ML20237L455
| ML20237L455 | |
| Person / Time | |
|---|---|
| Issue date: | 08/26/1987 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2515, NUDOCS 8708280111 | |
| Download: ML20237L455 (41) | |
Text
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TABLE OF CONTENTS E;
MINUTES OF THE 327TH ACRS MEETING JULY 9-11, 1987
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WASHINGTON, D.C.
I.
Chairman'sReport(0 pen).........................................
1 1
II. NRCSevereAccidentRiskAssessment(0 pen).......................
1 III.
IntegratedSafetyAssessmentProgram(0 pen)......................
2 IV. TMI-2 Core Removal (0 pen)........................................
4 V.
Safety Features in Foreign Nuclear Power Plants (0 pen)...........
6 VI. Control Room Habitability (0 pen).................................
7 VII.
Improved Safety Systems for Future LWRs (0 pen)...................
7
(
VIII. Liquid Level Indication in Nuclear Power Plants (0 pen)...........
8 a
IX. Review of Operating Events - Office for Analysis and Evaluation of Operational Data Activities (0 pen)...............................
8 j
X.
Executive Sessions (0 pen / Closed)................................
12 A.
Subconni ttee Reports (0 pen).................................
12 1.
Human Factors...........................................
12 4
- l 2.
Thermal-Hydraulic Phenomena.............................
13 j
3.
Advanced Reactors.......................................
14 l
l B.
Repo rts, Letters and Memoranda (0 pen ).......................
15 1.
Comments on Draft NUREG-1150, " Reactor Risk Referenc Document"...........................................e
{
19 l
2.
Comments on the Integrated Safety Assessment Program (ISAP).................................................
15 3.
Comments on Research into Continuous Containment L e a ka g e Mon i to ri n g.....................................
16 i
4.
Comments on Improved Safety for Future LWRs.............
16 5., Comments on the Embrittlement of Structural Steel.......
16 6.
Comments on Licensee Event Reports Pertaining to Control Room Habi tabili ty...............................
17 8708280111 870826 DESIGNATED ORIGINAL l
E "S ce m n ea 37 43g_
m
l i
N 1
327TH ACRS MEETING MINUTES 11 7.
Comments on International Cooperation on Research Rel ated to Radi ati on P rotection.........................
17 C.
Other Committee Conclusions (0 pen / Closed)...................
17 1.
New Members (C1osed)....................................
17 2.
Safety Research Program (0 pen )..........................
17 3.
Eme rgency Pl anning (0 pen)............................... 19 4.
Beaver Vall ey Power Sta tion (0 pen)......................
19 5.
South Texas Project (0 pen)..............................
19 6.
Planning Committee Meeting (0 pen).......................
19 j
D.
Future Activities (0 pen)....................................
19 1.
Future Agenda...........................................
19 2.
Futu re Subcommittee Activi ties.......................... 20 j
's -
'.s iii APPENDICES j
MINUTES OF THE 327TH ACRS MEETING JULY 9-11, 1987, WASHINGTON, D.C.
Appendix I
. List of Attendees Appendix II Future Agenda Appendix III Future Subconunittee Activities Appendix IV Other Documents Received l
i
&M. gi9 [
v a
Fed:ral Register / Vol. 52, No.124 / Monday Jttne 29, 1987 y Notites
. M $M 1
Advancement Section) to the National building rubble and contaminated soil to Thursday, }uly 8,1987 d(
Council on the Arts will be held on luly an adjacent disposal site, and
^*
14.1987, from 9 00 a.m.-5:30 p.m. in room stabilization of building rubble, a30 AJf-&45 A.M. ReparvAf ACRS l
714 of the Nancy Hanks Center,1100 contaminated soil, ore tailings, and ore.
Choirm n (OpenHThe ACRS h Pennsylvania Avenue, NW.,
residues on the disposal site. Kerr-
"d 'eport brie 0y Hems d,
Washington, DC 20500.
McGee's proposed plan and alternatives current interest to Committes Bc45 A.M-1c30A.AL NRCse,q :
This meeting is for the purpose of to the plan are discussed in the DSFES.
we Panel review, discussion, evalua tion, The DSFES is being provided to the Accident Po//cy (Open)-Consider -
and recommendation on applications for State Clearinghouse. Bureau of the proposed NRC Reactor Risk Reference financial assistance under the National Budget. Lincoln Tower Plar.a. 524 S.
Document (NUREG-1150) to be used in Foundation on the Arts and the Second Street. Springfield. Illinois 82706. implementation of the NRC severe Humanities Act of1965, as amended, The DSFES is also being sent to the accident policy statement.
including discussion of information Metropolitan Clearinghouse, la45 AAL-12:45PJE Afillstone given in confidence to the Agency by Northeastern Illinois Planning Nucleor Power Station, Unit I (Open}-
grant applicants. In accordance with the Commission,400 West Madison Street.
Discuss report of ACRS subcomadttee determination of the Chairman Chicago. Illinois 60606.
regarding the Integrated Safety published in the Federal Register of Copies of the Draft Supplement to the Assessment for this nuclear station.
February 13,1980 these sessions will be Final Environmental Statement Representatives of the NRC Staff and l
closed to the public pursuant to (identified as NUREG-0904, Supplement the licensee will participate as l
subsections (c)(4) (6) and (9)(B) of No.1) may be obtained by calling (202) appropriate.
section $52b of Title 5. United States 275-2060 or (202) 275-2171 or writing to 1:45 P.AL-2:15 PJf.: Future A CRS Code.
the Superintendent of Documents U.S.
Further information with reference to Government Printing Office, P.O. Box Activities (Open)-Discuss anticipated ACRS subcommittee activities and items this meeting can be obtained from Ms.
37082. Washington, DC 20013-7982.
Yvonne Sabine Advisory Committee Interested persons may submit written proposed for consideration by the full Committee.
Management Officer. National comments on the DSFES for the Endowment for the Arts. Washington, Commission s consideration to the U.S.
2:15 P.AL-4:00 P.AL: Foreign Nuclear DC 20500, or call (202) 682-5433.
Nuc r Regu a o PowerPlant Sofety features (Open/
Yvonne Sabine.
05 A
uel ClosedHDiscuss proposed ACRS report
(
Acting Director. CouncilondPonel C cle Safety Branch. Federal, State, and to NRC regarding safety features in I al agencies are being provided with I reign nuclear plants.
Operations. Notiona/ Endo wment for the Art.s.
lFR Doc. 87-146:2 Filed 6-06-67: 8.45 am]
copies of the document. All comments Portions of this session will be closed l
received by the Commission wi!) be as required to discuss information
" C N F58' "
made available for public inspection at provided in confidence by a foreign the Commission's Public Document source.
NUCLEAR REGULATORY Room in Washington, DC and the Local 4:15 Pht.-5:15 P.A1.: Improved Sofety COMMISSION Public Document Room in West forfuturelight WaterReactors Chicago, Illinois. Comments are due by (Open}--Discuss proposed ACRS
[ Docket No. 40 20611 August 17,1987.
response to NRC request for additional Upon consideration of comments information regarding the feasibility, Kerr McGee Chemical Corp.t submitted with respect to the Draft benefit, and cost effectiveness of Availability of Draft Supplement to the Supplement, the Comtnission's staff will systems recommended in its report of Final Environmental Statement f or the prepare a Final Supplement to the Final January 15,1987.
Subject:
ACRS Rare Earths Facility, West Chicago, Du Environmental Statement. the Recommendations on Improved Safety Page County,IL availability of which will be published for Future Light Water Reactor Plant era epster.
Design.
Pursuant to the National Environmental Policy Act of1969 and Dated at Silver Spnng Maryland, this ::nd 5:15 P.AL-6:00P.AL Preparation of the United States Nuclear Regulatory day of June,1987.
ACRSReports to lAe NRC(Open}-
Commission's regulations in to CFR Part For the Nuclear Regulatory Commission.
Discuss proposed ACRS report to the 51, notice is hereby given that a Draft 141and C. Rouse, NRC regarding control room habitability Supplement to the Final Environmental Chief, hel Cycle 5afety Brcnch. Division of in nuclear power plants.
Statement (DSFES) prepared by the F"
C
',di Friday, July 10,1987 o
and g
Commission s Office of Nuclear Maten,al Safety and Safeguards. related to the lFR Doc. 87-1487, Filed 6-26-87: 8A5 am) 8:30 A.AL-10 00 A.AL TAff-2 Core decommissioning of the Rare Earths ey em ww Pemoval(Open}-Briefing by Facility located in West Chicago, representatives of INEL regarding status Illinois,is available for inspection by the of TMI-2 core removal and examination.
public in the Commission's Public Advisory Committee on Reactor Document Room at 1717 H Street NW.,
Safeguards; Meeting Agenda 10:15-12W Noon: TVA Nuclear Performance Plan (Open)--Discuss Washington DC 20555.The DSFES is In accordance with the purposes of proposed TVA Corporate Nuclear also available for inspection at the sections 29 and 182b. of the Atomic Performance Plan and plans to restart Commission'a Local Public Document Energy Act (42 U.S.C. 2039,2232b), the TVA nuclear power plants.
Room in the West Chicago Public Advisory Committee on Reactor 100 P.AL-2:30 P.AL: Activities of NRC Library,332 E. Washington Street. West Safeguards will hold a meeting on July Office for Analysis and Evoluotion of Chicago. Illinois 00185. Kerr-McGee's 9-11,1987, in Room 1046.1717 H Street, OperationalData (Open}-Briefing and proposed decommissioning and NW., Washington. DC. Notice of th!s discussion regarding 1987 Case Studies stabilization plan involves demolition of meeting was published ir the [ederal and Startup Plant Study by the NRC the existing buildings, removal of Register on June 18,1987.
Office of AEOD.
f L
g Fedesal Regiawe / Vol 52. No.124 / Monalmy;. Jose 29s 1987 / Nd e 2:30 AM-4:30 AM: Safety Festuses se Executive Directoc as fas in advaam se submkted for Connaission Foreign Nuclear Power Plonis (Open/
practicablese that appropriate censideratiom, pursuant to section 19(b)
ClosedFDiscuss peopoesd ACES arrangements.can he made to. allow the of the Securities Exchange Act of19:M Report tu the NRC regarding safety necesury timedanng the meeting for
("Act"), proposed rule changes which fea tures ist fore'ign naclanc power pta nts.
such statements. Use of still motions would permit them to waive or modify Portiorrs of dat sessice noLL be closed picture and television camera + durmg certain listing standards for foreign
(
as n-ry to discuss infunnation this meeting may be limited to selected securities.* The Amex and the NYSE provided in conflalaese by a foreign portions of the meeting as determined,
believe that such a change is necessary i
sourca.
4:45AM-&JeRMrLiguidlevel by the Chairman. Information regarding because some foreign companies are j
Indication ia Naclear Power Plants the time to be set aside for this purpose-reluctant to list on their respective i
may be obtained by a prepaid telephons-exchanges due to the fact that some 1
(Open)--Discuss proposed ACES call to the ACRSExecutive Director; listing standards are either inconsistent action / comments regarding nuclear core l
liquid.levelindica tion in nucrear powes R.F. Fraley, prior to the meeting. In vieur with or contrary to the laws, customs or plants.
J the possibility that the schedule for practices of the company's home 5'J0 AM-63DRM ACRS ACRS meetings may be adjusted by the country.8 In particular, the Amex and Chairman as necessary to facilitate the NYSE contend that listing requirements Subcontmittee Activities (OpenFHear conduct of the meeting, persons relating tocertain corporate governanca anddiscussreports of ACRS planning to attend should check with the procedures and interim earning reports subcommittees regarding thermal-ACRS Executive Director if such may unduly inhibit foreign companies hydraulic phenomena in nuclear power plants and related research activities rescheduling would result in major fro n listing on the exchanges.
inconvenience.
and proposed topics forselected safety Accordingly, the Amex and NYSE.
research reports.
t have determined in accordance with proposals would permit them to waive os Subsection 10(d) Pub. L 92-463 that it is modify certain of their listing standards Satunday; }uky 11.1987 necessary to close portions of this for foreign companies,when it can be B 30 A.M.-12:00 Noon: Preparation of meeting as noted above to discuss shown that the foreign company's ACRS Raports (Open/ClosedHDi""
nformation provided in confidence by ar procedure is based on the laws, customs proposed ACRS comments / reports to foreign source (5 U.S.C. 552b(c)(4)) and or practices of its home country, the NRC regarding met ters considered nformation the release of which would 6" f sais during.this meetinaa represent a clearly unwarranted Portions of this session will be; closed invasion of personal privacy (5 U.S.C.
The Amex and NYSE proposals would 552b(c)(6)),
permit the Exchanges to consider a as necessary to discuss information providad in ennnan" by a fareign Further information regarding topics foreign company's compliance with the to be discussed. whether the meeting laws. customs and practices of the source
- 100 AM-1:15EMt Appointment of has been cancelled or rescheduled, the country ofits domicile in determining New A CRSMembers (Open/ClosedH Chairman's ruling on requests for the whether the company has complied with Discuss status of selection regarding-opportunity to present oral statements the otherwise applicable listlag candidates for appointment to the and the time allotted can be obtained by standards.* The Amex and NYSE have ACRS.
a prepaid telephone call to the ACRS Portions of this session will be closed ExecuWe Ditector, Mr. Raymond F.
pr p sal was n ticed in a release which also to discuss information the release of Fraley (telephone 202/834-3285),
requesied comment on inues rei.ed by the Nysir and Amen proposals and on a simitar proposat which would represent a clearly between 8:15 A.M. and 5:0a P.M.
submitte t by the Neuonel Associauon of Secnnuss unwarranted invasion of personal Dated: June 23.1967 poim p.s45o 3,s,.secunnes Exchange Act privacy.
John C. Hoyle, Release No. 23460 (July 1s.1986). 51 FR 61s (Fila No. SR.NYSE-as-14)(~luly Release").
1:15 PM-2'30 PM:Prepointion of Advisory Monogement Officer.
- In discussms listms standards for foreign ACRS Reports (Open/ClosedH (m Doc. 87-14657 Filed 6-26-8h 8.45 am[
- l*l['* ^"" N'hd he. Y es?l fero Continue preparation of reports to NRC comp,
e,tus.o coo 73,o.oi.as Pegarding items considered during thit secunties issued by naa.U.S. companies. Neither the meeting; Ames or NYSE define these terms m their rules.
Amex and NYSE. however, have represented to Portions will be clooed as necessary SECURITIES AND EXCHANGE
@$s"li',
" dlEU d"c*o*mNny'. " hey tcr dioevee biformation provided in COMMISSION n
confidence by a forsgn source.
wili gen,sity refer to the derimuon for -foreign 2:30 AM-J:00 AM.: Miscellaneous.
[Rhse nica. M NAH; Flie Ns. SR-Amea.
prwate issuer" mi forth in Rule at>4 under the Act..
(OpenHComplete discussion of matters
$6-4, SR-NYSE-46-14)
Rule 36-4 dertnes foreasn pnwate inaer as any corporauon incorporated or orsamzed undes the Considieredduriog this meeting.
Procedures.fue thecoeduct of mut SeH46evy Orgsniz.ations; Order laws of any foreren country and that does not have-(1) more than so pereeni of the outsiendios vouse perwnpadosiism ACRS aseetings wear Approving Proposed Rufe Cttanges try acurttwo of such ian.er be;d of record either pirblishadiin the Federal Registes c88 the Americart Stocir EW...v inc.
d""Y ' *' ush vouns tnist ceruncan or October aL lase (61113724t), in and New York Stock Exchan64.Inc.To NNtE[n'MhY'n"af, h tSe h'htb accordesco with these pnecedueens seek Amend the Exchanges' Listlog orricers ordueetors are U s. causens or rewdenis.
or writies aMinay be p---M, Standards for Foretgn Companfer tio more man so p cent w ow emeu of me i aus are loca+.d in the U.S6 or(n0 the busmess of the by s.nhean el the peabbe,receedy I. latroduction issuer is edanisted pnncipaHy in the M will be permitted Enh daismg,dasaar pess6eas o$ tha mearingwhen a The Americas f*dmex7 and !IIese' S
ano n m n are e
affare tra escript is.haisgiapL and questions York ("NYSE"lStock Para =-aav sismficant stiard fewei representaihm to there rq har malsadOErg kr"--E of th&.
(collect,,eh.ther"r SM ' have
'ap** "h"* '**, b. 6n coneet =m Am requirements rslatmg to wouas foe a Beard d Coaiarities.Ga-9'wa andStaIT.
PerwasAsieing la meine oomt s.r w premah w h i SM' riceu the amen c.=,m.cy c.de a seysc a aw mv Emenemme.m.samamme >- m.= on.eese:s samer statements should notify the ACES st m tit 2s (Fim e " r.mmas,,
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e C QtCu ~
'o UNITED STATES 8'
",g NUCLEAR REGULATORY COMMISSION
.l ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.
O, WASHINGTON, D, C. 20555
%...../
Revised: July 7 1987 SCHEDULE AND OUTLINE FOR. DISCUSSION 327TH ACRS MEETING July'9-11, 1987 WASHINGTON, D.C.
Thursday, July 9, 1987, Room 1046, 1717 H Street, N.W., Washington, D.C.
1) 8:30 - 8:45 A.M.
Chairman'sReport(0 pen) 1.1) Opening remarks (WK) 1.2)
Items of current interest (WK/RFF)-
2) 8:45 - 10:30 A.M.
SevereAccidentRiskAssessment(0 pen) 2.1)
Report of joint subcommittee meeting on July 8, 1987 regarding i
TAB 2----------------
NUREG-1150, Reactor Risk Reference-Document (WK/D0/MDH/RPS) 2.2) Meeting with representatives of NRC Staff 10:30 - 10:45 A.M.
BREAK
- 3) 10:45 - 12:45 P.M.
Integrated Safety Assessment Procram (0 pen) 3.1)
Report of subcommittee meeting on July 7, 1987 regarding the Integrated TAB 3----------------
Safety Assessment Program (ISAP) for Mil.lstone: Nuclear. Power Station, Unit 1 and Haddem Neck (DAW /RPS) 3.2) Meeting with representatives of the NRC Staff and.the licensee 12:45 -
1: 45 P.M.
LUNCH ~
4) 1:45 -
3:15 P.M.
TMI-2 Core Removal (0 pen) 4.1) Briefing by INEL representatives-TAB 4-----------------
regarding status ef TMI-2 core removal and examination (DWM/RKM)
- 5) 3:15 -
4:00 P.M.
FutureACRSActivities(0 pen)
TAE------------------- 5.1) Anticipated subcommittee activities (MWL/RFF)
TAB-------------------5.2)
Proposed items-for. ACRS consider-ation (WK/RFF).
5.3) EmergencyPlanning(DWM/EGI)
~
5.4) Low Temperature Radiation' EmbrittlementofSteel.(PGS/EGI) 4:00 -
4:15 P.M.
BREAK
327th ACRS Meeting Agenda,
l
- 6) 4:15 -
6:00 P.M.
Safety Features in Foreion Nuclear Power Plants (0 pen / Closed)
TAB 6--------------
6.1)
Discuss proposed ACRS report to NRC regarding safety features in foreign nuclear power plants (00/RPS)
(Portions of this session will be closed as necessary to discuss information provided in confidence by a foreign cource.)
Friday, July 10, 1987, Room 1046, 1717 H Street, N.W., Washington, D.C.
- 7) 8:30 -
9:15 A.M.
Control Room Habitability (0 pen)
TAB 7--------------
7.1)
Discuss proposed ACRS report to NRC regarding control room habitability in nuclear power plants (DWM/EGI)
- 12) 9:15 - 10:15 A.M.
ACRS Members (0 pen / Closed) i 12.1)
Discuss status of selection / appointment l
l of Committee members (WK/HWL/NSL)
(Portions of this session will be closed as necessary to discuss information the release of which would represent a clearly unwarranted l
invasionofpersonalprivacy.)
l l
10:15 - 10:30 A.M.
BREAK
- 9) 10:30 - 11:30 P.M.
Imoroved Safety Systems for Future LWRs (0 pen)
TAB 9---------------
9.1)
Discuss proposed ACRS action regarding inquiry from Chairman i
)
Zech cn the feasibility, I
benefit, and cost effectiveness of the systems noted in the ACRS letter dated January 15, 1987 (CJW/RKM)
- 10) 11:30 - 12:15 P.M.
Liquid Level Indication in Nuclear Power Plants (0 pen)
TAB 10---------------
10.1)
Discuss proposed ACRS action / comments regarding liquid level indication in nuclear i
power plants per recent event at i
Diablo Canyon (JCE/MME) l l
12:15 - 1:15 P.M.
LUNCH l
l l
327th ACRS Meeting Agenda 11) 1:15 - 2:45 P.M.
Office for Analysis an'd Evaluation of 4
Operational Data Activities (0 pen) j TAB 11 -------------
11.1). Briefing / discussion regarding 1987 Case Studies and Startup Plant Study by AEOD representa-tives (JCE/RKM) 2:45 - 3:00 P.M.
BREAK
- 13) 3:00 - 5:15 P.M.
ACRS Subcommittee Activities (0 pen)
I TAB-----------------
13.1) 3:00-3:3D: Human Factors Subcommittee -
l Report regarding Degree Requirements j
for Senior Operators (SECY-87-101)
(FJR/HA) 1 TAB-----------------
13.2) 3:30-4:15:
Thermal-Hydraulic Phenomena
{
i
.B&W MIST program and proposed action i
regarding ACRS report on the NRC Thermal-Hydraulic Research Program (DAW /PAB)
TAB-----------------
13.3) 4:15-4:45:
Proposed ACRS Reports on-Selected Research Topics - Discuss:
proposed candidate topics for ACRS reports (CPS /SD) 13.4) 4:45-5:15:. Advanced Reactor Designs'-
Report regarding DOE Advanced Reactor Designs (DAW /MME)
- 15) 5:15 P.M. - 6:15 P.M.
Preparation of ACRS Reports'to the NRC (0 pen) 15.1) Safety Features in' Foreign Nuclear Plants (WK/D0/RPS)
Saturday, July 11, 1987, Room 1046,'1717 H Street, N.W., Washington, D.C.
- 14) 8:30 - 12:00 Noon Preparation of ACRS Reports to the NRC
-j l
(0 pen / Closed).
J l
14.1)
Discuss proposed ACRS. reports regarding:
14.1-1) 8:30-9:30: Safety' Features in Foreign Nuclear Power Plants.
.(WK/D0/RPS)
'14.1-2) 9:30-10:00: Control Roor Habitability (DWM/EGI) 14.1-3) 10:00-10:45:
Integrated Safety Assessment Program' (DAW /RPS)
- .s -
327th ACRS Meeting Agenda -
14.1-4) 10:45-12:00:
Severe Accident Risk Assessment (NUREG-ll50).
(WK/D0/MDH/RPS)
(Portions of this session will be closed as necessary to discuss information provided in confidence by a foreign source.)
12:00 - 1:00 P.M.
LUNCH 16) 1:00 - 2:00 P.M.
Preparation of ACRS_ Reports (0 pen / Closed) 16.1) Complete preparation of ACRS reports noted above O
0ERiiEfi MINUTES OF THE 327TH ACRS MEETING JULY 9-11, 1987 The 327th meeting of the Advisory Committee on Reactor Safeguards, held at 1717 H Street, N.W., Washington, D.C., was convened by Chainnan. W.
Kerr at 8:30 a.m., Thursday, July 9,1987.
[ Note:
For a list of attendees, see Appendix I.
[Dr. Okrent did not i
attendthemeeting.].
l i
The Chairman said that the agenda for the meeting had been published.
l He identified the items to be discussed on Thursday. He stated that the
{
meeting was being held in conformance with the Federal Advisory Commit-l tee Act and the Government in the Sunshine Act, Public Laws92-463 and i
94-409, respectively.
He also noted that a transcript of some of the public portions of the meeting was being taken, and would be available in the NRC Public Document Room at 1717 H Street, N.W., Washington, D.C.
j
[ Note:
Copies of the transcript taken at this meeting are also avail-I able for purchase from ACE-Federal Reporters, Inc., 444 North Capitol l
Street, Washington,DC20001.]-
l I.
Chairman'sReport(0 pen)
[ Note:
R.
F.
Fraley was the Designated Federal Official for this j
portion of the meeting.]
1 The Chairman welcomed Dr. Martin Steindler to the ACRS meeting.
(Dr.
Steindler attended the meeting as an observer as his appointment as an ACRS member was not yet effoetive.)
The Chairman said that Dr. Okrent might not attend the meeting as Dr. Okrent's wife was ill.
Dr. Kerr reported that full-power licenses had been issued to Nine Mile Point, Unit 2 and to Braidwood.
II. Severe Accident Risk Assessment (Ocen)
[ Note:
M.
D. Houston was the Designated Federal Official for this portionofthemeeting.]
Dr. Kerr, Chainnan of the Subcommittee on Severe Accidents, reported on the joint meetings of the Severe Accidents /Probabilistic Risk Assessment Subcommittees on June 3 and July 8, 1987 held to review the draft NUREG-1150, " Reactor Risk Reference Document," that was issued for comment in February 1987.
He indicated that the Committee should prepare a report on NUREG-1150, either at this meeting or at the August meeting, and he requested that members provide him with any input for the report during the meeting.
Individual Committee members expressed their opinions of the RES report, and questioned its objectives, analyti-cal methods and regulatory applicability.
J. Murphy (RES) presented a general overview of the NUREG-1150 studies-of five plants.
He discussed the intended objectives, methodology, t
327TH ACRS MEETING MINUTES 2
I treatment of uncertainties, dominant accident sequences, and risk results.
The results were compared to the Safety Goal concepts for early fatalities, latent fatalities and large releases.
He briefly
]
discussed the current studies that will be: factored into the final i
version of NUREG-1150, tentatively scheduled to be issued in June 1988.
Mr. Ebersole questioned the method of discrete value. selection and of -
weighting factors applied in the analyses.
Dr. Shewmon questioned the challenge to containment indicated' by the direct containment heating (DCH) process.
He also questioned.the.high probability that was assigned to. the DCH process.. He expressed some-concerns about the. lack of organized sections in the report on insights and conclusions.
J. Murphy indicated that all of Volume I represented conclusions.
Mr. -Michelson expressed a concern that containment penetrations would not withstand the pressure levels displayed in.the reports.
He also questioned the. lack of consideration of external initiators.
Dr. Siess questioned the sensitivity studies and some of the arbitrary assumptions.
Dr. Kerr asked about the source term code package (STCP) and how it was used in the studies.
J. Murphy indicated that a modified STCP was used in most cases after the first 5-6 runs. The modifications were intended to overcome some of the known limitations in the published ve.sion of the STCP.
Dr. Moeller questioned the assumptions used for calculating early and late fatalities and the apparent discrepancies between NUREG-1150 and other studies at the New York Power Authority and at Oak Ridge Associ-ated Universities.
III.
IntegratedSafetyAssessmentProgram(0 pen)
[ Note:
R. P. Savio was the Designated Federal ~0fficial for this portion of the meeting.]
Mr. Ward reported on' the July 7,1987 meeting of the Subcommittee on the IntegratedSafetyAssessmentsProgram(ISAP).
The NRC Staff discussed the history of the development of ISAP.
The ISAP concept evolved from the Systematic Evaluation Program (SEP) and the Interim Reliability Evaluation Program (IREP).
The SEP was. started in 1977 to review operating plants against the regulatory requirements that had evolved since the particular reactor had been -licensed..An initial program involving 10 the Standard Review Plan (SRP) plants and comparing these plants against has been completed. The concept of IREP
327TH ACRS MEETING MINUTES 3
was developed from the TMI Action Plan and was intended to involve the performance of a relatively simple probabilistic risk assessment (PRA) which was to be used to identify risk outliers.
The experience with these two programs indicated resolution of safety issues would be more effectively achieved if they were addressed in an integrated, risk-based, and plant-specific review.
The ISAP program was initiated in late 1984 as a method for integrating ongoing plant safety improvements and licensing actions. The intent was to develop a procedure for combining, scheduling, and doing a risk-based evaluation of changes in plant procedures and equipment.
A plant-specific PRA and a review of plant operating experience were to be part of the ISAP.
The ISAP procedures involved an initial screening to sort out those topics which could be reasonably included in an ISAP-type procedure.
The sources of topics for the ISAP were general licensing
- actions, USI/ Generic
- Issues, licensee-initiated plant improvements projects and topics which arose from a review of the plant-specific PRA and a general review of plant operating experience.
A pilot program was developed involving Northeast Nuclear Energy Compa-ny's (NNECO) Millstone 1 and Haddam Neck plants.
NNECO has completed the ISAP evaluation for these two plants.
The NRC Staff has issued their draft Integrated Safety Assessment Report (ISAR) evaluation on Millstone 1 and expects to issue their draft evaluation for Haddam Neck
{
in August 1987. The NRC Staff is preparing a SECY paper describing the lessons learned in the pilot ISAP and giving recommendations for the l
i future use of ISAP. A draft paper has been prepared and is currently undergoing management review.
It is very likely that the NRC Staff will recommend that ISAP be made available in the future to licensees that wish to participate.
i NNECO oiscussed their experience with the ISAP process.
NNECO has proposed extending the program to Millstone 3 and Millstone 2.
They believe that the process provides an effective method for resolving safety issues in a cost-effective fashion and for interacting effective-I ly with the NRC Staff on licensing issues.
There was some discussion as to how the Millstone 1 and Haddam Neck ISAPs could be compared to the use of the Integrated Living Schedule l
l l
(ILS), and the use of the SEP process.
The type of evaluation which is performed in a SEP process is included in the ISAP process (Millstone 1 and Haddam Neck had already been through the SEP process).
Plants for which a SEP have not been performed would have to be handled differently in an ISAP.
The ILS allows a prioritization of safety issues but not j
the combining of, or dropping of, safety issues.
l A plant-specific PRA is an important part of the current ISAP process.
The Millstone 1 and Haddam Neck ISAPs used Level 1 PRAs with a limited treatment of external events and without any consideration of seismic risk.
There was discussion of the adequacy of this approach.
It was i
i l
1 l
l l
327TH ACRS MEETING MINUTES 4
suggested that the consideration of seismic events, a better treatment of external events, and upgrading the PRAs to a Level 3 (i.e., considera-tion of both core melt frequency and containment performance) ' were necessary for a complete evaie= tion of some ISAP issues.
IV. TMI-2 Core Removal (0 pen)
[ Note:
R. K. Major was the Designated Federal Official for this portion ofthemeeting.]
Dr. Don McPherson, DOE, introduced and showed audiovisual TV tapes of the TMI-2 core removal.
Dr. McPherson ie, a member of the Light Water Reactor Safety and Technology Office whkh includes a series ofl severe-accident technology programs of which TMI-2 is one.
The objective of the TMI-2 program is the transport and disposal of the core, as well as -
producing an accident evaluation that will allow an understanding of the sequence involved in core damage and the accident so rte term.
Mr. Phil Grant, EG&G, manager for the TMI-2 program for.D0E, presented a discussion of the defueling at TMI-2.
He noted the loose core debris or rubble bed on top of the core which represents about 35 tons of materi-al.
A central prev 1ously molten core mass, which now is characterized by a ceramic core, weighs about 30 tons. The bottom of the core is made up of the stubs of the 177 fuel assemblies.
This region of partial standing rods represents about 60 tons of core.
It is estimated that on the order of 20-25 tons of core debris migrated from the. original core 3
i region into the lower vessel head region.
It was noted that an eight-inch metallic layer of material coats the inside of the bottom vessel head.
This eight-inch layer is believed to be control rod material.
Current estimates are that 40% of the core debris has been removed.
Mr. Grant explained that the hardest regions of core to defuel will be those outside the core barrel, in the lower core support assembly, and the lower head.
To gain access to these regions, the remainder of the vessel internal structures must be avoided.
Currently the plan is not to remove the remaining core assembly attachment, but rather to attempt to vacuum out the debris.
Regarding shipping the core debris offsite, Mr. Grant mentioned that two shipping casks capable of holding'seven debris-filled canisters are in -
Twelve shipments have been made from TMI which represent about 80 use.
percent of the debris removed so far. Typically each cask holds between four and five tons of debris.
The casks are shipped by rail.
Current plans are to lease another cask with the anticipation that the rate of defueling will increase during removal of the intact partial fuel subassemblies.
Mr. Grant mentioned the problem of biological growth '(bacteria, algae, and Euglena) in the vessel.
This creates a lot of turbidity which slows the defueling process.
With the combination of problems and technical
I 327TH ACRS MEETING MINUTES 5
{
difficulties, it is believed that defueling activities will go well into 1988.
Slides of the defueling platform, the types of tools designed for defueling, and pictures of core debris were shown.
The tools were developed by EG&G and Westinghouse.
Also, during the photographing of the core debris, a major degradation in the core's former wall was discovered.
This represents a potential pathway that the molten core could have taken to flow into the lower vessel head.
Mr. Jim Broughton, EG&G, discussed the TMI-2 Accident Evaluation Pro-gram. The objectives of this program are:
to understand what happened during the accident, apply the understanding to resolution of severe accident and source term technical issues, and to transfer results of the program (both domestically and internationally) to government, I
nuclear industry and the public.
Reports which describe the accident I
scenario are scheduled for publication later this year and next year.
]7 as noted that significant amounts of the fission products,129I and w
Cs, were retained in the high temperature core debris and previously l
molten core materials. There was also a significant amount of ruthenium released from the fuel that was retained in metallic structures.
The molten core material is composed of a central homogeneous region of a previously molten ceramic. Through this region run veins of previous-ly molten metallic material.
This homogeneous region is encased in an agglomerate of rod stubs and remnants held together by a crust of previously molten material.
I EG&G has concluded that at least 35% of the core was molten during the accident.
About 20-25% of the core was oxidized and fragmented.
About 40-45% of the core was unaffected by the accident (the rod stubs at the bottomofthecore).
l Between 100 and 174 minutes into the accident, the water level in the vessel dropped below the height of the core.
The core began to heat up from the top.
The highest temperatures were in the central region of the core.
Fuel cladding oxidized and liquid fuel began to flow toward the bottom of the core.
At 150-160 minutes into the accident, the minimum water level was reached which was just above the second grid space, 20-25 inches above the bottom of the core.
The progression of molten material stopped here.
A crust was formed by the coolant water and a crucible with a partially molten slug of material was fonned.
Between 175-180 minutes a cooling pump was turned on and the water level rose to the mid-core elevation. The thermal and mechanical shock of the reflood shattered the oxidized fuel rods in the top of the core.
The molten central region remained liquid but a crust formed around it.
i There was no relocation of core material to the lower vessel head yet.
+
327TH ACRS MEETING MINUTES 6
Betweca 175 and 224 minutes, the central region of the consolidated mass continued to heat up.
This happened despite the fact that the water level was near the top of the original core height.. There was insuffi-cient surface area to remove, decay heat or.sftbilize' thermal conditions.
At 224 minutes the reactor coolant pressure increased to 300 psi.
Source range monitors showed an increase by O factor of two.
Supei-heated steam entered the cold ings.
It is. postulated that durin, the next tw minutes, N tons of malten corium flowed from the central region near the mytheast core ' periphery and relocated in the lower plenum.
Smaller f b s may have melted holes in the side of the core on the former wall, but it is believed the probable relocation route was through the core proper.
Evidince to confirm this should be in had in the next year.
Mr. Reed asked what the difference in accident sequen' e would have been if the core at TMI-2 had,been through several fueling cycles, rather than a new core.
It was believed that the initial melting sequence could have been faster due to greater fission product decay heat and more fission prodLcts would have been released.
However, after the rapid oxidation of the If rcalloy and its large heat input, the accident sequence would have been much the same.
At this time, EG&G is starting to use research results;to determine some of the unresolved severe actident and source term technical issues.
It is believed the results of the TMI-2 research will hue a significant impact on resolving existing uncertainties related to core damage progression, steam explosions and fission product release and transport.
V.
Safety Festu'res in Foreign Nuclear Power Plants (0 pen)
[Non:
R.
P., %vio was the Designated Federal Official for this portion of tne vaeting.} -
The ACRS continued its discussion.on the nw safety features being installed on sne foreign reactors. The topic was last discussed during the June 1987 ACRS meeting.
Mr. Michelsca and Mr. Wylie had visited with representatives of the Federal Republic of Germany after the June meetine and obtained information as to the design of the KONV0I reac-tors.
In addition A. Tabatabai has completed a comparison of the safety features recommended by the ACRS for consideration for future LWRs (see January 15, 1987 ACRS report) to the design features of the KONV0I.
It appeared that many of the features recommended for consideration by the ACRS are incorporated in the KONVOI design.
A draft ACRS report by Dr. Okrent was considered.
The report noted (in i
l response to Ct6f rman Zech's request for such information) that the KONV0I plants did incorporate the features which ACRS had recommended for consideration in future 1LWRs.
327TH ACRS MEETING MINUTES 7
q q
VI.- Control Room Habitability (0 pen)
[ Note:= E. G. Igne was. the Designated Federal Official for this portion of.the meeting.]
The ACRS Subcommittee on Occupational and -Environmental Protection 1
Systems'(OEPS) had been asked by the Comittee to do the following:
1 check. with AEOD. on as~ to whether they are studying air monitor l
failures with respect to control room habitability, and-i check on the Technical Specifications requirements on air monitor reliability.
Dr. Moeller reported to the ACRS on the OEPS Subcommittee meeting held j
on June 22-23, 1987 to discuss its charge.
He stated that AE0D has no fomal program on this matter and that an ACRS letter would be-useful..
Although the Standard Technical. Specifications requirements.specify a 3
1 time limit within which a defective air monitor must be repaired and returned to service, they do not epply to 'all nuclear power : plant licensees.
[The Comittee decided to send the letter modified so that its main thrust would ~be on improving air' monitor reliability.]
VII.
Improved Safety Systems for Future LWRs (0 pen)-
[ Note:
R. K. Ma of the meeting.]jor was the Designated Federal Official.for this portion 1
During this session the Comittee reconfirmed its intention to assign a subcommittee to consider the matter of the feasibility, benefit, and cost effectiveness of the systems noted in the Comittee's January 15, 1987 letter entitled "ACRS Recommendations'on Improved Safety. for Future 1
Light Water Reactor Plant Design."
This subcommittee will attempt to make an estimate of.the resources required to prepare a reply that the 1
Committee believes will respond to Chairman Zech's inquiry.
[ Note:
This task has been assigned to the Subcommittee on Future LWR Designs.
1 (C. Wylie, Chairman; Members C. Michelson, D. Okrent, G. Reed,' C. Siess; R. Major,StaffEngineer).]
The Comittee ' decided to send a' letter to Chaihnan Zech on Improved i
Safety for-Future ' Light Water P.eactors.
In this. letter, subsequently.
sent on July-15, 1987, the ACRS transmitted the name of'the FRG KONVOI plant as an example of an existing plant that incorporated the desirable features recommended for consideration in. their -letter of January 15, 1987.
Additional comments by Dr. Lewis to this letter stress that the items in the January 15, 1987 letter were recommended for consideration only and not necessarily for mandatory. incorporation into a future plant.
327TH ACRS MEETING MINUTES 8
VIII. Liquid Level Indication in Nuclear Power Plants (0 pen)
[ Note:
M. M. El-Zeftaw portion of the meeting.]y was the Designated Federal Official for this i
Mr. Ebersole briefed the Committee in regard to the seriousness of erroneous water level indications in reactor vessels and cited the Diablo Canyon event which occurred on April 10, 1987.
The Diablo Canyon, Unit 2 reactor experienced a loss of decay heat removal capabil-ity in both trains. The reactor coolant system had been drained down to the mid-height of the hot leg piping in preparation for the removal of the steam generator manways.
The heat removal capability was lost for 85 minutes, during which the reactor coolant heated from 87 F to boil-ing.
Steam was vented from an opening in the hecd, water was spilled from the partially unsealed manways, and the airborne radioactivity level was increased in the containment.
Tne plant had been shut down for seven days at the time and the containment equipment hatch had been opened.
Erroneous level instrumentation, inadequate knowledge of pump suction head / flow requirements, incomplete assessment of the behavior of the cir/ water mixture in the system, poor coordination between ccntrol room operations and containment activities all contributed to the event.
In addition, there was no reactor coolant system temperature indication available during this event.
An Augmented Inspection Team has investigated this event.
In addition, the NRR Staff has prepared a 50.54(f) letter requiring information from all PWR licensees pertaining to this issue.
The NRC Staff made a presentation on this event to the ACRS ac its 326th meeting held June 4-6, 1987.
i M r. Reed commented that the root cause of this "mid-loop" loss of shutdown cooling incident is that the design is deficient in not having j
water box vents on the primary boxes (hot leg side and cold leg side) of the steam generators. The Committee took no action at this time.
IX.
Review of Operating Events - Office for Analysis and Evaluation of Operational Data Activities (0 pen)
[ Note:
R. K. Ma l
of the meeting.]jor was the Designated Federal Official for this portion j
j i
The ACRS was briefed by the AEOD Staff on their review of operating j
events.
It was hoped that the Committee would get an idea of how the analysis of trends and patterns functions through discussion of events included as examples.
The level of AE0D analysis of operational events is expected to remain constant during the oncoming years.
Internally, AEOD is increasing emphasis on the human factors side of their analyses.
i AE0D is also exploring improved ways of communicating the lessons i
learned from their analyses to the industry.
i
327TH ACRS MEETING MINUTES 9
Mr. Jack Rosenthal discussed the mission of AEOD's Reactor Operating Analysis Branch.
The Branch annually reviews on the order of 3,000 LERs, and produces about 30 studies a year made up of technical reviews, engineering evaluations, and case studies.
Mr. Rosenthal noted that the Reactor Operating Analysis Branch has always been involved with questions of operator actions and procedures 1
as well as hardware related questions.
In the future, it is envisioned that a substantive amount of resources will be applied to human perfor-mance, procedures, and procedure inadequacies and the man-machine interface.
The diversity of work products was discussed.
A Technical Review is a quick LER search coupled with a little bit of analysis.
Ashort(about three pages) report is written that addresses a single concern.
An Engineering Evaluation takes several months to complete and is the next step up in product line.
It involves LER searches and some analytic work, and produces suggestions.
Suggestions are ideas AE0D feels have merit, but probably would not pass the backfit criteria.
A case study is the most elaborate product.
It requires a one person-year effort and results in a substantive report.
Recommendations in a case study could require backfitting.
Mr. Rosenthal discussed the effort that went into a case study on loss of decay heat removal.
Even after the report and its recommendations, i
I events continued at an unacceptable rate.
AE0D has written letters and advised the industry that further action must be taken, requiring assurance that decay heat removal can be maintained at plants.
Mr.
Jordan believes that their response to this issue appears appropriate.
Mr. Matt Chiramal presented an example of an engineering evaluation which dealt with " Auxiliary Feedwater Pump Trips Covered by Low Suction Pressure."
This study was initiated by an event that occurred at Millstone-3 on January 29, 1987.
The AFW pumps tripped repeatedly during surveillance testing. The trips were caused by pressure oscilla-i l
tions in the pump suction lines. After reviewing operating experience, i
three similar events at other plants were discovered.
The safety significance of this type of event is that a low suction pressure trip caused by pressure oscillations constitutes a common mode failure that could render the AFW system inoperable.
Licensees have taken two types of corrective actions. Some licensees have provided a time delay on the trip so that momentary low pressure does not actuate the trip.
Other licensees have removed the low suction pressure trip.
AE0D has recom-mended that NRR issue an Information Notice on this matter.
Mr. Sanford Israel discussed a study, " Loss of Pressure Control," that was initiated following an event at Salem on August 20, 1986. The event was caused by technicians shorting an instrument bus.
This started a series of events including a safety injection signal, plant trip and l
loss of station transformers. As the emergency buses fed by the diesels
327TH ACRS MEETING MINUTES 10 started to load, they did so in a sequence to respond to loss of offsite power and a safety injection signal.
In this sequence, component cooling water is not automatically loaded.
Without component cooling water, the operators became concerned over the cooling of the reactor coolant pump (RCP) seals.
A decision was made to turn off the RCPs.
1 Without the pressure difference created by the RCPs, the pressurizer
?
sprays would not function and pressure control became difficult.
The pressurizer became filled with water and one of two power operated relief valves (PORVs) was opened.
The second PORY had been isolated because it was leaking.
The third diesel was also down for repair and 3
this removed a power supply that prevented closing a valve necessary to l
isolate ECCS flow.
The safety significance of the loss of pressure control is that the steam generator tubes are challenged.
The failure would result in the loss of the primary system pressure boundary and could challenge i
containment if safety valves outside containment actuate.
Depressurization is also needed for natural circulation cooldown and i
4 feed-and-bleed situatiens.
f Currently, AE0D is planning to draft an Information Notice and send this report to NRR's Tech. Specs. improvement group. The report will also be sent to those responsible for the Standard Review Plan and the accident management program to make any appropriate changes.
Mr. Peter Lam, AE0D, presented the case study process using Service Water as an example.
Before this' effort is finished, 1-11 person-years will be spent on it.
Significant amounts of time are spent on the collection of data and on the analysis of data which looks at the sequence of cyents and causes.
Evaluations of incidents are also performed in ordar to study their safety significance.
The final direction of the study is decided by AE0D management.
The final step i
before finalizing the report is a peer review process by ~;he other NRC Offices. Recommendations contained in the report would have to pass the criteria in the backfit rule where a vigorous cost-benefit analysis may be required.
In the current service water case study, over 1000 events l
have been studied.
I Mr. M. Williams discussed the activities of the Trends and Patterns Analysis Branch.
This group has responsibility for maintaining the agency's data bases for operating experience.
Operating experience is collected, screened, and stored. This Branch also evaluates the quality of the NRC data base and the data bases maintained by industry (e.g.,
the Nuclear Plant Reliability Data System).
Reporting requirements are also subject to formal evaluation.
Recent initiatives by the Trends and Patterns Analysis Branch include increased and improved use of operating experience feedback to improve operational safety, and to evaluate corrective actions.
i i
1
', L 327TH ACRS MEETING MINUTES 11 l
J
<l 1
Dr. Lewis asked if AE0D can bring pressure to bear on the NRC Staff and the industry to correct a problem that has been identified, but which, it appears, is not receiving appropriate attention.. Mr. Jordan said AE0D could provide, and Mr. Jerdan agreed to provide, correspondence on the decay heat removal issue as an example of AEOD pursuing closure of an issue, i
Dr. Kerr requested that at a later session AE0D be prepared to' discuss the risk increase or decrease that might accrue if, instead of testing during plant' operation, testing is deferred until a plant is down--as he understands the Japanese do.
The Committee was briefed on AEOD efforts regarding operating experience feedback on new plants.
This effort was aimed at improving new plant startups and reducing the frequency of reportable events.
This effort has reached the draft report stage.
A Commission briefing is planned '
for August 4, 1987.
Mr. Denning, AE00, also discussed the study of operating experience of new plants.
He noted that information and preliminary conclusions were drawn from LER information, but that final conclusions were based on plant visits and discussions with utilities.
AEOD wants to make sure that experience at plants of a given type, or at a particular. utility, is fed back to new plants so new plant crews need not learn from their own errors.
Good performing plants were those with specific management traits. Recommendations in this report were obtained through interviews with utilities that were considered to be both good and bad performers.
The good performers named the reasons for their success; the poor performers listed the things they wished they had done.
This report is basically a reflection of ' industry practices.
Ms. Kathleen Black, AEOD, Non-Reactor Safety Group, briefed the Commit-tee on Human Factors studies. She discussed a completed study on events
{
involving the wrong unit / wrong train / wrong component.
She also. dis-cussed a study in progress that concerns events involving procedural errors during the time period 1984-1985.
The special study of wrong unit / wrong train / wrong component events was published in January 1984.
As a result of this study, Generic Issue-102, " Human Error in Events Involving Wrong Unit or Wrong Train " was created as part of the solution.
Generic Issue '102 is still open.
Study updates or data summaries were published in.1984,1985,1986, and 1987 Presently, data show no major change in the rate off occurrence of wrong unit / wrong train events with time.
The data do show that older plants have generally lower rates'than new plants.
Regarding the ongoing study of events including procedural problems, approximately 100 events involving procedures have been identified.
Preliminary observations include:
327TH ACRS MEETING MINUTES 12 j
I There is wide variation in frequency of such events from plant to l
plant; about 12% of plants reported about 44% of the events.
i About as many events involving procedures occurred when the reactor 1
was shut down as when the reactor was critical.
Only three events were ascribed to personnel failing to follow a J
procedure.
Procedural content or presentation were dominant problems.
]
X.
Executive Sessions (0 pen / Closed)'
A.
SubcommitteeReports(0 pen) 1.
Hu_ man Factors l
[ Note:
H. Alderman was the Designated Federal Official for this portion of the meeting.]
Chainnan Remick reported on the Human Factors Subcommittee meeting of June 14, 1987. He noted that J. Ebersole, W. Kerr, G. Reed and D. Ward were present at the subcommittee meeting.
i He remarked that t.he Advance Notice of Proposed Rulemaking, p(ertaining to degree requirements for Senior Reactor Operators SR0s) was issued on May 31, 1986.
The rulemaking, as dis-l cussed in the advance notice, requires bachelor degrees in l
engineering or physical science for Senior - Operators after January 1991.- Degree equivalents would not be acceptable.
There would be a " grandfather" clause to cover SR0s licensed l
I before January 1991.
The advance notice called for a concur-rent policy statement in conjunction with the rulemaking.
Further education was called for for R0s and SR0s.
1 Dr. Remick noted that there were 200 responses to the advance notice of proposed rulemaking. Coments included those of KMC (which represented a number of utilities),
professional reactor operators, licensed operators, and others.
Of these comments, 5 favored the degree requirements and 195 were opposed. The opponents to the degree requirement did not see its validity, believed it would block career paths, and thought it would be too costly.
The Subcommittee heard a presentation from Mr. John Gallagher of the Westinghouse Electric Corporation.
Mr. Gallagher's presentation centered upon the concept of providing a better information flow to the licensed operators as an alternative to the requirement of a degree. Dr. Kerr noted that he didn't find Mr. Gallaghers' arguments very convincing. He noted that Mr. Gallagher didn't present a convincing argument of a correlation between the supervisory center (the Westinghouse name for this concept) and the degree requirement.
Dr. Siess
327TH ACRS MEETING MINUTES 13 noted the proposed degree requirement would not promote professionalism as proposed. Dr. Remick noted that one of the bases for the proposed rule is to promote a career path into upper management for degreed _ operators with operating experi-ence.
The Subcommittee proposed that the Comittee write a letter regarding this topic during the August 1987 meeting.
2.
Thermal-Hydraulic Phenomena I
[ Note:
P. A. Boehnert was the Designated Federal Official for this portion of the meeting.]
D. Ward, Subcommittee Chairman, presented a briefing on high-lights of the Thermal-Hydraulic (T/H) Phenomena Subcommittee's meeting on June 18, 1987.
He stated the principal purpose of the meeting was to review the combined NRC/EPRI/B&WOG/B&W Integral System Test Program for B&W plants.
The Program centers around the MIST facility at B&W's Alliance Test Center. He said the overall program is proceeding fairly well but the original basic issues have not all been addressed.
Some surprises and new questions have been seen but it is hoped the outcome should improve the. codes.
In response to a comment from Mr. Michelson, Mr. Ward said that any surprises seen will be followed up at a future Subcommittee meeting.
Mr. Ward noted that beyond FY-88 the T/H research program becomes vague.
He said the Subcommittee, at ACRS behest, has begun a review of the RES T/H program, given the central importance of T/H vis-a-vis plant safety.
The lack of any real imediate concrete T/H concerns is causing the program vagueness noted above.
Mr. Ward said that the Subcommittee will hold 2-3 meetings with the goal of developing a "research requirements docu-ments" for ACRS consideration by the end of this calendar year.
Mr. Michelson referred to the Appendix K Rule revision.
Mr.
Ward indicated that the Subcommittee plans to review the RES proposed methodology for addressing best estimate code uncer-tainty and will bring this item before the ACRS at its Septem-ber meeting.
Mr. Reed said that T/H research should be focused on "real worl d" plant transients, especially for B&W plant designs, i
_______-________________-__-__A
327TH ACRS MEETING MINUTES 14 Referring to a particular set of experiments, discussed at the June 18 meeting, he indicated concern that the real problems (e.g., vapor block) are not being addressed.
Mr. Ward said further tests are planned and that Mr. Reed should press for this concern to be addressed.
1 3.
Advanced Reactor Designs
[ Note:
M. M. El-Zeftawy was the Designated Federal Official for this portion of the meeting.]
Mr. Ward briefed the Comittee regarding the three DOE-sponsored advanced reactor designs (MHTGR, PRISM, and SAFR).
The NRC Staff is currently reviewing the three designs at the conceptual stage.
The Modular High Temperature Gas-Cooled Reactor (MHTGR) is a 350 MWt reactor with a prismatic fuel concept. The plant uses four modular, steel vessel reactors in the side-by-side configuration, and supplies steam to two turbine generators.
The net plant electrical output is 558 MWe.
Each reactor module is housed in a vertical cylindrical concrete enclosure that is fully embedded and below grade.
The nuclear island portion consists of four reactor enclosures and adjacent structures that house fuel handling, helium processing, and other essential reactor service systems.
A comon control room is used to operate all four reactors and the turbine plant.
The design has no containment.
The design utilizes active systems for normal decay heat removal and reactor shutdown.
Passive means are provided as backup.
The Power Reactor Inherently Safe Module (PRISM) is a 425 MWt power reactor designed by General Electric.
The plant uses nine PRISM reactors.
The plant's combined power is 1245 MWe.
Each module is a pool type 1.MFBR design with its own interme-diate heat transport system and steam generator system.
It has a homogeneous core and metallic fuel design.
It is capable of breeding.
The small size of each module facili-tates the use of passive inherent self-shutdown and shutdown heat removal features.
The B0P is completely disconnected from the primary loop safety considerations.
The Sodium Advanced Fast Reactor (SAFR) is being designed by Rockwell International and it is a liquid metal producing 350 MWe per module.
Each site will have four 350 MWe modules.
Each SAFR unit is a pool-type design with passive decay heat removal.
The core design can accomodate either an oxide or metal fuel.
Each concept will use a building block approach with discrete increments of power generation called power packs.
The SAFR plant will be designed to be commercially
l l
327TH ACRS MEETING MINUTES 15 competitive with coal and LWR Plants by the year 2000 and beyond.
Mr. Ward indicated that the purpose-of the Subcommittee meeting (heldonJune 17,1987) was to familiarize the members with the DOE concepts and the NRC Staff review plans.
There is no ACRS letter or action required at this time.
As a result of the members' discussion, Mr. Ward expressed some concern regarding how the industry would maintain balance between prevention and mitigation for severe accidents and, specifically, how to apply it to the advanced designs.
Mr.
Ebersole commented that the severe accident policy is condi-tioned to the existing LWRs.
Mr. Ward indicated that GE, in their design of the PRISM concept, has made changes in the direction of simplification, inherent safety and maintainability with emphasis on breeding, which is quite different from what DOE claims to emphasize.
Dr. Siess indicated that the ACRS might get interested in the near future in the spent fuel disposal issue for advanced reactors.
[Dr. Steindler agreed that the spent fuel disposal issue is a serious one.]
Mr. Ward commented that for the advanced reactor concepts the inherent favorable response characteristics depend on the detailed design of the core, internals, and surrounding structures.
The detailed design has to be constructed and fabricated to exact specifications to maintain these charac-terf stics without modifications over 60 years (lifetime of the plant).
Yet, the applicant does not seem concerned about the effect of aging on materials, creeping, modifications, etc.
B.
Reports, Letters, and Memoranda (0 pen) 1.
Coments on Draft NUREG-1150, " Reactor Risk Reference Docu-ment,"
In a report to Chainnan Zech dated July 15, 1987, the Conunit-tee commented on this draft report in detail.
Certain ques-tions were raised and it was recommended that the Office of Nuclear Reactor Regulation begin an early examination of this report, both to apply its insights and to guide its further development.
2.
Coments on the Integrated Safety Assessment Proncam (ISAP)
In a report to Chairman Zech dated July 15, 1987, the Commit-tee commented favorably on ISAP and recommended that it be
327TH ACRS MEETING MINUTES 16 i
extended to other plants.
The Comittee made other detailed recommendations and requested that they have a further oppor-tunity to review the proposed future uses of ISAP.
3.
Coments on Research into Continuous Containment Leakage i
Monitoring I
In a report to Chairman Zech dated July 16, 1987, the Commit-3 tee comented that continuously monitoring the leakage from 1
containment may reduce the risk to operators and the public in severe accident situations and recomended that an investiga-i tion of the concept be initiated.
4 Comments on Improved Safety for Future LWRs 1
In a report to Chairman Zech dated July 15, 1987, in response i
to an inquiry by the Chairman, the Comittee suggested that plants designed, licensed, and constructed in the Federal Republic of Germany under the KONVOI process contain most of the features recommended in an ACRS letter of January 15,
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l S.
Comments on the Embrittlement of Structural Steel Dr. Shewmon discussed the embrittlement of structural steel based on surveillance samples used in the pressure vessel of the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory.
The samples have shown that the nil-ductility I
transition. temperature (NDT) of the material irradiated slowly I
at 120 F can rise significantly more rapidly with exposure to t
fast neutrons than would be expected from the available l
experimental work obtained in test reactors.
This appears to be due to:
- 1) a flux rate effect (a lower fast neutron flux embrittles more than the same fluence accumulated at a much higher flux in test reactors), and 2) the difference in l
temperature (550*F for comercial reactor pressure vessels versus 120*F for the HFIR).
This phenomenon suggests that steel structures outside the pressure vessel in comercial nuclear power plants may have embrittled where such behavior was not expected.
The NRC was asked to investigate this phenorrenon and take appropriate action.
The Comittee commented on this matter in a letter to the Executive Director for Operations dated July 15, 1987.
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327TH'ACRS MEETING' MINUTES 17 6.
Comments on Licensee Events Reports Pertaining to Control Room Habitability In a letter to the Executive Director for Operations dated July 15, 1987, the Comittee noted the large number. (over 500) of LERs involving control room habitability in the years 1984
. through 1986.
It was noted - that a large fraction involved malfunctions of air monitoring equipment.
These data suggest a problem with the reliability of such monitors.
The Comit-tee suggested several approaches which might be useful..
7.
Comments on International Cooperation on Research Related to Radiation Protection In a letter to the Executive Director for Operations dated July 15, 1987, the Comittee comented - favorably on the efforts of the Office of Nuclear' Regulatory Research to develop an international cooperative effort for the coordina-tion. of research on the biological effects and control of ionizing radiation.
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Other Committee Conclusions (0 pen / Closed) l 1.
New Members (Closed)
The Committee agreed to. start the search for new members. The Nominating Comittee was charged to prepare a slate 'of poten-tial candidates for the existing vacancy.
It was suggested that the Comittee consider generalists, earth scientists and former nuclear power plant operating personnel.
2.
Safety Research Program (0 pen)
[ Note:
S. Duraiswamy was the Designated Federal Official for this portion of the meeting.]
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Siess said that in our January 14, 1987 report to Congressman Udall, and also in our April 13, 1987' report to i
NRC Chairman Zech,(we proposed to provide more focused react safety research RSR) reports devoted to fewer research issues.
Recently, Dr. Kerr solicited the opinion of Dr. Siess on important research topics on which we can provide coments to the Congress and to the Commission.
In response. Dr. Siess prepared a memorandum, dated June 17, 1987,. to Dr. Kerr in which he identified the following as. potential ' candidates for RSR letters:
Thennal-hydraulic phenomena Human factors Severe accidents Water hammer 1.9...
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327TH ACRS MEETING MINUTES 18 Fire protection Technical Integration Centers Peer review process Dr. Siess stated that, of the total NRC research budget, the Thermal Hydraulic Research Program has been allocated about 16 percent and the Severe Accident Research Program has been allocated about 25 percent.
Since these two research areas thus consume about 40 percent of the research budget, he believes that we should comment on the appropriateness of the research proposed in these areas.
He said that preparing RSR letters on certain topics should be the responsibility of individual subcommittees.
However, the full Comittee can l
also identify important research issues, and then assign an issue to a subcommittee asking it to review and prepare comments for consideration by the full Comittee.
Dr. Siess said that detailed guidelines for writing RSR letters are included in his memorandum dated July 4,.1984.
He solicited coments on the suggested guidelines -contained in his memorandum. The Comittee members did not raise any objection to those guidelines.
Dr. Siess said that RSR letters should be audressed to the Comission, should include comments on the omission of certain important research topics, on deficiencies in the proposed research, and on achievements.
Dr. Siess stated that recently he heard that the Office of Nuclear Regulatory Research (RES) is supposed to tell the Office of Nuclear Reactor Regulation (NRR) about how to use j
the research results.
He is not sure whether RES can do this task effectively.
Dr. Kerr suggested that we should probably invite Dr.
T.
Murley, Director of NRR, to tell us how NRR plans to function under the new organization. We should also ask him to comment on the interaction between RES and NRR.
Dr. Siess suggested that sometime in the future we may want to address the role of NRC in performing safety research and its basic research program.
Dr. Kerr suggested that cognizant subcommittees look for potential candidates for RSR letters.
Mr. Michelson said that the Auxiliary Systems Subcommittee plans plans to review the status and preliminary results of the Fire Risk Scoping Study on the need for future fire pro-tection research that is being performed by the Sandia Nation-al Laboratories. After that review, the Subcommittee plans to
e 327TH ACRS MEET!NG MINUTES 19 prepare a letter, addressing the adequacy of the direction of i
the Scoping Study and also including coments and recommendations on the need for fire protection research.
Dr. 'Meeller said that the Waste Management Subcommittee plans to look at NUREG-1245, "Research Program Plan for High-Level Waste," and provide comments on the plan proposed.
3.
Emergency Planning (0 pen)
The Comittee decided to hear presentations during the August or September ACRS meeting by F. Rowsome and H. Spector on the use of sheltering versus evacuation in emergency planning.
4.
Beaver Valley Power Station (0 pen)
The Comittee agreed that it sees no need for ACRS review of evaluations by the licensee of the seismic capability of emergency power supplies and small equipment that are part of decay heat removal systems at the Beaver Valley Power Station or the potential for steam generator overfilling per ACRS coments in its report of November 13,.1974 5.
SouthTexasPmject(0 pen)
The Comittee agreed to defer consideration of further review of the resolution of issues at the South Texas Project until.
the ACRS receives the pertinent supplement to. the SER and receives a recommendation from Dr. Mark who-is chairman of the South Texas Project Subcommittee.
6, Planning Comittee Meeting (0 pen)
The Comittee raised no objection to the proposal for the Planning Comittee to meet October 22-24, 1987 to discuss the ACRS's role.
Mr.
Fraley requested members to identify specific additional agenda items and possible sites for the meeting.
D.
Future Activities (0 pen) 1.
Future Agenda The Comittee agreed on the tentative agenda for the 328th meeting, August 6-8, 1987, as shown in Appendix II.
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4 327TH ACRS MEETING MINUTES 20 1
2.
Future Subcommittee Activities A schedule of future subcommittee activities was distributed to members (see Appendix III).
The 327th ACRS meeting was adjourned at 12:40 p.m. Saturday, July 11, 1987 i
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APPENDICES MINUTES OF THE 327TH ACRS MEETING JULY 9-11, 1987, WASHINGTON, D.C.
Appendix I List of Attendees l
Appendix II Future Agenda l
Appendix III Future Subconnittee Activities i
Appendix IV Other Documents Received l
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. r APPENDIX I ATTENDEES - 327TH ACRS MEETING Thursday, July 9, 1987 l
l NRC Attendees R. Hernan, NRR C. Thomas, NRR 4
Public Attendees
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E. Fotopoulos B. Nevill, Doub & Muntzing P. F. Riehm, KMC S. Savage, NUS L. Con'nor, DSA i
M. Beaumont, Westinghouse C. Callahan, Associated Press W. B. Scott, PNL Invited Attendees 3
i NORTHEAST UTILITIES j
J. Quinn l
P. A. Blasidli J. H. Bickel M. S. Lederman R. Racich EG&G, ID, INC.
l J. M. Broughton j
P. J. Grant l
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i ATTENDEES - 327TH ACRS MEETING Friday, July 10, 1987 NRC ATTENDEES C. M. Trammell, NRR T. M. Novak K. Black PUBLIC ATTENDEES C. Le,we, NUS T. Ray, SERCH Licensing L. Connor, DSA J. Barker, Press R. Borsum, Babcock & Wilcox M. Beaumont, Westinghouse P. Kelley, Newhouse News B. Ham, USHR D. L. Neal, General Electric A. C. Schafer, URA D. Hampshire, Pacific Gas & Electric M. Simons B. Neville, Doub & Muntzing T. Imbi, JARI Y. Noguchi, Chubu EPC J. Peterson, Chubu Electric Power Company 1
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327TH ACRS MEETING MINUTES APPENDIX II FUTURE AGENDA August 6-8, 1987 Meeting with NRC Commissioners (D0/ DAW /RPS) Estimated time 2 hrs. -
Discuss ACRS report regarding implementation of NRC Safety Goal.
Human Factors (FJR/HA)
Es.timated time 21 hrs.
Discuss SECY-87-101, Degree Requirements f'or Senior Reactor Operators j
Emercency Planning (DWM/EGIf Estimated time 21 hrs. - Discuss pro-l po.sec ACR5 recommendations regarding changes in emergency planning.
TVA Nuclear Power Plant Performance (CJW/RPS) Estimated time 11 hrs. - Discuss proposed TVA corporate Nuclear Performance Plan and propo' sed restart of TVA nuclear plants.
Nuclear Plant Auxiliary Systems (CYM/SD) Estimated time 2 hrs. -
Discuss safety research for fire protection and systems interactions.
4 Regulatory Practices in Italy _ (ions from meeting with the Italian WK et al./TGM) Estimated time 3/4 hr. - Discuss member's observat Technical Comittee for Reactor Safety.
ACRS Subcommittee Activities Waste Management (DWM/0SM) Estimated time 3/4 hr. - Report of Subcommittee visit to. Nevada Test Site, including proposed Yucca Mountain repository site and University of Arizona hydrologic test site.
Nuclear Plant Incidents and Events (JCE/HA) Estimated time 21 hrs. - Be briefed and discuss recent operating events and incidents.
ACRS Future Activities (WK/RFF/MWL) Estimated time i hr. -
Discuss anticipated subcommittee / full Comittee activities.
New Members (HWL/NSL) Estimated time 3/4 hr.
Discuss candidates for appointment to the Comittee.
GDC-4 Environmental and Missile Damage (PGS/EGI) Estimated time li hrs. - AGR5 comments requested regarding proposed revision of GDC-4 to take leak-before-break into account.
Subcommittee meeting on July 24,1987.
Westinghouse Standard Plant (SP-90) (DAW /MME) Estimated time 3/4 hr. - Status report by the Staff regarding proposed review of this project.
Fire Protection (CYM/SD) Estimated time 2 hrs. - Scoping studies for research needs in this area.
APPENDIX II
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327TH ACRS Meeting September 10-12, 1987 l
Renewal of Nuclear Power Plant Licenses (CJW/RKM)
TechnicalSpecificationsImi[rovements(CYM/EGI)
License Requirements for Storage of Spent Fuel (CPS /0SM) 1 ECCS Evaluation Models (DAW /PAB)
I Generic Issue 23, Reactor Coolant Pump Seal Failure (DAW /PAB)
Qualification of Nuclear Power Plant Components (CJW/RKM)
Improved Safety Features for Future LWRs (CJW/RKM) - Response-to Chairman Zech's request (SRM dated - 4/22/87) regarding the cost, effectiveness, etc., of proposed systems.
October 8-10, 1987 Unresolved Safety Issue A-48, H Condenser Containments (JCM/MDH) ydrogen Control for BWRs and Ice B&WWaterReactorDesignAssessment.(CJW/RKM) i Effectiveness of NRC Programs which Address USIs and Generic Issues I
(CPS /SD)
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ACRS SUBCOMMITTEE MEETINGS Auxiliary Systems, July 23, 1987, 1717 H Street, NW, Washinc ton, DC (Duraiswamy), 8:30 A.M., Room 1046. The Subcommittee will ciscuss with the NRC research staff and the personnel from the Sandia National Laboratories the progress of the "Scopino Study" being performed by the Sandia National Laboratories for NRC on the need for future research in the fire protection area. Attendance by the following is a.nticipated, and reservations have been made at the hotels indrated for tfie night of July 22:
J Mr. Michelson DAYS INN Mr. Reed DAYS INN Mr. Ebersole DAYS INN Mr: Wylie DAYS INN Metal Components, July 24, 1987, 1717 H' Street,NW, Washington,DC(Igne),
8:30 A.M., Room 1046. The Subcommittee will review GDC-4 Amendment (leak-befnre break rule), research programs on dosimetry, irradiation effe. cts on pressure vessel materials (Regulatory) Guide.1.99, Revision 2, and other matters (e.g., drywell shell corrosion. Attendance by the-following is anticipated, and reservations have been made at the hotels indicated for the night of July 23:
Dr. Shewmon NONE Mr. Ward NONE Mr. Michelson DAYS INN IA Rodabaugh ANTHONY
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Thennal Hydraulic Phenomena, August 4, 1987, 17 U H Street, NW Washingtnn, DC (Boehnert), 8:30 A.M., Room 1046. The Subcommittee will: (1) review Te'velopment of Uncertainty Methodology for best estimate ECCS Codes, (2) review Status of the Generic Issue addressing Steam Generator / Steam Line l
Overfill Issues, (3) discuss the Status of the Water Hammer Issue, (4) discuss a potential issue regarding long-term core cooling given a LOCA,-
and (5) discuss proposed review of the NRC-RES themal hydraulic research Program. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of August 3:
Mr. Ward ANTHONY Mr. Wylie DAYS INN Mr. Ebersole DAYS INN Dr. Catton DUPONT PLAZA Dr. Kerr (tent.)
LOMBARDY Dr. Schrock ANTHONY Dr. Mark LOMBARDY Dr. Sullivan NONE Mr. Michelson DAYS INN Dr. Tien ANTHONY Mr. Reed DAYS INN Decay Heat Removal Systems, August 5, 1987, 1717 H Street, NW, Washingto h DC (Boehnert). The Subcommittee will review the resolution status for:
TI)GI23: "RCP Seal Failure", (2) GI 93: " Steam Binding of AFW Pumps, and (3)GI124: "AFW System Reliability." Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of August 4:
i Mr. Ward ANTHONY Mr. Wylie DAYS INN Mr. Ebersole DAYS INN Dr. Catton DUPONT PLAZA Mr. Michelson DAYS INN Mr. Davis HOLIDAY INN Mr. Reed DAYS INN APPENDIX III t
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. 328th ACRS Meeting, August 6-8,1987, !!ashington, DC, Room 1046.
Waste Management, August 17-19, 1987, 1717 H Street, NW, Washinoton, DC, (Merrill), 8:30 A.M., Reom 1046. The Sube'ommittee will review several pertinent HLW, LLW, and related research with the NMSS and RES Staffs.
Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the nights.of August 16, 17 and 18:
Dr. Moeller LOMBARDY
. Dr. Shewmon NONE Dr. Mark LOMBARDY Dr. Steindler NONE Dr. Penick NONE Auxiliary Systens, August 18, 1987,2717 H Street, NW, Washincton, DC (Duraiswamy), 8:30 A.M., Poom 1167. The Subcommittee will discuss the heating, ventilation, and air conditioning (HVAC) system malfunctions In addition, it will discuss problems and their impact on safety systems.
associated with instrument air systems, AEOD findings concerning the instrumert air system malfunctions and its recommendations to alleviate Attendance by the following is anticipated, and reservations this problem.
have been nde at the hotels indhated for the night of August 17:
Mr. Michelson DAYS INN Mr. Reed DAYS INN Mr. Ebersole DAYS INN Mr. Wylie DAYS INN Regional and I&E Programs, August 28, 1987, Region V, 1450 Maria Lane, The Subcorm11ttee will Suite 210, Walnut Creek, CA (Boehnert), 8:30 A.M.
review the activities under the control of the Region V Office. Lodging will be announced later. Attendance by the following is anticipated:
Mr. Ward (tent.)
Dr. Remick Mr. Wylie Mr. Michelson Dr. Moeller Mr. Reed Future LWR Desians, September 8, 1987, 1717 H Street, NW, Washington, DC The Subcommittee will discuss its reply IMajor), 1:00 P.M., Room 1046.
to the 4/22/87 Staff Requirements Memorandum regarding the feasibility, benefit, and cost effectiveness of selected and combined systers as recom-Attendanc? by the mended in the ACRS letter of 1/15/87 on Improve LWRs.
following is anticipated, and reservations have been made at the hotels indicated for the night of September 7:
Mr. Wylie DAYS INN Mr. Reed DAYS INN Mr. Michelson DAYS INN Dr. Siess ANTHONY Dr. Okrent ANTHONY
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g a Generic Items, September 9, 1987, 1717 H Street, NW, Washington, DC (Duraiswamy), 8:30 A.M., Room 1046. The Subcommittee will continue the discussion on the effectiveness of the programs that address generic issues and USIs. Also, it will discuss with selected licensees the contribution to plant safety resulting from the implementation of the resolved generic issues and USIs. Attendance by the following is anticipated, and reservations have been,Jnade at the hotels indicated for the night of September 8:
Dr. Siess ANTHONY Dr. Moeller LOMBARDY Mr. Ebersole DAYS INN Dr,' Retnick NONE Hr. Michelson DAYS INN
,Mr. Wylie DAYS INN l
329th ACRS Meeting, September 10-12, 1987 Washington, DC, Room 1046.
330th ACRS Meeting, October 8-10, 1987, Washington, DC, Room 1046.
Joint Waste Manag'ement and Quality and Quality Assurance, October 16, 1987, 1717 H Street, NW, Washington, DC (Merrill/Igne), 8:30 A.M., Room 1046.
The Subecmmittees will review OA Experience in Readiness Reviews as applied to nuclear power plants, HLW geologic repositories, and monitored retrievable storage (MRS) facilities. Lodging will be announced later.
Attendance by the following is anticipated:
Dr. Moeller Dr. Shewmon Mr. Reed Dr. Siess Dr. Kerr Mr. Wylie Dr. Remick Decay Heat Removal Systems, Date to be determined (August / September),
l Washington, DC (Boehnert). The Subconnittee will continue its review of the hRR Resolution Position for USI A-45. Attendance by the following is anticipated:
Mr. Ward Mr. Wylie Mr. Ebersole Dr. Catton Mr. Michelson Mr. Davis Mr. Reed Babcock & Wilcox Reactor Plants, Date to be determined (late sunner/early fall), Washington, DC (Major). The Subcommittee will continue its review
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of the long-term safety review of B&W reactors. This effort was begun during the summer of 1986; initial Committee connents offered on July 16, 1986 in a letter to V. Stello, EDO. Attendance by the following is anticipated:
Mr. Wylie Mr. Michelson Mr. Ebersole Dr. Okrent Dr. Kerr Mr. Reed Dr. Lewis Mr. Ward i
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/ t Auxiliary Systems, Date to be determined (September), Washington, DC i
i (Duraiswamy).
The Subcommittee will discuss the criteria used by the utilities to design Chilled Water Systems, associated regulatory requirements, and the criteria being',used by the NRC Staff to review i
the Chilled Water System' design. Attendance by the following is I
l anticipated:
)
Mr. Michelson
- Mr'. Reed Mr. Ebersole Dr. Shewmon Dr. Moeller i
Mr. Wylie I
Thennal Hydraulic Phenomena, Date to be detennined (September / October),
Washington, DC (Boehnert).
The Subcommittee will review: (1) the final version of revised ECCS Rule, and (2) the status of RES-proposed new integral test. facility. Attendance by the following is anticipated:
Mr. Michelson Dr. Catton Mr. Ebersole Dr. Schrock Dr. Kerr Mr. Sullivan Mr. Ward Dr. Tien Mr. Wylie 1
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GE Reactors (ABWP). Date to be determined (September / October), Washington, DC Fe'ge(Ma jor). The Subcommittee will review the status of activities rding the General Electric Advanced Boiling Water Reactor. Attendance by the following is anticipated:
Dr. Okrent Dr. Remick Dr. Kerr Dr. Shewmon Mr. Ebersole
'Mr. Ward i
Mr. Michelson Mr. Wylie j
Standardization of Nuclear Facilities, Date to be determined (October),
washington, DC (Alderman).
The Subcommittee will review the Staff SER and Chapter I of the EPRI Requirements document. Chapter II may also be dis-cussed.
Attendance by the following is anticipated:
Mr. Wylie Mr. Reed Mr. Michelson Dr. Siess i
l Diablo Canyon, Date to be determined (late November /early December),
Location to be detennined (Igne).
The Subcommittee will review the status
'i of the Diablo Canyon Long-Term Seismic Program. Attendance by the follow-ing is anticipated:
Dr. Siess Mr. Ebersole Dr. Page Dr. Maxwell Dr. Kerr Dr. Lewis Dr. G. Thompson Dr. Trifunac Dr. Moeller Dr. Scavuzzo
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., Joint Seabrook/ Occupational 8 Environmental Protection Systems / Severe Acci-dents, Date to be determined, Washington, DC (Igne/ Houston / Major). The Subcommittees will review Seabrook Emergency Planning and other related matters.
Attendance by the following is anticipated:
1 Dr. Kerr Mr. Reed Dr. Lewis Dr. Rpmick i
Dr. Mark Dr. Shewmon Mr. Michelson
.Dr. Siess Dr. Moeller Mr. Wylie Dr. Okrent Dr. Catton (tent.)
Seabrook Unit 1. Date to be determined, Washington, DC (Major). The Subcommittee wil) review the appifcation for a full power operating license for Seabrook Unit 1.
Attendance by the following is anticipated:
Dr. Kerr Dr. Moeller Dr. Lewis Mr. Michelson l
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APPENDIX IV ADDITIONAL DOCUMENTS PROVIDED FOR ACRS USE 327TH ACRS MEETING, JULY 9-11, 1987 Report of ACRS Consultant John Lep concerning comments on ACRS Subcom-mittees on Severe Accidents /PRA joint meeting of June 3, 1987 Memo For:
D. Okrant from A. Tabatabai,
Subject:
KONV0I Plants 4
Memo For:
R. Fraley from J. Hoyl,e,
Subject:
Staff Requirements -
Periodic Meeting with ACRS i
Memo Fo'r E. Jordan, from T. Murley,
Subject:
NRR Plans for Response to
. Lessons Learned from Diablo Canyon Loss of RHR Event of April 10, 1987 and Related Events, dated 6/2/87 l
AE00 Report " Loss of Decay Heat Removal Function at Pressurized Water Reactors with Partially Drained Reactor Coolant Systems" Memo for ACRS Members, from R. F. Fraley,
Subject:
Future Activities -
Items Proposed for the 328th ACRS Meeting (August 6-8,1987).
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Other Documents Slides:
" Reactor Risk Reference Document, NUREG-1150" by NRC NRC Staff presentation
" Integrated Safety Assessment Program" by NRC Staff. presentation
" Integrated Safety Assessment Program Presentation to ACRS" by Northeast Utilities TMI-2 Core Removal - by DOE Staff
" Review of Operating Events by AE0D Division of Safety Programs" by NRC Staff Documents:
Official Use Only Memo For: ACRS Members from Richard Major,
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Subject:
GRS Response to ACRS Questions Concerning Improved LWRs l-(
Memo For: ACRS Members, From: Richard Major
Subject:
"KWU t
Report with Bearing on ACRS Consideration of Improved LWRs" l
NUREG-1269, Loss of Residual Heat Removal System
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