ove reports sent to members 10/2/85).

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS OCTOBER 24 & 25, 1985 JOINT WASTE MANAGEMENT (MERRILL/IGNE)Moeller, AND METAL COMPONENTS Carbon, Etherington, Mark, Shewmon Cons.: Carter, Kassner, Parker, Steindler, Clark PURPOSE: To review: (1) High-Level Waste Progran: Programmatic Overview and Approach -- Products, Activities, and Schedules.

(2) Definition of High-Level Radioactive Wastes.

(3) NRC Staff's General Technical Approach to Identify Licensing Information Needs -- Overview of Performance Assessment Methodologies and Issues.

(4) Final Waste Form Package Reliability Generic Technical Position.

(5) NRC High-level Radwaste Fom and Container Materials Research and Technical Assistance Programs.

LOCATION: WASHINGTON, DC

'ACKGROUND:

What action is requested; by what date is it needed?

The topics to be reviewed in this meeting were arrived at by mutual discussion among ACRS, RES, and NMSS representatives to facilitate the ACRS in its oversight of the NRC High-Level Radioactive Waste Program.

, What will be done at this meeting?

Review subjects named above.

! _W hat would be the consequence of postponing this meeting?

L Lose advantage of timeliness and ensuring that programs are properly directed and properly providing meaningful results. No adverse consequences.

l _ PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

! 1. SECY-85-309, Advance Notice of Proposed Rulemaking,10CFR60 -- Definition of the i term "High-Level Radioactive Wastes," September 17, 1985 i 2. Final Waste Fonn Package Re11 ability Generic Technical Position. (To be presented ,

and distributed at the meeting.)  ;

3. Memo for D. Moeller fm P. Shewmon dtd. August 23, 1985,

Subject:

Review of Container Material for HLW, Battelle Columbus, August 22, 1985.

). NUREG/CR-3900, Volumes 1-4 (and later volumes if avaf f able), Long-Term Perfonnance

,V of Materials Used for High-Level Waste Packaging NRC Contract No. 04-82-015.

5 Status Report will be provided prior to meeting.

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/ ._N SCHEDULE OF ACRS SUBCOMMITTEE MEETING O

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEM ERS NOVEMBER 1, 1985 BEAVER VALLEY 2 (ALDERMAN) Wylie, (OCTOBER 31, 1985, site visit) Ebersole, Kerr, Remick PURPOSE: To review application for an operating license.

LOCATION: PITTSBURGH, PA BACKGROUND:

What action is requested; by what date is it needed?

Review application for operating license - ACRS letter desired November 1985 meeting What will be don'e at this meeting?

Visit the site and review readiness for Beaver Valley II for operating license.

What would be the consequence of postponing this meeting?

\ Slippage of license.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. SER received Oct 'r 11, 1985.

2. Tentative Schedule ant to members 10/2/85.

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, o SCHEDULE OF ACRS SUBCOMMITTEE MEETING I

(G DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS NOVEMBER 1, 1985 REGULATORY POLICIES (CAPPUCCI) Lewis, AND PRACTICES Michelson, Remick PURPOSE:

To discuss recomendations in SECY-85-208 and those made by Judge Cotter of the ASLBP, and OPE related to the establishment of an incident investigation organization within NRC.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Report to full Committee, November 1985 meeting (307th)

O What will be done at this meeting?

Review proposals mentioned above and prepare report to full Comittee.

What would be the censequence of postponing this meeting?

No report to the full Comittee.

l PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Memo, Cappucci to Lewis, August 28, 1985 with attachments.

2. Memo, Cappucci to RP&P Subcommittee, October 1, 1985 with attachments.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING.

b DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS NOVEMBER 4, 1985 REACTOR OPERATIONS (MAJOR)Ebersole,Kerr, (1 P.M.)

Michelson, Okrent.

Reed, Remick, Ward, Wylie PURPOSE:

To review recent operating experience.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Review operating experience for November 1985 ACRS meeting.

What will be done at this meeting?

Review recent operating experience.

.W hat would be the consequence of postponing this meeting?

None PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

Status report will be provided.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEPEERS NOVEMBER 6, 1985 LONG RANGE PLAN FOR NRC (MAJOR) Carbon, Lewis (p/t),Moeller, Remick (p/t), Wylie PURPOSE:

The Subcommittee will continue discussions on developing coments on a long range plan for the NRC. Topics under discussion are primarily technical issues related to the regulation of nuclear power plant safety and safety regulation over the next 5 to 10 years.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

The Committee expects to interface with the OPE /EDO effort on a Long-Range Plan.

,. .Also, under consideration are a series of White Papers dealing with selected topics.

What will be done at this meeting?

l i

Three interviews are currently planned: Fonner Comissioner, Peter Bradford; V. Stello, DEDROGR; J. Tribble, President, Yankee Atomic Co.

What would be the consequence of postponing this meeting?

Timeliness of effort would be effected. Would become out of phase with a parallel OPE effort on LRP.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

A Status Report will be prepared before the meeting.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE

, SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS NOVEMBER 6, 1985 JOINT RELIABILITY & PROBABILISTIC (SAVIO) Okrent, Kerr, (9:00A.M.) ASSESSMENT AND SAFETY PHILOSOPHY, Lewis (p/t), Mark.

TECHNOLOGY, AND CRITERIA Remick (p/t) Siess PURPOSE:

To (1) continue the review of the two-year trial use of the proposed Safety Goal Policy.

(2) review the NRC Staff proposed resolution for USI A-17, " System Interactions in Nuclear Power Plants," and (3) review the status of the ongoing NRC Staff work on steam generator overfill.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

ACRS coment on USI A-17 (before review by the CRGR) and ACRS coment on the EDO O October proposal forACRS revision tomeeting) at the the proposed Safety GoalNovember ACRS Policy (if not completed meeting at the What will be done at this meeting?

Review of the 3 items listed under " Purpose" above.

What would be the consequence of postponing this meetino?

Comment on the EDO paper has been requested before presentation of that paper to the Comission (currently scheduled for mid-November). Coments on USI A-17 could probably be deferred to the December ACRS without impacting on CRGR schedule.

, _ PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. EDO Paper on Safety Goal Policy - to be provided.

2. CRGR briefing package on USI A to be provided.

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. m SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. 8 MEMBERS NOVEMBER 12, 1986 PROCEDURES & ADMINISTRATION (FRALEY) Ward, Ebersole, Lewis, Siess PURPOSE:

The Subcommittee will continue discussion of the recomendations of the Panel on ACRS Effectiveness and reorganization of subcommittee assignments.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

What will be done at this meeting?

What would be the consequence of postponing this meeting?

O PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS NOVEMBER 18 & 19, 1985 MILLSTONE NUCLEAR POWER (SCHIFFGENS)Sheumon, (1:30 P.M. on 11/18 STATION UNITS 1-3 Moeller, Ward 8:30 A.M. on 11/19)

PURPOSE:

To review the Northeast Nuclear Energy Company's application for conversion of the T.ovisional Operating larm Operating License.License for Millstone Nuclear Power Station, Unit No. I to a Full LOCATION: WATERFORD. CONNECTICUT BACKGROUND:

What action is r'equested; by what date is it needed?

Subcommittee ACRS Conversion Meeting, December review 5-7, 1985. in time for full Committee consideration at the 308t p What will be done at this meeting?

Review the Conversion application and draft a letter.

What would be the consequence of postponing this meeting?

The NRC Staff schedule currently has license issuance on December 11, 1985 I don't think this schedule can be met.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. The Draft SER has been distributed.

2. APRAStudyhasbeenissued(wehaverequestedadditionalcopies).

3.

A Draft supplement to IPSAR (NUREG 0824) will be made available before the end of September.

4. A Draft ISAP report will be made available by mid-September.

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SCHEDULE OF ACRS SUBCOPNITTEE MEETING V

DATE SUBCOMMITTEE MEETING STAFF ENGR. 4 MEMBERS NOVEMBER 22, 1985 ECCS

' (80EHNERT) Ward, Ebersole, Michelson, Reed Cons.: Catton, Schrock, Sullivan, Theofanous. Tien PURPOSE: To continue the review of the proposed revision of 10 CFR 50.46 and Appendix K.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

m RES has scheduled submittal of revision proposal to Connission in December 1985.

What will be done at this meeting?

See above.

What would be the consequence of postponing this meeting?

Possible impact on submittal scheduled noted above.

PERTINENT DOCUMENTS AND THEIR AVAILABILITY:

1. Research report on justification for Rule revision.

2. SECY paper outlining proposed Rule revision.

3. Companion Draft Regulatory Guide.

(NOTE: 1-3 will be provided prior to meeting.

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O SCHEDULE OF ACRS SUBCOP9fITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. 8 MEMBERS NOVEMBER 25 & 26, 1985 HUMAN FACTORS (SCHIFFGENS) Ward, Michelson, Remick, Wylie PURPOSE:

The Subcomittee will complete its review of current reactor operator requalification procedures and initiate review of proposed final rulemaking on 10 CFR 55 and three related Regulatory Guides.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

The Subcomittee is to arrive at a consensus position on requalification with recomendations for Comittee consideration in relation to questions raised by Commissioner Asselstine.

The ACRS is comitted to respond to the Commission by September 24, 1985.

What will be done at this meeting?

1.

Conclude the review of current requalification procedures and prepare recomendations.

2.

Begin review of proposed new regulation in the area of licensing operators.

What would be the consequence of postponing this meeting?

We will go further off scMdule in our response to the Comission on this issue.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Proposed Final Rulemaking on 10 CFR 55,

2. Three related Regulatory Guides (1.134, 1.149, and 1.8).

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m SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS DECEMBER 2, 1985 (tent.) DECAY HEAT REMOVAL SYSTEMS (BOEHNERT) Ward, l (P.M. Only on 12/2) and Ebersole, Etherington, DECEMBER 3, 1985 Michelson, Reed t

Cons.: Catton, P. Davis PURPOSE:

Dec. 2 (tent.) - To discuss the issue of AFW reliability.

Dec. 3 - To continue the review of the NRR resolution position for USI A-45

" Shutdown Decay Heat Removal Requirements."

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

p f

No specific action date. Our review is consistent with NRR review schedule (See 3 below)

What will be done at this meeting?

Explore proposed NRR resolution position for this USI.

What would be the consequence of postponing this meeting?

Potential item.

impact on NRR resolution schedule and/or ACRS could become " critical path" What would be the consequence of postponing this meeting?

Potential loss of timely review of this USI.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

To be provided in near future.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING

! DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS DECMEBER 4, 1985 QUALIFICATION PROGRAM FOR SAFETY- (CAPPUCCI) Wylie, (tentative) RELATED EQUIPMENT Ebersole, Michelson, Reed, Siess, Ward Cons.: Lipinski PURPOSE:

To discuss the final resolution and implementation of USI A-46, Seismic Qualification of Equipment in Operating Plants (post public courent phase).

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

ACRS comments on the final resolution of USI A-46 following public comments.

What will be done at this meeting?

Review final resolution and implementation plan for USI A-46; prepare report to full Committee and suggested report to the Connission.

What would be the consequence of postponing this meeting?

Delay of ACRS comments to Comission.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

Letter fm K. Kniel to R. Fraley outlining public coments plans, dated July 23, 1985.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING

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DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS DECEMBER 10 & 11, 1985 ECCS (BOEHNERT) Ward, Ebersole, Etherington, Michelson, Reed, Wylie Cons.: Catton, Schrock, Sullivan.

Theofanous Tien PURPOSE:

Continued and review supporting of joint NRC/B&W Owners Group /EPRI/B&W Intergral System Test Program programs. A visit to the EPRI-sponsored SAI and SRI faci'lities supporting the work is also planned.

LOCATION: Palo Alto, CA BACKGROUND:

What action is requested; by what date is it needed?

No specific action date -- review is concurrent with NRC-RES program conduct.

What will be done at this meeting?

Continue above review and visit EPRI's supporting program test facilities.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

To be provided in near future.

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1 f'% SCHEDULE OF ACRS SUBCOMMITTEE MEETING G

DATE SUBCOMMITTEE MEETING STAFF ENGR. 8 MEMBERS DECEMBER 13, 1985 QUALITY & QUALITY ASSURANCE (MAJOR)Remick, IN DESIGN AND CONSTRUCTION Michelson, Okrent, Reed Siess, Ward, Wylie PURPOSE:

General discussion with the NRC Staff of such programs as CAT, IDVP, IDI, and readiness reviews to ensure quality in nuclear plant design and construction. Further, a discus-sion with with the Staff of their program to deal with allegations at the OL stage (e.g., Comanche Peak). Emphasis should be on comparing the resources required by the various programs and the effectiveness of the programs in assuring quality of plant design, constructi,on and readiness for operation.

_ LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

For infonnation; no due date What will be done at this meeting?

See purpose.

What would be the consequence of postponing this meeting?

None PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

Improving Quality and the Assurance of Quality in the Design and Construction of Nuclear Power Plants (NUREG-1055) r i

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7^N SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEM ERS JANUARY 8, 1986 SAFETY RESEARCH PROGRAM (DURAISWAMY)Siess, (CLOSED) Carbon, Kerr, Mark, Michelson, Moeller, Okrent, Shewmon, Ward PURPOSE:

The Subcommittee will discuss the NRC Safety Research Program and Budget for FY 1987 and the OMB initial mark, and gather information for use by the ACRS in its preparation of the annual report to the Congress on the related matter.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

ACRS needs to provide its report to the Congress during February.

What will be done at this meeting?

V A draft ACRS report will be prepared. (Also see purpose.)

What would be the consequence of postponing this meeting?

Preparation of the draft ACRS report will be delayed until February.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Commission's budget request to the OMB (Blue Book). (Sent to all ACRS members and staff on September 30,1985.)

2. OMB mark (will be available during December 1985).

m A-30

SCHEDULE OF ACRS SUBCOMMITTEE MEETING s

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MDEERS FEBRUARY 12, 1986 SAFETY RESEARCH PROGRAM (DURAISWAMY)Siess, (Tentative) Carbon, Kerr, Mark, Michelson, Moeller, Okrent, Shewmon, Ward PURPOSE:

The Subcommittee ACRS will discuss the OMB final mark and a final draft of the report to Congress.

LOCATION: WASHINGTON, DC BACKGROUND:

i What action is requested; by what date is it needed?

ACRS needs to finalize its report during the February full Committee meeting.

What will be done at this meeting?

See purpose.

"" ~ "'" "' '"' ' """'" ' ' '*" "'"' " '" "

(DACRSreporttoCongresswillbedelayed.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Draft ACRS report to the Congress.

2. OMB final mark.

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G SCHEDULE OF ACRS SUBCOP9tITTEE MEETING i DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED WESTINGHOUSE WATER (lateNovember) REACTORS (CAPPUCCI)Ebersole.

Etherington, Michelson, (CLOSED) Shewmon, Siess, Ward, Wylie Cons.: Davis PURPOSE:

To begin PDA review of Westinghouse Advanced PWR (RESAR SP/90).

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

ACRS letter on PDA approval by 11/86.

> \

What will be done at this meeting?

, Begin reviewing design modules.

What would be the consequence of postponing this meeting?

Delay in the completion of ACRS PDA review.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. RESARSP/90StandardPlantDesign(50-601).

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. A MEMBERS TO BE DETERMINED HUMAN FACTORS (SCHIFFGENS) Ward, (NOVEMBER)

Reed, Remick, Wylie Cons.: Gimy PURPOSE:

To explore methods for deciding what actions should be automated in nuclear power plant operation.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date ?s it needed?

Mr. Ward asked researchers from the University of Illinois to make a presentation to the Subcomittee.

What will be done at this meeting?

What would be the consequence of postponing this meeting?

No serious consequences from postponement that I can see.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

None at this time.

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--e .

SCHEDULE OF ACRS SUBCOMMITTEE MEETING J

DATE SUBCOMMITTEE MEETING STAFF ENGR. 8 MEM ERS TO BE DETERMINED FORT ST. VRAIN (McKINLEY) Siess, (NOVEMBER / DECEMBER) Carbon, Ebersole, Kerr, Remick, Ward PURPOSE:

To review helium circulatory bolting failures, PCRV tendon corrosion, control rod drive failures, control rod cable replacement, reserve shutdown material replacement, PSC management improvement actions, equipment qualification, and recent operating experience.

LOCATION: Near LONGMONT, C0 BACKGROUND:

What action is requested; by what date is it needed?

No action requested of ACRS,; ACRS would be perfonning its oversight function of monitoring major maintenance, operation, QA, and management functions.

. What will be done at this meetig O\ (seeabove)

What would be the consequence of postponing this meeting?

Appearance of ACRS disinterest in major management problems at a unique operating (?)

power reactor.

PERTINENT FUBLICATIONS AND THEIR AVAILABILITY:

1. Memo from J. C. McKinley to C. P. Siess dated March 27, 1985,

Subject:

" Management i at Fort St. Vrain."

l l

O e34

i SCHEDULE OF ACRS SUBCOMMITTEE MEETING O DATE SUBCOMMITTEE MEETING STAFF ENGR. 8 MEMBERS TO BE DETERMINED RELIABILITY AND PROBABILISTIC (SAVIO) Okrent, Kerr, (FALL) ASSESSMENT (tentative) Ebersole Lewis, Mark, Michelson, Siess, Ward, Wylie Cons.:

PURPOSE:

To review the PRA for Millstone 3 (not an OL critical path item).

LOCATION: To be determined BACKGROUND:

What action is reouested; by what date is it needed?

Review of the Millstone 3 PRA; the meeting is to be scheduled after the completion of the NRC Staff's review of the PRA (estimated to be by the end of May 1985).

There is no ACRS action date.

What will be done at this meeting?

Review of the Millstone 3 PRA for intermation.

What would be the consequence of postponing this meiting?

ACRS has stated that this review need not be completed prior to full power operation.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Millstone 3 PRA (distributed).

2. NRC Staff report on the results of the NRC/LLNL review of the Millstone 3 PRA 1

(expected by the end of May 1985).

O or . . ._ ._

7% SC!gL".E OF ACRS SUBCOMMITTEE MEETING

\

\

DATE SUBCOMMITTEE MEETING STAFF ENGR. 8 MEMBERS TO BE DETERMINED SOUTH TEXAS 1 & 2 (EL-ZEFTAWY) Mark, (JANUARY)

'Axtmann, Ebersole.

Lewis, Michelson, Siess PURPOSE: To review Houston Lighting and Power Company's application for an OL.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Issue an ACRS " full power" OL letter; at the January 1986 ACRS meeting.

What will be done at this meeting?

Subcommittee OL review in time for Comittee consideration at the January 1986 ACRS n meeting.

What would be the consequence of postponing this meeting?

Possible delay of South Texas full power operation.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

To be provided later. The NRC Staff anticipates publishing the SER on, or about.

November 15, 1985.

O t e 36

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

SCHEDULE OF ACRS SUBCOMMITTEE MEETING O.

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED SCRAM SYSTEMS RELIABILITY (80EHNERT) Kerr, Ebersole, Ward, Wylie Cons.: Davis, Lee, Lipinski PURPOSE:

To discuss scram implementation breaker reliability for B&W and CE plants, and the status of ATWS Rule effort.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed? N/A What will be done at this meeting? See Purpose What would be the consequence of postponing this meeting?

No significant consequences PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

To be provided.

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l X SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE 1 SUBCOMMITTEE MEETING STAFF ENGR. 8 PEMBERS TO BE DETERMINED CE NUCLEAR PLANTS (80ENNERT)Wylie.

Ebersole, Kerr, Lewis, Reed PURPOSE:

Discuss the issue of rapid depressurization for CE plants without PORVs.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

ACRS assigned issue to CE Subconnittee as a result of discussion of issue at recent meeting.

What will be done at this meeting?

L4egin discussion of the subject issue to see what action if any ACRS should take.

What would be the consequence of postponing this meeting?

No consequence

_ PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

To be provided

&n -_ +---

(

\

October 12, 1985 ACRS Members ACRS Staff ACRS Fellows The meeting to be held on November 5,1985 on CE/Palo Verde was inadvertently left out. Please attach or include this sumary of the meeting to your Revised List of Subcommittee meetings dtd. 10/12/85.

4 CE/PALO VERDE, November 5, 1985, 1717 H Street, NW, Washington, DC (Boehnert/ Houston), 8:30 a.m., Room 1046. The Subcomittee will review: (1)

Arizona Nuclear Power's application for Palo Verde. Unit 2, and (2) portions of CE's design of decay heat removal system.

Mr. Ebersole Mr. Ward Mr. Michelson Mr. Wylie Mr. Reed O

A-39

SCHEDULE OF ACRS SUBCOMMITTEE MEETING (hv)

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS NOVEMBER 5, 1985 CE/PALO VERDE (80EHNERT/ HOUSTON)

Ebersole, Michelson, Reed, Ward, Wylie PURPOSE: (1) To review Arizona Nuclear Power's application for Palo Verde. Unit 2.

(2) To review portions of Combustion Engineering's design of decay heat removal systems.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

The Committee requested a review of startup experience on Palo Verde. Unit 1 prior to fuel load at Unit 2. Unit 2 fuel load is tentatively scheduled for November 25, 1985.

What will be done at this meeting?

Review the startup experience on Palo Verde. Unit I and resolution of two major by problems: (1) loss of auxiliary spray in pressurizer, and (2) loss of offsite power.

Also review CE's design of decay heat removal systems.

What would be the consequence of postponing this meeting?

Delay the licensing schedule for Unit 2.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Memo from D. Houston to J. Ebersole dated October 9, 1985,

Subject:

Items for Consideration Regarding the Committee Position on Palo Verde Unit 2.

O

/-40

6 APPENDIX IV TVA PRESENTATION TO THE ACRS l I

TENNESSEE VALLEY AUTHORITY PRESENTATION TO ADVISORY COMMITTEE ON REACTOR SAFEGUARDS ON CHANGES TO THE TVA NUCLEAR ORGANIZATION OCTOBER 10,1985 9

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MAN AGER OF POWER & ENGINEERING H. O. PARRIS EXECUTIVE ASSISTANTS w

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CENTRAL STAFFS I I I I I I OFFICE OF OFFICE OF OFFICE OF OFFICE OF OFFICE OF PROJECT MAN AGEMENT ENERGY USE POWER OPERATIONS ENGENEERneG CONSTRUCTIOps NUCLEAR POWER MAJOR PROJECTS Effects,e Prior to July S.1985

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. ,. I POWER AND ENGINEERING ORGANIZATION (SUPPLY AND USE)

GENERAL MANAGER .

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MANAGER POWER & ENGINEERING (SUPPLY & USE)

R.C.STEFFY,JR.

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STAFFS I I DIVISION OFI DIVISION OF POWER ENERGY:

ENERGY USE &l OPERATIONS DEMONSTRATIONS DISTRIBUTOR'

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= OPERAT*ONS AND ENGINEERING SUPERINTENDANT M MAINTENANCE

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OBJECTIVES FOR THE TENNESSEE VALLEY AUTHORITY NUCWAR PROGRAM

= CONSOUDATE OUR NUCLEAR RESOURCES UNDER A TIGHTLY FOCUSED U

• ACQURE, RETAIN, AND TRAIN MANAGEMENT TALENT TO EFFECTIVELY MANAGE OUR NUCLEAR ACTIVITIES.

j ESTABUSH PRORTIES SO THAf WE UMIT OUR AC11VITIES TO THOSE THAT WE HAVE THE CAPABluTY TO EXECUTE IN AN EXCELLENT MANNER.

• DEVELOP A TEAM OF EXPERENCED AND QUALFIED KEY PERSONNEL TO P LEADERSHP AND DIRECTION TO OUF1 NUCLEAR PROGRAM.

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u u UA"AM MANAGER PROJECT MANAGEMENT NUCLEAR OPERATIONS ENOINEERING AND CONSTRUCTIOtl ORGANIZATION C.C.WASON W. R. SROWN (NEW POSITION) J. P. DARLING STAFF 1

I E l MANAGER SROWNS FERRY OFF Q OFFICE OF SEQUOYAH WATTS SAR

- NUCLEAR SERVICES SITE OsRECTOR SITE OIRECTOR SITE DIRECTOR

( ACTING) J. W. HUTTON J. A. COFFEy H. L. ABERCROMSIE R. W. C ANTRELL W. T. COTTLE CHARLE S BONINE. JR.

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SUMMARY

OF MAJOR CHANGES TO THE TENNESSEE VALLEY AUTHORITY O NUCLEAR POWER PROGRAM 1b

• SINGLE MANAGER PLACED IN CHARGE OF NUCLEAR PROGRAM {

• MANAGEMENT TEAM ESTABLISHED TO PROVIDE OVERALL DIRECTION

, TO PROGRAM

- CONSOLIDATED RESPONSIBILITIES FOR OPERATIONS UNDER A SINGLE MANAGER

- CONSOLIDATED RESPONSIBILITIES FOR ENGINEERING / CONSTRUCTION.-

UNDER SINGLE MANAGER ELEVATED THE MANAGER OF QUALITY ASSURANCE

- ELEVATED THE MANAGER OF NUCLEAR LICENSING AND ELIMINATED RESPONSIBILITIES NOT DIRECTLY RELATED TO, LICENSING

- CORPORATE ENTITY ESTABLISHED TO SET POLICY AND GUIDE TOTAL NUCLEAR PROGRAM TO A COMMON GOAL 1

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• OBTAINED (OBTAINING) ADDITIONAL TALENT TO PROVIDE EXPERIENCE / LEADERSHIP TO THE PROGRAM

l. - ESTABLISHED AND FlLLED MANAGER, NUCLEAR OPERATIONS

- ESTABLISHED AND FILLED SPECIAL ASSISTANT TO THE MANAGER, NUCLEAR ,

' OPER ATIONS

- RECRUITING FOR THE POSITIONS OF:

MANAGER, OFFICE OF ENGINEERING AND CONSTRUCTION SITE DIRECTOR, BROWNS FERRY NUCLEAR PLANT i OTHERS O

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IIRC ACTIVITIES REGARDING A AG MENT O

NRC ACTIVITIES REGARDING TVA MANAGEMENT SECY 85-231 IDENTIFIED CONCERNS (06/28/85)

TVA PROVIDED DRAFT ORGANIZATIONAL CHANGES (08/29/85)

TVA PRESENTATION ON MANAGEMENT CHANGES (09/06/85)

SALP MEETING WITH SENIOR NRC EXECUTIVE BOARD (09/10/85)

COMMISSION MEETING ON TVA (09/12/85) 50,5f4(F) LETTER ADDRESSED CORPORATE ACTIVITIES (09/17/85)

TVA REQUESTED MEETING TO ADDRESS RESPONSES (MID OCTOBER)

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/-50

, m oix v1 Egg eoison enesenrariou n '

U Toledo Edison Presentation to ACRS october 10,1985 Joe Williams, Jr. Introduction Senior Vce President Nuclear John Wood Eventinvestigation Mechanical / Structural (Equipmentinvestigation) c.ngineeringManager SushilJain Decay Heat Removal SeniorNuclearEngineer Jacque Lingenfelter Safety Review and Restart Test Operations Engineering Program 1

Manager Joe Williams, Jr. Closing Remarks O

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Reassignment of PEP and SALP Improvement Program Activities High Priority-Will Receive Commensurate Emphasis and Resources:

m Prepare Detailed Position Descriptions for New Organization u Merit Review and Salary Administration Program _

i a Configuration Management .

u Management Training l

a Management By Objectives a Fire Protection m Nuclear Mission Procedures m QA Awareness Program u Non-outage Work Prioritization l = STA Assume interim EDO Function l

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Configuration Management a Component / System data base a System descriptions / design basis a Validated vendor manuals a Contro1 of drawings and manuals a Accurate spare parts allowance O

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O V i Equipment Concems Group A a Steam Feedwater Rupture Control System a Auxiliary Feedpump Turbinos e Auxiliary Feedwater Turbine Trip & Throttle Valves a Main Steam Headers a Main Feedwater Startup Control Valve a Auxiliary Feedwater Pump # 1 Suction Supply a Main Steam Valve MS-106 m Turbine Bypass Valve a Safety Parameter Display System Group B

= Main Feedpump Turbine a Auxiliary Feedwater Valves AF 599 and AF 608 l m Pilot Operated Relief Valve a Nuclear Instrumentation Neutron Source Range Detectors l

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(MFPTJ Concern: Overspeed tripping of MFPT 1-1 initiated a plant runback.

Findings: Failed circuit board capacitorin i General Electric controlsystem.

Corrective 1. Replace faulted board.

Actions: 2. Check and test control circuits for both MFPT 1-1 & 1-2.

Generic None-problem is specific to MFPT i

implications: control circuits.

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O Auxiliary Feedwater Valves AF 599 and AF 608 Concern: Valves failed to open on demand after closing earlier-would have prevented auxiliary feedwater flow.

Findings: Motor operators on valves were not properly adjusted allowing valves to " torque out".

1 Corrective 1. Readjust AF 599 and AF 608.

Actions: 2. Evaluate and readjust other motor operated valves.

! 3. Test valve operations.

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4. Provide new maintenance procedures.

J Generic: Applicable to other motor operated implications: valves

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Pilot Operated Relief Valve (PORVi Concern: During transient PORV failed to close properly after third opening

-closure of the block valve isolated the PORV and it reseated.

Findings: No physical evidence found to explain improperclosure-foreign materialin pilot cannot be ruled out-performance similarto industry experience.

l Corrective 1. Testing of valve - old/new.

Actions: 2. Add acoustic monitor flow indication light on PORV

, control panel.

3. Change PORV annunciator light from white to red.

4. Improve panellabeling of solenoid open/close switch.

l 5. Provide for PORV exercising l

during shutdowns.

l Generic None-no valves of similar l Implications: design.

O A - (o 3

O NuclearInstrumentation Neutron Source Range Detectors

Concern

Prior to event NI-1 was inoperable and NI-2 failed during transient - previous problems had been experienced.

Findings: Ni-1-inadequate grounding of shield found at preamp due to paint and lack of starwashers.

NI-2-intermittent failure of containment penetration cable center conductor.

I Triax cable connectors also found I degraded in each detector string.

Corrective 1. NI-1-proper ground established .

Action: 2. NI 2-using spare penetration. .

3. Replacing / refurbishing connectors as required.

Generic Preventeive maintenance program I implications: needed for source range, intermediate range, and power range connectors. ,

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O Decay Heat Removal Reliability improvement Program  ;

Task Force Effort a Chartered to review all systems used for decay heat removal.

Main Feed and Steam AFW SUFP SFRCS _

Feed and Bleed s identified changes to improve operational reliability and to reduce complexity of

! O' SFRCS.

I a Broad Membership a Experience in design, engineering, .

operations. L sincluded outside expertise:

MPR Associates Babcock and Wilcox  ;

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O Decay Heat Removal Reliability Task Force Objectives:

a Reduce frequency of demand for emergency decay heat removal.

m Reduce number of automatic system responses required to initiate auxiliary feedwater.

m Reduce potential for common mode failure.

m Evaluate diverse and redundant means of decay heat removal.

Goal:

a Provide equipment recommendation that would improve reliability of systems used for decay heat removal. Specific improvements for the AFWshould eventually achieve SRP reliability criteria.

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AFW/SFRCS Reliability Reduction of spuriousinitiators:

a Filter existing steam generator level signals.

siimprove SFRCS powersupply performance.

m Remove main steam and main feedwater isolation on SG low level.

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O AFW/SFRCS Reliability AFW initiation to SG-improvements:

m Valve motor operator improvements.

m Main flowpath valve reductions.

m Provide hot steam lines to AFW pumps.

m SFRCS panel revision.

m PGG governor a Remove /resize pump suction strainers.

m Suction transfer O

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l Actions Conceming Motor Driven Feed Pump Pre-Start Up Install new motor driven feed pump prior to startup.

New pump design features:

a Provides 100% capacity auxiliary feedwater flow, a Pump discharge aligned to the auxiliary feedwater headers during normal full power  :

operation. -

a Pump suction normally from the condensate storage tank.

m Pump capable of being started from the Control O Room.

m Pump motor can be supplied from either emergency diesel generator following a loss of offsite power.

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Longer Term Decay Heat Removal Reliability improvements Feed and Bleed:

a Primary system depressurization improvements.

• Additional high pressure injection capability.

AFW/SFRCS:

a Reconnect replaced startup feed pump.

m increase margin between SFRCS trip and ICS low level limit.

m improved AFW level control.

• SFRCS logic revision to further minimize isolation.

a Control Room " mimic" panel for finalized AFW/SFRCS.

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2) Cutler Hammer Guard With Clear Piastic Sliding Window Cat # E30ER32 (Red)

A .- - .. . -__ _ - _ _ -

O System Review and Restart Test Program Program Objectives For systems important to safe operation the objectivesincluded:

i eidentify significant or recurring maintenance and operations problems.

s identify testing required to assure that systems will perform their specified functions.

m Conduct a test program to assure that the systems are fully functional.

O 4-73

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't System Review Program Specific Systems included  :

Group 1 Reactor Coolant System High Pressure injection Core Flooding System Decay Heat Removaland Low Pressure injection Containment Spray System Containment Emergency Ventilation

. Containment Air Cooling and Hydrogen

Control Makeup and Purification System i Group 2 Electrical 125/250 VDC(includes Battery Room H&V) j Electrical 4.16 KV System (13.8/4.16 KV Transformers)

Electrical 480 V Distribution (includes inverters and Required Transformers)

Electrical 13.8 KV System (includes Startup and Auxiliary Transformers)

Emergency Diesel Generators (includes "Q" Fuel Oil Tanks and Diesel Room Ventilation)

Instrument AC Power (includes inverters and Required Transformers)

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System Review Program Specific Systems included (Cont'd)

Group 3 Anticipatory Reactor Trip System Control Rod Dr.ve Control System Incore Monitoring (includes Core Exit TC)

Reactor Protection System Steam and Feedwater Rupture Control System ,

Safety Features Actuation System j integrated Control System Security System Group 4 Control Room Normal and Emergency H&V Systems Station and instrument Air Station Fire Protection '

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O Steam Feedwater Rupture Control System (SFRCS)

Concern: Spurious SFRCS actuation closed both main steam isolation valves and isolated steam to main feedpump turbines.

Findings: Turbine trip caused pressure  ;

oscillations which SFRCS detected as low steam generatorlevel. Level pressure tap was made more sensitive due to transmitter changeouts.

, Corrective ' Add electronic filtering to signals.

l Action:

Generic increase in sensitivity / response can implications: result due to transmitter changeouts.

installing filtering in Reactor

Protection System flow transmitter circuitry.

t I

i ,

P O

A-81

O ,

Auxiliary Feedpump Turbines Concern: Both auxiliary feedpump turbines tripped on overspeed - this prevented supply of waterto steam generators. ,

Findings

Condensation in long steam inlet lines disrupts proper turbine control.

Corrective 1. Keep lines hot with steam to Actions: greatly reduce waterformation.

2. Improve governor controis.

Generic None-no other quick start steam implications: driven turbines.

4-82. .-

O Auxiliary Feedpump Turbine Trip and Throttle Valves i Concern: Operators experienced problems resetting the valves-delayed ,

initiation of auxiliary feedwater to  !

steam generators.

Findings: Procedures and prior training not sufficient.

Corrective 1. Provide improved hands-on training.

. Actions

2. Provide placards andlocal  :

indicators on T&TV to help operators.

3. Enhance communications between O pump rooms and from pump rooms to Control Room.

Generic Other crucial operator actions performed l Implications: locally. Covered by Operator Actions '

review.

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_ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ , . _ __ . _ . . . . _,,.., _ _ _ __,____._m,_,__.__

O Main Steam Headers Concern: After closure of main steam isolation valves, pressure control problems were experienced in the main steam headers.

Findings: Manualactuation of atmospheric

! vents valves (AW) caused large

-l pressyre drop in header # 1 - AW  :

i controlcircuitry on header # 2is a lesser concern. Switch contacts corroded on ICS module.  :

O Corrective Actions:

1. Full check-out and adjustment of AW control circuitry.

2. Testing of main steam safety valves and refurbish as needed.

Generic Switch contacts being evaluated on implications: otherICS modules.

1 l

\

O ,

O l

Main Feedwater Startup Control Valve Concern: Operators were uncertain of status of control valve SP-7A due to blown light bulb.

Findings: Valve operated properly - technician ,

inserted incorrect voltage lamp during event.

Corrective Provide additionalinformation to Action: operators.

Generic None-no significant findings.

Implications:

O g

us

O I

Auxiliary Feedwater Pump

1. 1 Suction Supply Concern: Pump suction transferred from normalto backup watersupply about 20 minutes after reactor trip. '

Findings: Noimpact to steam generator-

, transientlow suction pressure caused transfer.

Corrective 1. Remove / replace strainers.

. Actions: 2. Revise transfer switch setpoints.

l 3. Provide time delay.  ;

l Generic Other pump suction transfer implications: systems.

6 0  :

A-86 ,

e O

I l Main Steam Valve MS-106 Concern: Valve positon indication recorded as  ;

closed to not closed to closed in about one-third the expected time-this valve is used to admit steam from steam generator #1 to auxiliary feedpump turbine # 1.

Findings: Motor _ operator on valve was not properly adjusted.

Corrective Readjust and test valve.

Action:

Generic Other motor operated valves.

l Implications:

l l

t O

M7 - - _ - - - .

i l

1 O

Turbine Bypass Valve Concern: Pneumatic actuator assembly cracked and failed during cooldown operations several hours following reactor trip.

Findings: Internalvalve components became l disengaged and caused hammer blow forces _which damaged actuator.

Corrective 1. Repair damaged valve.  ;

l Action: 2. Repair steam traps and drains.

i 3. Refurbish otherturbine bypass l valves.

4. Revised operating procedure to assure proper drainage of headers.

l Generic Applies to both turbine bypass valve Implications: headers.

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! Safety Parameter Display System (SPDS) i Concern: Both SPDS Control Room display devices were inoperative during event -

they areintended to be used by the i

operators during transients.

Findings: Bad fiberoptic cable and faulty terminations on data transmission cable._

Corrective 1. Use spare cable.

Action: 2. Correct terminations.

j 3. Replace obsolete terminal.

Generic None-no other fiber optic systems.

implications:

c I

I O.

+ 89

O System Review and

Restart Test Program

System Review Methodology Five system review groups.

8 Headed by Toledo Edison engineering personnel.

m Supported by highly-qualified industry '

representatives.

Selected documentation review a LER's, DVR's

NPRDS Data -

l MWO's FCR's I

HED's i TAP Reports O a Focused interviews of operations and maintenance personnel.

a Evaluation / decision making guided by consistent, , ,

specific criteria / review process. '

s Preparation of suggested corrective actions.

m Overview and decision by a designated independent system review group, i

A-90

O System Review and

, Restart Test Program Test Program Review l m Each group will review their respective system design functions to assure that each function has been appropriately tested by the existing test

program.

, s identified concerns wiff be documented and  !

i recommended test outlines developed.

m Independent system review group will provide oversight and will approve the test outlines.

a As appropriate, new or revised tests will be O developed, approved and conducted under the direction of a joint test group in accordance with existing procedure and test programs.

• System review group and the Independent System Review Group will review tests performed and will assume responsibility for rectifying problems.

i l

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I ACRS BRIEFING l9 October 10,1985

+

lY Followup Actions Resulting From l The Davis-Besse June 9,1985 Event 3 m

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55 i "R

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1 l

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l O O INTRODUCTION

i.

• Review of Restart Evaluation items; and NRC

} Generic Technical Actions Following June 9,1985 Event l (Currently being Evaluated by NRC Staff) i e Restart Evaluation items Derived from Staff Concerns in 50.54(f) Letter Issued 8/14/85 l

i k - 50.54(f) Letter Based in NRC Team Findings in NUREG-1154 and Other NRC Concerns i

W e TED Response to 50.54(f) Letter Received 9/12/85 I

e Staff Will issue Restart Safety Evaluation

- Current Schedule Target - Dec.1985

- Commission Briefing I

o L .

O o--

l RESTART EVALUATION ITEMS

• Derived from Staff Concerns in 50.54(f) Letter l - Response to the EDO Memorandum of 8/5/85 i - Based Largely upon the Findings of the incident

, Investigation Team 4 e Concerns to be Addressed by Toledo Edison Co. Response 4 - Completion of Event Investigation l

l - Plant Specific Findings

- Programmatic and Management issues I

l l .

i

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l O

• O O~

I i

j i

RESTART EVALUATION ITEMS i

~

e Completion of Event investigation

- Completion of the investigation of Equipment Malfunctions and Operator Errors h

t i - Determination of Root Causes of the Malfunctions i

% and Errors and Implications to the Restart of

q the Plant

! - Corrective Actions Needed to Assure the Reliability j of Systems Which Mitigate Loss of Main Feedwater l Events '

l i

l

\ -

i

~ ~

l o ,

o .

o--

i RESTART EVALUATION ITEMS (CONTINUED)

I e Concerns Directly Related to June 9,1985 Event

! Evaluate Licensee's Response to Concerns identified j in NUREG-1154:

! (1) Adequacy of Loss of Feedwater Analysis (2) Adequacy of Design / Operation of SFRCS i

b (3) Physical Security and Administrative Features i

(4) Role of STA t (5) Reliability of AFW Containment isolation Valves 6 and Other Safety-Related Valves i (6) Adequacy of ENS Reporting i (7) Reliability of AFW System, Pumps, and Need i for Diverse Pump i

1 O

RESTART EVALUATION ITEMS (CONTINUED)

(8) Reliability of PORV (9) Adequacy of Control Room Instrumentation A and Controls k

y (10) Inability to Place Startup FW Pump in Service from Control Room (11) Resolution of Other Equipment Deficiencies (12) Adequacy of Procedures for " Drastic" Action (13) Adequacy of Safety System Testing I -

l ,

O

~

O O~

l' RESTART EVALUATION ITEMS t

(CONTINUED) e Evaluate Licensee's Response to Additional NRC Concerns

- Adequacy of Procedures, Equipment and Training for Restoring Equipment for LOF Mitigation .

A - Adequacy of Programs to Resolve Likelihood of i

)  % inadvertent Isolation of AFW to Both Steam og Generators .

- Installation of Diverse Drive AFW Pump

- Other ESF Systems Adequacy in Light of Single

! Failure Vulnerabilities identified in SFRCS and l AFW Systems l

i l

h O

O 9

~

O RESTART EVkUATION ITEMS O~

i

.(CONTINUED)

!

• Evaluate Licensee's Response to Management and l Programmatic Concerns

- Adequacy of Management Practices

- Adequacy of Maintenance Program improvements Ai

- Adequacy of the implementation of the Performance Enhancement Program i  %

j q - Adequacy of the Resources Committed to the

! Davis-Besse Facility for the Investigation

! of the Event, Resolution of Findings and

{

Conclusions, and Longer Term Actions

)

i 1

e May Be Additional Restart issues as a Results of Continuing ,

Staff Review i

i I

i i

?*

.O .

O

~

O~i NRC GENERIC TECHNICAL ACTIONS

~

e immediate Generic issues - None e Short Term Generic issues b - Potential Inability to Remove Decay Heat Because of Questionable Reliability of AFWS N

h - Adequacy of Emergency Procedures, Operator Training and Available Plant Monitoring Systems for Determining Need to initiate Feed and Bleed '

Cooling l - Physical Security System Constraints which could j - Deny Timely Access to Vital Equipment 1

- Prioritization of Short Term issues is Nearly Complete

~

i i

l NRC GENERIC TECHNICAL ACTIONS (CONTINUED) e Potential Long Term Generic issues .

j (1) Availability and Role of STA ,

g (2) Actions to improve Reliability of PORV, and Need s for Failure Mitigation -

D (3) Adequacy of Requirements for SPDS Availability N

(4) Need for Plant-Specific Simulator (5) Adequacy of Safety System Testing (6) Re-Evaluate NUREG-0737 Item II.E.1.1' (AFW System Reliability) l l

4

O

~

O

• O' NRC GENERIC TECHNICAL ACTIONS (CONTINUED)

(7) Adequacy of Maintenance Requirements (8) Adequacy of Single-Failure Aspects of Steam h Line/ Feed Line Break Mitigation Systems

\

(9) Effects of Loss of Feedwater On OTSG

} $ (10) Thermal-Hydraulic Aspects of Loss of Feedwater Event on Reactor Vessel (11) Re-Examine PRA-Based Estimates of Core Damage Resulting from Loss of Feedwater Other Additional issues, as identified I

i bm

O o D" .' .

Program Relationships Source Term Analysis l ubb BMi-2104 l

l Uncertainty Estimates SANDS 4-0410 l 1

Status of Validation l ORNLITM-8842 I  % Source Term l Changesin N Reassessment Study ,1 = Source Term-Based D l Regulations

{f Containment Working Group Loads l

( NUREG-1079 g

l '

Containment Performance Working Group 1f l @$

m>

NUREG-1037 l 1 C[n i

DE l Severe Accident l gy l APS Review Risk Rebaselining/ l Severe Accident *5

> Regulatory Risk Reduction j  %(

Program (SARRP) l 5E E,

• I

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I go 8E l M>

54 G9 E

d w I -

l

APPENDIX IX s' . NRC PRESENTATION - STATUS OF IMPLEMENTA-TION OF SEVERE ACCIDENT POLICY

,',n (a)

NRR STAFF PRESENTATION TO THE ACRS

SUBJECT:

STATUS OF THE IMPLENEtiTATION OF THE U.S. NUCLEAR REGULATORY COMMISSION'S SEVERE ACCIDENT POLICY O DATE: OCTOBER 11, 1985 PRESENTER: ZOLTAN R. ROSZTOCZY PRESENTER'S TITLE / BRANCH /DIV: CHIEF, RESEARCH Afl0 STANDARDS C00R0!NAT!0ft BRANCH, DST, NRR PRESEffTER'S NRC TEL. N0.: 492-4221 O

7' {

l STATUS OF THE IMPLEMENTATION OF THE U.S. NUCLEAR REGULATORY COMMISSION'S SEVERE ACCIDENT POLICY I. THE COMMISSION'S POLICY STATEMENT ON SEVERE ACCIDENTS II. EVALUATION OF THE REFERENCE PLANTS III. EXAMINATION OF INDIVIDUAL PLANTS IV. RESOLUTION OF OUTSTANDING ISSUES V. CHANGES IN RULES AND REGULATORY PRACTICES O' - SOURCE TERM RELATED CHANGES

- POTENTIAL SEVERE ACCIDEiiT RELATED CHANGES 6 9 O

A-to r

Q-g THE COMMISSION'S POLICY STATEMENT ON SEVERE ACCIDENT U

INTRODUCTION ON THE BASIS OF CURRENTLY AVAILABLE INFORMATION, THE COMMISSION SEES NO PRESENT BASIS FOR IMMEDIATE ACTI RULEMAKING BECAUSE OF SEVERE ACCIDENT RISK.

THE COMMISSION PLANS TO FORMULATE AN APPROACH FOR A SYSTEMATIC SAFETY EXAMINATION OF EXISTING PLANTS TO DETERMINE WHETHER PARTICULAR ACCIDENT VULNERABILI PREST.NT AND WHAT COST-EFFECTIVE CHANGES ARE DESIRA THE COMMISSION ENC 0URAGES THE DEVELOPMENT OF STANDAR DESIGNS THAT REALIZE SAFETY BENEFITS, FOR EXAMPLE:

GREATER SIMPLICITY, SLOWER DYNAMIC RESPONSE TO UPSET CONDITIONS AND PASSIVE HEAT REMOVAL FOR LOSS-0F-COOLA ACCIDENTS.

POLICY FOR NEW PLANT APPLICATIONS

' NEW DESIGNS ARE ACCEPTABLE IF THEY MEET THE FOLLOW

- COMPLY WITH CURRENT REGULATIONS

- RESOLVED ALL USIs AND MEDIUM- AND HIGH-PRIORITY GSIs

- COMPLETED A PRA

- STAFF REVIEW 0F THE DESIGN HAS BEEN COMPLETED WITHIN 18 MONTHS OF PUBLICATION OF THE POLICY STATEMENT, NRC WILL ISSUE GUIDANCE ON THE FORM, PURPOSE AND ROLE THAT PRAs ARE TO PLAY IN SEVERE ACCIDENT ANALYSIS AND DECISI MAKING FOR BOTH EXISTING AND FUTURE PLANT DESIGNS AN MINIMUM CRITERIA PRAs SHOULD MEET.

A-ios

m.

THE COMMISSION'S POLICY STATEMENT ON SEVERE ACCIDENTS, (CON'T.)

POLICY FOR NEW PLANT APPLICATIONS, (CON'T.)

A DECISION WILL BE MADE ON WHETHER TO ESTABLISH NEW PERFORMANCE CRITERIA FOR CONTAINMENT SYSTEMS AND, IF S0, WHAT THESE SHOULD BE.

POLICY "0R EXISTING PLANTS DURING THE NEXT TWO YEARS THE COMMISSION WILL FORMULATE A SYSTEMATIC APPROACH FOR THE EXAMINATION OF EXISTING PLANTS, INCLUDING THE DEVELOPMENT OF GUIDELINES AND PROCEDURAL CRITERIA, WITH AN EXPECTATION THAT SUCH AN APPROACH WILL BE IMPLEMENTED BY LICENSEES.

~

THE MOST COST-EFFECTIVE OPTIONS FOR REDUCING VULNERABILITIES WILL BE IDENTIFIED AND A DECISION WILL BE REACHED CONSISTENT WITH THE COMMISSION'S BACKFIT POLICY.

' ANY GENERIC DESIGN CHANGES THAT ARE IDENTIFIED AS NECESSARY FOR PUBLIC HEALTH AND SAFETY WILL BE REQUIRED THROUGH RULEMAKING.

A-/o7 t

EVALUATION OF THE REFERENCE PLANTS l

(  !

1 RES HAS COMPLETED THE FIRST PHASE OF THE SEVERE ACCIDENT RESE GC!! PROGRAM, WHICH RESULTED IN BETTER UNDERSTANDING 0F PHYSICAL PHENOMENA ASSOCIATED WITH SEVERE ACCIDENTS, AND DEVELOPED METHODOLOGY FOR THE ANALYSIS OF SEVERE ACCIDENTS, RES IS PRESENTLY APPLYING THE NEW KNOWLEDGE AND METHODOLOGY FOR THE ANALYSES OF FIVE REFERENCE PLANTS:

PEACH BOTTOM, GRAND GULF, SEQUOYAH, ZION, AND SURRY. THE RESULTS, IN TERMS OF PREDICTED CORE DAMAGE FREQUENCY, RADI0 ACTIVE MATERIAL RELEASE FROM CONTAINMENT, AND PUBLIC RISK WILL BE PUBLISHED IN THE SUMMER OF 1986.

THE RES EFFORT WILL ALSO ESTIMATE UNCERTAINTIES ASSOCIATED WITH THE REFERENCE PLANT ANALYSES AND WILL DISPLAY THEM IN THE REPORT (NUREG-1150). THE UNCERTAINTIES WILL BE USED (1) FOR COMPARING THE RESULTS AGAINST BOTH DETERMINISTIC AND PROBABILISTIC CRITERIA, (2) COMPARING CALCULATIONS AGAINST OTHER CALCULATIONS, FOR EXAMPLE IDCOR CALCULATIONS

. . VERSUS NRC CALCULATIONS, (3) SEARCHING FOR VULNERABILITIES

- IN PLANT DESIGN AND PLANT OPERATION, AND (4) PRIORITIZATION OF FUTURE EFFORTS.

PARALLEL WITH THE RES EFFORT, IDCOR ON BEHALF 0F THE NUCLEAR INDUSTRY HAS ANALYZED FOUR OF THE FIVE REFERENCE PLANTS. THE IDCOR RESULTS HAVE ALREADY BEEN PRESENTED TO NRC, AND WERE DOCUMENTED. BASED ON NRC COMMENTS AND NEW INFORMATION AVAILABLE FROM THE RESEARCH PROGRAMS, IDCOR WILL UPDATE THE REFERENCE PLANT EVALUATIONS BY EARLY 1986.

IDCOR IS ALSO CONSIDERING INITIATION OF A TASK TO ASSESS UNCERTAINTIES INHERENT IN THE IDCOR ANALYSES.

O A-/08

5 EVALUATION OF THE REFERENCE PLANTS, (CON'T.)

O

• NRR IS INITIATING AN EVALUATION OF THE IDCOR REFERENCE PLANT ANALYSES TOGEThER WITH RES'S ANALYSES OF THE SAME PLANTS (AUDIT CALCULATIONS). RES WILL FACILITATE THIS EVALUATION BY PROVIDING INFORMATION TO NRR AS IT BECOMES AVAILABLE, STARTING IN OCTOBER 1985.

IN THEIR PRESENT FORM BOTH THE RES AND THE IDCOR ANALYSES ADDRESS ONLY INTERNAL EVENTS. AN APPROACH FOR TREATMENT OF EXTERNAL EVENTS (INCLUDING SEISMIC EVENTS) AND FOR THE COMBINATION OF RISK FROM INTERNAL AND EXTERNAL EVENTS NEEDS TO BE DEVELOPED AND EXECUTED FOR THE REFERENCE PLANTS.

O e e A - (09

,-ss EXAMINATION OF INDIVIDUAL PLANTS THE NRR REVIEW 0F THE REFERENCE PLANTS HAS TWO G0ALS:

(1) EVALUATE THE PERFORMANCE OF THE REFERENCE PLANTS WITH RESPECT TO SEVERE ACCIDENTS, AND (2) BASED ON THE EXPERIENCE GAINED FROM THE EVALUATION OF THE REFERENCE PLANTS DEVELOP GUIDANCE AND CRITERIA FOR THE SYSTEMATIC ASSESSMENT OF INDIVIDUAL PLANTS. THE ACTUAL ASSESSMENT OF INDIVIDUAL PLANTS WILL BE DONE BY THE LICENSEES.

IDCOR IS CURRENTLY DEVELOPING SIMPLIFIED METHODOLOGY WHICH, TOGETHER WITH THE REFERENCE PLANT ANALYSIS, WILL PROVIDE MEANS FOR ASSESSING INDIVIDUAL PLANTS. THE IDCOR PROPOSED METHODOLOGY WILL BE SUBMITTED FOR NRC APPROVAL EARLY IN 1986.

NRR WILL REVIEW THE IDCOR NETHODOLOGY TO BE USED FOR INDIVIDUAL PLANTS, AND WILL ISSUE ITS EVALUATION AND

(]

APPROVAL CONCURRENT WITH THE ISSUANCE OF THE GUIDELINES AND CRITERIA MENTIONED AB0VE.

i- -

IN THE MEANTIME, RES WILL BE WORKING ON THE MOST IMPORTANT

' TECHNICAL ISSUES AND WILL PROVIDE UPDATED INFORMATION ~

PRIOR TO ISSUANCE OF THE GUIDELINES, CURRENTLY SCHEDULED FOR THE SUMMER OF 19E.7.

THE OVERALL EFFORT IS A CLOSELY COORDINATED EFFORT BETWEEN NRC AND IDCOR. THERE ARE PERIODIC MANAGEMENT AND TECHNICAL MEETINGS BETWEEN IDCOR AND NRC (TYPICALLY ONE MEETING PER MONTH).

O o

k- (f0

-, i RESOLUTION OF OUTSTANDING ISSUES O

• CORE MELT PROGRESSION, IN-VESSEL HYDROGEN GENERATION, FISSION PRODUCT AND AEROSOL RELEASE - DIFFERING MO BEHAVIOR OF FUEL AND CLADDING DURING A MELTDOWN LE  ;

WIDELY DISPARATE PREDICTIONS OF CORE TEMPERATURES, FISSION PRODUCT CHEMISTRY AND THE AVAILABILITY OF STEAM IN lH  !

CORE REGION.

THESE PHENOMENA IN TURN AFFECT PREDICTIONS OF IN-VESSEL HYDROGEN, IN-VESSEL RELEASE OF FISSION (

PRODUCTS, AND THE INITIAL CONDITIONS FOR EX-VESSEL CORE-CONCRETE INTERACTIONS. GIVEN THE LACK OF RELEVANT i

DATA, THESE ISSUES WILL NOT BE RESOLVED IN THE NEAR TERM, AND THE INTEGRAL ANALYSES WILL HAVE TO ACCOUNT FOi UNCERTAINTIES.  !

RETENTION AND REVAPORIZATION OF FISSION PRODUCTS I REACTOR COOLANT SYSTEM - NRC CONCERNS ABOUT THE Vl O 0F SIMPLIFIED IDCOR CORRELATION FOR AEROSOL DEPOSITIO WILL BE ADDRESSED BY FURTHER ANALYTIC WORK, CONTINUED COMPARISON WITH APPLICABLE DATA AND POSSIBLY BY C t l WITH THE NRC TRAP-MELT 2 CODE. NRC ENDORSES IDCOR'S INCLUSION OF REVAPORIZATION IN THEIR FISSION PRODUCT :

l

- ANALYSES, BUT NRC BELIEVES THAT THE IDCOR MODEL PREDIC'TS REVAPORIZATION T00 EARLY AFTER CORE NRC MELT.

AND IDCOR WILL COOPERATE ON INTEGRAL ANALYSIS TO ASSESS THE IM OF THIS RESULT, l EX-VESSEL FISSION PRODUCT AND AEROSOL RELEASE AND DEPOSITION - ANALYTICAL MODELING OD CORE / CONCRETE  !

INTERACTION - NRC AND IDCOR DISAGREE ON SEVERAL OF T ,

BOUNDARY CONDITIONS AND MODELS WHICH AFFECT FISSION PRODUCT RELEASE DURING CORE-CONCRETE INTERACTIONS. THESE INCLUDE INITIAL CORE TEMPERATURE, EXTENT OD DEBRIS DISPERSAL AND MODELS FOR HEAT TRANSFER FROM THE CORE O

4 - ///

?

?

O RESOLUTION OF OUTSTANDING ISSUES, (CON'T.)

h DEBRIS, COMPARISON AMONG CODES AND WITH RECENT CORE-CONCRETE DATA WILL BE PERFORMED. THE AEROSOL DEPOSITION CORRELATION DISCUSSED AB0VE ALSO APPLIES TO CONTAINMENT AND SECONDARY BUILDINGS, FINALLY, THE WIDE DISPARITY BETWEEN NRC AND IDCOR ESTIMATES OF FISSION PRODUCT RETENTION IN SECONDARY BUILDINGS WILL BE REDUCED BY CHANGES IN THE ASSUMPTIONS USED BY BOTH PARTIES, SCRUBBING EFFICIENCY OF SUPPRESSION P0OLS AND ICE CONDENSERS - SUPPRESSION POOL BYPASS IS THE MAIN CONCERN, PLUGGING 0F LEAKAGE PATH BY AEROSOL HAS BEEN QUESTIONED, MORE EVIDENCE IS NEEDED TO JUSTIFY PLUGGING CORRELATION -

ANALYTICAL MODELS USED TO PREDICT FISSION PRODUCT RETENTION IN ICE BEDS NEED TO BE VERIFIED, l

CO G INMENT PRESSURE LOADS - THE SUSCEPTIBILITY TO DIRECT HEATING FAILURE IN HIGH PRESSURE SEQUENCES WILL BE ADDRESSED ON A PLANT SPECIFIC BASIS, BY DETERMINING WHETHER THE REACTOR CAVITY GE0 METRY IS CONDUCIVE TO DISPERSAL OF AEROS0LIZED CORE DEBRIS. THE MORE COMPLEX

- PROBLEM 0F HYDROGEN IGNITION AND FLAME PROPAGATION WILL REQUIRE DETAILED STANDARD FROBLEM CALCULATIONS AND COMPARISON OF IDCOR MODELS AND RESULTS WITH THOSE OF NRC'S HECTR CODE, CONTAINMENT FAILURE MODES - CONSIDERABLE UNCERTAINTY EXISTS AS TO HOW RAPIDLY LEAKAGE WILL GROW AS A FUNCTION OF PRESSURE AND TEMPERATURE, THE MODE BY WHICH A CONTAINMENT WILL FAIL (CRACKS VS, LARGE HOLE), AND THE LOCATION AT WHICH FAILURE WILL BE INITIATED, INFORMATION IS NEEDED ON THE BEHAVIOR OF GASKETED PENETRATIONS, ELECTRICAL PENETRATION ASSEMBLIES, STRUCTURAL RESPONSE OF CONCRETE CONTAINMENTS, AND FISSION PRODUCT DEPOSITION IN SECONDARY CONTAINMENT, A - //2-.

t

[

, RESOLUTION OF OUTSTANDING ISSUES, (CON'T.)

(3 U')

EQUIPMENT PERFORMANCE - METHODOLOGY TO PREDICT ENVIRONMENTAL CONDITIONS DURING SEVERE ACCIDENTS IS NEEDED. TEMPERATURE, TIME AT TEMPERATURE, PRESSURE, RADIATION AND SUBMERGENCE ARE THE MAIN PARAMETERS.

AEROSOL AND PARTICLE DEPOSITION ON EQUIPMENT AND RADI0 ACTIVITY OF WATER POOLS CAN AFFECT EQUIPMENT PERFORMANCE. EQUIPMENT NEEDED TO MITIGATE SEVERE ACCIDENTS AND THE TIME INTERVAL FOR WHICH EQUIPMENT ARE NEEDED FOR SHOULD BE IDENTIFIED. EXPECTED EQUIPMENT PERFORMANCE SHOULD BE ASSESSED BASED ON QUALIFICATION OF EQUIPMENT.

TREATMENT OF EXTERNAL EVENTS - FIVE RECENT PRAs ADDRESSED EXTERNAL EVENTS. CONCLUSION: RISK FROM EXTERNAL EVENTS IS COMPARABLE TO THE RISK FROM INTERNAL EVENTS. THIS CONCLUSION IS BEING QUESTIONED. UNCERTAINTIES ARE LARGE, TREATMENT OF INTERNAL AND EXTERNAL EVENTS IS NOT CONSISTENT. TWO POSSIBLE APPROACHES ARE BEING CONSIDERED FOR REFERENCE PLANTS: (1) USE INFORMATION FROM EXISTING

' PRAs WITH EXTERNAL EVENTS, TRANSFER KNOWLEDGE TO REFERENCE

' PLANTS, AND (2) PERFORM A LIMITED EXTERNAL EVENT ANALYSIS FOR THE REFERENCE PLANTS.

ASSESSMENT AND QUANTIFICATION OF UNCERT.*INTIES -

UNCERTAINTIES OF SEVERE ACCIDENT ANALYSIS HAVE NOT YET BEEN EVALUATED AND PRESENTED IN A REASONABLY COMPLETE AND SUBSTANTIATED MANNER, A METHODOLOGY NEEDS TO BE DEVISED THAT IS SIMPLE EN0 UGH TO PERMIT TREATMENT OF ALL SIGNIFICANT UNCERTAINTIES AND PROPAGATES THESE UNCERTAINTIES TO THE RESULTS OF THE ANALYSIS. IF DATA OR INFORMATION ARE MISSING, ESTIMATES SHOULD BE USED AND THE COST AND BENEFIT OF OBTAINING DATA SHOULD BE tSSESSED.

A -N 3

RESOLUTION OF OUTSTANDING ISSUES, (CON'T.)

DEVELOPMENT OF NEW SOURCE TERMS FOR REGULATORY APPLICATIONS - THE VARIOUS REGULATORY USES OF SOURCE TERNS REQUIRE DIFFERENT FORMS OF SOURCE TERMS. AT MOST, THREE FORMS OF SOURCE TERMS ARE BEING CONSIDERED: (1) DETAILED SOURCE TERM CALCULATIONS FOR INDIVIDUAL PLANTS, (2) USE OF TABLES OR PROCEDURES FOR PLANT TYPES, AND (3) A SIMPLE, B0UNDING SOURCE TERM APPLICABLE TO N0ST PLANTS. THESE FORMS HAVE TO BE DEVELOPED BEFORE THE REGULATIONS CAN BE CHANGED.

i O

b o e e O

A-HV

O CHANGES IN RULES AND REGULATORY PRACTICES b

SOURCE TERM RELATED CHANGES SOURCE TERM RELATED REGULATORY REQUIREMENTS AND PRACTICES WERE REVIEWED. TEN AREAS WERE FOUND WHICH CURRENTLY USE SOURCE TERMS AND COULD BENEFIT FROM THE NEW KNOWLEDGE.

THE VARIOUS REGULATORY USES OF SOURCE TERMS REQUIRE DIFFERENT FORMS OF SOURCE TERMS. FOR EXAMPLE, THE RESOLUTION OF SAFETY ISSUES RELATED TO SEVERE ACCIDENTS IS WELL SUITED TO DETAILED SOURCE TERM CALCULATIONS FOR INDIVIDUAL PLANTS. HOWEVER, EQUIPMENT QUALIFICATION NEEDS A SIMPLE SOURCE TERM, EASILY AVAILABLE TO EQUIPMENT MANUFACTURERS AND APPLICABLE TO ALL PRESENT AND FUTURE PLANTS.

NRC WILL EVALUATE AND GROUP THE TEN AREAS OF APPLICATIONS OF SOURCE TERMS, AND WILL DEVELOP PRACTICAL SOURCE TERMS FOR EACH GROUP. THE NEEDED FORMS OF SOURCE TERMS WILL BE DISCUSSED WITH THE NUCLEAR INDUSTRY (AIF). WE FORESEE, AT MOST, THREE FORMS OF SOURCE TERMS: (1) DETAILED SOURCE

' TERM CALCULATIONS FOR INDIVIDUAL PLANTS, (2) USE OF TA'BLES OR PROCEDURES APPLICABLE TO PLANT TYPES, AND (3) A SIMPLE, B0UNDING SOURCE TERM APPLICABLE TO ALL PLANTS. APPLICANTS AND LICENSEES WILL BE ABLE TO OPT FOR ANY ONE OF THE ESTABLISHED FORMS FOR EACH USE OF THE SOURCE TERM.

4 A - // T

CHANGES IN RULES AND REGULATORY PRACTICES, (CON'T.)

SOURCE TERM RELATED CHANGES NRC IS PRESENTLY APPLYING THE DETAILED SOURCE TERM METHODOLOGY FOR THE ANALYSES OF THE FIVE REFERENCE PLANTS. THE REFERENCE PLANT APPLICATIONS WILL CALCULATE NOT ONLY THE AMOUNT AND TIMING OF FISSION PRODUCT RELEASE FROM THE CONTAINMENT, BUT WILL ALSO PREDICT THE SOURCE TERM SEEN BY EQUIPMENT LOCATED INSIDE CONTAINMENT.

SPECIAL ATTENTION WILL BE PAID TO THE QUANTIFICATION OF UNCERTAINTIES ASSOCIATED WITH THE PREDICTION OF SOURCE TERf1S.

THE REFERENCE PLANT ANALYSES WILL BE THE BASIS FOR ARRIVING AT THE SIMPLIFIED FORM OF THE SOURCE TERMS, PROGRESS MADE IN DEVELOPING SOURCE TERMS WILL BE REVIEWED T PERIODICALLY AND WILL BE COMPARED AGAINST THE INFORMATION s

NEEDED TO INITIATE CHANGES IN CURRENT REGULATIONS AND REGULATORY PRACTICES. CHANGES WILL BE INITIATED AS S00N AS THE AVAILABLE INFORMATION WARRANTS IT, e e A NC

CHANGES IN RULES AND REGULATORY PRACTICES O POTENTIAL SEVERE ACCIDENT RELATED CHANGES THE GUIDELINES AND CRITERIA PREPARED FOR THE VARIOUS PLANT TYPES WILL BE REVIEWED TO IDENTIFY GENERIC REQUIREMENTS AND GUIDANCE THAT NEED TO BE INCORPORATED INTO THE REGULATIONS, EXISTING OR NEW REGULATORY GUIDES, AND THE STANDARD REVIEW PLAN.

SPECIAL ATTENTION WILL BE PAID TO CONTAINMENT PERFORMANCE CRITERIA. THE NEED FOR AND THE BENEFIT TO BE GAINED FROM THE ISSUANCE OF CONTAINMENT PERFORMANCE CRITERIA WILL BE EVALUATED. IF THE NEED EXISTS, CRITERIA WILL BE FORMULATED AND INCORPORATED IN THE REGULATIONS OR REGULATORY PRACTICES, AS APPROPRIATE, g

~

EXPERIENCE GAINED FROM PRAs TOGETHER WITH THE SEVERE ACCIDENT GUIDELINES AND CRITERIA WILL BE USED TO ESTABLISH THE PURPOSE AND ROLE THAT PRAs ARE TO PLAY IN SEVERE ACCIDENT ANALYSIS AND DECISION MAKING FOR BOTH EXISTING AND FUTURE PLANT DESIGNS. GUIDANCE AND CRITERIA ON THE

' ROLE OF PRAs IN NEW PLANT APPLICATIONS WILL BE ISSUED BY FEBRUARY 1987. TO THE EXTENT APPROPRIATE, THESE GUIDANCE AND CRITERIA WILL BE INCORPORATED INTO THE REGULATIONS OR REGULATORY PRACTICES, A -// 7

.l REGULATORY AREAS TARGETED FOR SOURCE TERM RELATED CHANGES (J

1. RESOLUTION OF SAFETY ISSUES RELATED TO SEVERE ACCIDENTS -

PREVENTION AND MITIGATION OF SEVERE ACCIDENTS, MAINTAINING CONTAINMENT INTEGRITY IN CASE OF SEVERE ACCIDENTS

2. CONTAINMENT PERFORMANCE REQUIREMENTS - LEAK RATE TEST, UNDETECTED BREACH OF CONTAINMENT INTEGRITY

3. ENVIRONMENTAL QUALIFICATION OF EQUIPMENT IMPORTANT TO SAFETY - QUALIFICATION REQUIREMENTS FOR DESIGN BASE ACCIDENTS

4. EMERGENCY PLANNING - ONSITE PLANNING, 0FFSITE EMERGENCY PLANNING ZONES, GRADED RESPONSE

5. ACCIDENT CONSEQUENCES AND INDEMNIFICATION - DEFINITION OF

(\ AN EXTRAORDINARY NUCLEAR OCCURRENCE, EXTENSION OF THE PRICE-ANDERSON ACT

6. AIR FILTRATION AND OTHER FISSION PRODUCT ATTENUATION SYSTEMS - CONTAINMENT SPRAY SYSTEMS, RECIRCULATING AIR

' FILTERS, CONTROL ROOM AIR FILTERS AND FILTERED BUILDING EXHAUSTS WERE DESIGNED FOR ELEMENTAL IODINE AND FOR INSTANTANEOUS RELEASE - AUTOMATIC ACTUATION OF PWR SPRAY SYSTEMS

7. ACCIDENT MONITORING AND MANAGEMENT, ONSITE AND OFFSITE INSTRUMENTATION - DIAGNOSTIC CAPABILITY OF 0FFSITE MONITORS

8. OFFSITE CONTAMINATION AND REC 0VERY - PRESENT BASIS IS WASH-1400, PREDISTRIBUTION OF POTASSIUM IODIDE nv

I.'

REGULATORY AREAS TARGETED FOR SOURCE TERM RELATED CHANGES, (CON'T.)

9. SAFETY ISSUE EVALUATION - PRIORITIZATION OF GENERIC ISSUES IS PRESENTLY BASED ON WASH-1400

10. SITING - EXPLICIT CONSIDERATION OF SEVERE ACCIDENTS IN SITING, USE OF SOURCE TERM IN SITING O

1. 9 1

l O

n-09

L 1 APPENDIX X l PRINCIPAL FEATURES OF NEW SOURCE TERM  !

ANALYTICAL PROCEDURE l I

tilli l t

6 1

PRINCIPAL FEATURES OF l

t NEW SOURCE TERM ANALYTICAL PROCEDURE  !

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R. O. MEYER 1 i

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{ ACRS MEETING OCTOBER 11, 1985 g  :

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WASH-1400 Assumptions / Limitations Very Limited Data Base for Fission Product Release i from Fuel j

j No Credit for Fission Product Deposition in the RCS i

) lodine Assumed to Transport as 1 2 l h j

i (F Very Limited Data on Fuel Melt Progression and Hydrogen Generation t Crude Empirical Model for Containment Transport of

!N

?

Vapors and Aerosols i

!. No Data on Core-Concrete Interactions i

I

ScurCo Torm Codo Suito Fission Product Thermal Hydraulic O i Tr ne<<

i i sea vie, i

l ORIGEN l MARCH I Fission Pr,oduct Inventory in Fuel Overall l Behavior of El == = Reactor Coolant System,

• I l Molten Core, and "1a sl l Containment l

gl CORSOR l g l 8 l

usl o Release Retained from Fue!

l

, in Fuel l l l sI i i l l 5I J i I I TRAPMELT $,  ! MERGE I I Reactor Coolant System Transport and meu Detailed Temperature, Pressure, and Flow in l

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( l Retention 5 I Reactor Coolant System I

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Release from "" """ Detailed Core- 1 Nl E

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.y I l and interactions E

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El I 2 NAUA, SPARC, ICEDF g

l l Containment Transport l

and Retention l l l

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i L____ __ ___ _ _ _ .I 1 r l

l O Release of fission products to the environment: Source Term A< z 2- . . .

L O O O L

Fission Product Deposition in the RCS i

i Fission Product

! Deposition Processes g j Vapor Condensation l on Sudaces b Chemical Reaction

with Surfaces '

l Y Aerosol Formation s4'% l u , . . , e ,, ,

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Aerosol Agglomeration , __  ?!'

and Growth Gravitational Settling l"

) Deposition on Walls I i

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Containment Deposition Processes Fission Product Deposition Aerosol Agglomeration and Growth  :.

Gravitational Settling Deposition on Walls d.

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3 Major Advances in t

Source Term Technology i

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• Fission Product Chemistry h

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• Retention in Reactor Coolant System (RCS) ,

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• Core Melt Progression Y

p

• Aerosol Behavior in Containment

• Fission Product Release from Fuel

• Core-Concrete Interactions J

• Containment Pressure Loads i '

l, CODE VALIDATION REVIEW REVIEW 0F THE STATUS OF VALIDATION OF THE COMPUTER CODES USED IN THE SEVERE ACCIDENT SOURCE TERM REASSESSMENT STUDY (BMI-2104)

T.S. KRESS ET AL. ORNL/TM-8842 h

O  !

INDIVIDUAL REVIEWS OF ALL 10 CODES

SUMMARY

APPRAISAL l

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l UNCERTAINTY STul)Y UNCERTAINTY IN RADIONUCLIDE RELEASE UNDER

. SPECIFIC LWR ACCIDENT CONDITIONS.

R.J. LIPINSKI ET AL. SAND 84-0410 1985 t

l SURRY: STATION ELA.CK0UT (TMLB')

SURRY: SMALL LOCA (S2D)

O GRAND GULF:

ANTICIPATED TRANSIENT WITHOUT SCRAM (TC) i 1

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G INPUT SELECTION FROM MANUALS 1 I MARCH 3 (MARCH,CORCON,CORSOR) i i TRAP-MELT 3

! VANESA (TRAP-MELT 2 MERGE)

if NAUA/

u SPARC/ a ICEDF I f SOURCE TERM CODE PACKAGE l

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Major Areas of Uncertainty I

i

• Natural Circulation in the Vessel -

• Core Melt and Hydrogen Generation In-Vessel Releases from Fuel 4

• Retention and Revaporization in RCS W

• Core-Concrete interactions -

o a Scrubbing by Pools and Ice Compartments l

Containment Pressure Loads -

• Containment Failure Modes 0

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4 CALENDAR YEAR 1984 1905 1988 Igg 7 I I I l 1

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from the Fuel and

.h Aerosol Generation d ,

vap irat on of mt Fission Products in the RCS l

5. Fission Product Flefease A and Aerosol Generation a 4.c_,

from the Core-Concrete Interaction

8. Scrubbing Efficiency l

of Suppression Poole p and Ice Compartments

7. Containment A Pressure Loads 1 M

8. Containment Fallure A Modes a l

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Figure 7.1 Important milestones in research programs that are addressing major technical areas of uncertainty.

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- g NUREG-0956

SUMMARY

EVALUATION o THE BMI-2104 SUITE OF CODES REPRESENT THE STATE OF THE ART IN 1983-1984.  :

o THE CODES ARE FULLY OPERATIONAL, THEY HAVE BEEN PEER-REVIEWED AND EXTENSIVELY DOCUMENTED.

o THE CODES PRODUCE BEST-ESTIMATE (1.E., UN3IASED)

RESULTS ALTHOUGH UNCERTAINTIES REMAIN LARGE.

o MAJOR AREAS OF UNCERTAINTY HAVE BEEN IDENTIFIED AND RESEARCH IS BEING CONCENTRATED IN THESE AREAS, o THE BMI-2104 SUITE OF CODES REPRESENTS A MAJOR ADVANCEMENT IN METHODOLOGY SINCE WASH-1400.

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s Conclusion 1. The BMI-2104 Suite of Computer Codes 8

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Represents a Major Advance and Can Be Used to Replace the Reactor Safety Study ,

Methods. '

lL Conclusion 2. Principal Omissions and Oversimplifications in  ;

l the Reactor Safety Study Methods Have Been i ,g

Corrected. q l

Conclusion 3. Remaining Areas of Uncertainty Have Been Identified and Indicate Areas of Research that 7

q Should Be Pursued. -

l Conclusion 4. The New Analytical Procedures Have Been l Extensively Reviewed (including an APS l I

Review) and Documented. .; .

1 Conclusion 5. The Analytical Procedure is Complex, and i

Successful Application Requires,a Thorough Understanding of the Problem.

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i Source Term insights Conclusion 6. New Source Terms Have Been Calculated for Selected Accident Sequences, and These '

Sequences Have Provided a Sufficient Test of the Computer Codes. .

Conclusion 7. For Most Accident Sequences, the Largest Single Factor Affecting Source Terms is Containment Behavior.  %

9 Conclusion 8. Source Terms Were Fcund to Depend Strong- C ly on Plant Design and Construction Details, - \

%l l

Thus Making Development of Useful Generic Source Terms Difficult. =

Conclusion 9. New Source Terms for Many Accident Sequences Were Found to Be Lower'than Those in the Reactor Safety Study, but Some Were Larger. Therefore, Generalizations Are ,

inappropriate.

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} Updated Risk Estimates for the Surry Plant l Using WASH-1400 Accident Frequencies '

I1

. Early Fatalities Latent Fatalities l Analytical Method (per reactor year) (per reactor year) -

WASH-1400 Source Terms 4.0 x 10-8 1.8 x 10-2 WASH-1400 Containment Evaluation .

j BMI-2104 Source Terms 1.1 x 10-er S.7 x 10-s WASH-1400 Containment Evaluation I

, f s

BMI-2104 Source Terms 3.1 x 10-er. 3,4 x 10 4 2 \

Containment Reevaluation Q

a

, Sources of Uncertainty i

Event Frequencies .

l Source Term Analytical Procedures -

Containment Behavior i Consequence Calculations

• Uncertainties Will Be Taken into Account in NUREG-1150  :

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! Risk insights '

I i , .

j Conclusion 10. A Comparative Risk Appraisal for the Surry o l Plant Using BMI-2104 Source Terms and a ];

' Containment Reevaluation Shows a '

Reduction in Estimated Risk Compared with <

the Reactor Safety Study.

Conclusion 11. For the Other Plants, Further Analyses Need to Be Made Before Any Conclusions Can Be [s; l

Drawn About Changes in Estimated Risk, t

' and Significant Reductions May Not Be X Found in all Cases.

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i l Continuing ResearCh I

, Conclusion 12. Research Programs that Address the l Remaining Major Areas of Uncertainty in the l-Source Term Technology are Currently in ,

Place and Being Pursued by the NRC.

i l

Conclusion 13. A Major Conclusion of the American Physical ,

j Society Study Group Confirms the NRC Staff 3 i Position that Source Term Research Must Be Continued in Order to Complete the N

W i; '

Regulatory Actions Being Considered.

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.' Recommendations Recommendation 1. The New Source Term Analytical i

Methods Should Be Used to Aeevaiuate Regulatory Practices While Additional Confirmatory Research is Being

,, Completed.

)i Recommendation 2. The Source Term Code Package is the

); Recommended Tool for NRC Analyses.

Additional insights Can Be Obtained  %

m with the NRC's Detailed Mechanistic. N Codes and Their Experimental Data \

l ,

Bases. . %

Recommendation 3. The Source Term Code Package Pro-vides Best-Estimate Results (i.e.,

Without intentional Blas). Close ,

Coupling Between the Research Effort -

and the Regulatory Effort Will Bs Required in Assessing Uncertainties .

and Evaluating Technical issues. .

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4 EMERGENCY PLANNING ie i

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, SEVERE LOW FREQUENCY NATURAL PHENOMENA i E;9

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0 BEFORE DECEMBER 1931 -

LIMITED CONSIDERATION OF EARTHOUAKES (UP TO SSE) ON A CASE BY CASE BASIS

O DECEMBER 1981 - SAN ON6FRE DECISION 0

AUGUST 19J4 - DIABLO CANYON DECISION - s- Q ~fU A 0 DECEMBER 1984 -

PRCPOSED RULE PUBLISHED IN FEDERAL REGISTER 0 JULY 1935 -

D ACRS MET WITH COMMISSION AND STATED THAT "WE...SEE NO TECHNICAL REASO FOR THE EXCLUSION OF EARTHOUAKES FROM THE NATURAL PHEN 0f1ENA CONSI IN 0FFSITE PLANNING."

0 SEPTEMBER 9, 1985 -

STAFF PRESENTATION TO COMMISSION PROPOSING LIMITED CONSIDERATION OF SEVERE, LOW FREQUENCY NATURAL PHENOMENA.

THE SAN ONOFRE NiD DIABLO DECISION AND THE PROPOSED RULE ALL STATED THAT THE POTE 4

COMPLICATING EFFECTS OF EARTHOUAKES ON EMERGENCY PLANNING NEED NOT BE CONSIDEP,ED.

1

ALTERNATIVES O PROMULGATION OF THE PROPOSED RULE i

'h s 0 LEAVING THE ISSUE OPEN FOR ADJUDICATION ON A CASE BY CASE BASIS

!N O PROMULGATE A FINAL RULE WHICH LIMITS THE CONSIDERATION OF THE COMPLICATING EFFECTS OF SEVERE LOW FREQUENCY NATURAL PHEIGENA ON EERGENCY PLANNIdG

g . .

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PUBLIC C0ffENTS I

i 0 61 PUBLIC COMENT LETTERS RECEIVED i

0 25 LETTERS FAVORED THE PROPOSED RULE l3 0 34 LETTERS OPPOSED THE PROPOSED RULE-9 WERE IN PETITION FORM WITH 94 SIGNATURES 0 2 TAKING NO OBVIOUS POSITION

!Y JAPM, FRANCE, SWEDEN, GERMANY a TAIWAN ALL STATED THAT THE POTENTIAL COPPLICATING

! O l EFFECTS OF EARTHQUAKES WERE NOT SPECIFICALLY CONSIDERED IN THEIR EPERGENCY PLANS.

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I i RECDPM M ED RULE CHANGE l 0 "THE PLMS SHALL ASSURE THAT THE FOLLOWING CAPABILITIES EXIST RELATIVE TO THE i COMPLICATING IPFACTS OF SEVERE, LOW FREQUENCY NATURAL PHEN 0ENA CHARACTERISTIC 0F

! THE SITE. Id ADDRESSING THE FOLLOWING CAPABILITIES THE LICENSEE SHALL ASSUME l- THAT THE SEVERE NATURAL PHE0 NOMEN 0N HAS DISRUPTED NORMAL COMfiUNICATION AND ROAD NETWORKS."

{

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CAPABILITIES iT FW i

l - SEVERE, LOW FREQUENCY NATURAL PHENDENA ASSUfiPTI0dS L

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! RECOM]IDED CAPABILITIES 1 PROPOSED IMPLEENTATION OVERALL THRUST IS IQ CDDIFY FIFXIBILITY IN ORDER TO D0 THE BEST YOU CAN FOR ANY

!' NATURAL PHEWOMEHA, l 0 ABILITY TO TRANSPORT PERSONNEL BACK INTO THE PLANT - COULD BE IMPLEENTED BY

]%\

i AN AGREEENT BETWEEN A LICENSEE AND A TRANSPORTATION PROVIDER.

!t 0 ABILITY TO C0 m uMICATE PLANT STATUS WITH OFF-SITE AUTHORITIES - COULD BE IMPLEENTED BY REDUNDANT AND DIVERSE TRANSMITTING CAPABILITIES, 4

!' 0 RECOMEND THAT STATE AND LOCAL 60VERifENTS IDENTIFY ALTERNATE ROUTES OF TRAVEL I' MD METHODS FOR DETERMINING WHETHER TO SHELTER OR EVAUCATE - FOR MOST SEVERE

' NATURAL PHENDENA SHELTERING WOULD BE THE MOST VIABLE AND PREFERRED PROTECTIVE ACTION - DO THE BEST YOU CAN UNDER THE CIRCUMSTANCES.

l 4

I

l SEVERE WEATHER AND EXTERNAL EVENTS (NATURAL) i I

't ,

IN CURRENT EMERGENCY PLANS 1

l i

{ 1. EMERGENCY PREPAREDNESS FRAMEWORK

• ib 2. EMERGENCY ACTION LEVELS '

!w t lQ 3. ACTUAL EVENTS j 11 . OFFSITE RESPONSE (FEMA)

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EMERGENCY PREPAREDNESS FRAMEWORK i

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FORMATION OF EMERGENCY RESPONSE ORGANIZATIONS (PREESTABLISHED DIVISION OF RESPONSIBILITIES) j COMMUNICATIONS AMONG THEM i

\ ,

N PREDETERMINED ACTIONS BASED ON PLANT CONDITIONS 1& (DECISIONMAKING, COORDINATION, AND IMPLEMENTATION)

I C)

TRAINING AND EXERCISES I

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10 CFR PART 50 APPENDIX E.IV.C AND NUREG-0654

, JNITIATOR CLASSIFICATION ACTIONS

1. EARTHQUAKE FELT INPLANT OR DETECTED NOTIFICATION OF LICENSEE ON STATION INSTRUMENTS 2,43 UNUSUAL EVENT NOTIFY

2. 50 YEAR FLOOD OR LOW WATER, TSUNANI, AUGMENT STAFF I HURRICANE SURGE, SEICHE

's- STATE & LOCAL

\T 3. ANY TORNADO ON SITE PROVIDE FIRE OR SECURITY

- ASSISTANCE

4. ANY HURRICANE

1. EARTHQUAKE GREATER THAN OBE ALERT LICENSEE

2. FLOOD, LOW WATER, TSUNAMI, HURRI- NOTIFY ACTIVATE TSC, OSC CANE SURGE SEICHE NEAR DESIGN LEVELS BRING EOF TO STANDBY

3. ANY TORNADO STRIKING FACILITY DISPATCH ON-SITE TEAMS DOSE ESTIMATES

4. HURRICANE WINDS NEAR DESIGN BASIS LEVEL STATE & LOCAL BRING EBS TO STANDBY DISPATCH OFF-SITE TEAMS

!.-.-- 5.-_. b- . C) 1 NUREG-065tl (CONT'D)

INITIATOR CLASSIFICATION ACTIONS

! 1. EARTHQUAKE GREATER THAN SSE i SITE AREA EMERGENCY LICENSEE

!' 2. FLOOD, LOW WATER, TSUNAMI, HURRI- NOTIFY '

l' CANE SURGE, SEICHE GREATER THAN ACTIVATE TSC, OSC AND EOF l

DESIGN LEVELS DISPATCH ON-SITE AND OFF-SITE TEAMS DOSE ESTIMATES

3. SUSTAINED WINDS OR TORNADOES IN EXCESS i 0F DESIGN LEVELS STATE & LOCAL i

ACTIVATE LMS S

ACTIVATE PRIMARY RESPONSE +

N CENTERS DISPATCH EMERGENCY PER-9 SONNEL AND MONITORING TEAMS '

PUT MILK ANIMALS OUT TO

• 2 MILES ON STORED FEED i

ANY EXTERNAL EVENTS WHICH COULD CAUSE GENERAL EMERGENCY LICENSEE l MASSIVE COMMON DAMAGE TO PLANT SYSTEMS NOTIFY

~

i RESULTING IN CORE MELT, LOSS OF PHYSICAL CONTROL OF FACILITY, LOSS OF 2 0F 3 FIS- ACTIVATE ALL CENTERS

! DISPATCH ALL TEAMS SION PRODUCT BARRIERS WITH POTENTIAL' I LOSS OF THIRD, OR HIGH DOSE RATES ESTI- DOSE ESTIMATES j MATED OFFSITE.

STATE 8 LOCAL "

i ACTIVATE LMS RECOMMEND SHELTERING OR

EVACUATION j PUT MILK ANIMALS OUT TO 10 <

t MILES ON STORED FEED l

t I

5i:1 [.T EXISTING EVALUATION CRITERIA RELATED TO EXTERNAL EVENTS FACILITIES AND EQUIPMENT LICENSEE PROVISIONS TO OBTAIN METEOROLOGICAL, HYDROLOGIC, SEISMIC DATA PROTECTIVE RESPONSE M LICENSEE PROVISION FOR EVACUATION ROUTES FOR ON-SITE PERSONNEL INCLUDING ALTERNATIVES FOR INCLEMENT \

WEATHER, TRAFFIC AND RADIOLOGICAL CONDITIONS g

STATE AND LOCAL ORGANIZATION PLANS FOR DEALING WITH POTENTIAL IMPEDIMENTS SUCH AS SEASONAL IMPASSABILITY OF ROADS FOR USE AS EVACUATION ROUTES AND CONTINGENCY MEASURES O O O

APPENDIX XIV PROBABILISTIC ESTIMATES OF EXCEEDING SEISHIC DESIGN LEVELS

/

G G PROBABILISTIC ESTIMATES OF EXCEEDING SEISMIC DESIGN LEVE i e 50TH PERCENTILE LLNL (10 SITES) OTHER (6 SITES) EPRI (9 SITES)

OBE 0(10-2 To 10-3) 0(10-2 To 10 3) 0(10-3 To 10-4)

SSE 0(10-3) To 10-4) 0(10-3 To 10'4) 0(10-4 To 10-5) l 2xSSE 0(10-4 To 10-5) 0(10-4 To 10- 5) 0(10-4 To 10-6) ,

4xSSE 0(105) To IC-6) -

0(10-5 To 10-7) e ISTH PERCENTILE-FACTOR OF 5, 10 OR MORE LESS THAN 50%

6 85% PERCENTILE-FACTOR OF 5 TO 20 MORE THAN 50%

I I u e EPRI RESULTS ARE PRELIMINARY TEST RESULTS. CHANGES PARTICULARLY IN GROUND MOTION MODEi.S MAY BE FORTHCOMING.

ONLY MAGNITUDE 5 OR GREATER CONSIDERED, Sock-97r AssoteEO e

LLNL RESULTS MAY CHANGE AS A RESULT OF ADDITIONAL FEEBACK.  ;

e OTHER RESULTS INCLUDE 4 PRAS. RESULTS ARE SOMEWHAT LESSl THA LLNL.

e ESTIMATES ARE BASED UPON EXPERT OPINION AND JUDGEMENT FOR RANGES WELL BEYOND EXISTING DATA SETS WITH LITTLE UNDERSTANDING OF CAUSATIVE MECHANISMS

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APPENDIX XV SECY 84-320

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August 9, 1984 POLICY ISSUE s m 86320 (InfOrmatiOn)

For: The Commissioners From: William J. Dircks Executive Director for Operations

Subject:

NRC STAFF COMMENTS TO ENVIRONMENTAL PROTECTION ACENCY (EPA) ON THE SCIENCE ADVISORY BOARD REPORT ON PROPOSED EPA STANDARD FOR MANAGEMENT AND DISPOSAL OF SPENT NUCLEAR FUEL, HIGH-LEVEL AND TRANSURANIC WASTE (40 CFR PART 191)

Purpose:

To inform the Commission of comments of the NRC staff on the EPA Science Advisory Board's . . ort. ,,

Discussion: In January 1983, the Environmental Protection Agency (EPA) formed a subcommittee of its Science Advisory Board to review the technical basis for the proposed 40 CFR Part s , ) 191, Environmental Standards for the, Management and C/ Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes. The Subcommittee prepared.a report that contains a number of findings and recommendations which the EPA is considering incorporating into the final version of 40 CFR Part 191. On May 8, 1984, the EPA published in the Federal Register a notice of availability of the report, and encouraged the public to comment on it (49 FR 19604, Enclosure 1). Since the report contains j recommendations affecting the Commission's ability to ' )

implement the standard, and other matters on wh'ich the 1 Commission and staff had previously commented, the staff commented to EPA on these matters. A copy of the staff comments is contained in Enclosure 2.

T

Contact:

D. Fehringer/J. Linehan, WMRP 427-4177

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The Commissioners r-L On July 19, 1984, Acting Chairman Roberts wrote to i Administrator Ruckelshaus, suggesting a possible resolution of the jurisdictional issue regarding the l procedural and assurance requirements contained in  ;

EPA's proposed HLW standard. If the recommendations of  !

Acting Chairman Roberts' memorandum and of the staff's ' '

comments on the Science Advisory Board report are adopted by EPA, the staff considers that all of its concerns with the EPA HLW standard would be resolved. i

.N b 111am J. Dirck A

E cutive Director for Operations w.

Enclosures:

1. 49 FR 19604 - Commissioners, SECY, OGC and OPE only. I

2. Staff Comments-b

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PAltomare KGoeller DFehringer & r C JLinehan a t Central Docket Section (LE-130) LBHigginbotham EConti U. S. Environmental Protection Agency meH a HJMiller ATTN: Docket No. R-82-3 RRBoyle l Washington, DC 20460 SCoplan 1 JKennedy i The Nuclear Regulatory Comission (NRC) staff is pleased to respond to the l request by the Environmental Protection Agency (EPA) for comments on its -

Science Advisory Board's review of the proposed EPA standards for management and disposal of spent nuclear fuel, high-level and transuranic radicactive wastes (40 CFR Part 191). The NRC staff shares the concerns of the Science Advisory Board (SAB) regarding the implementability of the proposed standards, and several of our comments address this concern. Our specific comments are listed below. In each case the SAB's recomendation is first presented, followed by the NRC staff's comment.

EXECUTIVE

SUMMARY

RECOMMENDATIONS A. The Standard

1. The Subcommittee recommends that the release limits soecified in Table 2

-- of the proposea standaras ce increasea by a factor of ten, tnereoy causing a related ten folo relaxation of the prooosed societai cojective (oooulation

_ risk of cancer).

The NRC staff considers that the proposed release limits can be achieved provided that the implementation concerns expressed in the NRC's formal comments on the proposed standards (comment letters dated May 10 and 11,1983) are resolved. (Specific issues of concern are discussed below.) Thus, from the point of view of implementation of the standards, the NRC staff would not consider it necessary to increase the proposed release limits.

B. Uncertainty and the Standard

1. We recommend that the probabilistic release criteria in the draft standard be modified to reaa " analysis of repository performance snale demonstrate tnat there is less thar a 50s cnance of e.vceecino tne Taale c limits, modified as is appropriate. Events wnose mecian frecuency is less tnan one in one-thousand in_10,000 years need not ce considerea."

2. We recontend that use of a cuantitative probabilistic condition on the

, modified Tabla 2 release ilmits ce made decencent on EPA's aoility to provide evidence that such a condition is oractical to meet and will not lead to serious impediments, legal or otnerwise, to the licensing of nign-level waste h-l63

Wi , .ej c :WMRP :WMRP :RES :JWM :RES :DWM

DJFehringer :HJMiller :JLinehan :Econti :MJBell :PCcmella :REBrowning

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406.3.3/DJF/84/06/15/0 ,vt J .; n

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geologic repositories. If such evidence cannot be provided, we recommend that EPA acoot oualitative criteria, such as tnose suggested by tne NRC.

The NRC staff believes it is important to recognize a distinction between a standard of performance and the quality of evidence that the standard has been met. A quantitative probabilistic standard of performance can be appited in aaministrative proceedings. However, when it comes to determining the evidence-that should be admitted, or in evaluating the probative force of that evidence, different considerations come into play. These matters involve judgments which are qualitative. We would regard it as arbitrary and inappropriate to rule, as an absolute principle, that a certain median frequency of occurrence should always be the basis for considering particular events. And while we could apply a preponderance-of-evidence standard of proof (in place of the suggested 50%), we regard the reasonable assurance standard as being more restrictive and better suited to protection of public health and safety. Ac;ordingly, the NRC staff would take exception to the above recommendations. Moreover, it is noted that the recommendation are more matters of implementation than environmental standards as such. It may nevertheless be possible to meet the intent of the .

SAB recommendation in an appropriate manner that would be implementable in a license review. The NRC staff considers that the revised wording of the containment requirements set out on page three of this letter would

" > achieve this goal.

x

)

~ C. The Time Frame . 10,000 Years and Beyond

1. We recommend that EPA retain the 10,000-year time ceriod as the basis for cetermining tne aceouacy of recository performance. We celieve tnat use of formal numerical criteria limited to this accrox1 mate time cerloa is a scientifically acceotable regulatory aporoacn.

2. We recommend that the process of selection of sites for discosal systems ,

also take into account potential releases of radioactivity somewnat oeyona 10,000 years. Particular attention sncula ce focused on potent 1al reieases of long-lived alona-emitting racionuclices end tneir cecay orocucts.

The NRC staff agrees, in principle, with these recommendations. We note, however, that the process of site selection is assigned by the Nuclear Waste Policy Act (NWPA) to' the Department of Energy (00E), and that DOE's siting guidelines include a criterion requiring comparative evaluations of the performance of alternative sites for 100,000 years, fully satisfying the SAB recommendation. Since criteria for site selection are not " environmental standards" as envisioned by the NWPA or by Reorganization Plan No. 3, EPA should not add such a requirement to its standards.

Y'" N '

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406.3.3/0JF/84/06/15/0

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MAIN REPORT RECOMMENDATIONS A. Uncertainty and the Standard

2. We believe that recository designers will find it ouite difficult and perhaps excessively excensive to cemonstrate witn reasonable or nign assurance tnat the levels of orotection sougnt by EPA in the draf t standara nave oeen met.

The following wording has been suggested, in recent discussions with the EPA staff, for the containment requirements:

191.13 Containment Recuirements (a) Disposal systems for spent nuclear fuel or high-level or transuranic wastes shall be designed to provide a reasonable expectation, based upon performance assessments, that the cumulative releases of waste to the accessible environment for 10,000 years after disposal from all significant m.

t processes and events that may affect the disposal system shall: (1) have a likelihood of less than one chance in 10 of exceeding the quantities calculated according to Table 1 (Appendix A), and (2) have a likelihood of less than one chance in 1,000 of exceeding ten times the quantities calculated according to

[o '7able 1 (Appendix A).

v (b) Performance assessments need not provide complete assurance that the ~

requirements of 191.13(a) will be met. Because of the long time period involved and the nature of the events and processes of interest, there will inevitably be substantial uncertainties in projecting disposal system performance. Proof of the future performance of a disposal system is not to be had in the ordinary, more short term, sense of the word. Instead, what is required is a reasonable expectation, on the basis of the record before the implementing agency, that compliance with 191.13(a) will be achieved. .

The NRC staff considers that this revised wording of the containment requirements will alleviate the concerns of the SAS, and will be implementable in a licensing review.

7. We find that an acoroach to the EPA standard emoloying " individual dose limits" (considering some " maximally exposec inciviauai," or alternativeiy scme " average exposea individual") woula in oractice make the standara difficult to meet witn nign assurance for very lona times for any repository concept currently under active consideration. However, we recommend tnat for tne first 500 years, tne EPA stancara emoody an extremely low iikelinood that increases in radioactivity acoroacning tne limits aliowea oy tne EPA arinking water standaras will occur in octable weli water drawn trcm any A - /W  !

lWh. - je :WMRP :WMRP :RES :0WM :RES  :

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___.....___:.... __..__.:DWM .___.......

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6/ /84 :6/ /84 :6/ /84 :6/ /84 :6/ /84 :6/ /84 l c j

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-s 406.3.3/DJF/84/06/15/0 4

M. 6 2C4 x-v well adjacent to the site of the repository. For lancer time ceriods, we recommena tnat EPA rely on tne assumotion tnat stancaras similar to tne present arinxing water stancaras will exist to protect groups of individuals.

The NRC staff agrees with this recomendation. We particularly note tne SAB's use of the phrase " adjacent to the site of the repository". We think it is clear that the SAB intends that groundwater protection requirements be applied only to groundwaters beyona the geologic barrier which serves as part of an overall repository system (e.g. , groundwaters beyond the " controlled area" as defined in 10 CFR Part 60). We believe that any groundwater protection requirement adopted by EPA should be applied as intended by the SAB.

E. Assurance Reouirements

1. We recommend that the assurance reouirements, as amended by this recort, be rulesuomitted package, as a Federal Raaiation Guidance cocument in succort of tne EPA s.

The NRC staff agrees with this recomendation, and notes that the NRC's formal comments on the proposed standards identified the assurance requirements as being inappropriate for an environmental standard and recommended that they be The NRC staff also agrees with the recomended specific changes O(eleted.

Reccamendations E.2-E.7, not listed here) if these provisions are to be published as Federal Radiation Protection Guidance.

Reccgnition of the similarity between several of the proposed assurance requirements and the corresponding provisions of the NRC's regulation, 10 CFR Part 60, suggests a possible alternative solution. This would take the form of modification of Part 60 if and as appropriate so as to incorporate the principles of EPA's proposed assurance requirements. In view of these changes, EPA would delete the assurance requirements frem the stancard. Thus, problems ,

of overlapping jurisdiction between the two agencies (and potentially conflicting requirements) could be avoided. The NRC staff will continue discussions with the EPA staff to explcre the feasibility of this approach.

L. Limit Standard to Mined Geologic Recositories

1. We recomend tha't the acolicability of Suboart B of the crocosed standards be explicitly restrictec by EPA to disposal in mined aeolog1c reposit]rles.

The NRC agrees with this recomendation for the reasons cited by the 3AB and because of the direction in NWPA to develop standards for disposal in mired geologic repositories.

h fY

W .ejc :WMRP :WMRP :RES :DWM :RES :DWM

DJFehringer :HJHiller :JLinehan :Econti :MJBell :PComella :REBrownin

.:__........-_:___.........:........____:.......__...:.....____ __:..___.......:.....__..g ..

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406.3.3/0JF/84/06/15/0

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' N._./

N. High-Level Radioactive and Transuranic Wastes Definitions

1. We recommend that EPA's definition of high-level radioactive wastes be consistent with that set fortn in the Nuclear waste Policy Act (NWPA) ano coordinated witn the oefinition used by tne Nuclear Regulatory Commission (NRC).

The NRC staff agrees with this recommendation and notes that recent working drafts of the EPA standards have contained an appropriate definition.

We appreciate the opportunity to comment on the SAB report and will be pleased to consult with the EPA staff on these comments or other matters that will assist in publication of the final standards.

Sincerely,

, . . T .- .:i t7 (t3Aira J. I'%U '

m.

Robert E. Browning, Director Division of Waste Management Office of Nuclear Material Safety and Safeguards

/

n T

\v) l cSee previous concurrence page. p/. jh ~fI -

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REFERENCEABL M Y TERMS 655w-DESIGN AUTHORITY (SAPS) TERM 5+ Years GESSAR II B.3.b(1)(i) 8 I

i 5 Years h

, CESSAR-F B.3.b(1)(ii) '

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APPENDIX XVII ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE

1. Memorandum, W. Kerr, Member, to ACRS Members, State of Nuclear Power Safety, October 12, 1985

2. Memorandum, S. J. S. Parry, ACRS Senior fellow, to D. Okrent, ACRS Member, Report on the Identification and Analysis of Probable Areas of Serious Contcr: tion or Uncertainty During the Licensing of a Geologic Repository for the Disposal of High-Level Radioactive Waste, September 19, 1985

3. Letter, H. E. Collier, Jr. , Chairman, High-Level Radioactive Waste Disposal Subcommittee, Science Advisory Board U.S. Environmental Protection Agency (EPA) to W. D. Ruckelshaus, Administrator, U.S.

EPA, Science Advisory Board Review of EPA's proposed environmental standards for the management and disposal of spent nuclear fuel,

-m high-level 7nd transuranic radioactive wastes (40 CFR 191),

/

) February 17, 1984 v

4. Executive Summary, Report on the review of Proposed Environmental Standards For the Management ar.d Discosal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes (40 CFR 191),

High-Level Radioactive Waste Disposal Subcommittee, Science Advisory Board, U.S. EPA, January 1984

5. Partial Listing of FY 1985 ACRS Reports Indicating NRC/NRC Staff Response, October 12, 1985

6. Certified Minutes of Joint Meeting of Waste Management and Site Evaluation Subcomittees held on June 18-19, 1985 m

1

/- / 69'