ML20214A461
| ML20214A461 | |
| Person / Time | |
|---|---|
| Issue date: | 05/15/1987 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2489, NUDOCS 8705190464 | |
| Download: ML20214A461 (57) | |
Text
. . . .
.m S. ; . iH Q y:j u b La J u CW TABLE OF CONTENTS MINUTES OF THE 322ND ACRS MEETING
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FEBRUARY 5-7, 1987 WASHINGTON, D.C.
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- I. Chairman's Report (0 pen) .................................. 1 II. Naval Reactors (Closed) ................................... 1 i
III. Advanced Reactors (0 pen) .................................. 1
- IV. Safety Research Program (0 pen) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 V. Quantitative Safety Goals (0 pen) .......................... 7 i
VI. Surry Nucl ear Station , Uni t 1 (0 pen) . . . . . . . . . . . . . . . . . . . . . . 10 ,
VII. Reactor Operating Experience (0 pen) . . . . . . . . . . . . . . . . . . . . . . . 12 VIII. Executive Sessions ........................................ 19 A. Reports, Letters and Memoranda ....................... 19
- 1. Report to Congress on the NRC i Reactor Safety Research Program and B udget fo r F Y 1988 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
- 2. Testing of c.harcoal Adsorption Capacity ......... 19 l
- 3. Naval Reactors Moored Training Ship Demonstration Project ........................... 19
- 4. Naval Reactors Analytical Studies . . . . . . . . . . . . . . . 19 i B. Subcommittee Reports ................................. 19
- 1. WasteManagement(0 pen)......................... 19
- 2. Seve re Accidents (0 pen) . . . . . . . . . . . . . . . . . . . . . . . . . 24
- 3. Regulatory Policies and Practices (0 pen) ........ 25
- 4. Planning (0 pen)................................. 25 C. Other Committee Conclusions .......................... 26 1 1. GE Advanced BWR ................................. 26
- 2. Standardized Plants ............................. 27
- 3. Electri cal Surge Protection . . . . . . . . . . . . . . . . . . . . . 27
! 4. Containment Performance ......................... 27 E
8705190464 870515 PDR ACRS Cortiflod D 7-ACRS-2489 PDR _u -
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D.. ' Future Activities..................................... 27
- 1. Future Agenda ................................... 27
- 2. Future Subconni ttee Activi ties . . . . . . . . . . . . . . . . . . 27 i
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APPENDICES TABLE OF CONTENTS 322ND ACRS MEETING FEBRUARY 5-7, 1987 Appendix I List of Attendees Appendix II Tentative Agenda Appendix III Future Subcomittee Activities Appendix IV Other Documents Received 1
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- Federal Register / Vol. S2. No. 21 /.Enday, February 4,-1987fWodces ha $181 l Advisory Committee on Reactor Safeguards; Subcommittee on 3 5:30pm.-e15p.m.rStandardised. , .(Open)-Discuss .
A WaclectPlants(Open)-Oiscuse - .C position /commerds seassdingtivm Standardization of Nuclear Facilities; proposed ACRS commento reganiing Postponed Meeting - -
improvements in standardized nuclese:
proposed ACRS solela the yN of plants
- radioactive waste mamapspentg.'
The Federal Register published o' n -- % -
liisposal. . . ~~t gjgp.m.-edspma Cortoinment 100p.di.45tM*'[NtIk; .. '
Friday. January 23.1987 (52 FR 2632) -
I contamed notice of a meeting of the 1beformange(Open)-Discussproposed Subcommittee Adifvityl(j p eis/M)-- ~ ;
ACRS Subcommittee on Standardizatio. n . NRC Staff resoludon SRV Discharge une Breakin the , . ;..:
of Generic Issue 81. - Hear and discuss re3 of Nuclear Facilities to be held on' . {
Wednesday. February 11.1987. 8:30 a.(m . Airspace of Mark I and Mark II - T ACRS subemittee ActMtlet faq[trding Suppression Pools, and other pool safety rels ted and regulatory activities of the NRC .
' Was ingt n. Thi eting"has been #*
postponed until further notice.
. . . y a-s of thinr'sessioniett!be' closed I Friday, February 5.1987 - -
' as recessary to discuss Proprietary t Dated;lanuary 28.1987 43044.-ILW a.maNuclearFacility Information applicable to the mtter ;
Morton W. Wrkin. being discussed. {
Opemting Experience (Open/ Closed)- _ _
Assistant Erecurire Directorfor'Pm/ect Discuss recent operating experience and
' Procedures for the'c'o nduc(of arid events at nuclear power plants. participation in ACRS meetings were
[FR Doc.47-1982 Filed t-Jo-67; 8:45 am] ' Portions of this session will be closed published in the Federal Register on
~ 880" " ~" as necessary to discuss Proprietary October 20. toes (51 FR 37241).In Information applicable to the facility eccordance with these procedures oral being discussed. !
Advisory Committee on Reactor or written statements may be presented 11:15canA215pm.-2Wp.ma Surry by members of the public, recordings Safeguards; Meeting Agenda NuclearpowerStation Unit 2(Open)-- wi'l be permitted only during those Briefing and discussion regarding portions of the meeting when a In accordance with the purposes of planned resolution of recent feedwater sections 29 and 182b. of the Atomic transcript is being kept, and questions h'ne failure at this facility. may be asked only by members of the Energy Act (42 U.S.C 2039. 2232b), the . . 2.2p.m.-2:15 p.m.r Edwin I. Hatch Advisory Committee on Reactor Co:nmittee, its consultants, and Staff.
NuclearPlant-Briefing regarding AIT Persons desiring to make oral '-
Safeguards will hold a meeting on -
evaluation of recent incident in which statements should notify the ACP.S February 5-7.1987, in Room 1048.1717 H spent-fuel cooling water was lost.
Street NW. Washington, DC Notice of 2:15pm.-2 Jap.ma Future Activities . Executive Director as far in advance as th!s meeting was published in the (Open)-Discuss anticipated ACRS practicable so that appropriate -
Federal Registar on January 21.1987 arrangements can be made to allow the
' subcommittee activities anditems Hursday February 5.1987 -
Pr posed for consideration by the full necessary time during the meeting for Committee. such statements. Use of still, motion 8:30 a.m.-d:40 a.m.rReport ofACRS 2:30p.mA:JOpaaNRCReactor picture and televisica cameras during Chainnan (Open)--The ACRS Chairman SafetyReseamh Pmgmm (Open)- this meeting may be limited to selected
- will report briefly regarding items of portions of the ueeting as determined Discuss proposed ACRS report to the current interest to the Committee. U.S. Congress regarding the proposed by the Chairman. Information regarding 8
- 40 a.m.-10:45 a.m.r Naval Reactors NRC Safety Research Program for FY ' the time to be set aside far this purpose Troemng Facility (Closed)-Consider 1988. may be obtained by a prepaid telephone proposed operation of a training facility
.4:J0pm.-6:30p.ma Quantitative call to the ACRS Executive Director, l for naval nuclear propulsion plant R.F. Fraley, prior to the meetir.g. In view personnel. Safety Coals (Open)--Discuss propcsesi I
- NRC Staffimplementation plan for the of the possibility that the schedule for '
This session will be closed to discuss NRC Policy Statement on Quantitative ACRS meetings may be adjusted by the classified information. Safety Goals. Chairman as necessary to facilitate the conduct of the meeting. persons Rea r 0 pen)-D s of Saturday, February 7,13s7 planning to attend should check with the
?
established technology and
~ a:Joas41.2 ama Prepamtion of ACRS Executive Director if such standardization of DOE non. water A CRS Reports (Open/Close d}-Discusa resched ing would result in malor -
cooled nuclear power plants. proposed ACRS reports to the NRC and cony once.
1;oop.mA.mp.m.:NRCSofety . the U.S. Congress regardingitems
. Reseamh Pmgmm (Open)-Discuss
.I have determined in accordance with considered during this meeting. A .
subsection 10(d) Pub. L 92-463 that it is proposed ACRS report to the United -
proposed report on electrical surge .. . . necessary to close portions of this States Congress regarding the proposed protection in nuclear power plants wilj. meeting sa noted above to discuss ,
. NRC Safety Research Program for FY 1968.. . '. elso be discussed. . ..
Portions of this session will be closeu
, information that involves Proprietary J:00p.mA:30p.m.: Quantitative .y information (5 U.S.C 552b(c)(41) and a' s necessary to discuss Proprietary classified information (5 U.S.C
. Safety Cools (Open)-Discuss proposed ~
NRC Staff plan for implementation of Information and Classified Information ' . 552b(c)(1)) applicable to the facility applicable to the matters being being discussed.
the NRC Policy Statement on discussed.
Quantitative Safety Goals. .
Further information regarding topics .
11:15 a.m.-1245p.ar.: Advanced 4:45pm.-5;JOp.m.: Radioactive to be discussed. wkether the meeting _
Boiling WaterReactor(0 pen)-Discuss .
Waste Management andDisposal has been cancelled or rescheduled. the l major issues associated with the scope Chairman's rullrig on requests for the (Open}-Briefing proposed by NRC ACRS activities Staff regarding of and basis for the licensing review of . opportunity to present oral statement!
in the . . ' this facility. !
regulation of radwaste management and ' 1:45p.m.-2.mp.maRadioactive . and the time allotted can be obtained by l disposal. ; - .
- a prepaid telephone call to the ACRS
, , Waste Management andControl +
Executive Director. Mr. Raymond F.
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/s2 Federal Register / Vd. 52. N2 21 f Mond:y, Fabruary 2,1987 / N tices l
[Fraley (telephone 202/634-3285),evaluated; or (2) create the possibility of Written" comments may be submitted between a:15 a.a and 5:00 p.a a new or differem kind of accident from by mailto the Rates and Pmeedures Dated:Jenaary 27.19s7 any accident previously evaluated: or (3) Branch. Division of Rules and Records, lohn C. Hoyle, la olve a sigmficent reduction in a Office of Administrat>on. U.S. Nucleer Advisory Comma. tee Management Officer.
margin of safety. Regulatory Co-taina. Washington, y, g_ ,, g, gg y ,gg ,gnggg egg, g,
[G Doc. a7-198 Mad 16 835 am] pmposed changes to the TSs with publication date and page number of
'" * *CC ' " "8 respect to the three criteria given in to this Federal Registar notice. Written CFR 50.92 and have concluded that the comments may also be delivered to goocent me, so.34sj proposed changes would not involve a Room 4000. Maryland National Bank I significant hazanis consideration as Building.7735 Old Georgetown Road.
.e Toledo Edison Co. and the Cleveland fouows: Bethesda. Maryland, from 8:15 a.m. to Electric tiluminating Co.; Consideration 1. ne accident conditions and 5.100 p.m. Copies of the written
'l of lasuance of Amendment to Facility assumptions are not affected by the comments received may be examined at l Operating Ucense and Proposed No proposed TS changes. The proposed Us the NRC Public Document Room.1717 H Sigolficant Hazards Conalderation continue to meet the safety function of Street NW., Washington DC. The filing
.( Determination and Opportunity for the Limiting Condition for Operation of requests for hearmg and petitions for Hearing and Surveillance Requirements, leave to intervene is discussed below.
- m. 888 " By March 2. m se Ucusus may The U 3 iny !ve a signifi ant increase in the g -M. Nucieer Regulat Iis Ib Cami file a request for a hearms with respect probability or consequences of an to issuance of the amendment to the cg'""8 **",* 'cc reviously evaluated (to CFR subject facility operating licenae, and
,, d enaeb g cht issued to Toledo Edison Company and any person whose intere t may be
- 2. No station equipment is modified by ne Cleveland Electnc Illuminating these changes and the station batteries a ec e y 8pmce g an , o Company (the licensees), for geratim * * ' par % ate as a party in b and distribution system will conthme to pmcee&.ng must & a Wm peWon cf the Davis-Besse Nuclear Power be tested and inspected to ensure the Station. Unit No.1. located in Ottawe *** ' " * " * ' '9"'**'"*
operability. Therefore, no new or Conty, Ohio. hearing and petitions for leave to different kind of accidst from eny ns amedment wald mvise Techrucal SpeciScation (TS) Sections previously anal ed is created.
Therefore, the c anges would not creete with the Commission 'W Rus of s .D '##
3.8.2.3. 4.3.2.11, 4.8.2.12. and 4.8.14.1 the possibility of a new or different kind * * " "" ##"8 and the associated Basis Section 3/4J in of accidtnt from any previously e ga s 1 2. U a cccordance with the licensees' analyzed (to CHL 50.92(c)(2)).
reqant fw a heanng w peMm fw application im amends-f dated 3.ne operability of the A.C. and D.C. leave to intervene is filed by the above January 21. E ness W Sectima date, the Commission or an Atomic power sources and associated relate to the D.C. electrical power distrhtion systems will continue to be Safety and bwnsing Board, designated distrihutinn system during plant demonstrated to ensure sufficient power by 6e Nnnman w by de Chamnan operation and when shutdown.% is available to supply the safety.related f the Atomic Safety and ucensmg proposed revisions would (1) cbsmae eqmpment required for (1) the safe Board Panet win rule on 6e mqant certain nomenclature used in the ne to and/w petition, and the Secmtary w the shutdown of the station. (2) the designated Atomac Safety and Lcensing tha specific equipment designations mitigation and control of accident used at Davis-Besse, and (2) revise the B nid willissue a notice of hearms or conditions within the station. (3) the an appmpnate arda.
Surveillance Requirements for D.C. maintaining of the station in the i g
Distributian using the guidance of the shatdown or refueung condition for As required by 10 CFR 2.714, a h mndel tuhl specifications for extended periods of time and (4) Petition for leave to intervene shall set st tion hatteries Lasued by the NRC on . Instrumentation and control capability forth with particularity the interest of ,
July 18,1981. The NRC gnid== was for monitoring and maintaining the the petitioner in the proceeding and how j based upon Regulatory Calde 1.129, station's status. Therefere. the changes that interest may be affected by the i F:b:uary tr8, and IEEE Standard 450- would act involve a significant results of the proceeding.% petition l 1980. De current Davis-Bessa TS seblius in the margin of safety (1g . should specificaDy explain the reasons j Survebance Requirements are based on CHL 50.92tet(37). why intervention should be pennitted j e:rlier revisions of these documents. De Commission agrees with the with particular reference to the Before lasuance of the proposed licensees' evalastion regarding the following factore:(1)% nature of the license amendment, the Commissian application of the criteria of to CFR petitioner's right under the Act to be will have made findings required by the 50.92, and proposes to determine that made a party to the proceeding: (2) the Atomic Energy Act of1954, as amended the proposed changes likely do not nature end extent of the petitioner's J (the Act) and the Commission's involve a significant hazards property, financial, or other interest in l regulations. consideration. the proceeding; and (3) the possible no Commission has made a proposed determination that the amendment The Conunission is seeking public comments on this rvycad effect of any order which may be entered in the proceeding on the .
)
request involves no significant hazanis determination. Any comments received petitlener's interest.W petition should cca.'L tion. Under the Commisslorfs within 30 days after the date of also identify the specific aspect (s) of the
. sedations in to CFR 50.92, this means publication of this notice wiH be subject matter of the proceeding as to thawi.L of the facility la considered in maldag any finel which petitioner wishes to intervene.
- d. veth the proposed . determinetion.W Gommission will net Any person who has filed a petition for asesidment would not (1) involve a normally make a Snal determanshoa. leave to intervene or who has been .
signfBcant inceese in the probability er unless it receivese rewest for a admitted as a party may amend the I consequences of an accident previously hearing. . petition without requesting leave of the Oe. . .
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UNITED STATES l' n NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS O, WASHINGTON, D. C. 20555
$,4 Revised: February 2, 1987 i
SCHEDULE AND OUTLINE FOR DISCUSSION l 322ND ACRS MEETING FEBRUARY 5-7, 1987 4 WASHINGTON, D. C.
Thursday, February 5, 1987, Room 1046, 1717 H Street, NW, Washinaton, D.C.
i 1) 8:30 - 8:45 A.M. Report of ACRS Chairnan (0 pen)
- T 1T Opening statement (WK) j 1.2) Items of current interest (WK/RFF) l 2) 8
- 45 - 10:45 A.M. Naval Reactors Training Facility (Closed) 2.1) ReportofACR5 Subcommittee (WK/PAB) 2.2) Meeting with representatives of NRC Staff &
DOE / Naval Reactors
[ Note: This session will be closed to discuss
! Classified Information per 5 U.S.C. 552b(c)(1)]
10:45 - 11:00 A.M. PREAK
- 3) 11:00 - 12:00 Noon Advanced Reactors (0 pen) 3.1) Report of ACRS subcommittee regarding
- advanced DOE nonlight water reectors
- (MWC/MME) j 3.2) Meeting with representatives of the NRC
- staff and DOE, as appropriate
] 12:00 - 1:00 P.M. LUNCH
- 4) 1:00 - 3:00 P.M. ReactorSafetyResearch(0 pen) 3 4.1) Discuss proposed ACRS report to the U.S.
! Congress regarding the proposed NRC
- Safety Research Program for FY 1988
! (CPS,etal./SD,etal.)
, 5) 3:00 -
4:30 P.M. Quantitative Safety Goals (0 pen) 5.1) Discuss proposed NRC Staff plan for
! implementation of the Commission Safety Goal i
Policy Statement (D0/RPS) 5.2) Meeting with NRC Staff, as appropriate l
4:30 - 4:45 P.M. BREAK i
e 322nd ACRS Meeting Agenda .
- 10) 4:45 - 6:45 P.M. Surry Nuclear Station, Unit 1 (0 pen) 10.1) Report of ACRS subcommTttee chairman (Metal Components) regarding the failure of the feedwater system pump suction line and related system malfunctions (PGS/EGI:JCE/HA) 10.2) Report by representatives of the NRC Staff and the licensee regardinp this incident
322nd ACRS Meeting Agenda Friday, February 6, 1987, Room 1046, 1717 H Street, NW, Washington, D.C.
- 9) 3:30 - 11:15 A.M. Reactor Operatino Experience (0 pen / Closed).
9.1) Report of ACR5 subcommittee chairman re-recent events at nuclear power garding(JCE/HA) plants 9.2) Meeting with representatives of the NRC Staff
[ Portions of this session will be closed as necessary to discuss Proprietary Information applicable to the facility being discussed.]
11:15 - 11:30 A.M. BREAK
- 6) 11:30 - 12:30 P.M. Radwaste Manacement and Disposal (0 pen) 6.1) Comments by ACRS Subcomittee Chaiman (DWM/OSM) 6.2) Briefing by and discussion with repre-sentatives of NMSS regarding proposed ACRS role in the NRC licensing and regulation of radwaste management and disposal activities (DWM/OSM) 12:30 - 1:30 P.M. LUNCH Future Activities (0 pen)
- 11) 1
- 30 - 1:45 P.M.
11.1) Discuss anticipated subcomittee activities
, (MWL) 11.2) Discuss proposed ACRS activities (WK/RFF)
- 12) 1:45 - 3:45 P.M. Reactor Safety Research (0 pen)
_ 12.1) Discuss proposed ACRS report to the U.S.
Congress (CPS, et al./SD, et al.)
- 13) 3:45 - 5:45 P.M. Quantitative Safety Goals (0 pen) 13.1) Discuss proposed NRC plan for implementation of the NRC policy statement on Quantitative Safety Goals (D0/RPS)
- 7) 5:45 - 6:15 P.M. Standardized Nuclear Plants (0 pen) 7.1) Discuss proposed ACRS report on improvements in standardized nuclear plants (JCE/HA) i i
l
I l
322nd ACRS Meeting Agenda 8) 6:15 - 6:45 P.M. Containment Performance (0 pen) 8.1) Report of ACR5 member and discussion regarding proposed priority for resolution ,
of Generic Issue 61, SRV Discharge Line '
Break in the Airspace of Mk I and Mk II Suppression Pools and consideration of other pool bypassing mechanisms (CYM/PAB) 1 O
h
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O, l-1 322nd ACRS Meeting Agenda f j
f Saturday, February 7, 1987, Room 1046, 1717 H Street, NW, Washington, D.C.
l 14) 8:30 - 11:30 A.M. Preparation of ACRS reports (0 pen / Closed) 14.1) 8:30-9:30: NRC Safety Research Program ll (CPS /et al./5D et al.)
14.2) 9:30-10:00: Naval Training Facility
! (WK/PAB) i 14.3) 10:00-10:45: Implementation of Quantitative Safety Goals (preliminary
! report regardina proposed plans for implementation - Tentative)
(D0/RPS) 14.4) 10:45-11:00: Electrical Suroe Protection for Nuclear Power Plants (CJW/MME) l 14.5) 11:00-11:30 (tentative) - Operating experience in nuclear plants (as needed)
- 15) 11:30 - 12:45 P.M. GE Advanced BWR (0 pen) 4 15.1) Discuss major issues regarding the licens-4 ing review (LBA) for this standard plant design (D0/RKM) ,
12:45 - 1:45 LUNCH j 16) 1:45 - 2:00 P.M. Radwaste Management and Disposal (0 pen) j 16.1) Discuss ACRS position regarding proposed ACRS role in the regulation of radwaste
! management and control (DWM/OSM) l 17) 2:00 - 3:00 P.M. ACRS Subcommittee Activities (0 pen) 17.1) Hear and discuss reports of ACRS j- subcommittees on:
- 17.1-1) 2
- 00-2:30: Severe accidents -
l Implementation Plan (WK/MDH)
Subcommittee meeting on i
- December 19, 1987 1 17.1-2) 2
- 00-2:45 Regulatory Policies and Practices - Subcommittee meeting on Jan. 14,1987(HWL/GRQ) 17.1-3) 2:45-3:00: ACRS Planning Sub-l committee meeting on Jan. 9, 1987 !
); (WK/RFF) 4 I
, [O m mam=mn LaU u u du I
, FOIA EXEMPTION b(6)
MINUTES OF THE 322ND ACRS MEETING-FEBRUARY 5-7, 1987 4
I The 322nd meeting of the Advisory Comittee on Reactor Safeguards, held at i 1717 H Street, N.W., Washington, D.C., was convened by Chairman W. Kerr at 8:30 a.m., Thursday, February 5,1987.
[ NOTE: For a list of attendees, see Appendix I. Mr. C. Michelson and Mr.
- G. Reed did not attend the meeting; Mr. C. Wylie did not attend on February 5 -
and Dr. P.G. Shewmon did not attend on February 6 and 7.]
The Chairman noted the existence of the published agenda for the meeting, and
! identified the items to be discussed. He noted that the meeting was being held in conformance with the Federal Advisory Comittee Act and the Govern-
! ment 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 j was being taken, and would be available in the NRC 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 Reports, Inc., 444 North Capitol ' Street,
, purchase from ACE-Federal Washir.gton, D.C. 20001.]
I. CHAIRMAN'SREPORT(0 pen)
[ NOTE: R.F. Fraley was the Designated Federal Official for this portion of themeeting.]
! The Chairman noted that the NRC Staff had issued its report to the Comission i on the Chernobyl Accident. He said that copies will be provided to the ACRS.
I The Chairman reminded the Members that any coments on the long version of i the Wingspread Meeting Sumary should be provided to Dr. McCreless.
II. NAVAL REACTORS (Closed)
See Classified supplement.
III.ADVANCEDREACTORS(0 pen) l [ NOTE: M.M. El-Zeftawy was the Designated Federal Official for this portion
- of the meeting.]
- Dr. Carton - stated that the Subcomittee on Advanced Reactor Designs met on February 4,1987 in Washington, D.C. to review the use of proven technology and standardization in the Department of Energy (DOE) advanced non-LWR t designs. In ' addition, the Subcomittee discussed a draft Comission paper
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322ND ACRS MINUTES r .
prepared by the NRC Staff regarding standardization of advanced reactor l Dr. Carbon indicated that DOE and its contractors are currently designs.
developing the designs of three advanced reactors (one gas-cooled, and two liquid metal). In early 1984, 00E. requested the NRC to provide information to the designers on requirements for the licensability of the design. It was requested that this guidance be prov:ded to the designers early in the design prior to any formal application. The ACRS agreed and urged early !
interaction. The NRC agreed and a schedule - for collaborative effort was established. Preliminary Safety Information Documents (PSIDs) have been prepared by DOE and its contractors on each of the three designs and were submitted to the NRC for review in September (HTGR) and November 1986 (LMRs).
The PSID is basically a description of the conceptual design, including proposed licensing criteria and safety analysis to illustrate plant response to accident conditions. PRA and a description of the supporting R&D programs are also to be provided for each of the designs. The output of the NRC -
Staff's review of the PSIDs would be a Safety Evaluation Report (SER) on each concept. The SER will provide guidance on the licensing criteria to be applied and the potential of the designs to meet those criteria. -
The NRC Staff intends to prepare and submit Commission papers on three topics: (a) Standardization, (b) Severe Accidents, and (c) Containment.
In accordance with the Comissioner's advanced reactor policy statement, which states that the ACRS should be involved early in the review, the Staff has asked the Subcommittee to review and provide comments on each of the three topics. In addition, the Staff has asked the Subcommittee to review other key issues (e.g., non-safety grade control room, control of multimodu-lar plants, and proposed use of metal fuel instead of oxide fuel for LMRs).
Dr. Carbon emphasized that there are large differences between the new conceptual designs being presented and conventional LWRs.
Mr. Tom King, Section Leader, Safety Program Evaluation Branch /NRR, stated j that the three DOE sponsored advanced reactor concepts have as their objec-4 tive the develop:nent of a standardized plant design which would be submitted j to the NRC for design certification and approval. Mr. King indicated that ,
- the advanced designs have unique characteristics, such as:
I l
The concentration of safety functions in the nuclear island of the
- plant, i
l The use of modular reactor designs, including extensive shop i fabrication of the modules, and provisions for staggered on-site i module installation and operation,
- Less operating experience to support the designs as compared to the
! LWR, and i
i Varying degrees of design detail planned for submittal for certi-i fication.
l Due to these unique characteristics, the NRC Staff is raising several key issues regarding what the NRC should require in order to certify a new
322ND ACRS MINUTES .
reactor concept. These issues can be stated in the form of questions as follows:
(1) What plant systems, structures and components should be reviewed to be able to certify the design?
(2) What level of design detail should be provided for review on those systems, structures and components provided for certification?
(3) What level of operating experience, existing technology and sup-porting R&D is required to support certification (i.e., is a prototype plant required to be built and operated prior to design certification)?
(4) What information should be provided to allow flexibility in the design certification for variations in plant size (i.e., number of modules)?
(5) Is a manufacturing license (10 CFR 50, Appendix M) required prior to shop fabrication of reactor modules.
Mr. King pointed out that the Staff's resources (in NRR and Research) are approximately 8 FTEs and $1500K of Technical Assistance (TA) in FY 87 and 5 FTEs and $900K (TA) in FY 88.
Dr. Carbon mentioned that the written material which was presented to the Subcomittee on February 4,1987 indicated that the topic for discussion was standardization. However, after discussion with the Staff, it was recognized that the topic is certification rather than standardization.
Dr. Okrent stated that there is a procedural problem with the Staff asking for Subcomittee's verbal comments and not for the full Comittee's coments.
Dr. Kerr concurred with that and stated the only formal mechanism by which the ACRS has for providing comments to the Comission is by writing letters; it is not through discussion. Dr. Kerr also mentioned that it would be a mistake if the NRC Staff after discussion with the ACRS concludes or indicates to the Comission that an ACRS position has been drawn.
Dr. Mark urged the Staff not to carry over General Design Criteria (GDC) terms that do not apply to the new designs.
l Mr. King briefly described the Modular High-Temperature Gas Cooled Reactor (MHTGR). The MHTGR plant design is a 350 MWe standard reactor module being developed in conjunction with Gas Cooled Reactor Associates, Stone & Webster and Bechtel. The concept uses prismatic fuel. There are four modular, steel vessel reactors in a side-by-side configuration that supplies steam to two
. turbine generators. The net plant electrical output is 558 MWe. Each i reactor module is housed in a vertical cylindrical concrete enclosure that is l fully embedded and below grade. The nuclear island portion consists of four reactor enclosures and adjacent structJres that house fuel handling, helium i
processing, and other essential reactor service systems. A comon control l room is used to operate all four reactors and the turbine plant. The design l has no containment. A confinement system is used in the design. The design l
s 322ND ACRS MINUTES : .
i l
) utilizes active systems for normal decay heat removal and reactor shutdown.
Passive means are provided as backup for accomplishing these functions.
Dr. Kerr questioned the statement made by the NRC Staff that there are no safety-related systems in the balance of plant (BOP) and there is no inter-i action between B0P and the nuclear island. Mr. King responded by stating j that there are certain interface requirements that the BOP has to meet in j order to keep the nuclear island or the safety-related portions of the plant ,
i within the design descriptions.
i l' Dr. Okrent expressed some concern regarding the reactor vessel reliability t and steam generator reliability including structural and multi-tube failures.
! Mr. King stated that f*>e NRC Staff does not have a position yet on the 4 reliability issue. However, a leak in the vessel is not catastrophic to this advanced design. The plant will remove decay heat and will shut down. In j terms of steam generator integrity, it can withstand tube ruptures varying i from one tube to the full inventory of feedwater with inherent shutdown i characteristics sufficient enough to compensate for the loss. Mr. King also stated that the Brookhaven and Oak Ridge National Laboratories are exploring the advanced designs over a wide range of accident conditions beyond what DOE
- is promising in their accident analysis. DOE's plan for the MHTGR i certification is not contingent upon a prototype reactor module or the first
] connercial plant being built and tested prior to receiving design j certification.
Mr. King briefly described a 425 Wt modular liquid metal reactor called the
, Power Reactor Inherently Safe Module (PRISM). General Electric (GE) Co. is ;
I the lead design organization. It is a pool design liquid metal reactor. The
- reactor and steam generator are located below grade. The plant uses nine ,
l PRISM reactors. The combined power output of the plant is 1245 MW. The
! containment vessel is located close to the reactor vessel. It has passive j decay heat removal and shutdown systems. The BOP is completely discontected i from the primary loop. It has a 60-year module life and it is plannad to
- utilize the idea of a safety demonstration test to support certification and p j licensing.
i Mr. King briefly described a 900 Et liquid metal standard reactor module >
called the Sodium Advanced Fast Reactor'(SAFR). Rockwell International (RI) i is the lead design organization. Sodium is the primary coolant, with a pool type primary system and passive decay heat removal system. Each module is ;
designed to produce 350 We. RI envisions four 350 We modules per site. -
- The reactor vessel and steam generators are above grade in this concept. It .
{ has heterogeneous core design and two loops per module. The reactor guard l
! vessel and the reactor head are the containment boundary. There is a second
! containment boundary above the reactor head. It has 60-year module life. RI
! claims that radioactive releases during accident conditions will be low 1 L enough so that offsite evacuation plans are not required. RI's intent is to ,
! request design certification on all systems, structures and components except !
site specific items.
l Mr. King pointed out that the standardization / certification of advanced
! reactors poses several unique issues not faced in the standardization of l current generation of LWRs. Criteria for resolution of five issues are being i proposed by the NRC Staff as follows:
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322ND ACRS MINUTES .
Issue #1 - Extent of design to be certified -
Prefer that complete plans be submitted Staff could review for certification less than the complete plans provided the following were met:
(1) Sufficient information is included in the application to allow co pletion of a PRA and safety analysis.
(2) Compliance with interface requirements is verifiable through inspection, testing, previous experience or analysis. Re-liability verification must be based on previous experience or testing.
(3) Certified portion of the design includes all systems, structures and components important to safety.
(4) Representative design for the non-certified portion of the plant is provided as an example of how interface criteria can be met.
Issue #2 - Level of design detail to be certified -
Prefer final design information Staff could review for certification less than final design infor-mation provided the following were met:
(1) The level of design detail provided is sufficient to allow development and review of a PRA and safety analysis.
, (2) The level of design detail provided is sufficient to support procurement, construction and operation of systems, structures and components that meet the performance and reliability characteristics assumed in the PRA and safety analysis.
(3) To aid in the review of the PRA and safety analysis, a representative design for those portions of the plant not finalized is submitted as an example of how the final design will look.
Issue #3 - Prototype Testing -
A prototype plant should be built and tested prior to design certification by NRC unless the following can be demonstrated:
(1) The performance of each safety feature of the plant has been demonstrated via previous experience or full scale testing.
(2) Sufficient performance data exist on each safety feature of the plant to validate safety analysis analytical tools over a full range of operating and accident conditions, including plant lifetime.
1 322ND ACRS MINUTES' '
l (3) System interaction effects among the plant's safety features have been properly accounted for.
Issue #4 - Design certification Options - l 1
A design certification with plant module size / options is acceptable 1 provided that these options are described in the application, l including: (1) variations in or sharing of common systems, (2) variations in interface requirements, and (3) variations in system interaction.
To ensure safe operation of those modules already on line, the PRA and safety analysis should assess each of the options, including any restrictions during the construction and startup phase.
Issue #5 - Manufacturing License (ML) -
Extent of proposed shop fabrication appears to be equivalent to fabrication of major components, not complete plant.
ML is not required unless an essentially ready to operate plant is shop fabricated.
Dr. Okrent expressed some concern that a wide range of information relevant
. to safety and risk will not be obtained from building.and operating a proto-
. type design. The Staff agreed with Dr. Okrent's concern. Dr. Kerr also
' agreed with Dr. Okrent and stated that the information that could be obtained from building and operating a full-scale prototype would be useful in the
- commercial possibility issue rather than in the safety issue.
Dr. Okrent asked if the Staff is departing from previous policy practices regarding principal issues (e.g. , Defense-in-Depth). Dr. Siess pointed out that this is a very important question, however, the main thrust of the Staff's presentation is the draft Comission paper on standardization.
Mr. King stated that the ACRS review and feedback on the issues and proposed staff positions associated with advanced reactor standardization are being sought at this time to allow consideration of ACRS coments prior to 'present-ing a recommendation to the Comission. Only verbal feedback is desired.
Timing of a recomendation to the Comission is currently under review.
IV. SAFETY RESEARCH PROGRAM (0 pen)
[ Note: Sam Duraiswamy was the Designated Federal Official for this portion ofthemeeting]
On Thursday, February 5,1987, the Comittee discussed Draft 1 of the ACRS report to the Congress on the proposed NRC. Safety Research Program and Budget for FY 1988. The Comittee proposed several changes and suggested clarifica-tion for improvement in various parts of the report. Subsequently, Draft 2 was prepared, incorporating the coments and suggestions made by the Comit-tee, and it was discussed and approved by the Comittee on Friday, February 6, 1987.
322ND ACRS MlNUTES .
During the discussion, the Comittee decided not to include the comments relating to the testing of Charcoal Adsorption Capability in the report to the Congress. Instead, it decided that coments on this issue should be transmitted to the EDO in a separate letter. Accordingly, Dr. Moeller agreed to prepare a letter to the EDO on this matter.
The Comittee also decided to leave the option open for providing coments to the Congress on the recomendations of the National Research Council's Comittee on Nuclear Safety Research included in the report entitled, "Revi- ,
talizing Nuclear Safety Research."
Dr. Carbon comented that he does not believe that the Congress would be interested in the kind of detailed comments and recommendations that the ACRS has been including in its reports. In his opinion, the ACRS should not provide detailed comments to the Congress. The ACRS should try to find out l what Congress needs to know about the NRC Safety Research Program and prepare i a report addressing only those items.
i l Dr. Siess stated that for the past several years he has been trying to i convince the Comittee that providing detailed coments to the Congress-does '
] not seem to be very useful to that body; that is one of the reasons for his
- proposal to stop writing annual reports to the Congress on the overall NRC '
I Safety Research Program. Instead we should try to provide more focused ,
{
reports devoted to important issues.
Indicating that each year the Congress has been asking the NRC to provide information on the contributions of the research results to regulatory requirements, Dr. Carbon suggested that the ACRS try to provide its opinion to the Congress on the applications of research results.
Dr. Siess stated that in one of its previous reports the ACRS tried to address this issue and found out that it was difficult to do. The fact that each research program should have an endorsement by the user offices implies that the results of the research will eventually be used by those offices..
Dr. Carbon reiterated his point that the ACRS should come up with a different fonnat for its report to the Congress and try to address only those issues in which Congress is interested. If this is done, he believes that Congress will pay more attention to the ACRS coments and recomendations.
V. QUANTITATIVE SAFETY G0ALS (0 pen)
[ NOTE: R.P. Savio was the Designated Federal Official for this portion of the meeting]
Dr. Okrent stated that the purpose of this discussion was to begin the
- process of developing ACRS comments on a Safety Goal Policy implementation plan. The NRC Staff has developed a proposal for an implementation plan
~
(Ref: Memorandum from V. Stello to the Comission, dated February 2,1987) and has discussed this proposal with the Comission (Ref: Comission meeting on January 8,1987).
322ND ACRS MINUTES - 8-
. The Commission has asked that the ACRS develop comments and plan to discuss i this subject with them in the near future.
Dr. Okrent stated that he believes that the ACRS should give consideration to i this matter and try to issue connents at the March,1987 ACRS meeting. He ;
indicated that the following topics should be considered: '
(a) The ACRS should address the adequacy of the NRC Staff's proposed approach to the implementation of the August, 1986 Safety Goal Policy. Some options on the positions ACRS could take are:
Accept the NRC Staff proposal for near term implementation and evaluation.
Reject the NRC Staff proposal and request that the NRC Staff develop an alternate proposal or that the NRC Staff work with the ACRS in developing an alternate proposal.
' Reject the NRC Staff proposal and state that the ACRS will develop an alternate plan.
(b) The ACRS should decide if the implications of the various
! value/ impact proposals are understood and, as necessary, develop more information as to what these proposals would mean in terms of the resolution of safety issues. The Committee should also evaluate the Safety Goal Policy statement as to " reasonable assurance" of no core melt and the thinking which has lead a number j of foreign countries to adopt new safety features different from those that have been developed in the US and issues which need to be considered.
! (c) The ACRS should consider if the $1000/ person-rem criterion i
adequately considers public health and off-site costs. In particular, secondary costs and loss of societal resources need to be considered.
(d) The ACRS should make reconnendations as to what is needed in the way of a containment performance objective, and management and op- ,
, erations goals. l
- (e) The ACRS should decide if the NRC Staff's proposed definition of a l large release is adequate and, if not, make recommendations as to
, what it should be.
(f) The ACRS needs to make recommendations as to what differences, if
. any, there should be in safety goals for existing and future plants.
. (g) The ACRS needs to consider further the treatment of uncertainties and the NRC Staff proposed guidelines (a factor of 3 for changes in core melt frequency and a factor of 10 for risk), and to provide coninents on these issues.
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322ND ACRS MINUTES '
l (h) The NRC Staff will be using the background for the development of i NUREG-1150 in the Safety Goal Policy implementation. The i NUREG-1150 and the Severe Accident work currently does not include j
external events. The ACRS needs to consider the implications of this fact in the use of this work in a Safety Goal Policy implementation.
It was noted that the NRC Staff is proposing that a large release be defined 1
as one that results in one or more fatalities beyond the site boundary.
Other definitions which have been suggested are more than 25 rem to a indi-
- vidual offsite (EPRI advanced reactors work) and 5 rem at the site boundary (Connissioner Asselstine).
There was lengthly discussion as to how the Severe Accident Policy implemen-tation, the proposed Safety Goal Policy implementation, and the backfit rule were related. It would appear that the first major application of the Safety Goal Policy Statement will be in the implementation of the Severe Accident 4
Policy. It is not clear as to how the Individual Plant Evaluations, as -
J proposed by IDCOR, will provide -the necessary qualitative information for j implementation of the Safety Goal Policy.
I Dr. Kerr stated that he believed that the NRC Staff should consider whether j the Severe Accident Policy and Safety Goal Policy implementations could be combined.
i Dr. Lewis noted that he believed that it was a mistake to treat a Safety Goal Policy as a regulatory hurdle for individual plants.
There was some discussion of the NRC Staff proposal to use a factor of 3 (for a change in core melt frequency) and a factor of 10 (for a change in risk) as indicators of the significance of the improvement associated with a proposed
, change. The NRC Staff stated that their intent was to use this factor as a
" rule of thumb" for deciding if the merit of the proposed changes' should be I tested under the value/ impact rule in the implementation " matrix". There was I some concern that this " rule of thumb" would eliminate worthwhile improvements. There also appeared to be concern that the " rule of thumb" would in practice end up being a threshold criterion.
Dr. Lewis questioned the basis of the NRC Staff for the selection of the factors of 3 and 10. Statistics and the results of the NUREG-1150 work are used to justify these parameters in the NRC Staff proposed implementation l plan. Dr. Lewis stated that he believed that this was another example of the l NRC Staff's lack of understanding of and frequent misuse of statistics.
9 There was some discussion on the treatment of uncertainties and the use of
) expert opinion in NUREG-1150. - A briefing on this matter was subsequently 1 l scheduled for the March ACRS meeting. The NRC Staff will provide written ;
i material before the March ACRS meeting discussing the manner in which the !
1 NUREG-1150 results will be used in the Safety Goal Policy implementation and i the treatment of uncertainties in NUREG-1150. The-NRC Staffs' planned use of i the results of NUREG-1150 in the Safety Goal Policy implementation will also
- be discussed at the March ACRS meeting.
! It was noted that a possible method for implementing the Safety Goal Policy i would be to use the insights gained from PRA and the guidance contained in i
322ND ACRS MINUTES .
l i the Safety Goal Policy to evaluate the regulations. Where the regulations '
were judged to be deficient, they would be changed and the appropriate backfits addressed for all plants.
VI. SURRYNUCLEARSTATION, UNIT 1(0 pen)
[ NOTE: E. G. Igne was the Designated Federal Official for this portion of themeeting.]
Dr. Shewmon, in his introductory remarks, provided a review of the Surry Unit 2 catastrophic rupture of a feedwater line that occurred because large areas
- of the pipe and elbow downstream from a tee were corroded / eroded to less than l half the initial thickness, with some areas being under 1/4 the original 4 thickness. He then described why it happened, when and where similar breaks j might occur, and its relevance to the leak-before-break (LBB) concept.
Discussions with experts indicate that, at the temperature and pressure of this line, single phase flow exists. Thinning has been observed in carbon
- steel lines with low oxygen water in steam generators and secondary water circuits. British, French, and German organizations have published ' papers on this subject, and there is one continuing project on this matter at MIT which ,
is supported by utilities. Single phase corrosion / erosion is caused by the t oxidation of carbon steel under conditions where the oxide that forms is
! dissolved and swept away, rather than forming a protective film. The process
! is temperature dependent and reaches a maximum at about 300*F-(the tempera-
! ture of the feedwater line at Surry). The corrosion / erosion process is 4 accelerated by high velocity flow and turbulence, and decelerated by things that stabilize the oxide, such as higher oxygen or pH, or alloying the pipe i material with small amounts of chromium. Failures of the type that occurred at Surry Unit 2 are quite uncommon in this country and heretofore designers ll have paid little attention to high fluid velocity or turbulence in. such lines, aside from their concern for energy efficiency. Thinning is readily l
- detected by ultrasonic techniques, however these lines were not normally '
1 inspected.
In Dr. Shewmon's impression, pipe wall thinning is worst (to date) at Surry !
because of the use of smaller diameter lines (which gives rise to higher bulk velocity) and a tee with a square corner design (which gives rise to higher
, turbulent local velocities) rather than a smooth blended tee or a takeoff at l 45 degrees, as has been used in other plants. Variation in water chemistry
- is a factor to be considered, but often the operator is limited in his
- options if he is to avoid problems with other materials in the feedwater
system. Low alloy steels, like 1/2 percent r.hromium, would also help elimi- l
- nate the corrosion / erosion problem. )
i l With respect to LBB, Dr. Shewmon stated that this event is not relevant ,
- because LBB is only applied to lines that are regularly inspected and have a j
- reliable record of performance. Most of the candidate lines for LBB are made i
- of stainless steel or are stainless clad, which eliminate the corrosion /-
i erosion problem. Regular inspection, in LBB targeted lines, will also guard against surprises as found in Surry.
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322ND ACRS MINUTES .
Mr. Ebersole noted his concerns about the operational implications of the Surry pipe failure. He noted that the fire suppression systens were actuated by water spray during the incident.
During his presentation, Mr. V. Panciera said that the C0 system and the Halon system were both actuated by water spray. The CO,2 system actuated above the control room and, since C0 7 is heavier than air, found its way into the centrol rocm. Some of the operating personnel felt sick and giddy from the CO-. The Halon system actuated but the Halon concentration (7-10%) did not hafe any effect on the personnel. Mr. Panciera noted the problems with the feedwater pump discharge check were a combination of erosion and corro-sion. The discs are held by two pins. Erosion and corrosion of these pins allowed the seat to separate from the body.
Mr. Panciera stated that the pipe failed catastrophically when the pipe walls thinned to the extent that global stresses in the pipe exceeded the yield strength at normal internal pipe pressures. Ductile failure in the pipe wall occurred near the weld when the stresses in the pipe exceeded its ultimate strength. An initial failure caused the pipe to leak for about 3-5 seconds before the crack became unstable causing a catastrophic failure. He stated that, upon inspection, it was observed that significant general thinning occurred in the elbow due to corrosion / erosion mechanisms.
Mr. J. Wilson, VEPC0, presented a brief overview of the utility recovery actions. Because a similar feedwater configuration existed in Surry 1, it was shut down for inspection. Inspections at Surry 1 and 2 of susceptible locations with possible corrosion / erosion thinning found some thinning but not as severe as the failed elbow in Surry 2. Some thinned sections were replaced and others that were within the code specifications were left alone.
The metallurgical and stress evaluations were reported to be nearly complete and a report on this matter will be issued by VEPC0 in March 1987.
Mr. R. Bosnak, NRR, discussed the generic implications of the Surry feedwater line failures. The following are some of the major recommendations:
Arbitrary changes in water chemistry should not be made at this time until further studies into the effects of pH and an oxidizing environment on materials have been completed. Investigations of local fluid dynamics in various pipe configurations and their effects on erosion rates should also be completed.
Available information should be presented to the utilities so that they can undertake pipe wall measurements at susceptible locations ano determine the need for periodic monitoring and corrective action.
The ASME should provide guidance to designers via the pressure piping and nuclear codes and standards on erosion, erosion /corro-sion and cavitation in single and two-phase systems.Section XI ,
and groups active in aging and life extension need to determine I minimum allowable pipe wall thickness.
Dr. B. Sheron, NRR, presented the Surry p1 failure action plan. The plan involves actions by IE (J. Rosenthal), NRR R. Bosnak), and RES (C. Serpan). '
322ND ACRS MINUTES .
IE will issue an Infonnation Notice (IN) on this event by February 9., .1987, and update IN 80-106. In addition, IE will compile plant inspection data and distribute it to NRR and RES. IE will also prepare an accident discussion paper in collaborate with OSHA and State authorities.
NRR will prioritize the issue using PRA methodology and issue an user-need letter from NRR to RES. NRR will also monitor industry activities in this area.
RES will review relevant research in this area and prepare a GDC-4 discussion paper presenting the impact of the Surry event on the broad scope leak-before-break rule.
The NRC long-range plan will incorporate the following elements:
Analysis - collect, review, and analyze results of IE, NRR, and RES actions.
Regulations - examine adequacy of existing regulations and propose rulemaking where needed.
Inspection - assess need for increased inspection or other IE actions and provide new industry guidance where needed.
Research - evaluate available knowledge consistent with safety significance and regulatory needs and initiate new research if needed.
VII. REACTOROPERATINGEXPERIENCE(0 pen)
[ NOTE: H. Alderman was the Designated Federal Official for this portion of themeeting]
Mr. Ebersole remarked that he had met with Mr. Jack Rosenthal and his staff in Bethe:da on Tuesday. About 15 events were reviewed which had been screened from a much larger list which had accumulated in the last two months. Six events were selected for presentation to the Coninittee.
Mr. Ebersole reminded the Conmittee of the mandate from Chairman Zech to pay considerable attention to reactor operating experience and he noted that we spent fairly little time at it. He remarked that we are going to have to evaluate whether we are approaching this in a proper manner. Mr. Ebersole then turned the meeting over to Mr. Rosenthal of the Office of Inspection and Enforcement.
Mr. Rosenthal listed the six events to be discussed. These were: A briefing on the Hatch augmented inspection team, updates on the drywell shell corro-sion at Oyster Creek, the Byron event, the Sequoyah/ Watts Bar ice condenser problem, and two events at Brunswick 2. The first Brunswick event concerned Diesel Generator Room cooling, and the second concerned reliability of motor operated valves.
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l 322ND ACRS MINUTES .
l Mr. Ebersole commented on the Hatch event. He noted it wasn't significant but the failure of the operators to recognize what was happening is indica-tive of what might happen in much more sensitive areas.
F. Cantrell AIT Leader at Hatch, Hatch Augmented Inspection Team Mr. Cantrell remarked that Georgia Power did a very commendable job in responding to this event.
He noted that if the spent fuel pool had drained all the way down, there would still be a foot and a half of water shielding the irradiated fuel. The water did not drain all the way down.
There were numerous precursors to this event that could have alerted personnel on how to prevent it from happening. This included the number of times that they had had to refill their spent fuel pool. This was not reccgnized as a problem.
Dr. Moeller asked it d..r: m m alann for spent fuel pool water level. Mr.
Cantrell responded "yes" and when it sounded the operator would go up and refill the pool.
Mr. Cantrell pointed out that for several weeks the spent fuel pool was being filled approximately once a shift due to evaporation and due to leakage in the cleanup system.
The initiating cause for this event was a defective air pressure regulator.
To compensate for the defective regulator, the supply valve was throttled back. An operator thought this valve looked like it was supposed to be closed. Since it was not tagged to be left open, he closed it.
The seals for the transfer canal are all supplied from a single air supply, about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the air supply was turned off the leaks started.
Mr. Cantrell noted that the leakage detection alarm system didn't work because the plant staff left the drain valve open the last time it was calibrated.
Mr. Cantrell discussed the irradiated control blades on short hangers that were in the pool . The water dropped to about two feet above the control hangers. There was a possibility of having these hangers exposed. Georgia Power calculated that the exposure rate from these on contact would be about 10,000 R/Hr. If the water level had dropped to expose these hangers, then Georgia Power would have let the blades drop into the pool to reduce the radiation level.
Mr. Ebersole pointed out the lessons to be learned from this event. He remarked it was the procedural or the administrative aspects of this accident that are significant. He noted sloppy design, sloppy review and sloppy attention to maintenance.
Update on Oyster Creek, Drywell Shell Corrosion - R. Hermann, NRR This event concerned local wastage of the drywell shell in the sand cushion region. Firebar D insulation is installed next to the shell and the
322ND ACRS MINUTES .
sand cushion is outside the Firebar D. The insulation was installed to provide an inside form for placing concrete and is compressible so that an air gap between the shell and the concrete will result. This allows thermal expansion for the shell. The sand cushion serves as a spring.
It is estimated that the sand cushion got wet fairly early in life. The corrosion is a general wastage type. There aren't any indications of pit-ting. It is believed that the corrosion occurred due to sulfates and chlorides that probably leached from the insulation.
The original plate shell thickness was 1.15 inches. The corrosion has been on the average about 3/10 of an inch per year; this reduced the shell thickness to about 0.7 inch.
There was an analysis done to see if the shell was okay and met the code requirements. The shell still met the code requirements. The only thing that is in question on Oyster Creek now is the corrosion rate. The question is whether the corrosion rate is 20 mils per year from the time it first started or is it 50 mils per year from 1980. The staff is assuming that the rate has been 50 mils per year since 1980. The licensee is planning on a mid-cycle inspection to verify what the rates were.
Mr. Hermann pointed out that one unique feature regarding Oyster Creek was the lack of a cover on the top of the sand cushion. There are drains at the bottom of the sand pocket which are fairly widely spaced. There may be some areas where water would not be adequately drained.
Dr. Moaller pointed out that if asbestos insulation was present in the area and it becomes wet, it then deteriorates and becomes readily airborne. This could represent an occupational health hazard.
Dr. Okrent commented about the combination of problems that could cause deterioration of buried piping and tanks at nuclear power plants. He noted the lack of scientific and systematic thought of what problems might arise.
He cited the potential of an earthquake affecting a weakened set of pipes needed for a vital system.
Loss of Component Cooling Water at Byron - W. LeFave, NRR (Followup)
Mr. LeFave noted that at the time of the event the component cooling water pumps (CCWP)werebeingshifted. This resulted in a pressure surge which led to the relief valve opening. The relief valve stuck open.
The flow through the relief valve drained the service tank supplying the CCWP and the CCWP tripped on a low surge tank level.
The low pressure discharge automatically started the standby pump which then drained the surge tank out the same break. The pump then tripped on the low l level of the surge tank. At this point, the leak was isolated. l Mr. LaFave pointed out that at many plants this would be a severe transient because you would end up with a loss of reactor pump seal cooling. In the case of Byron, the component cooling water system doesn't supply cooling 1
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, 322ND ACRS MINUTES 1 i
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< water to the reactor pumps. The component cooling water system only supplies F the RHR pumps and heat exchangers. The safety loads are minimal.
- In response to a question from Dr. Moeller, Mr. LaFave stated that the j Essential Service water supplies the cooling water for the charging pumps.
In summation, Mr. LaFave stated that, after a closer look, the component cooling water configuration at Byron proved not to be a problem.
l Unanalyzed Design Condition for the Ice Condenser - Sequoyah/ Watts Bar - J.
Giitter, IE l Mr. Giitter listed three potential problems that the unanalyzed design condi-j tion poses:
I 1. A potential for decreased ECCS sump inventory following a LOCA I
- 2. Potential for ice condensor bypass -
- 3. Potential for loss of one or both air return paths.
4 He noted that the unanalyzed design condition is caused by water accumulation 3 or pooling in the air return fan pit.- The fan pit is located below the operating deck.
Mr. Glitter postulated the following scenario: !
j Following a LOCA, the steam will displace the air in the lower compart-i ment. The steam-air mixtures will rush up, force the lower inlet doors
{ of the ice condenser open where the steam will be mostly condensed as it ,
l goes up through the ice condenser and the high containment pressure will '
i cause the containment spray to actuate af ter about 10 seconds. After about a 10 minute delay, the air return fans will start.
l
- Water from the containment sprays will drain down on the operating deck.
- This water should be channelled to the refueling canal where it would go i through the drain line to the ECCS sump where it is available for
} recirculation.
4
{ The design problem at Sequoyah and Watts Bar is that the curbing which j caused the water to flow to the refueling canal had been removed. The
! curbing was removed for operational reasons since it was an obstruction 4 without realizing the significance of its removal. Removal of the curbing allows the spray water to flow into the air return fan pit. The i water level in the pit could rise up to the operating deck, cover the
- inlet and cause the fan to fail.
i Mr. Giitter discussed the significance of the air return fans. He noted that
- only one of these was below the operating deck. He remarked that the air i return fans limit the potential for hydrogen accumulation in the upper dome 1 area. They also further scrub fission products and reduce the steam. .With the air return fans running, the containment is cooled mort quickly and this allows the containment to depressurize more rapidly.
1
- In response to a question from Dr. Okrent, Mr. Glitter replied that there
{ were two air return fans at Sequoyah and they were located on each side of a
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322ND ACRS MINUTES .
the ice condenser. One was above the operating deck and the other is re-cessed below the operating deck.
Mr. Ebersole pointed out that without the air return fans, the containment pressure would not decrease.
l Mr. Glitter agreed. He noted that the containment spray doesn't have too much effect on reducing containment pressure. It is only after the air return fans go on that the containment pressure is reduced by about 2 psig i very quickly. He also remarked that the air return fans have double the i
required capacity. Loss of both fans would result in approaching the design basis pressure for containment.
3 In terms of risk, Mr. Glitter stated, that the real importance of the air return fans is to prevent hydrogen accumulation in the lower compartment of the ice condensor containment.
Mr. Ward asked if the different pressure transient would affect. the ECCS performance?
I I
Mr. Glitter replied that the static head above the seals around the air return fans and the divided X-seal will cause a differential pressure that
! those seals aren't designed for. As a result, TVA has postulated that these seals will fail and water on the operating deck will drain through them and go into the lower accumulator room.
Until the level rises to about 12 feet, the water is being diverted from the refueling path down into the ECCS sump.
i The result of this is the ECCS sump level will be below that which would
- result in automatic transfer from the injection mode of ECCS to the I
recirculation mode of ECCS. The operators would have to do a manual trans-i fer. But, based upon TVA's calculations, it is above the level at which you will get vortexing in the ECCS sump and subsequent cavitation of the LPCI pumps.
Mr. Glitter remarked that TVA is considering installation of curbing on the operating deck and they are also considering installing drain lines in the air return fan pit.
Mr. Ward asked to the extent this reactor has Upper Head Injection (UPI) is
- there any risk associated with this event?
! Mr. Rosenthal replied that the UHI system is important earlier in the event than the time scales that we are talking about.
Brunswick, Emergency Diesel Generator Room Cooling Design Deficiency -
Ernie Sylvester, NRR
! Mr. Sylvester discussed a design deficiency that was discovered during the i course of a PRA review of all plant systems at Brunswick. It was found that all the HVAC dampers that supply air to the four diesel generator rooms failed closed on loss of the non-safety-related instrument air. As a result, i
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322ND ACRS MINUTES 1 1
l all four diesel generator rooms, which are shared by the two units, would lose cooling air.
With the loss of cooling air, there is a potential that all four diesel generators would stop running due to overheating of system controls.
Dr. Carbon asked how long it would take for the diesel generators to stop.
Mr. Sylvester responded that the licensee was doing an analysis on how long it would take, but it would occur.
Mr. Wylie asked if it was a comon instrument air line. Mr. Sylvester replied that it didn't matter if it was comon or not, all six compressors would stop on loss of offsite power.
Mr. Ebersole pointed out that we better go back and look at these comon features of the diesel generator rooms. He remarked that he thought we are going to find lots of additional problems.
Dr. Kerr asked how long the diesel generator would run?
Mr. Steven Floyd, Carolina Pcwer and Light, responded that their preliminary analysis indicates a 24 degree temperature rise in about 30 to 60 minutes.
He noted that they didn't believe that the diesel would stop running. He emphasized there would be adequate time to take corrective action during this time. The corrective action would be to open the door to the diesel generator rooms.
There was a brief discussion regarding air receivers on the instrument air system. Mr. Rosenthal remarked that he thought it would only be a matter of minutes before the receiver would go down.
Dr. Okrent comented on the vast efforts exper.eed on systems interaction and the conclusion that he felt the staff perceived that systems interaction doesn't seem to represent great risk and that the amount of money you would have to spend to look for them would be vastly greater than what you might reduce in risk.
Dr. Moeller pointed out that he was impressed by the number of system inter-actions involving air supply systems, instrument air systems, chilled water systems, air cleaning, air monitoring and air ventilation systems. He suggested asking AE00 to consider, on a systems interaction basis, analyses of the systems he had mentioned.
Dr. Carbon asked about the study by the Brunswick people that picked up this systems interaction problem. He asked: Why did they do it? How extensive was it and whether it was done on their own initiative?
Mr. Sylvester replied that it was done on the licensee's initiative. They decided to do a PnA review of all plant safety systems. Mr. Floyd, Carolina Power, added that ie study cost about 2 million dollars.
1 322ND ACRS MINUTES .
r
! Brunswick 2, Scram with Complications, - E. Sylvester, NRR l t
l.. Mr. Sylvester remarked that after a scram on January 5th, there was a loss of high pressure coolant injection capability due to a stuck open water in- 1 jection valve and a- partial loss of the RCIC injection capability due to a stuck open bypass valve. The case of both events was equipment failure. In ;
i the case of the HPSI system, it was motor burnout. In the case of RCIC, it '
was a failure of the valve internals. ;
As a result of these two post-scram failures, there was a potential for loss 4
of all high pressure coolant injection. The automatic depressurization ,
e system, the core spray system, and the low pressure coolant injection mode of l PHR were still available but as it turned out were not needed.
i
- Mr. Ebersole observed that at this point could you say, however, that the use of these systems would have been rather objectionable in that you would suffer a pressure or temperature transient?
Mr. Sylvester responded that he thought that this was fair to say, and they
- didn't rush in to use them.
i Mr. Sylvester noted that the initial event was a turbine trip that resulted j in a reactor scram from 100 percent power.
- i After the scram, the following events occurred:
i i
o Low reactor water level signal. It went down to low-low level momentarily. l o Signal long enough to open steam admission valves to the HPCI r turbine, and the RCIC turbines, causing the turbines to roll.
o The low-low level signal didn't lock in long enough for the in-jection valves to open on either system. So the HPCI and RCIC j systems operated in a mini-flow recirculation back to the j suppression pool as the level came back up. l o The reactor was being fed, at this time, by main feed pumps on '
a their pump coast-down curve.
- o As the pressure peaked at 1119 psig, five SRV's opened automatical- ;
i ly.
o Between the swell from the SRV's popping and the main feed continu- :
ing, a high level signal was received, which secured the HPCI and i RCIC systems. By shutting down the turbines, the steam emission
} valves closed on both turbines and the entire system shut down. j
$ Mr. Sylvester discussed the operations when the reactor level alternately decreased and increased. The bottom line was that the HPCI was out of
- connission and only partial RCIC flow was getting into the vessel due to one l
[
of the RCIC valves being half open. The operators manually switched valves to permit full RCIC flow into the vessel, i Mr. Sylvester discussed the failure mode of the RCIC valve. It is a rising
! stem globe valve. A motor turns a threaded collar which raises the stem as
! long as the stem is not allowed to rotate. In this case, the anti-rotation j device failed due to the set screw not going into the stem.
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322ND ACRS MINUTES .
The control room indication is based upon :novement of the stem by counting turns of the stem. In this event, the stem turned but the globe didn't move.
Mr. Ebersale pointed out the weakness of indirect measurement of whatever parameter you are interested in.
Mr. Ebersole noted there were two subcomittees that could pick up Dr.
Noeller's concerns with air problems and Dr. Okrent's concerns regarding buried piping and tanks.
VIII. EXECUTIVE SESSIONS A. Reports, Letters, and Memoranda
- 1. Report to Congress on the NRC Reactor Safety Research Program The Comittee completed its report to the Congress on the proposed NRC Safety Research Program and Budget for FY 1988.
- 2. Testing of Charcoal Adsorption Capacity
, The Comittee prepared a memorandum to the EDO commenting on the research program regarding the testing of charcoal adsorption capacity.
- 3. Naval Reactors Moored Training Ship Demonstration Project l The Comittee prepared a classified report to the NRC on the Naval Reactors Moored Training Ship Demonstration Project.
- 4. Analytical Methodology Used in Evaluation of Risk by Naval Reactors The Comittee prepared a classified report to the Director, Naval Nuclear Propulsion Program, regarding analytical method-ology used in the evaluation of risk by DNR.
B. Subcomittee Reports
- 1. WasteManagement(0 pen)
[ NOTE: Owen Merrill was the Designated Federal Official for thisportionofthemeeting]
Dr. Moeller reported to the full Comittee on nuclear waste topics discussed with the NRC Staff during an agenda planning session on January 21, 1987, which was held in preparation for the Waste Management Subcomittee meeting to be held on February 19-20, 1987 He indicated that Dr. Shewmon, and Dr.
Parry, ACRS Fellow, were also in attendance.
I With regard to the NRC Staff's high-level waste (HLW) activ-ities, Dr. Moeller discussed the following topics:
4 (1) Federally Funded Rosearch and Development Center (FFRDC),
which is to provide research support to the NRC Staff in
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this area. Proposals from various ccmpetitors for the FFRDC are due by February 10, 1987.
(2) The validation of models being developed (on a coopera-tive international basis) for assessing the performance of a geologic repository.
(3) Also on a cooperative international basis, studies of a range of media for potential high-level waste repos-itories. The Federal Republic of Gemany is working on salt, Italy is working on clay, and France, Switzerland, Sweden, the United Kingdom, and he believes, Japan, and Canada in response to a question by Dr. Mark about Canada) (are working on granite. He said that because of the U.S. interest in salt, the NRC Staff would like to get more of the other countries working in this area.
(Note: The U.S. is working on salt, tuff and basalt media).
(4) The definition of HLW, which the Congress included-in the Nuclear Waste Policy Act of 1982 (NWPA) as something that is highly radioactive and requires permanent isolation.
NRC's efforts are to come up with an official definition that is consistent with, but more definitive than, the Congressional definition. Dr. Siess asked if Congress' defining it made it definitive. Dr. Moeller stated that, because of the congressional definition, the NRC Staff is very concerned that certain groups may separate their wastes into categories, that only require permanent isolation or are not highly radioactive and therefore need not to be handled as HLW. He further stated that spent fuel is clearly defined as HLW and the initial product from a chemical reprocessing system is also so defined. Mr. Ebersole asked about used control rods, to which Dr. Moeller responded that they are not high-level, but low-level waste, as are also resins from the cooling water purificaticn system. To which, Dr. Parry added, "Or class C, depending on what system."
(5) The NRC Staff's overall assessment of a HLW repository, which they are trying to look at as a system, as well as component by component. They have many questions on which they want advice, such as:
(a) How do you evaluate a component? One at a time?
(b) How much of the total goal or how much of meeting the EPA Standard can be assigned to a single compo-nont?
They)havealsodecidedthat,iftneDepartmentofEnergy (00E , the applicant, identifies some components as l adequate, but says certain other components do not need I to be looked at, the NRC Staff will require the applicant i
[ i j ,,' 322ND ACRS MINUTES '
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l to look at all components to be sure they do not repre-sent a detriment or a negative ~ factor in terms of the other components that DOE.is looking at.
f (6) The NRC Staff's research on waste package corrosion,
! including the effects of stress corrosion cracking and '
t the potential effects on corrosion by biological i activity.
l In response to Dr. Kerr's question, "What biological l activities?" Dr. Moeller explained that the NRC Staff is I wondering if certain types of biological activity could i cause an acceleration or change in corrosion rates. As
- another illustration. he cited the situation at Hanford j . where they observed that certain biological activity j within the soil produces chelating agents, which totally 1 change the behavior of certain radionuclides. In the case of plutonium, for example, plants which for years
) did not take up plutonium were suddenly found to be j taking it up. Further discussion of this topic ensued
- between Drs. Moeller, Kerr, Mark and Parry.
j (7) Scaling factors, i.e., how do you scale from short-term (1-2 years) experimental tests to long-term (1000 to
~
1 j 10,000 years)systemsoperations.
l (8) The NRC Staff is looking to the Waste Management Subcom-mittee (which should really be to the full ACRS) to help t them in:
(a) Assigning priorities to the key issues they face.
l (b) Deciding whether those key issues shall be handled j by their writing a generic technical position or by j rulemaking.
l (c) Providing guidance on their total HLW program, both i research and operations,' including the license of j DOE to build and operate the repository, ,
i i With regard to the NRC Staff's low-level waste (LLW) activ-
] ities Dr. Moeller named and/or discussed the following topics: >
\
(1) LLW long-range plan through 1993, on which the NRC Staff
! asked for comment and feedback.
. (2) Providing guidance to the states and compacts (grcups of
! states) for their implementation of the Low-Level Radio-
- active Waste Policy Amendments Act of 1985 (LLRWPAA), ;
i.e., the preparation and submission of applications to
- the states for the construction and operation of LLW 4
disposal sites.
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322ND ACRS MINUTES .
The NRC Staff has developed and will soon publish two guidance documents which parallel similar documents currently used in licensing nuclear power plants. They are: (a) Standard Review Plan (NUREG-1200) and (b) Standard Format and Content (NUREG-1199). The NRC Staff wants ACRS coments on the initial issues of these documents and on subsequent annual revisions as they occur.
Dr. Remick asked for clarification on the licensing of LLW sites in agreement states. In the ensuing discussion among him, Dr. Moeller and Dr. Parry, it was brought out that the agreement states will license the LLW sites in their respec-tive states, with the distinction that the party that applies for the license will not be the party to issue the license.
Although the guidance documents are provided by the NRC to give the level of detail the applications should have, and to ensure a uniform approach to the licensing process from one state (or compact) to another, the agreement states may or may not use them, as they choose.
Dr. Mark asked if the guidance documents are written as if one were thinking only of shallow land burial, to which Dr.
Moeller responded that they are looking not only at shallow land burial but at all of the viable alternatives to it. Dr.
Parry commented that the NRC Staff feels that low-level burial is fully acceptable and a perfectly appropriate way to go.
Dr. Moeller stated that consideration is being given to having an NRC resident inspector at various disposal sites (although it was not clear whether it was meant at all LLW disposal sites or just at NRC-regulated sites). A discussion followed among Dr. Moeller, Dr. Kerr, Dr. Parry and Dr. Siess which clarified the issue that, although there are currently NRC representatives at two proposed, as yet unlicensed sites (the Hanford BWIP and Nevada sites), both of these sites are HLW sites whereas the subject being discussed was LLW sites. Dr.
Remick asked, "How about WIPP7" Is there one (i.e., an NRC site representative) there?" Dr. Parry said that "NRC has no position there," and Dr. Moeller added, "WIPP is not to be licensed." (Note: WIPP is DOE's Transuranic and Low-Level Waste Isolation Pilot Plant (for defense wastes) near Carlsbad, NM; hence, it does not come under NRC regulation).
Dr. Moeller gave a brief report on the DOE /NRC Licensing Support System (LSS), which was also a topic of discussion at the January 21, 1987 agenda planning session. He stated that the NRC, with DOE 's cooperation and funding, and with the advice of an advisory committee, is setting up the LSS. l (Note: Such an advisory committee is yet to be formed; notice 1 of intent to form "An Advisory Committee for Negotiated i I
Rulemaking(,"
18, 1986 51 FR was published 45338). The in the Federal Committee Register would on December be composed of organizations representing the major interests affected by the ;
rule, viz., DOE, NRC, States and Indian Tribes, in particular. 1
1 j ,. 322ND ACRS MINUTES !
4
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) Public comment on this issue was to have been provided by .
Dr. Moeller said that the LSS will be a February 17, 1987.
- computerized data base of technical infonnation that will,
! according to the cited Federal Register notice, be used for
! the management of records and documents for the HLW licensing procedure.
l Dr. Moeller than called upon Dr. Parry to coment on the LSS.
! Dr. Parry made a few preliminary remarks about DOE's recent
! issued Draft Mission Plan Amendment (dated January 1987) which
! proposed a 5-year delay in their program, which, he said,
! would extend the period of time that NRC has to consider the -
i matter. He said that the basic value of the LSS, which is j essentially a data retrieval and document retention system, is that it will provide an essentially instantaneous capability
< for participants in the hearing process to do their own
- discovery process, thus eliminating the necessity for manual i retrieval of documents and the copying of them, which is required because of the legalities of providing facsimiles for
- the hearing system. The system being developed by the Staff will have the capability of holding all (some 16 million) the
- expected documents in their entirety, with addenda, notes, i
appendices, figures, tables, etc., so that the participants can search on their own and thus relieve the Staff of this
- burden. Dr. Moeller added that an interested party will be able to call up any individual word that is in any document.
- Dr. Remick asked if intervenors might claim that they don't j have the money to buy the terminals and printers to access '
> that information, and they are therefore being taken unfair ;
j advantage of. Dr. Parry replied that, once the requirements !
i '
for the system are set, it is hoped that DOE will pay for the l entire system and will provide the terminals, etc., for the authorized participants, who in turn, must put all their ,
documents into the system in a prescribed fonnat.
Dr. Moeller briefly discussed what was to have been the main topic for discussion with the NRC, Staff at this meeting, but the Staff was unable to come, i.e., an advisory comittee for the NRC high-level waste program. The question for discussion
, was how best for the NRC Staff to obtain independent advice on j the HLW activities comparable to what the ACRS provides for
{
nuclear power plant problems or questions. He said that an
- effort will be made to place this matter on the agenda for the i
March ACRS. The NRC Staff wants feedback from tne full !
l comittee on this matter, in writing. :
1
! Dr. Moeller mentioned that, in response to the minutes of the i
! Planning Subcomittee and to Mr. Ebersole's and others' '
! requests, the Waste Management Subcomittee has work underway
{ on approaches for evaluating the risk associated with the i high-level repository as compared to a nuclear power plant.
! He provided all members present' with a copy of a draft letter i that he and Dr. Parry had prepared, and requested that the members provide him with coments in it, in writing no later i l
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i 322ND ACRS MINUTES .
than the 323rd ACRS meeting, March 5-7, 1987. Dr. Siess commented that he had read it and said that what comes out of it, although it's not explicit, is the distinction (not in terms of risk) between health effects and perceptions.
Mr. Ebersole asked Dr. Moeller a question about where he, Dr.
of having a Monitored Retriev-Moeller, sees (MRS) able Storage the potentiality facility (in view of the DOE's proposed 5-year postponement of the schedule for the first HLW geologic repository, which indicates in a rather positive way that an integrated MRS facility, if approved, would be built in Tennessee to receive a maximum of 15,000 MTU spent fuel in the interim). He stated his concern about the attitudes of concerned citizens in the state, some adamantly con, some enthusiastically pro. Realizing the lack of real. hazard that comes about from stored fuel, he said, "I even wonder if the state has the prerogative to influence the installation of such a facility in view of its low risk." Dr. Moeller an-swered that he believed the congressional law for an MRS does not give the state that right, to which Dr. Parry added that.
he does not believe the states have any legislative mandate to raise an objection with Congress about the location of an MRS as they do on the location of a geologic repository. Dr.
Parry further stated that the extension of the schedule for the repository almost absolutely mandates the development of an MRS somewhere.
Dr. Remick asked if DOE's MRS proposal had gone to Congress or if it was stalemated. Dr. Parry answered his question by answering that it was a very complex situation where Tennessee was trying to file an injunction with the courts on the same day that DOE was trying to submit the proposal to Congress, but he didn't know who got there first 50 the situation is not known. He added, on a more serious note, that the MRS capaci-ty will probably eventually have to be expanded because even now the time extension that is in the mission plan amendment is not adequate for a detailed site characterization, specif-ically at the BWIP (Hanford) site -- it is only two and a half years.
Dr. Moeller concluded his report by emphasizing that "we would very much appreciate coments on this initial first rough cut, comparing risk of the repository and of the nuclear power plant, because we need your guidance."
- 2. Severe Accidents (0 pen)
[ NOTE: Gary Quittschreiber was the Designated Federal Offi-cialforthisportionofthemeeting]
Dr. Kerr reported on the December 19, 1986 meeting of the Severe Accident Subcommittee, which met to review an NRC generic letter on how the NRC would implement the Severe l
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322ND ACRS MINUTES .
Accident Policy Statement and its investigation of outliers.
He suggested that the Members review the copies of the Consul-tants reports resulting from the December 19th meeting. The Implementation Plan was not complete at the time of the December 19th meeting and could not be judged as to its value.
The next meeting of the Subcomittee wn!1 occur about April, 1987. It was not obvious to the Subcomittc? that the Severe Accident Policy Implementation Plan was in agreement with the Safety Goal Policy.
Dr. Kerr noted that there were differences in tte Individual Plant Evaluation Model viewpoints of the NRC Staff and the IDCOR group. The NRC Staff wants to be more quaititative in its evaluation, while IDCOR wants to look at systens and their weaknesses. There is no general agreement on the approach to be used.
- 3. Regulatory Policies and Practices (0 pen)
[ NOTE: Gary Quittschreiber was the Designated Federal Offi-cial for this portion of the meeting]
Or. Lewis reported on the January 14, 1987 meeting of the Regulatory Policies and Practices Subcomittee which met with representatives of !NPO and NUMARC to obtain industry input into it's review of the problem associated with the NRC's current regulatory process of building and operating nuclear power plants. The Subcomittee members tentatively agreed that a sensible path to follow for the continuation of this current review is to meet with present and past managers of ,
the NRC to exanine the NRC's development of the structure of the regulatory process. One question to be examined is - does the NRC regulate to a threshold or to make plants as safe as the market will bear.
Dr. Lewis discussed the plan of action the Subcomittee intends to take, by talking to past and present senior NRC representatives at its next meeting. lie had talked to Comis-stoner Zech about the current Subcomittee's plans for per-fonning this review and Comissioner Zech agreed to this course of action. Dr. Lewis suggested that the Subcomittee would meet in the next month or two to continue its review of the matter.
- 4. Plannina
[ NOTE: R.F. Fraley was the Designated Federal Official for thisportionofthemeeting]
g 322ND ACRS MINUTES .
There was a brief discussion of how the ACRS will acccmodate in FY 1988 a reduction in its manning level of three FTE's (full time equivalents).
l C. Other Comittee Conclusions (0 pen)
- 1. General Electric Advanced BWR (NOTE: Richard K. Major was the Designated Federal Official forthisportionufthemeeting]
Dr. Okrent began a discussion of how the ACRS was to proceed with the review of the General Electric Advanced Boiling Water Reactor. He noted that one question before the Comittee is the degree of detail that must be provided prior to granting a final design approval. It was pointed out that the ACRS Subcomittee on the Standardization of Nuclear Facilities was exploring this question. Questions concerning the degree of documentation and the amount of detail necessary to produce a l standard plant desig n is a primary concern in the ACRS review l uf the EPRI ALWR Utility Requirements Document. A safety l evaluation report on the EPRI effort is expected in April, l 19d7.
It was agreed that the General Electric Reactor Plants Subcom-mittee and the Subccmittee on the Standardization of Nuclear '
Facilities should hold joint meetings to focus the interrela-tionship between the EPRI-ALWR, GE ABWR, and Standardization Policy Statement reviews.
Dr. Okrent inforned the Comittee that there were significant open issues, both technical and procedural, pending in the ABWR final design approval review. It is unclear how issues such as severe accidents, safety goals, and new instrumenta-tion will be handled. It is also unclear how backfit questions and the definition of review areas will correlate to the licensing basis agreement sought by General Electric.
Dr. Okrent also said it was ambiguous how ACRS comments on Improved LWRs will be factored into the current reviews.
Three hours were requested to discuss the GE ABWR during the March 1987 meeting. As preparation for this session, an ACRS follow was asked to compare the Comittee's report of January Iti,1987, "ACRS Recomendations on Improved Safety For Future 1 Light Water Reactor Plant Design" to the EPRI, "ALWR Utility Requirements Documents," and to information available on the GE ABWR. The Cemittee wanted to learn about dif farences '.
between their recomendations and plans by EPRI and General Elvetric highlighted. Discussions in March will conter on areas that are not addressed by EPRI and GE.
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o 322ND ACRS MINUTES j 0
- 2. Standardized Nuclear Plants The Comittee agreed to defer for an indefinite period the preparation of a report on the basis for standard nuclear plant improvements proposed by Mr. Ebersole. Members proposed that this topic be deferred until a standard plant is proposed for licensing.
- 3. Electrical Surge Protection for Nuclear Power Plants Mr. Wylie drafted a letter during the 321st ACRS meeting (January 8-10, 1987) expressing his concern regarding pro-tection from electrical surges at nuclear power plants. The l Comittee decided not to review this letter at this time and
' assigned this issue to the Reliability Assurance Subcomittee (chaired by Mr. Wylie) for further discussion.
1 4 Containment Performance The Comittee agreed to defer until the March meeting the discussion of proposed priorities for the resolution of Generic !ssue 61. SRV Discharge 1.ine 8reak in the Airspace of MK! and Mk !! Suppression Pools and consideration of other pool bypassing mechanisms.
D. Future Activities
- 1. Future Agenda The Comittee agrood on tentative agenda items for the 323rd ACR$ Heeting, March 5-7,1987(seeAppendix!!).
- 2. Future Subcommittee Activities A schedule of future subcocnittee activities was distributed tomembers(seeAppendix!!!).
i
! The 322nd ACRS recting was adjourned at 11:00 a.m., Saturday, February 7, 1987.
1
s O
I APPEflDICES TO MilluTES OF 322fl0 ACRS MEET!f1G FEBRUAR/ 5-7, 1987 O aus-ane9 O
APPENDICES O TABLE OF CONTENTS 322ND ACRS MEETING FEBRUARY b-7, 1987 Appendix I List of Attendees Appendix !! Tentative Agenda
' Appendix III Future Subcomittee Activities Appendix IV Other Documents Received O
O
,0)
\
~~' 325 326 327 328 329 330 331 332 323 324 321 ACRS MEETING DATE: [
~ ~
,(977 AT TENDEES_
V Dr. William Kerr, Chairman
, V Dr. Forrest J. Remick, Vice Chairman M
Dr. Max W. Carbon Mr. Jesse C. Ebersole
[
Dr. Harold W. Lewis Dr. Carson Mark .
%f
Mr. Carlyle Michelson 7 -
' Dr. Dade W. Moeller Dr. David Okrent Mr. Glenn A. Reed Dr. Paul G. Shewmon P M.A ,Q ( 7 Dr. Chester P. Siess d
V Mr. David A. Ward i
Mr. Charles J. Wylie _
% d gf i
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an.#
l
J I^ ATTENDEES 322ND ACRS MEETING 1
1 Thursday February 5,1987_
l John Trotter, NUS Corp.
W. Bickford, Battelle N. W. Brown, General Electric j- B. Genett, General Electric j A. J. Neylan, GA R. T. Lancet, Rockwell i
C. Guild, Doub & Muntzing
)
J. Bruning, Rockwell Enternational i M. Beaumont Westinghouse 4
j P. F. Riehm, KMC D. Mears, GCRA P. Peir, DNL W. Buckford, PNL
' H. L. Miller, Va Power
! J. Wilson, VA Power J. Mc Avoy,VA. Power 1 E. Fotoponlos Serch Licensing, Bechtel H. M. Fonrecilla, VA power C. A. Guild, Doub & Muntzing j J. D. Hegner. VA Power
' R. Huston, AIF i W. A. Cross , Self i L. Hinkle, W. Labs I r 1
Reps, from DOE i l i
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i PUBLIC ATTENDEES S
322ND ACRS MEETING I Thursday, February 5,1987 R, Hernan NRR/PPAS ,
D. Scaletti, NRR/PBSS J. R. Hall, NRR/DPLB H.,Berkow,NRR/DBSS A. Lynch, Jr., NRC/PBSS ,
I J. E. Rosenthal, IE D. Terao, NRR ,
H. Reponan, NRR B. Shwn, NRR R. Bosnak, NRR
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' ' ATTENDEES 322ND ACRS MEETING Friday, February 6, 1987 M.W. Ebert, NUS Corp.
S. D. Floyd, Carolina Power & Light R.Borsum, BUW W. Buckford, PNL L.Connor, DSA P.F. Riehm, KMC L. Peeters, SAIC M. Specter, Wash Post D. Weaver, McGraw Hill G. Brown, Stone & Webster
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.s NRC ATTENDEES 322ND ACRS MEETING
' Friday, February 6, 1987 Eric Weiss, IE J.E.Rosenthal, IE R.A. Hernar.n, NRR i
l G.W. Rivenbark, NRR R. W. Hernann NRR J.G. Gitter, IE E. Sylvester, NRR
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s APPENDIX A O FUTURE AGENDA MARCH ACRS MEETING
- Quantitative Safety Goals - Discuss proposed NRC plan for implementation of NRC quantitative safety goals.
i Safety Responsibility at Nuclear Facilities - Discuss pro-posed ACRS report to NRC regarding the responsibility for safety at nuclear power plants.
- Fitness for Duty - Briefing regarding application of NRC rule on fitness for duty of nuclear power plant operations personnel.
- Nuclear Power Plant License Renewal - Briefing regarding proposed NRC policy regarding extension of nuclear power plant licenses.
- Radwaste Management and Disposal - Briefing regarding pro-posed ACRS participation in the NRC regulation of radio-active waste management and disposal. ;
- Risk from Radwaste Management - Discuss proposed ACRS comments f
( regarding comparison of the risks associated with radwaste
% management and disposal compared to other nuclear related risks.
- Safety Features in Foreion Nuclear Power Plants - Discuss safety features in foreign nuclear power plants which are different frem those in facilities licensed by NRC. j Pressure Suppression Containment - Discuss proposed NRC resolution of steam relief valve discharge line failures I in the airspace of Mark I and Mark II type containments
< and other methods of suppression pool by-passing.
GE Advanced Boiling Water Reactor - Discuss proposed licens-ing basis agreement for the review of the standardized advanced boiling water reactor being proposed by the General Electric Company.
Improved Light-Water Reactors - Discuss features proposed in the EPRI requirements for improved standardized LWRs compared to the ACRS report dated January 15, 1987 on Improved Safety for Future Light Water Reactor Design.
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- Future Activities -- Discuss anticipated ACRS subcommit- l tee activities and items proposed for consideration by the full Comittee.
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- New ACRS Members -- Discuss qualifications of candidates l
(' proposed for consideration as nominees for appointment to !
l the Comittee (tentative).
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) 322'O ACRS '4EETING 2 LATER ACRS MEETINGS April BWR Pipe Crack Guidance -- Coments on incorporation of public coments into NUREG-0313, Rev. 2, regarding monitoring repair of pipe cracking in BWR systems. NRC Staff report expected in early March with Subcomittee meeting in April.
Safety of P&W Reactors -- ACRS coments requested regard-ing NRC Staff review of BWOG evaluation of the long-term safety of B&W reactor plants. NRC Staff report oa BWOG evaluation expected by mid-February 1987. Subcommittee meeting needs to be scheduled.
Meetina with NRC Comissioners -- Improved Safe 1iy for Future Light-Water Reactors - The Comittee's report dated 09/15/87 was suggested by the ACRS as an item for discussion during the March ACRS-COMM. meeting.
The Comissioners have agreed but have requested that this session De deferred until the April (324th)
ACRS Feeting.
April /May
- Radiation Damaae -- ACRS coments recuested recarding Regulatory Guide 1.99, Rev. 2, Radiation Damage.
Revised Guide expected by early April 1987. Subcom-mittee meeting date to be selected.
Station Blackout -- ACRS coments requested regarding proposed NRC rule on station blackout. Proposed rule expected during April 1987. Subcomittee meeting needs to be scheduled (High priority).
! NRC Severe Accident Policy -- ACRS coments requested regarding proposed NRC Staff plan to implement the NPC Severe Accident Policy Statement. Proposed NRC Staff plan expected by mid-April 1987. Subcommittee meeting date to be selected (defer to April /May meeting).
TVA Management Problems -- Discuss proposed resolution of
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TVA management problems. Revisions to the TVA Nuclear Performance Plan and the associated NRC Staff evaluation
- are expected in February. A subcommittee meeting is not planned.
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3 322ND ACRS MEETING June / July General Desian Criterion 4 Environmental and Missile Damage -- ACRS Comments recuested regarding proposed changes in GDC-4 regarding design of pipe whip re-straints in nuclear plants. Revised rule expected in about 3 months. Subcommittee meeting date to be selected (High priority). ,
August NRC Policy Statement on Technical Specifications --
Briefing / discussion of proposed changes in NRC Policy Statement on Technical Specifications per ACRS report dated July 15, 1986.
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l ACRS SUBCOMMITTEE MEETINGS
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Human Factors, February 18, 1987, 1717 H Street, NW, Washington, DC (Alderman), 8:30 A.M., Room 1046. The Subcommittee will review " Safety Conscience" concept at utilities. Attendance by the following is antic-ipated, and reservations have been made at the hotels indicated for the night of February 17:
Dr. Remick NONE Mr. Ward NONE Mr. Ebersole DAYS INN Mr. Wylie DAYS INN Mr. Michelson DAYS INN Mr. Kruesi ANTHONY Waste Panagement, February 19 and 20, 1987, 1717 H Street, NW, Washington, D; (Merrill), 8:30 A.M., Room 1046. The Subcommittee will review the T611owing nuclear waste management topics: (1) Rulemaking for the definition of high level wastes (HLW), (2) Implementation of the HLW Five Year Plan, (3) HLW Geologic repository performance allocation, (4)
Assessing compliance with the EPA standard for a HLW Geologic Repository, (5) Hydroloav of geologic repositories (domestic and international programs),f6)NRC'swastepackagecorrosionresearchprogram,(7) Guidance documents for low level waste (LLW) Shallow Land Burial (Standard Review i Plan, and Standard Format and Content Guide), and (8) Long range plans for LLW Program through 1993. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the nights of February 18 and 19:
Dr. Moeller LOMBARDY Dr. Dillon LOMBARDY Dr. Carbon ANTHONY Dr. Kastenberg LOMBARDY Dr. Mark LOMBARDY Dr. Krauskopf COSMOS CLUB Dr. Remick NONE Dr. Steindler NONE Dr. Shewmon NONE Dr. Parker LOMBARDY 323rd ACRS Meeting, March 5-7, 1987, Washington, DC, Room 1046.
Regional and I&E Programs, March 12, 1987, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M.- 1:00 P.M., Room 1046. The Subcommittee will c6ntinue its review of the activities of the Office of Inspection and Enforcement. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of March 11:
Dr. Remick NONE Mr. Reed NONE Mr. Michelson DAYS INN Mr. Wylie DAYS INN Dr. Moeller LOMBARDY G
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j Metal Components, April 2, 1987, 1717 H Street, NW, Washington, DC (Igne),
Room 1046, 8:30 A.M. The Subcomittee will discuss: (1) Beaver Valley, l Unit 2 Whipjet Program, first application of GDC 4 broad scope rule, (2)
NUREG-0313, Revision 2 with public coments, (3) other related matters,
' e.g., presentation on double-ended-guillotine-break by Savannah River Lab.
Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of April 1:
Dr. Bush NONE Dr. Shewmon NONE Dr. Kassner NONE Mr. Michelson DAYS INN NONE Mr. Rodabaugh NONE Mr. Ward Mr. Bender NONE Babcock & Wilcox Reactor Plants, April 8, 1987, 1717 H Street, NW, Washington, DC (Major), Room 1046, 8:30 A.M. The Subcommittee will continue its review of the long-term safety review of B&W reactors.
This effort was begun during the sumer of 1986; initial Comittee coments offered on July 16, 1986 in a letter to V. Stello, EDO. Attendance by the following is anticipated, and reservations have been made at ther hotels '
indicated for the night of April 7:
Mr. Michelson DAYS INN Mr. Wylie DAYS INN ANTHONY Mr. Ebersole DAYS INN Dr. Okrent j NONE Dr. Kerr LOMBARDY Mr. Reed l Mr. Ward NONE Dr. Lewis HYATT l
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324th ACRS Meeting, April 9-11, 1987, Washington, DC, Room 1046.
Regulatory Policies and Practices, May 26, 1987, 1717 H Street, NW, j Washington, DC (0uittschreiber), Room 1046, 8:30 A.M. The Subcommittee will continue its current review of the nuclear plant regulatory process, and will review the NRR policy for nuclear plant license renewal. Lodging will be announced later. Attendance by the following is anticipated:
Dr. Lewis Dr. Siess Dr. Kerr Mr. Ward Dr. Remick Mr. Wylie Advanced Reactor Designs, Date to be determined (April), Washington, DC (El-Zeftawy). The Subcommittee will review DOE advanced non-LWR designs Attendance regarding the treatment of severe accidents and source terms.
by the following is anticipated:
Dr. Carbon Dr. Remick Dr. Shewmon Mr. Ebersole Dr. Mark Dr. Siess .
Mr*
j Mr. Michelson l
Dr. Okrent -
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'O Advanced Reactor Designs, Date to be determined (April), Washington, DC (El-Zeftawy). The Subcommittee will review DOE advanced non-LWF; designs regarding the containment issue. Attendance by the following is anticipated:
Dr. Carbon Dr. Remick Mr. Ebersole Dr. Shewmon Dr. Mark Dr. Siess Mr'. Michelson Mr. Ward Dr. Okrent Mr. Wylie AC/DC Power Systems Reliability, Date to be determined (April), Washinoton, DC (El-Zeftawy). The Subcommittee will review the proposed Station ETackout rule (USI A-44). Attendance by the following is anticipated:
Dr. Kerr Mr. Michelson Mr. Ebersole Mr. Wylie Oc. Lewis
. Severe Accidents, Date to be determined (April /May), Washington, DC (Houston). The Subcommittee will discuss the research plan intended to resolve the source term uncertainty areas and review the Expert Panels O5 assessment of these programs. Attendance by the following is anticipated:
Dr. Kerr Dr. Shewmon Dr. Mark Dr. Siess Dr. Okrent Severe Accidents, Date to be determined (April /May), Washington, DC (Houston). The Subcommittee will continue the review of the proposed generic letter for Individual Plant Examinations (IPEs) as part of the NRR Implementation Plan for the Severe Accident Policy Statement. Attendance by the following is anticipated:
Dr. Ker Dr. Shewmon Dr. Mark Dr. Siess Dr. Okrent Thermal Hydraulic Phenomena, Date to be determined (2-day meeting, April /
May) INEL, Idaho Falls, ID (Boehnert). The Subcommittee will review: (1) the Final ECCS Rule and associated documentation, (2) uncertainty method-ology to be applied to review the new BE ECCS code models, and (3) the TIC activities at INEL. Attendance b,* the following is anticipated:
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Mr. Michelson 1 Dr. Catton Mr. Ebersole
- Dr. Schrock Dr. Kerr ! Dr. Sullivan r ; Dr. Tien Mr. Ward
( Mr. Wylie
r'~'s Decay Heat Removal Systems (tentative), Date to be determined (April /May),
Washington, DC (Boehnert). The Subcommittee will continue its review of Attendance by the following is the NRR Resolution Position for USI A-45.
anticipated:
Mr. Ward Mr. Wylie Mr. Ebersole Dr. Catton Mr. Michelson Mr. Davis Mr. Reed Standardization of Nuclear Facilities, Date to be determined (May),
Washington, DC (Alderman). The Subcommittee will discuss requirements of the EPRI Advanced Light Water Reactors Program. Attendance by the following is anticipated:
Mr. Wylie Mr. Michelson Dr. Kerr Dr. Siess Regional and I&E Programs, Date to be determined (May), Region IV, Arling-ton, TX (Boehnert). The Subcommittee will continue its review of the activities under the control of the Region IV Office. Attendance by the ss following is anticipated:
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) Dr. Remick Mr. Reed Mr. Michelson Mr. Ward Dr. Moeller Mr. Wylie Joint Seabrook/ Occupational & Environmental Protection Systems / Severe Acci-dents, Date to be determined, Washington, DC (Igne/ Houston / Major). The Subcommittees will review Brookhaven National Laboratory's draft report of the Seabrook Emergency Planning Sensitivity Study. Attendance by the following is anticipated:
Dr. Moeller Dr. Remick Dr. Kerr Dr. Siess Dr. Mark Dr. Catton (tent.)
Seabrook Unit 1, Date to be determined, Washington, DC (Major). The Subcommittee will review the application for a full power operating license for Seabrook Unit 1. Attendance by the following is anticipated:
Dr. Kerr Dr. Moeller Dr. Lewis Mr. Michelson
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q 322ND ACRS FULL COMMITTEE HEETING (V) FEBRUARY 5-7, 1987 OTHER DOCUMENTS Meeting Notebook Containing:
- 1. Project Status Report on Advanced Reactor Designs.
NRC Activities
- 2. Memorandum from M. El-Zeftawy to Dr. M. Carbon; subject:
Related to the Comissions Policy on the Regulation of Advanced Nuclear Power Plants, December 17, 1986.
- 3. SECY-86-368, subject: NRC Activities Related to the Comission's Policy 10, 1986.
on the Regulation of Advanced Nuclear Power Plants, December Stan-
- 4. Draf t Comission paper from V. Stello to Comissioners, subject:
dardization of Advanced Reactor Designs, undated.
- 5. Project Status Report on Implementation of Safety Goal Policy, January 28, 1987.
- 6. Working Copy /Sumary of January 7,1986 Meeting of21, ACRS 1987.Subcomittee on Safety Philosophy, Technology and Criteria, January O 7. Memorandum from V. Stello to Comissioners, subject: Safety Goal Imple-mentation Status, January 2, 1987.
- 8. Project Status Report on Radwaste Management and Disposal, undated.
Staff Requirements -
- 9. Memorandum from S. Chilk to V. Stello, subject:
SECY 86-192, " Sponsorship of a Federally Funded Research and Development Center (FFRDC) for Waste Management Technical Assistance and Research (SECY-85-388)", November 5,1986.
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- 10. Draft SECY, subject: Advice to Comission on Waste Management Program, undated.
- 11. Draft Coments on Draf t SECY above,1/30/87.
- 12. Project Status Report on Priority Assignment of Generic Issue 61.
" Highlights of NRR's
- 13. Memorandum from E. Igne to ACP.S Members, subject:
Technical Meeting to Discuss Generic Implications of the Surry Feedwater Pipe Failure, January 27, 1987.
- 14. IRS Report on Feedwater Line Break Due to Severe Pipe Wall Thinning at Surry, Unit 2.
- 15. Listing of Planned Meetings of ACRS Subcomittees,1/30/87.
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- 17. Draft ABWR Licensing Basis Agreement, undated.
- 18. Working Copy of Minutes of January 14, 1987 Meeting of the ACRS Subcom-mittee on Regulatory Policies and Practices,1/16/87.
Meeting Handouts:
Handout #1 including:
- 19. Project Status Report on NRR proposal generic letter for Individual Plant Examinations as part of the implementation Plan for Severe Acci-
. dent Policy Statement.
- 20. Minutes for Deccmber 19, 1986 Meeting of the ACRS Subcommittee on Severe Accident.
- 21. Consultant Reports on the above described 12/19/86 Meeting.
- 22. Draft Generic Letter, subject: Individual Plant Examinations for Severe Accident Vulnerabilities, undated.
Handout #2 including:
- 23. Introductory Remarks by P. Shewmon on the Surry Pipe Break, m
Handout #3 including:
- 24. Sumary of February 3,1987 pre-briefing and handouts.
Handout #4 including:
- 25. Memorandum from R. Fraley to ACRS, subject: Anticipated Topics for the 323rd ACRS Meeting, February 5, 1987.
Handout #5 including:
- 26. Draft Report on Qualitative Comparison of Risks Associated with the Operati% of a HLW Repository vs. a Nuclear Power Plant.
Handout #6 including:
- 27. Document on Transitional Licensing Support System, undated.
Handout #7 including:
- 28. Draft Summary of January 9, 1987 Meeting of the ACRS Subcommittee on Planning.
Other Handouts:
- 29. Viewgraphs on Certification of Advanced Reactor Designs.
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' ' ~ ~ -y gressman Udall's Questions on the NRC Safety Research, February 2,1987.
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'- 31. Viewgraphs on NRC's Surry Action Plan.
- 32. Memorandum from R. Bosnak to T. Speis, subject: Report on the Outcome of the NRR Technical Meeting Held on 1/15/87 to consider the Generic Implications of the Surry Feedwater Line Failure, January 30, 1987.
- 33. Viewgraphs on VEPC0 presentation on Surry pipe failure of December 9, 1986.
34.. Viewgraphs on I&E presentation on Surry pipe failure.
- 35. Recent Significant Events at Oyster Creek, Byron 2, Sequoyah/ Watts Bar 4
and Brunswick 2.
- 36. Viewgraphs on Release of December 3,1986 from Spent Fuel Pool at Hatch.
- 37. Memorandum from F. Bernthal to L. Zech, subject: Comments on Staff Implementation of Safety Goal Policy, January 22, 1987.
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- 38. Draft Report by ACRS Fellows concerning Regulatory Actions in Various Countries to Deal With Severe Core Damage,1/28/87.
- 39. Memorandum from R. Savio to D. Okrent, subject: Comments on Safety Goal l Policy Implementation, February 3,1987.
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