ML20043D834
| ML20043D834 | |
| Person / Time | |
|---|---|
| Issue date: | 12/15/1988 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2614, NUDOCS 9006110211 | |
| Download: ML20043D834 (54) | |
Text
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TABLE OF CONTENTS-MINUTES OF THE 344TH AC64 NEETING
{
DECEMBER 15-16, 1988
. Chairman,'s Re' ort (0 pen)'.............................................
1 I.
p II.
SodiumAdvancedFastReactor(SAFR)(0 pen)..........................
1 1
~
III. ContainmentSystems(0 pen)...........................................
5 i
IV.
Equipment Qualification (EQ)-Risk Scoping Study (0 pen)..............
10 V.
NRC Quantitative Safety Goals (0 pen)................................
11 I
VI.
Meeting with Director, NRC Office of Nuclear Regulatory-Research (0 pen)......................................................=14 VII. Review of NRC/RTES Code Scaling Applicability and Uncertainty-(CSAU)
Hethodology(0 pen)...................................................'18 VIII.-Meeting with Director, Office of Governmental and Public A f f a i r s ( 0 pe n ).....................................................
21 IX. Operator Requalification (0 pen)......................................
23 X.
Nuclear Power Plant Operations - (0 pen)................................ 25 XI. ExecutiveSessions(0 pen).............................................27 A.
Subcommittee Reports.............................................
27 i
B.
Reports, Letters and Memoranda...................................
28 1.
EquipmentQualificaten(EO)-RiskScopingStudy.............. 28 C.
Other Committee Conclusions......................................
28 1.
Vogtle Electric Generating P1 ant.............................
28..
2.
Implementation of, Severe Accident Policy for Future Reactors.x28 3.
Application of Leak-Before-Break Technology..................v 28 4.
B W S te am Gene ra to rs......................................... 29 D.
Future Activities.................................................
29 I
- 1. ' Future Agenda................................................
29
- 2.. Future Subcommi ttee Acti vi ti es............................... - ~ 29 Supplement (OFFICIALUSEONLY):
Election of Officers / Appointment of New
//67 grpf)
Members Figutt 1 - p. 20a Figure 2 - p. 26a
-DESIGNATED ORIGINAL
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corung 3rd/pp 6
900611o211 es121c, o
POR ACRS
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2614 PDC
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11 3
. APPENDICES MINUTES OF THE 344TH ACRS MEETING l
DECEMBER 15-16. 1988!
l t
1 I.
Attendees II.
Future Agenda III.
Future Subcommittee Activities i
IV.
Other Documents Received' i
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l'edarel Rosieln / W1. Es, No, 2.39 / Tuesday. December 13, 1988 / Notloos 3.-
80141 rw me Scleet Rasvletory comminton.
Rescorch (Open)-Briefing and 8- ~
which would repreunt o' clearly g
A. ceive.
discuulon of itema of mutual interest.
unwortantedInvaelon of personal oitector, hojwt Daectorole -IV. Divislon of.
- 30-!!Do Noon: Enetyency Core.
Asocier hWerJe--tII. IV, VeadSpecial Coollag Systems (Open)--Cominente by pglvecy lg US.C s6sb(c)(6)) to discuee AguloidoA ACRS subcommittee chairman regarding Information provided to confidence by a Anbete alpcee/%cicerAeoctor w
proposed NRC Code Scaling...
forelgn source l6 US.C. 662b(c)(4)). and c
FR Doe, e6-assas Nd 13-tMs. t45 eml Applicab!!1ty and Uncertalnty (CSAU) to discuse ProprietaryInformation appucable to saattere bdng considered enAme caos reewun F. valuation Methodol toposed for use with best. estimate
~ evaluation-16 UA.C. 662b(c)(4)).
modele. Meeting with representative.'of
%rinfamaum toPg#
Advisory Comminee on Reactor NRCStaK.
to be disomud, who en k
Safeguards;Revisedleoeun0 Agende JA.StM230PNs Astum ACAS
- has been omelled or moneduled in occadance wHh the pu of Acf/v/t/se (Open)-Discuse anticipated Channaan's ruling on requeste for the subcommittee activities and discuse OPportmity to preont oral statemente -
Secueno as and latiof es tomic F.norgy Act (42 UAC. 2030. 2232b), the items proposed for conalderation by the and the time allotted ce, be obtainedby Advloory Committee on Reector full Committee, including Split of a prepaid telephone call to the ACRS Safeguards willhold a meeung on responalbilities between ACRS/ACNW.
Execouve Director. Mr. Raymond F.
December 16-17,1988, in Room p-114*
2:#-J:00P.Afa Meelling with D/Mcfor, Fraley (telephone 901/492-4049).
79a0 Norfolk Avenue,Bethesda Md.
A@ffaire GominmentalandP@
between 8:18 A.M. and 500 p.m.
Notice of th!: insettag was published in
/ Closed)-Brie a rdt e
ee
,eE h FederalRegleter on December 1-Refsted M ISSR Encha Safe -
fmestion. -
r/ ' ',,. Ma C. Hope, @m h 1998.
at.
(Portione ot thle seselon wdi belolooed AdWeery0m". "O#aon -
Miwoday,Decem6esa saga Roova A.
as necomagy to disome.by a fossign saformanen PR Da swa apa4sma se aml-
- it FaspNorfo& Avenue Delheede, AfD providedin conhdence eaAsse coes meme 2"!.ss**@e2*a'r ".7 ". '_*" ^e"gh,gg py,b,,eeee 8:atHH8 AM.tChoirmon1r Cammente
- mger n jg, oj, a
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'a**' "a *** *a* **
ne ini.e ca m_Am m eema-d
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...m.
D, ee,,e M.,,,eeee coi.nti Fast Repeder(SAFR)(Open)-Repod of
,equuncaum m.uda y(D,a a *o oow a sheltman regardtag Examinar Standard 901).-
ACRS Finseng of NogagnlReentimpact e
the sowiewof lofstandarcised 4,43.g;3p pp,,yyej,af pp op,,qfone:
s eol6ar pie d aedonMed mposimb The U.S.NeoleerRegulator{la a
~mNitC)and DOE.gpe63. Meeting '
- ~. '
Naclear Stationcoes power oedl@.. J Assonledon(the -'t:* " :" 1,leseenc
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e
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g transtea,,,, p,,,
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+,-.kee sotws. neck sr. sens noen5. '
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r shainiosa q pqmed' y ff4 FalpNorfo&s m.q% add
'o,f b e
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withaP'***twFirSA*. eeting, aagego A.Agagk@ie!fACRI
[ tion, Unite 1 and 3 loca talle&ver n.=
eppmpdate-ome= for er s=(aos.d>-.
- 00480PRJ Dlecuselon and election of ACR8 Comi,, penne,ivania.
chairman and vice chairman for CY 1969 tal Asessement Qualiflootton.Alsh Study and Member.staarge for ACRS n)-Continue meetlag discussion of poemt Quatinentloosu,a,a Scoping Planning Subcommitta and report Identiflootton ofPrupponedAction regarding A The " eed amendments would
(
t Membm. ppointment of New acied'e [contenamat teolaneet opdne.
8 aci4uPer- '.' 9mm pue ee ganwinb dosedto'discia. and wetsht leeded check valmanot Informatloa the releas,e of ivbich would(. subjeel to Type C tesd M '.
conetttute a des unwarranted test regelremente of SpeedRostfee 4A2.1.
peoposed NRC _+,
tavadce of petvaa Noone t t g.): c ACRS ' otheegpte.
by oomtalementg
- r. n
?,
esp.
Agut(Open)-Diecu;e pop;ose(d ACRS
%_ x1forBWR9a&1.
+.
regdeBon and the kal o
spoets.
SpedRce6oniff% m. sot affected 5,7_
oontelrunent.
with -.1 C Staff. asp,,".
r representatives of J.80420PRa ACASSv6commillee bw amadments.
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,g,g,,,, g, CeoAs(h @, ;_,
NI
- 1. Obmee
' ",, InternettonalSeminaronhlityand 4.84 RJ tdlag wahDegeeneut 4
=
We apphoeuon deled, une as.
T ed@ @' g y"% L ## died W -
^
ley. Meeung with repeesentatives of (Open ne ud changes to the TS are ag)-Discues one ofinternet/
othe N..RC 8%es a J,os,,nnece,thelift testaosabeshown etace L-M*
'."*F88h.ofA,mnue.8dAced.,.
Subeecuon told) P.I.98-468 that H le involved. Performin6 manecessary teete ea,yfo, u,d,ee A
hacessary to clam gf des-increases resource tures afofries[
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.'Fedo'ral ReglUer' / Vof. 63.'No. 235
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qualifications of nominees proposed for' -
r-B:45 o.m-1A15 oa: Operator i
election as Committee officere for u
- 6. - We principal alternative would be to Requal/ficot/on progtom (Open)--
CalendarYear1980.,
g '-
4.- deny the requested amendment.%Is 11riefing tegarding leuono learned from
%Is seulon willbe dosedi'YEcun 7. woual not reduce environmental implementsUon of revised operefor o
' W. resultinreducedoperationalflexibility,' qualification methodology,[ Draft -
Information the release of wh8ch would-
, 7.i. impacts of plant operation and would Examiner Standard 001).
represent a clearly unwarranled. b ' '
inveelon cf personalpelvscy,
~
. sax on.-22:Mpa.: Sodium f, Alt:rnouve Use of Resowces%1s action does notinvolve h use of 'AdvancedFast1:soc ACAS Asports (Open)-Discuss Open}-Revicw RJ5 cm-!f.50Nooli Preparot/on of' L
of proposed standardise
-J~
's, any twources not previously considered nuclear plant. Representatives of the proposa nports to NRC regarding 4 *
, - in the 'Tinal Environmental Statement e Department of Energy and the NRC Staff lesu Related to the Operation of the Fort '
ygp,a.27;gpp. Acas l
w!!! participate, Agencies andPersons Considted quhotfafff hthy
{u t v{#
Calhoun Unit 1", dated August 1978.
m (Open}--Review and comment on NRC.
status of acuvities assigned to cognizant
%e NRC staff reviewed thelicensee'ssponsored Equipment Qualification Risk subcommittees including thermal.
P request and did not consult other Scoping Study including consideration h draulic phenomena.E. Fermi plant Ny egencies or persons.
(37 m $n louveSofoty uji llty a En
[ Finding 3f Nisignificant impact a
k heCosimissionhasdeterminednot* L. Cools (O n)-ReviewandcommentonNRC Staff an f v
,gse(Oimn % oun Y
ta prepare an environmentalimpact '
mentation of C Quantitatin '
~ review and discuselon of 4 a.4Ar.,
ro Q.'
state:r.ent for the proposed license' '* 6.
ety Ooels...,,3, 2.
r kry ph
. '.r.w @.
ares fde coMof ande '
caendment.
Sased upon the forgoing Friday, th=aher 16,1ses parucipetionla ACRS aneetings were environmental assessment, we conclude, g.ga g g,, % gcgg published in the Federal Register'on *;*-
that the proposed actica wiu nothan a Activlues (Open)-Discuss anticipated October 37.1988 (53 FR 43487.in
- and toples accordance with then p uses, om!
significant effect on the quality of the ACRS subcommittee activitby the fuu h"**"'"* "*'"ls with respect to th!e proposed for consideration or written statements may be presented L'
by members of the public,secordings Conunittee For further detai RCO o.m-10.2 as.rNuclearSosty willbe permitted only during those -
~ acuori, see the application for V.mendment dated September 2.1988, as Reseemh (Open) -Briennaby and porties of the muung when a discussion with the Director. Office of transcriptis pVandquedone, i
~)s:pplemented Nowmber 22,1908, whichcre evallable for public inspection at the NuclearRe snaybe asked membereof the Commiselon's public Document Room, regarding puts of Gw ufe res e h Committee,its cons tants. and Staff.,
and 2120 L Street.NW., Washington, DC
[m o interest to the A persons desiring to make oral statements should notify the ACRS 30558 and at the W. Dale Clark 1.1 acy$bre F.xecutin D! meter as far in advance as -
its South tSth Street. Omaha,Ne J&f5 cm-12:1 comment.
eview' e Scaling Applicabiliff ~ practicable so that appropriatt'llN the A
80103.
....k Coolig(
arrangements can be made id"a Deted at Rockvtue, Maryland, this soth day on repos Uncertainty proposed for use with necessary time during the meeting for er of November,19es.
best. estimate ECCS evaluation models, such statements.Use of still, motion -
For m Nuclear Regulatory Commlulon.
N5pm-2:4 m:US-USSR picture and televlolon cameros during paulW.o'connor.
&choge of oudli Ope:/.
thle routing may be limited to selected Dhoctor. Pro /ect Distomde--IV.
Closed)-Brie rega reement Portl*a' 'I th' ***tlas ** det*rala*d Acil IWyde e/Aeoefor. / ate!//.IV. Fond to exchange safety 4 slated ormation oy the Chairman. Information regarding specderAe/ sets, ofpos e/Nucher Asocear related tq the design, operation, etc. of the time to be.pt aside for this purpoos 4
Aggudsedan.
pit Dec.es-aesso riled 1s e est aes em)
- Portions of this session willbe clEsed may be obtained b'y a prophld telephone nuclear realtine.1.
can to the ACRS Executive Director.Mr.'
e u ms onesra w s.m.
as necessary to discussinformaHon prior to the meeting.
Rappond F.Fraley,bility that the providedin confidence by a foreign
/
In view of the posel schedule for ACRS meetinge may be.
source.
y Advloory Committee on Reactor 2.45pa-4:45 par Contolament adjusted by the Chaltman as necessary Safeguerds; Meeting Agenda Systems (Open)~ Review and comment to faciutate the conduct of the meeting, la accordance with the purposes of on recommendations for containment persons planning to attend should check sections to and 182b.of the Atomie '
performance requirements and specific with the ACRS Executive Directorif H
Energy Act (et U.S.C. 2039,2232b), the
. aspects of the BWRMkIcontainment.
such rescheduling would result in major 4:45 on.4isp.m. ReoeforOpemt/g Advisory Committee on Reactor inconvenience.
and discuselon of essondaTned frorE**^1have deter ga a n,teAng pxper[,,c,(
8g F Sefogeards will hold a aneting on ;g,,
subsection 10(d) Pub' t. 92J463 that it is December 15-17.1988. In Room P-t14.'
792c Norfolk Avenue. Bethesda. Md.
1ower osculation transient at the necessary to close rtionsof this aSalle Nuclear Power Station.
Nopce ellbis meeting ytas published in.% Pdrtions of this session willbe closed meeting as noted
~g I-e
,r 1>,.
the Federal Register on Octolier 30,'1986.
information'the' hEof cli ik' as necessary to discuss Proprietary represent a clearly unwarranted S'
%ursday December 15,1908 Information related to this matter, invaston of personal privacy (5 U.S.C. :
"~ ' Yurday, December 17.190s 652b(c)(6)] to discuseinformation
.M-830 cm445 om: Comments by Sa ACRSChatraron(Open)--The ACRS
&Joam400 om:Selecilon of ACRS prwidedin confidence by a foreign Chaltman will report briefly regarding.
Officen (Closed}-Discuss source (5 U.S.C. 552b(c)(4)]. and to items of current interest
.. n.
5
- 4~
- 44
e y
el
,.49366 roderel Register / Vol. 63, No. 235 / Wednesday, December 7,1988 / Notices l
i discuss Proprietary Information of tests, conducted in 1982,1985 and requirements pending the NRC staffs l
applicable to matters being considele'd,,, 1987, to meet the acceptance crlieria for review of the licensee's amendment (5 U.S.C. f,52(b)(4)).
the "As Found" condition was due to request. In order to complete its teview Further Information regarding ics,
excessive combined leakage from in an expeditious manner, yet allow for g'
.4 to be discussed, whether the meet several contelnment isolation valves, l
e the licensee concluded that the rnost public comment, the NRC is processing has been cancelled or rescheduled, the licensee's amendment proposal on Chairman's ruling on requests for the eIfective approach 19 ellminate the.
u ex! gent basis undet the provisions of p6 opportunity to prese91 oral statementsN exce' 81#1eWagehPto'linplement a,.
and the time allotted can be obtained by Corrective Action Plan (cap) ustng Before (ssuance of the proposed a prepald telephone call to the ACRS guidance given in Informstlon Notice Ehecutive Director, hit. Raymond F.
85-71 dated August 22,19as. In this CAP license amendment, the Commission Fraley (telephone 301/492-4049).
the licensee determ'ned that 33 will have made findings required by the between 8.15 a.m. and 5100 p.m.
contalnment isolation valves, which Atomic F.ncrgy Act of 1954, as amended (the Act) and the Commlulon's Date: December 2,1968 previously were identified as havin8 regulations.
W C. Wyg excenive leakage, should be replaced ne Commluton has made a proposed Adr/sory Comm/ nee Alanctemc.W O//ker.
(21 during the 1988 refueling outage and ll R Doc. so-28122 rded 12-6-88. 645 am) 2 during the 1990 refueling outage).%e determination that the amendment
- Q COM ?"* "
12 valves scheduled to be rep! ced request involves no significant harards i
during the 1990 refueling outage ha ve consideration. Under the Commission's acceptable leakage rates based on the regulations in 10 CFR 50.92, this meana (Docket No.40 3331 test performed during the 1988 refueling that operation of the facilityin accordanca with the proposed James A. Fitzpatrick Nuclear Power e art of the CAP, the licensee h*gn't pro a$1b or Plant;Conalderation ofissuance %f replaced the f(PCI turbine exhaust line n
e in Amendment to Facility Operating manual block valve to the suppression consequences of an accident prevlo sly ucense end Proposed No Signif6 cont charnber(23-HPI-11).TS 4J.Alf and evaluated; or (2) create the possibility of Hazarda Conalderation Determination Section IV.A of10 Cm Part So, a new or different kind of accident from
~
cnd r';-;-C lor Hearing Appendix j require that following an accidentprevioual evaluated;or(3) i
%e U.S. Nuclear Regulatory replacement of a component which is involve a e nificant reduction in a I
"*'8 " 'I' IY' Commisston (the Commission)is hartof theprimarycontainment considering issuance of an amendment undary,either a Type A Type B,or
%ese proposed changes do not T[fected, must be conducted and thepc C LRT, se applicable for the area incman h probaWty or to Facility Operating License No. DPR-consequences of an accident previously 60, issued to the Power Authority of the State of New York (the licensee, for appropriate acceptance criteria met, evaluated.The containment leakage N cperationofJames A.FitzPatti Since an Isolation volume for the rates suumed In the Final Safety
) County,NewYork. Nuclear Power Plant, located in Oswego resulting welds on the primary Analysis Report (FSAR) require that the containment side of the valve could not valves which perform containment l
By application dated November.9, be atlained, the licensee conducted 1005 leolation functions, as well as the 1988, the licensee requested that the radiography and dye penetrant tests on Primary containment itself, exhibit primary containment leak rate test the welds to verify the structural superiorleak rate characteristics When requirements described in Technical Integrity of the welds,in lieu of a Type the licensee found that the limit was A, B, or C test.
Inquently being exceeded, a cap was Specification (TS) Section 45.A.2.a(10)
Based on an evaluation of the initiated.%e cap involved a deta!!ed and Section (J.A1(be amended for the 1988 refuelln6 outage on an emergency licensee's CAP, the alternate testa analysis of the causes for exceeding the 7p basis under the provisions of to Cm
' performed to ensure system Integrity, allowable limit, detennination that the I
60.9t(s)(5).%e appucation stated that and the implementation of an improved primary cause was valve seat leakage, these 75 changes were necessary to
. valve malatene'n' ace pmgram, en identification of the valves which were allow plant startup from the 1968 exemption in the requirements of causing the problems, determination of refueling outage without performing a Section III..A.8(b) and Section IV.A of the best method to correct the problem W
Type A primary containment integrated. Appendix l to to CFR Part So wasissued valves,andimplementationof the lied rete test (It.RT) or a Type A, B, or C to the licensee by letter dated November resulting plan to ensure that the leak
-g 16,1968. The exemption was noticed on limits are not exceeded in the future. It leak rate test (LRT)following November 25,1988 (53 FR 47784).
was determined that over time some of.
L replacement of the high pressure coolant Whenit was recognized that the these valves exhibited gradual injection (ILPCI) system turbine exhaust line manual block valve, as explained licensee had inadvertently failed to -
degradation to the point where their identify that a TS amendment would be combined seat leakage rate, when L
- below, l
Section 4J. Ale (10) of the TS and required in addition to the exemption, added to the leakage rate resulting from
~
the licensee submitted the nece%ary the previous Type A test, caused the l
Section llLA.e(b) of Appendix j to 10 amendment request dated November 9, limit to be exceeded. This resulted in the hj CFR Part 50 require that if two 1968. Based on an evaluation of the determination that many valves needed Fv'
~
consecutive periodicType A tests amendment application (which is to be replaced, some during the 1988 Y'
(it.RTs) fall to meet the acceptance t
virtually identical to the exemption), a refueling outage and other during the f'
l criteria, a Type A test must be temporary walver or compliance from 1990 refueling outage. Allof these valves
%'i '
pstformed at each plant shutdown for the provisions of TS Section 45.A.2.s(to and Section 43.Alf was issued by the ) were tested prior to the end of the
.M; refueling or approximately every to t
outage with satisfactory results. Using O,
oonths. whichever occurs first, until two NRC staff to the licensee by lettet dated this program, the intergrity of the 1
consecutive Type A teste meet the November 18,1968. This allowed plant primary containment has been restored E)j '
i ecceptance criteria.When it was startup from the refueling outage so that it is reasonable to assume that i
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determined that the cause of the fatture. without compliance with these 73 the design lealage rate limits of the
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'o UNITED STATES g
8 NUCLEAR REGULATORY COMMISSION o-
}
ADVlsORY COMMITTEE ON REACTOR SAFEGUARDS 1
0 WASHINGTON, D. C. 20666 4,,,,,
Revised: December 12, 1988 i
SCHEDULE'AND OUTLINE FOR DISCUSSION 344TH ACRS MEETING
-DECEMBER 15-17, 1988 BETHESDA, MARYLAND Thursday, December 15, 1988. Room P-114. 7920 Norfolk Avenue Bethesda, Md.
- 1) 8:30 - 8:45 A.M.
Chairman's-Coments (0p(WK) en) 1.1) Opening remarks 1.2) Items of current interest-(WK/RFF) i
- 2) 8:45- - 10:45 A.M.
SodiumAdvancedFastRe' actor ('SAFR)-
(0 pen 2.1) ) Report of ACRS subcomittee chairman regarding the review-of.
this ty
- plant-(pe of standardized. nuclear initialsession)-(DAW /MME)
(
)
2.2) Meeting with representatives of
~NRC and DOE 10:45 - 11:00 A.M.
BREAK 3)11:00 12:00 Noon ContainmentSystems(0 pen) 3.1) Comments by ACRS subcomittee -
chairman regarding proposed HRC-recommendations for containment performance and improvements for:
BWR Mark I containment (DAW /MOH) 3.2).Meetingwithrepresentativesof.
NRC Staff, as appropriate 12:00 -
1:00 P.M.
LUNCH
- 3) 1:00 2:00 P.M.
ContainmentSystems(0 pen) 3.3) Continut meeting / discussion of o
Containment Systems:
- 4) 2:00 --
4:15 P.M.
Equipment-Qualification-Risk Scoping :
_(3:00-3:15-BREAK)
)
4.1 Report of ACRS subcomittee ls
('~'l:
comment of this Scoping Study? *
(CJW/SD);
4.2) Meeting with representatives of NRC Staff, as appropriate:
.1
I ia i
,I
-l2:-
)
- 5) 4:15 -
6:30 P.M.
NRCQuantitativeSafetyGoals(0 pen)
{
b.1) Remarks by ACRS subcomittee i
chairman regarding proposed NRC i
Staff plan for implementation of j
NRC's Safety Goal Policy 1
(DAW /MDH) 5.2) Meeting with representatives of-the NRC Staff, o appropriate l
Friday. December 16. 1988. Room P-114. 7920 Noyolk Avenue Bethesda Md.
I
- 6) 8:30 -
9:30 A.M.>
Meeting wht Director. NRC Office of-Nuclear Requlatory Research LOpen).
i 6.1) Brie"ing and discussion of items ofmutualinterest(CPS /SD) 7)(9:30 12:00 Noon-Emergency Core Cooling ' Systems (0 pen) 10:0010:15-' BREAK) 7.1) comments by AcR5 subcomittee '
chairman regarding proposed NRC CodeScaling(Applicabilityand Uncertainty CSAU) Evaluation Methodology _ proposed for use
'(
)
with best-estimate ECCS evaluationmodels(DAW /PAB) a 7.2) Meeting with representatives of NRC Staff i
7.3) Discuss ACRS-report to NRC 4
.(tentative)-
8)12:00 12:30 P.M.
FutureACRSActivities(0 pen)'
8.1) Discuss anticipated subcomittee activitiesL(MWL/RFF) 8.2) Discuss items proposed for con-sideration by the full. Comittee-(WK/RFF) 8.3) Split of responsibilities betweenACRS/ACNW-(WK/RFF)-
9 12:30 1:45 P.M.
LUNCH 9)1:45 3:00 P.M.
Meeting with Director. Office of Eovernmental and Public-Affairs (open/ closed) 9.1) Briefing-regarding US-USSR Exchange of Safety-Related Information
(
1
3 l
.1 l (Note: Portions of.this session-
.will be closed as-necessary to discuss information provided.in confidenceby'aforeign-source.)
3:00 3:15 P.M.
BREAK 10)3:15-4:45 P.M.
OperatorRequalificatio^(0 pen) 10.1) Briefing regarding lessons-learned from implementation of revised operator requalifica-l tion methodology (DraftExaminer-Standard-601)_
(FJR/HA)-
11)4:45-6:30 P.M.
NuclearPowerPlantOperatior.s:(0 pen)
[
11.1) Briefing regarding action taken in response to'LaSalle Nuclear Station core power oscillation-transient j
(WK/PAB)'
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Saturday, December 17, 1988, Room P-114, 7920 Norfolk Avenue :Bethesda, Md.
12)8:30 9:00 A.M.
Election of ACRS Officers,for CY;1989 (closed)'
12.1) Discussion and' election of ACRS-chairman and vice-chairman for CY1989(WK/NSL): _.
12.2) Discussion and election of Member-at-Large for ACRS i
PlanningSubcommittee-(WK/NSL) 12.3) Report regerding Appointmen.t of New Members (Note: This session will be closed to l
discuss-information the release of which would constitute +a clearly unwarranted invasion of personal-privacy.)
i 13)9:00 12:00 Noon PreparationofACRSReport'(Open)-
1 13.1)
Discuss-proposed ACRS report on:
l
-13.1-1) Equipmei,c' Qua lifica-tion Risk-Scoping l:
Study (CJW/SD).
t L-12:00 1:00 P.M.
LUNCH
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14)1:00 2:00 P.M.
ACRS Subcomittee Activities i
(0 pen)
L 14.1) Report of the December 1, 1988 l
meeting regarding International-
- Seminar on Quality and Quality Control-(CPS /EGI)'
14.2) Visit to E. Fermi Nuclear Plant (WK/PAB) 15)2:00 3:00 P.M.
ACRSActivities(0 pen).
15.1) Discuss areas of-interest /
activity in W. Kerr memo i
dated 10/4/88 l
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4 k
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O9"m7 h j i i y y *j L(wM. l (j,
0 MINUTES OF THE
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344TH ACRS MEETING L
DECEMBER 15-16, 1988 The Advisory Committee on Reactor Safeguards (ACRS) met on December 15-16, l
1988 at 7920 Norfolk Ave., Bethesda, Md. The purpose of this' meeting was to i
conduct. the discussions and perform the actions described in the attached agenda. The meeting was chaired by Dr. ' Kerr.
J All. of the discussions were held in s open session except: for brief dis-cussions during which the ACRS officers for CY :1989 were elected and the -
appointment of new members was discussed.
1 A transcript of selected portions of the meeting was kept and is available in the NRC Public Document Room. -[ Copies.of the transcript are also' avail-l able for purchase from the Heritage Reporting Corporation, 1220 L St., N.W.,
Washington, D.C.
20005.]
I.
Chairman'sReport(0 pen) 1
[ Note:
Mr. R. F.
Frale portion of the meeting.) y was the Designated Federal Official for this-
~
L Dr. Kerr began the meeting with a brief sunnary of the planned agenda and L
the procedures under which the meeting discussions were being conducted.
Dr. Kerr noted that Mr. Lockard would be, retiring in the near future and i
expressed the Committee's appreciation for Mr. Lockard's excellent work.
II. SodiumAdvancedFastReactor-(SAFR)(0 pen)
[Dr. M. El-Zeftawy was - the Designated Federal Off*,cial-for this portion of themeeting.)
Mr. Ward, Advanced Reactor Designs Subcommittee-Chairman,ibriefed the Committee regarding the-Subcommittee's discussions 'of. the SAFR design.- He' l
indicated that the -NRC Staff has reviewed a-preliminary safety information document (PSID) which was provided by 00E. The Staff's' review is considered a preapplication review for 'the purpose of-providing guidance early;in the design process on the acceptability of the SAFR design.wThe issuance.of the r l-draft safety evaluation report'(SER) does not constitute'.anLapproval of the, s
L SAFR' design.
The. Staff's review has -been performed under -the guidance"of-the Commission's advanced reactor Lsevere accident, safety goal, and stan-dardization policy statements. Mr. Ward indicated that he expects the ACRS will write a report to the Commission on this subject at -the January 11989-ACRS meeting. Mr. Ward also noted that DOE has made the selection between.
SAFR, and Power Reactor Inherently _ Safe cModule -(PRISM) designs and_ has selected the PRISM concept with GE as the-prime contractor.
Mr. Landry, - NRC/RES, indicated that the staffs review is expected to be completed by January 1989 with the CRGR review in Januaryf l989 and reconnene w -
dations to the Commission by end of January 1989.v y
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344TH ACR$ MEETING MINUTES ;
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l The SAFR has been designed by Rockwell International in cooperation with i
Bechtel, Inc. and Combustion Engineering, with Argonne National Laboratory i
providing major analytical and testing support. The SAFR conceatual design i
utilizes one or more indewndent power paks with each power pa c consisting of four modules.
The incividual modules produce 900 MWt (350 MWe).
The i
design consists of a sodium-cooled reactor system that transports heat via the primary coolant through two intermediate coolant loops to two steam generators (i.e., a two-loop design).
The power pak mactor system employs
)
a compact pool-type design.
It is fueled by a stainless-steel clad. sodium-i bonded metallic alloy of U-Pu Zr.
The design relies on passive reactor shutdown and decay heat removal systems.
i SAFR will be designed for an SSE of 0.3g and OBE of 0.lg. DOE believes that on this basis the design will be acceptable for about three-quarters of the j
currently identified potential sites.
DOE has proposed a sithg design basis source term based on radioactive materials released from melting a single fuel subassembly rather than the traditional TID-14844 releases.
No conventional containment building and no requirements for preplanned off.
l site emergency evacuation are proposed.
l Dr. Remick questioned the seismic specifications for the SAFR design (OBE =
l 0.lg and SSE = 0.3g) from the perspective of the NRC regulations for the establishment of the OBE at 50% of the SSE value.
Dr. Siess noted that l
there has been ea attempt to eliminate the 50% requirement based on probabi-Hstic analysir, j
Mr. Landry stated that the overall objective of the SAFR program is to develop a conceptual design that minimites plant cost'and maximizes inherent safety. Other objectives are minimum potential for severe accidents and the elimination of the need for off-site evacuation planning by demonstrating low risk. The SAFR concept proposes fewer systems, components, and struc-tures classified as safety related than is the practice with licensed LWRs.
l The main control room and balance-of-plant items as well as many other items associated with the nuclear island (such as diesel generators and cooling water systems) are specified as comercial industrial grade.
i Two shutoown (scram) systems are utilized in SAFR, neither ofawhich arei classified as safety-grade.
The Automatic Plant Trip Systems (APTS)- can drive in all six primary control rods, and can interrupt power to -the electromagnetic latch and drop three secondary control rods into the core.
In addition, the three secondary rods can be dropped in by the Self-Actuated ShutdownSystem(SASS). The SASS is based on a magnet with a curie point at about 1050'F -(higher than operating temperatures).
The secondary safety rods are released whenever the core outlet temperatures exceed the curie point temperature.
Mr. 6. Van Tuyle, Brookhaven National Laboratory (BNL). presented results of t
analyses that were perfomed by(LOF), transient overpower (TOP), and unpro *
- BNL'. 'The - accidents : analyzed are ;1oss;of wm o
4 heat sink (LOHS), loss of flow tected single-pump-seizure accidents.
For the LOHS, the-feedwater pumps
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384TH ACRS MEETING MINUTES t I
providing water to both of the two steam generators are assumed to lose power, causing the steam generator to dry out, with the resulting loss of i
heat rejection.
BNL assumed loss of nomal cooling and of the Direct i
Reactor Auxiliary Cooling System (DRACS), with the outside surface of the l
reactor vessel well insulated. The caleviation indicated that the combina-Air Cooling System (RACS) y feedback from radial expansion and the Reacto tion of negative reactivit heat removal will prevent damage to the reactor.
If RACS air flow is stopped, major fuel damage starts to occur in about 18 4
hours.
Sodium boiling occurs in about 36' hours.
In another scenario, an LOF event is initiated by an instantaneous loss of power to the primary, the intermediate loop, and the steam generator pumps.
Scram does not occur.
The inertia 11y controlled coastdown of the primary pump is modeled by an initial six-second flow ' halving" time.
The fuel temperature increases as-the flow decreases, and, as a result, the power level decreases.
The negative reactivity feedbacks decrease power but the core remains at an elevated temperature. Scram must occur to bring the plant to cold shutdown.
i It is believed that the operator would have sufficient time to take action j
before significant core damage occurs.
For the TOP event, the calculations indicated that the radial expansion is the largest of the negative feedback mechanisms. BNL concluded that no fuel l
damage is expected during this event.
BNL analyzed an unprotected single pump seizure event in which one of the two centrifural pumps would seize during full.-power operation.
The other pump continues to operate and the plant protection systems fail to scram.
The seizure of one pump causes a drop in primary system flow impedance.
As a result, the unfailed pump)will experience a large flow increase (up to 128% of its rated condition and will cavitate.
The reactivity feedback reduces the power level to a point where the maximum full centerline temperature and the maximum sodium temperature in ' the core are - within acceptable limits. The conclusic1 is that this event would be mitigated by the reactivity feedbacks in the core, with no fuel damage or requirements for imediate operator action.
Overall conclusions are:
- 1) the SAFR passive cooling via RACS is effective and fault tolerant, 2) the SAFR inherent shutdown systems are similar to'.
Power Reactor Inherently Safe-Module '(PRISM) design, and 3) transients result in higher temperatures in the SAFR design and, as a result, control rod driveline expansion is enhanced.
Mr. Landry described some of the SAFR design events that were considered in the NRC staff's review.
These events are intended to bound the LMR design basis accidents, as well as some beyond design basis accidents, with margins to account for uncertainties.
These bounding events are also expected to provide conservatism in selecting a suitable site source tem.
i
- The staff has concluded that the SAFR designches' the : following' general) w safety advantages:
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344TH ACRS MEETING MINUTES i l
A slow response to core heat-up events
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1 Inherent beneficial reactivity feedback effects associated with the
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fuel and core expansion of the metal fuel pins Characteristics which will make it possible to demonstrate by test the significant safety features and perfomance of the plant over a wide j
range of events.
The staff also concludes that the design has potential vulnerabilities, such i
as:
)
Toe large positive reactivity effects associated with sodium voiding j
Th1 use of a relatively new metallic alloy in a sodium-bonded fuel debign, with the potential for relocation of the fuel following melting or eutectic femation and resulting reactivity-induced power excur-sions; The Staff's conclusion is that the SAFR design has the potential to achieve a level of safety at least equivalent to current generation LWRs provided the design and research and development needs are resolved, r
Dr. Kerr questioned the validity of the high reliability the NRC Staff is associating with nonsafety-grade systems.
Dr. Siess' questioned the Staff's argument regarding the reliability of safety-grade systems vs. nonsafety-grade systems.
He added that it is' not clear, and has never been demon-t strated to the ACRS.that safety-grade components are more reliable than nonsafety-grade components.
He noted that sometimes the only difference i
t'etween the safety-grade and nonsafety-grade components is different quality assurance requirements.
Dr.. Kerr wondered if the long time that this SAFR design allows for the operator to take action could increase the likelihood = that the operator might do the wrong _ thing.
u Mr; Carroll commented' that the term " inherently safe" is" confusing' and misleading.
Dr.. Remick agreed and added that use of the tem " walk away' reactor is similarly confusing.
Mr. Michelson expressed concern regarding the definition of external. events and the fire hazards from sodium, especially from pipe-break events in piping around one inch in diameter.
Dr. Remick expressed concern regarding analysis of the wetting of'the inside surface of the reactor vessel in sodium spill-over accidents.
He, thinks that: this. type,of accident has not been analyzed satisfactorily.
Dr..
Shewson' shared > the' concern and ~ added that ' consequences +cf; sodium rleakage 2 from the reactor vessel should be evaluated. <
WW:*
Mr. Ward commented that fuel rod bowing has been inadequately analyzed.
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344TH ACRS MEETING MINUTES III. ContainmentSystems(0 pen)
[ Note: Mr. Gary Quittschreiber was the Designated Federal Official for this portionoftheMeeting.[
Mr. Ward provided a brief report on the December 6,1988 Cantainment Systems Subcommittee meeting noting that a consultant report from that meeting was available. A draft teport had been prepared and distributed to the members.
He requested that members review this draf t and provide coments to him prior to the January 1989 ACRS meeting.
l Dr. Themis Speis, Deputy Director for Generic Items of the Office of Nuclear i
Regulatory Research, provided a condensed version of his December 6th subcomittee meeting presentation on the Mark 1 Containment Improvements a
Program.
This program was initiated as a result of concerns about the ability of Mark I containments to perform adequately during some severe accidents.
The Plant Examination (IPE) program is intended to complement the Individual It focuses on the ability of the containment to with--
stand severe accident challenges. Dr. Speis noted that the primary objective is to determine what actions, if any, should be taken to reduce vulnerabili-ty of containments to severe accident challenges.
Early efforts will be focused on Mark I containments.
Several early containment failure modes are associated with,large scale core melt.
The problem is exacerbted by the small size of these containments.
Members of the Comittee questioned the staff on the co1clusion that molten core material would exit the cavity and attack the liner. The staff stated i
that there was no -door seal (dam) which would delay the exit of the core material from the vessel cavity.
Mr. Michelson indicated that he knew of some plants that had such a door seal.
The staff is collecting information from the utilities as to the design configuration of individual Mark 1 plants.
It did not appear, however, that the staff has asked for this 1
information.
in response to a question from Mr. Michelson concerning whether the staff considered putting a dem on the flat floor below the Peach Bottom vessel,,
o Dr. Speis said that the staff does not see much benefit in doing this.
Mr.-
Beckner said they may not have specifically looked at this but their generic conclusion is that there are problems with the technical feasibility of this approach to mitigation. The material may be undercut, or if it is-a ceramic it may shatter upon contact with the core material.
He also believes the construction in an operating reactor will result in significant; serem exposure.
t Dr. Speis discussed insights obtained from the 12 PRAs which wr,re performed for Mark I containments (6 b dominant accident initiators (y industry and 6 by the NRC) in the area of. station blacko removal). PRAs have predicted wide variation in the accident frequency for w 1
different plants.
It appears that implementation of venting procedures can o
reduce the core melt frequency for the TV sequence (loss of the ultimate
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I 344TH ACRS MEETING MINUTES t heat sink with containment failure prior to core melt) by an order of magnitude or more.
In response to a question from Dr. Kerr concerning whether a similar survey of accidents for Mark 11 and III containments would lead to the same dominant accident initiators, Dr. Speis said he would obtain this infomation for the Comittee.
dr. Speis sumarized the discussions carried out at a workshop held on Februery 24-26, 1988 with about -150 representatives of industry, research groups NRC staff, and the public.
There was a variety of views on the i
probabil:ty of liner melt-through.
However, there was general agreement that water in the drywell is useful to delay / prevent she'1 failure and to reduce fission product releases.
Industry emphasis was on prevention.
The industry position is that the backfit requirements should be plant specific, i
and subject to the backfit process..Dr. Speis' discussed the staff approach in this area.
An effort is made to achieve a balance between accfdent i
prevention to reduce the likelihood of an accident occurring..and mitigation to reduce the challenge to containment and the magnitude of radioactive releases to the environment.
This approach produced the following recom-mendations for the Mark I containments:
- 1) Accelerate implementation of the station blackout rule (ATWS implementation will be essentially complete by January 1989).
2)' Require an alternate water supply for drywell spray and vessel injection, with a pumping capability that is independent of both nomal and emergency AC power.
- 3) Requirehardened(i.e.,abletowithstandsevereeccident pressures) venting capability from the wetwell.
3
- 4) Require enhanced ADS reliability with additional power and/or a
nitrogen supplies and improved cable reliability.
- 5) Require the implementation of the improved EPas (Revision 4 of t,he BWROwnersGroup).
Dr. Kerr noted that-the Station Blackout Rule must not provide for an f
adequate electrical power supply if the modification proposed in Item 2 was needed.
Mr. Thadani noted that Item 2 goes well beyond the scope of the Station Blackout Rule but that-a utility could combine their station black-out modifications into the proposal.
Dr. Speis suggested that it may be possible for many plants to use the modifications provided as a result of the Station Blackout Rule to take care of this matter.
Dr. Kerr-suggested that the staff reexamine the Station Blackout Rule to assure that the AC power supply was sufficiently reliable.
Mr. Michelson questioned the staff as to whether new equipment installed as a
result of the Mark I
Containment-Improvement Program - would be-r
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344TH ACRS MEETING MINUTES,
environmentally qualified for the expected accident tnvironment conditions.
Dr. Speis said the equipment would be qualified 1o function during an accident.
l In response to a question from Dr. Kerr concerning whether the Station l
Blackout Rule implementation was considered in the evaluations which con-i cluded that the TW sequence was an important contributor, Mr.- Thadani said that it was not.
The Station Blackout Rule has not yet been implemented, and backfits will vary from plant to plant. Some plants may not be affected very much since they may already be meeting most of the requirements of the rule. Mr. Thadani added that the risk-associated with the W sequence is not driven by the availability of electrical power. Mr. Beckner stated that j
the TW sequence can be considered separately.
In response to a question from Mr. Carroll concerning how one assures that i
water gets into the vessel cavity for those plants that do not have a ' door seal," Mr. Soffer said that their observation at Peach Bottom was that two j
spray headers had spray angles of 180 degrees and that the water drains into the sump at the center of the pedestal.
Mr. Michelson added that he be-1 lieves that the pedestal area would eventually fill with water.
Dr. Speis discussed several industry efforts which have already been pro-posed and/or are taking place in this program, including the following:
l
- 1) The BWR Owners Group has proposed Revision 4 to the EPGs which includes venting of containment. The NRC staff has recently approved this revision.
- 2) Vemont Yankee is planning changes during their 1989 refueling outage associated with this program.
- 3) Pilgrim has developed a safety enhancement program which implements several recomendations from this program, including the addition of i
a hardened vent frm the torus to the stack, adding a third onsite diesel, adding a backup nitrogen supply for ADS, additional inerting, and use of fire protection diesel pumps.for decay heat removal.-
)
Dr. Speis discussed the cost / benefit analysis of the suggested improvements, i
noting that the major benefit is core melt reduction resulting from venting.
from slow overpres-Venting can also prevent containment failure resulting$48 million to $176-sure.
The industry cost estimates range from about million with the result that the cost / benefit does not always meet. the
$1,000 per man-rem criteria. However, the staff believes that some of the enhancements can be justified from an engineering analysis.
Consideration is being given to implementing some changes via rulemaking.
Dr. Speis asked the Comrdttee to provide coments to the Comission from a technical perspective as to whether the proposals recomended by the staff are appropriate.
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344TH ACRS MEETING MINUTES,
i Dr. Speis provided the following NRC staff conclusions and recomendations:
- 1) The proposed enhancements provide substantial increase in the overall protection of public health and safety, i
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- 2) The proposed enhancements are generally cost beneficial.
3)
It is proposed to implement the enhancements via rulemaking, j
- 4) Confinnatory research should be performed in the areas of phenomena relevant to in-vessel and ex-vessel accident progression, the effect of cavity water on the probability of liner melt-through, and the associated source terms.
Dr. Speis mentioned several coments prodded by the CRGR as a result of their review of the proposed enhancements.
Mr. Wayne Hodges, Chief of the Reactor Systems Branch in the Office of Nuclear Reactor Regulation, discussed the Emergency Procedure Guidelines (EPGs) provided in the latest Revision 4 (recently approved) as they relate to the Mark I Containment Improvement Program. The criteria used to develop a
the EPGs for BWRs are symptom-based (not on specific scenarios) and specify appropriate actions for all emergencies including severe accidents.
The EPGs are not limited by licensing or design basis assumptions.-
They apply i
to plants as currently built using all available plant equipment, and not just safety-related equipment. Operators are guided to take the best action possible, including using nonsafety systems.
Mr. Hodges stated that the j
guidelines are developed by the BWR Owners and not the NRC.
l Mr. Hodges discussed the major improvement 3 that Revision 4 incorporated, including the restructuring and simplification of the general fonn of the guidelines, and significant changes in the guidelines for ATWS mitigation, which acknowledge that large power oscillations might occur. Revision 4 has also added hydrogen control guidance for Mark I and Mark 11 containments and has provided improved containment venting criteria.
4 4
Mr. Hodges discussed the improvements in Revision 4 in the area of ATWS 4
mitigation.
He noted that alternate rod insertion guidelines have been restructured and simplified, and that the reactor pressure vessel water level control band has been extended to below the top of the active fuel 4
during ATWS. There have been changes made to lower the reactor vessel water level limit which initiates isolation.
This keeps the condenser available as a heat sink for a longer time.
Mr. Hodges discussed containment venting as a means to prevent core melt-and reduce dose.
This strategy was discussed in NUREG-0737 (Item I.C.1) and-required procedures which allowed for events with multiple failures and operator errors.
The BWR Owners' Group developed EPGs to comply with this early requirement by calling for venting to prevent failure of the
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344TH ACRS MEETING MINUTES i containment.
It later became evident that venting was an effective means for preventing core melt, and for mitigating the consequences of core melt..
the possibility that In response to questions from Mr. Michelson concerning(through low-pressure leakage of steam to equipment areas during venting ducting) might make things worse instead of better, Mr. Hodges said that the i
staff feels that if the ducting is ruptured recovery may be complicated.
The staff believes that this is still better than allowing core melt to occur.
Installation of the hardened vent in the Mark I program is being aroposed to prevent this problem.
Current EPG guidelines do not require a iordened vent.
Mr. Thadant added that the staff is fimly convinced that the Mark I plants need a hardened vent to take full advantage of the bene-l fits of venting. He noted that the licensee for the Pilgrim Plant estimates that the most likely use of the vent would be for prevention before any fuel damage occurred.
Mr. Hodges stated that earlier versions of the guidelines recommended that venting be performed at the time thGt the primary containment design pres-sure limit was reached, ihe new guidelines recommend venting prior to reaching that limit, heither set of guidelines specifies how long to leave the vent open. This will be defined in the plant-specific procedures, i
In response to questions from Mr. Michelson concerning what assurance the NRC has that the EPGs can actually be implemented with the equipment and procedures in place, Mr. Hodges said at this point ther do not have any I
assurance.
The NRC has audited the implementation of the EPGs and found cases where the implementation was not satisfactory. Mr. Thadant added that this is a problem with EPGs in general, and not just for the Mark I plants.
l Mr. Hodges noted that the NRC has audited the implementation of the EPGs for 15 plants (including both PWRs and BWRs) over the past several months and his found equipment-based implementation problems in all of them.
i Fr. Hodges discussed the changes in the primary containment pressure limit f.PCPL) from earlier revisions of tW EPGs.
Revisions 2 and 3 required opening vents upon reaching the PCPL, which was twice the design pressure, Kevision 4 provides better guidance for determininfi PCPL.
The PCPL is set i
l at a pressure limit which bounds the pressure capab< lity of the containment.
l This may be determined by the containment vent valve operability limit, the steam relief valve operability limit, or tha reactor pressure vessel vent l
valve operability limit.
Mr. Hodges discussed advantages and disadvantages of containment venting.
Advantages include maintaining \\ow reactor pressure vessel pressure to allow core cooling by low pressure systems, cooling of the core with systems external to containment and removal of decay heat by pool steaming.. use of the pool for scrubbing of fission products, and prevention of containment failure due to overpressure.
Disadvantages of venting included the possi-bility of 6 easing dose if not coordinated with evacuation, possible loss of sors equipment due to the harsh environment caused by venting, the
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344TH ACRS MEETING MINUTES 1 potential for inadvertent venting, and the potential for the bypassing of pool scrubbing due to improper venting.
Mr. Hodges stated that the NRC has taken the position that the decision on venting should be made by the senior utility manager on site at the time venting is needed.
Dr. Speis promised that the staff will update the Committee on this program at the January 1989 ACRS meeting.
IV. EquipmentQualification(EQ)-RiskScopingStudy(0 pen)
I. Note:
Mr. Sam Duratswamy was the Designated Federal Official for this portionofthemeeting.)
Dr. $1ess. Acting Chaiman of the Reliability Assurance Subcouaittee, stated that for several years the Office of Nuclear Regulatory Research (RES) of l
the NRC has funded EQ research to study the methods for qualifying safety-related electrical equipment and to demonstrate their survivability during and following design basis accidents (DBAs) that produce harsh environments.
The EQ research was teminated by the end of FY 1986.
At that time, the ACR$. in its February 16, 1986 report to the Congress and in its June 11, 1986 report to the Comission~ recomended that the EQ research be funded to assess the survivability of electrical equipment under hostile environmental conditions retulting from accident, includie severe accidents. = In n-sponse, the RES Staff stated that they plan.to pefom a risk-based priori-tiration study on EQ in FY 1987 to detemine the neM for further research in this area.
Acordingly, the EQ-Risk Scoping (Study was it.itiated and per-fomed by the Santia National Laboratories SNL).
He stated that the purpose of the EQ-Risk Scoping Study was to use the information from exist-ing PRAs and assess the risk significance and risk uncertainties associated with current EQ requirements for safety-related electrical equipment.
Dr. Siess stated that the Reliability Assurance Subcommittee discussed the EQ-Risk Scoping Study during the meetings on December 16, 1987. - June 14.-
1988, and December 12, 1986.
This matter was also presented to the full Comittee by representatives of RES and SNL during the 339th ACRS meeting, July 14-16, 1988.
Since a set of peer-review coments was provided to the ACRS the day before the July-1988 ACRS meeting, the Comittee-decided to.
t defer action on this matter pending a detailed review of the peer review-comments and the associated SNL responses by the Reliability Assurance '
Subconmittee.
Accordingly, the Reliability Assurance Subcommittee met with i
the representatives of RES and SNL on December 12, 1988 and discussed the peer-review coments and SNL's msponses to them.
Dr. Siess stated that the peer review was performed by the following person--
nel:
l Mr. Kenneth Canady. Duke Power Company Mr. George Siiter, Electric Power Research Institute
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l 344TH ACRS MEETING MINUTES !
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i Mr. Andrew Wolford, Idaho National Engineering Laboratory Mr. Sal Carfagno, Franklin Research Center l
The peer-review. panel members met twice with representatives of SNL to
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discuss their coments on the preliminary results and conclusions of the Scoping Study.
The peer-review coments were generally favort ble.
In response to the peer-review coments, several clarifications and changes have been made to the draft report of the Scoping Study.
Dr. Sieis stated that, in his opinion, the review of the Scoping Study by the petr-review panel has contributed to the credibility of the conclusions and th! quality
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of the final report of the Scoping Study.
l Mr. Michelson asked who decided to do a peer review of the EQ-Rhk Scoping Study. Dr. $1ess stated that the Staff is trying to do peer review on most of the studies similar to this one.
He is not sure.whether there is a j
general RES policy related to peer review.
He suggestina.that when the i
l Comittee meets with Mr. Beckjord, Director of RES, on December 16, 1988, it may want to find out about the RES policy regarding seer review.
l After further discussion, the Committee discussed the coments and recommen-dations proposed by the Subcommittee on the EQ-Risk $ coping Study and approved them, with minor changes, for transmittal to the Commission.
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V.
NRC Quantitative Safety Goals (0 pen) l
[Nate: Mr. Gary Quittschreiber was the Designated Federal Official for this portionofthemeeting.)
The Comittee held a discussion with Mr. Wayne Houston, Director of the-Division of Safety Issue Resolution in the Office of Nuclear Regulatory Research, on the staff's presently proposed plan for implementation of the Safety Goal - Policy.
The plan -being proposed to implement the program is said to clarify the role that safety goals, the quantitative objectives, and the use of PRAs will have in future regulatory ' decisions. It includes discussion on implementation of the Backfit-Rule.
The NRC staff's resolu-tion of ACRS recomendations made in reports on May 13.1987 April 12, 1988, and July 20. 1988 was also discussed.
Mr. Houston noted that the staff is and has been using PRAs to make deci-sions, but does not have any clearly stated practices, goals, or criteria as.
to what purpose PRAs serve in the regulatory process.
He hopes that the implementation plan will bring this into clearer focus.
He suggested that the hierarchy for use of the safety goals would be use6 first by-the design..
ers, then by the operators, and, last of-all, by-the regulators.
With' regard to regulation, he suggested that.the quantitative objectives incor -
porated in the safety goal should be regarded as targets for generic regula-tory requirements and not as criteria for individual licensing decisions.
L Mr. Houston discussed the words ' substantial-increase in overall protection" as stated in the Backfit Rule. as a statutory standard and suggested that'
- i 344TH ACRS MEETING MINUTES 12 this could be used in conjunction with the Safety Goal and PRAs in the context of developing regulatory analyses which detemine if e " substantial increase
- is achieved.
The staff's proposed plan provides a discussion of a potential approach to developing a relationship between the Safety Goal i
Policy's quantitative health objectives (QHO) and the Backfit-Rule defini-tion of an adequate protection standard.
Mr. Houston suggested that the l
relationship might fall between two and ten times the QHO. The Backfit Rule discussion does state that compliance with the Comissions rules, regula-tions, and positions are presumptive evidence of meeting an adequate protec-tion standard. However, under the Backfit Rule, no matter how safe a plant is, if the cost benefit criterion can be satisfied by a p m posed change, it ccn be required.
Mr. Houston stated that the specification of.a core damage frequency as a safety goal objective provides a goal against which one can measure and make 4
design judgments.
The NRC staff's proposed plan will recomend to the Commission that the existing policy statement be sup>1emented to include design objectives for core damage frequency and a spec <fic definition for a large release, l
Mr. Houston discussed recomendations by.the OGC for dealing with averted onsite costs.
The staff will continue to use the $1,000 per person-rem criteria (for dose received within 50 miles from the site) and will. recom-mend that averted onsite costs not be treated as a benefit but rather as a factor which offsets the utility's costs.. When used in this way )(appearing in the numerator of the cost / benefit equation as a negative cost it often does not make a large difference when dealing with perspective modifications since it affects only the cost side of the ratio and not the benefit side, l
Mr. Houston described the principal elements of the NRC staff's recomenda-tions in the plan:
- 1) establishing a hierarchy of quantitative objectives
- 2) reviewing PRAs to assess effectiveness of regulatory requirements, 3} -
integrating risk reduction modifications and testing proposed modifications, i
,and 4) using subsidiary goals for generic safety issue resobtion in con-junction with less than full scope PRAs.
The staff is prv,.osing.a five-
!? vel hierarchy, starting with the qualitative safety goals and working down to tne last level.of regulatory requirements.
Mr. Ward noted one signifi-cant difference from the ACRS proposal was that the staff -is not estab-11shing guidelines for defense-in-depth. The staff is dealing with the core 3
melt issue by prevention and has not established criteria which set goals for sitigation.
In response to questions from Mr. Michelson concerning the scope of the PRAs being suggested, Mr. Houston said they are - recomending full scope PRAs which.'onsider both external and internal events and exclude only sabotage l
and other fuel cycle matters.
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j 344TH ACRS MEETING MINUTES 13-L In response to questions from Dr. Kerr concerning allocation of fractions of risk to certain safety issues Mr. Houston noted that the staff is already doing this for core melt frequency but not with regard to a containment-performance design objective.
The staff's uefinition of a large release is e release that has a potential for causing an off-site early fatality which is said to be a " natural 4
threshold" effect.
The staff believes that this avoids the problem of arbitrarily choosing a release of a given number of curies wh9ch has no direct relationship to a particular consequence.
The staff believes that j
the word " potential" has a significant meaning in this definition.
The staff has defined the term " core damage" as the potential threshold assoc 1-l ated with the loss of adequate' core cooling.
Mr. Houston noted that the major thrust of the safety goal policy is defined in the plan as being directed to light water reactorst however, many of the principles could be applied to advanced reactor designs.
4 ACRS members questioned Mr. Houston on the terms and frequencies being pro-l posed by IAEA, EPRI, and others and whether there was any need for them to be consistent with the NRC's proposal.
Mr. Houston discussed some problems associated with having different requirements in the areas of siting, emer-gency planning, and need for containment when core damage frequency is very low.
He suggested that establishing lower targets for' future plants might result in better designs.
He believes that one cannot have as much confi-
)
dence in a PRA for new unproven designs as compared to a PRA on an operating plant.
In addition, as the number of plants increases one will need ' lower i
j risk per reactor to have the same overall risk.
Mr. Houston discussed the information which can be obtained' from existing J
PRAs and the use of this infonnation in PC-based codes. Dr. Kerr questioned the ease of using these codes by those other than the PRA analysts.-
Mr. Houston stated that he believes NUREG-1150 will play a dominant role.in the assessment of regulatory requ::*ments over the next several years.
Mr. Houston stated that he believes the staff does plan to deal with the i
question of whether one can' give a probabilistic interpretation of the l
. meaning of " credible."
He noted that this and in the use of the Standard Review Plan. question has come up in hearings-The staff proposes to deal with-this term in such a way as not to change the large release guideline definie tion.
The staff intends 'to use partial scope PRAs in those cases where a PRA ex-1sts for a particular type event and will use the same numerical objectives for the overall safety goal. This gets away from allocation of risk.
The Committee decided to try to write a report on this matter at its January.
1989 meeting. - No further NRC Staff presentations were requested for the-January 1989 meeting, i
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344TH ACRS MEETING MINUTES '
VI.
Meeting with the Director of the NRC Office of Nuclear Regulator.y Research (0 pen) i
[Mr. Sam Duraiswamy was the Designated Federal Official for this portion of themeeting.]
Mr. Beckjord, Director of the NRC Office of Nuclear Regulatory Research (RES), discussed the items given below.
RES Reorganization Mr. Beckjord stated that the RES reorganization became effective on July 17, i
1988. The main objectives of the reorganization are to:
l Consolidate the efforts related to implementing the Commission's Severe Accident Policy and the resolution of Unresolved Safety Issues (USI) and Generic Safety Issues.
Restructure severe accident research so as to provide earlier input to the decision-making process on severe accident issues as well as longer 1
term confinnatory research needed for closure of severe accident-issues.
Make most effective use of limited RES resources..
Clarify the responsibilities of the RES Deputy Directors.
Mr. Beckjord stated that, under the current organization, the four divisions i
of RES fall into two major categories:
Divisions Responsible for Research l
Divisic( of Engineering Division Of Systems Research These d; visions will come under Dr. Ross Deputy Director for Research.
Divisions Responsible for Resolution of Issues. and' Development of Rules and Requirements 1
Division of Safety Issue Resolution Division of Regulatory Applications These divisions will come under Dr. Speis, Deputy Director for _ Generic Issue Resolution.
Mr. Michelson asked which Division is responsible for research related to equipment qualification and fire protection.
Mr. Beckjord responded thate such research will be the responsibility of the Division of Engineering.: +:
u
~'
344TH ACRS MEETING MINUTES In response to a question from Dr. Siess about manpower allocations Dr.
Ross stated that the Division of Engineering and the Division of System Research have about 60 people each, and the Divisions of Safety Issue Resolution and Regulatory Applications have about 50 people each.
Dr. $1ess noted that, under the reorganization, although resolution of USIs and generic issues has been consolidated under one division, prioritiration of generic issues is handled by the Advanced Reactors and Generic Issues Branch under a different division. He asked whether the same personnel work on both advanced reactor issues and generic issues. Mr. Beckjord responded that these issues are handled by two separate sections in that branch.
Mr. Michelson commented that the Staff seems to be reluctant in addressing which 9eneric issues will be a p11 cable to future plants.
Dr. Speis re-spondec that the Staff recenti had a two-day workshop to discuss several matters, including the applica 111ty of generic issues related to future plants.
Mr. Michelson stated that in the resolution of each generic issue the Staff should specify clearly whether that specific issue will be applicable to future reactors.
Dr. Speis said that he would discuss athis matter in detail with the ACRS at a future meeting.
Dr. Siess requested a copy of the summary report related to the workshop held recently on advanced reactors. Dr. Speis agreed to send a copy of that report.
In response to a question from Dr. Siess, Dr. Speis stated that implementa-tion of the resolution of USIs and generic issues is a joint effort between RES and NRR.
Status of Im)1ementation of the National.Research Council's Recommendations on Revitaliz' ng Nuclear Safety Research Mr. Beckjord stated that, as ' requested. by - the Commission, the Nation' l a
Research Council's Committee (ad hoc) on Nuclear Safety Research performed a Study regarding the future course of nuclear safety research in the U.S.-
The conclusions and recommendations of the study are documented in a report entitled " Revitalizing Nuclear Safety Research," dated December 8,1986.
There are several recommendations specifically addressed to RES, some to the EDO, and some to the Commission.
Mr. Beckjord discussed the recommendations specifically addressed to RES and the status of their implementation:
The NRC should bring in high-cal Ber researchers to bolster management.,
Mr. Beckjord stated that.to accommodate this recommendation he tried to n ;
hire an experienced research person.
He made job offers-to 11 quali-; -
fied individuals, but none of them accepted the -job. He then tried to
s' l
344TH ACRS MEETING MINUTES accomodate this recommendation by means of visiting fellowships and staff exchanges.
Under the visiting fellowship program, he had one orofessor from Stanford University to work 'on the severe accident program, one professor from MIT to work on PRA studies, and a person from Brookhaven National Laboratory to work on various aspects of severe accidents.
He believes that this approach has been very con-structive and brought some outside experience and new ideas.
Dr. Siess asked about the role of the people who came to RES under the visiting fellowship program. Mr. Beckjord responded that it depends on the neids of RES.
The people who came on board thus far helped in reviewing the existing research, planning of new research, and perfom-ing some PRAs.
The NRC should consider separating the function of stap. lards develop-ment and research.
Mr. Beckjord stated that this recommendation was not accepted either by RES or by the Commission. He believes that they have to have addition-al people and resources to accommodate this recommendation effectively.
Owing to budget constraints, he does not believe it is possible to get additional people for RES.
The NRC should develop a cogent philosophy of safety research.
l Mr. Beckjord stated that RES had already prepared a statement of research philosophy and was approved by the Commission in May 1988.
It is included in NUREG-1325, Disposition of Recommendations of the i
, National Research Council in the Report Revitalizing Nuclear -Safety l
Resear:h, and also in the NRC's-Five-Year Plan.' He stated that, in this statement of philosophy, RES has set forth the key principles that should govern -the definition, planning, conduct.- use, and closure of i
nuclear regulatory research projects.
Dr. Siess asked whether the statement of research philosophy has been applied so far in making decisions, settling arguments, or assigning research priorities. Mr. Beckjord stated that~1t is being applied.
4 ine NRC should' establish a research pro ram planning process involving all of the relevant offices within the 1 RC. as well as reoresentatives from industry and the university research community acting as partici-pating advisors.
l Mr. Beckjord stated that to accommodate this recommendation they have established research review groups that include -re>resentatives from i
various offices of the NRC. ' hese groups meet periocically to plan and develop research programs to support NRC programs and strategies. -They also plan ' to, meet with industry groups. such - as. EPRI, IDCOR,n and -
NUMARC,: twice a year to discuss coordinated research efforts. :
i i
344TH ACRS MEETING MINUTES,
The NRC should impanel an independent advisory group reporting to the Director of RES.
Mr. Beckjord stated that the Commission has approved the creation of an independent advisory consnittee called the " Nuclear Safety Research l
Review Comittee "
RES had already met with this Comittee twice.
This Comittee plans to establish some subcomittees to look at various research programs in the areas of human factors, severe accidents, etc.
Dr. Kerr asked about the membership of the subcomittee for severe accidents.
Mr. Beckjord responded that it consists of the following members:
Inc. (Chainnan)
Salomon Levey, S. Levey, University Richard Wilson, Harvard i
Cordell Reed, Comonwealth Edison David Morrison IIT Research Dr. Siess asked whether RES has a general policy for setting up peer-review panels to review research results.
Mr. Beckjord responded that they have a general policy that requires contractors performing re-search for the NRC to publish the results in scientific journals.. As far as peer review of other research studies are concerned, they do not have a written policy.
However, he strongly encourages the use of.the peer-review process.
Mr. Beckjord discussed briefly other recomendations of the National Research Council related to creating a fair and competitive process for contracting research, and performing more research at universities.
He
)
stated-that in FY 1988 they spent about $13 million for research at universities and private contractorst in FY 1989, they expect to spend about $16 million.
Dr. Kerr asked' how much money is spent at universities for research.
Mr..Beckjord.and Mr. Bartlett comitted to provide this information'at-a later date.
Status of implementation of the National Research Council's Recommenda-tions on Human Factors Research Mr. Beckjord stated that in response to another request by the Consnis-ston, an ad hoc comittee of the National Research Council performed a
(
study on the need for additional research in the human factors area.
The results and conclusions of this study are documented in a report entitled " Human Factors Research and Nuclear Safety," dated Februery 19, 1988.
In sumary. the National Research' Council' recomended ;that tN. NRC.m >
facilitate the capability for conducting human factors research 'oy:;
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L; 344TH ACRS MEETING MINUTES Staffing and maintaining continuity of the program at the branch 1
level Adopting a systems-oriented or socio-technical perspective to the research Utilizing independent peer reviews to enhance the quality of research products Establishing improved mechanisms for transferring research results to user communities by means of annual written. reviews and a bibliographic search system.
Increasing the timely transfer of knowledge to industry.
In addition, the National Research Council recommended that research be conducted in the following major areas:
Human-system interface design Research on the personnel subsystem (training, qualifications, etc.)
Humanperformance(humanerror)
Management and organizational performance Studies on the regulatory environment.
Mr. Beckjord stated that the human factors program plan and the Staff's responses to the recommendations of the National Research Council are included in SECY-88-141. " Human Factors Initiatives and Plans." dated May 23, 1988. Subsequent to reviewing the contents of SECY-88-141, the Comnission directed the Staff to address several issues related to the NRC's human factors programs.
The Staff's responses to the issues raised by dated Octoberthe-Connission are included in SECY-88-294
' Human Factors,
- Program, 13, 1988.
Mr. Kaufman stated that an updated document related to human factors programs is being prepared and is expected to be submitted to the Commission during January 1989.
The Comnittee suggested that the ACRS Subcommittee on Human Factors review the human factors program plan and other related issues.
VII.
Review of NRC/RES Code Scaling Applicability and Uncertainty (CSAU)
Methodology (0 pen)
[ Note:
Mr. ' Paul Boehnert was the Designated Fedaral Official for this portionofthemeeting.]
l l
s.
i 344TH ACRS MEETING MINUTES Mr. Ward, Chairman of the Thermal-Hydraulic Phenomenon (T/H) Subconunittee, briefly summarized the history of the CSAU effort. He noted that CSAU was l
developed to evaluate the effectiveness of the best-estimate ECCS codes.
The T/H Phenomena Subconnittee last discussed this issue at a meeting on December 7, 1988.
Mr. Ward said the Subcommittee believes the CSAU program is a success and requested consents from ACRS members at the conclusion of this discussion.
Mr. Ward said he would provide a draft letter on this topic for the Conunittee's consideration during the January meeting.
Dr. Shewmon requested information during this presentation that addresses how much margin would be available by using the CSAU methodology as measured against the Appendix K PCT. He said that recent reactor vessel fluence data indicates that some plants' vessel design life may be shortened due to higher-than-expected fast neutron (pressurized thermal shock limitations)-
fluence.
He wondered if the additional margin available could offset the
-l j
expected design life penalty.
Mr. Ward said the margin obtained by use of best-estimate ECCS codes could indeed be used to increase meer peaking, thus reducing the fluence at the vessel wall.
Te quantify tie margin, one must do a best-estimate ECCS evaluation. The r;AU methodology is an accept-able method for use in the required uncertair.cy analyses.
Dr. Zuber, RES, described the development of the CSAU methodology and its -
application to an LB LOCA code calculat19n.
He also identified the members of the Technical Program Group (TPG) that developed and applied' the CSAU method.
Dr.1. Catton has acted as an observer to the TPG on behalf of the ACRS.
While' the TPG effort consumed 117 man-months of time, Dr. Zuber indicated that > to apply.the technique to, some other area would probably :
require only 36-48 man-months.
The objectives of the CSAU methodology are to:
1.
Provide a technical basis for. quantifying uncertainty within' the context of the revised ECCS rule.
2.
Provide an auditable, traceable, and practical method for combining quantitative analyses and expert opinion to arrive at a computed _value of uncertainty.
3.
Provide a systematic and comprehensive approach for:
a.
Defining scenario phenomena b.
Evaluating code applicability c.
Assessing code " scale up" capability d.
Quantifying code uncertainty related to:
Code and experiment inaccuracies j
_ Code scale-up capabilities Plant state and operating conditions.'
l
344TH ACR$ MEETING MINUTES Representatives of RES said that the CSAU was reviewed during its develop-the ACRS and an international peer review group chaired by N.
ment by (MIT).
Todreas The TPG also developed three simple physical models designed to calculate:
Dr. Zuber indicated that the central reason for developing these simple models was to provide & practical means of summarizing the information/ knowledge obtained via CSAU. He said simple physical models are the best method for such a knowledge transfer to future users.
These physical models were not developed for use in the licensing process.
RES has calculated the LB LOCA PCT bound, using CSAU to be 1572'F at the 955' certainty limit.
In response to Dr. Shewmon, Dr. Zuber indicated that the BC appreach provides about 600'F margin in PCT vis-a-vis the "old" Appendix K requirements.
The key sumary points noted by Dr. Zuber were:
We have an extensive exaerimental data base for LB LOCAs.
We have systems codes tiat are very detailed.
We have a methodology that meets the three objectives stated above.
We understand the physics of an LB LOCA very well.
The LB LOCA issue is resolved.
Dr. Zuber said the " moral" of the LOCA research effort is that given: 1)an extensive experimental data base from well-scaled facilitiet. and well-designed tests, and 2) close integration of experiments and analyses, closure of the LB LOCA issue was possible.
Dr. Zuber also indicated that CSAU could be applied to other parts of the severe accident research effort.
The details of the CSAU methodology were provided. The method is subdivided into three subelements:
- 1) scenario requinements and code capability, 2) assessment and ranging of parameters, and 3) sensitivity and. uncertainty analysis.
Details of the above subelements were reviewed.
Figure 1 outlines. the methodology subdivided pursuant to i.he subelements and d s summarized below.
' Requirements and CapabOities" in which, scenario modeling requirements are identified and compared to computer code capabilities to detenmine their ap111cability to the particular scenario and to identify poten-tial lim'tations.
Assessment and Ranging of Parameters" in which code capabilities to calculate processes important to the scenario are assessed against experimental data to determine code accuracy, scale-up capability and ranges of parameter variations needed for sensitivity studies.
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" Sensitivity and Uncertainty Analvsis" in which the effects of indi.
vidual contributors to the total uncertainty are obtained, and for which the propagation of uncertainty through the transient is properly accounted.
4 Dr. Zuber str6ssed that the key advantages to use of CSAU is that the methodology is systematic, comprehensive, traceable, and auditable.
The j
last two characteristics are important for its use by a regulatory agency.
in response to cuestions from Mr. Ward as to the possible benefits of this research, Dr. Zu>er stated that, as an example, using BE LOCA analyses could j
result in a savings of about $8 billion to utilities owning Westinghouse reactors.
RES representatives said that the CSAU method will be applied to the scenar-io of an SB LOCA in a B&W plant, using the RELAP-5/M00-3 code..
in response to questions from Drs. Remick and Kerr concerning applying CSAU methodology to severe accident research, Dr. Zuber indicated that the use of CSAU methodology would drive the researchers to focus on issues / phenomena that are of central importance to resolution of their concerns.
l Dr. Ward said he would provide the ACR$ with a draft letter on this topic for the Comittee's consideration during its January meeting.
4 Vill. Meeting with the Director, Office of Governmental and Public Affairs (0 pen)
[ Note:
Mr. Herman Alderman was the Designated Federal.0fficial for this wrtion of the meeting.)
i i
Mr. Harold Denton, Director of the Office of Governmental and Public Af-i fairs, described some of the recent NRC contacts with the Soviet Union. Mr.
i Denton reminded the Committee of his previous briefing in which he told the Comittee about his first trip to the Soviet Union.- He noted that 'the i Soviets made a reciprocal visit to the United States in the fall of 1987 and at that time both parties agreed to proceed to attempt to develop an agree-i ment covering civilian nuclear power safety.
Chairman Zech and Alexander Pxotsenko signed, in April 1988, an agreement to cooperate in the field of civilian nuclear-power plant safety. One of the-features of that agreement was the establishment of a joint coordinating committee which selects specific topics in which coopnation would take place and develop the framework under which the cooperation would occur.
Mr. James Taylor was named chairman of the U.S.
delegation and Dr.
PonomarevStepnoy was named chairwan of the Soviet delegation.
The first meeting of the joint committee was held in the Soviet Union in August 1988.
Mr. Taylor discussed the results of. that meeting.
He noted that they had set out to select topics where they could be of benefit to the Soviet Union l
I l
344TH ACRS MEETING MINUTES and the United States.
One area of cooperation that was agreed to was to exchange inspections, on a trial basis, at operating nuclear stations.
Dr. Shewmon asked if the Soviets have a set of regulations comparable to the U.S. regulations. Mr. Taylor replied that the Soviets are developing a body of regulations.
Mr. Taylor said that the Soviets are very interested in having the U.S.
participate in a joint analysis of the level of safety of a Soviet nuclear power plant design. An agreement has been reached whereby the Soviets. will provide the safety analyses for Zaporsia and the U.S. will provide the complete safety analysis for the South Texas plant.
One other area of exchange was in the area of nuclear safety research. The United States outlined the broad areas of safety research conducted by the NRC.
The Soviets, in turn, provided elements of their ongoing research.-
The objective was to determine if there were amas of possible joint coopera-tion in research that could be carried out under the US/ USSR agreement.
j Of interest to the United States was the area of radiation embrittlement and in-place vessel annealing.
The Soviets have annealed several reactor vessels in place.
An area in which the Soviets expressed considerable interest was the work that the United States has perfonned on fire safety.
Mr. Taylor noted that the Soviets had a fire at their Ignalena plant.
The i
fire was in the cable area and affected some of the reactor control cir-cuits. Mr. Taylor said the NRC expects to get more of the details concern-ing this fire in the future.
Dr. Denton said the Soviets have' formed a ministry called the State Commit-tee for the Safety of Nuclear Installations.
It employs about 900 people-and was fonned from existing organizations.
He noted that the Soviets do i
not have guides, rules, or regulations comparable to those which the United.
States has developed over the past two decades.
Mr. Taylor noted that the Soviets are interested in how the United States conducts the decision-making process on backfits and in - severe accident analysis. The perfonnance of and use of PRAs are also of interest.
The NRC and DOE will both participate in discussions with the Soviets on the health-and environmental effects of Chernobyl.
It is expected that a large body of information on the biological effects will be - developed.
The Soviets are trying to set up the mechanisms to give health and physical examinations to the population that has been involved.
The United States -
hopes to share in the biological and environmental infonnation that will be-derived from this study.
Chainnan Kerr asked to what extent the dose to the people involved in Chernobyl was measured.- Mr. Taylor replied that he did not know how accu-rate the exposure records were.
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4 I
l 344TH ACRS MEETING MINUTES 1 i
Mr. Taylor said that the Soviets and the United States had discussed how the United States works to gather operational data and attempts to benefit from r
4 operational data and events.
i The Soviets have had considerable experience in the erosion / corrosion area.
They are willing to share their experience and data with the United States, i
Mr. Edward Shomaker has been appointed as the NRC project manager to handle interagency coordination.
Chaiman Kerr asked if the ACRS could participate in the April meetings with the Soviets. Mr. Taylor replied that the ACRS would be welcome to partici-pate with the working groups or in discussions of any topics that are on the agenda.
Dr. Remick asked about the effects of the recent Soviet earthquake on nuclear reactors.
Mr. Taylor replied that there were tro reactors in.the area and they are both operating.
He said the Soviets had. announced that these two reactors will be shut down in two years and that the shutdown was related to concern as to possible effects of the earthquakes.
IX. OperatorRequalification(0 pen)-
['iote:
Mr. Heman Aldeman was the Designated Federal Official for this portion of the meeting.)
Dr. Remick noted that when Part 55 was modified one of the modifications l
was that reactor operator licenses had to be renewed every six years instead of every two years.
Requalification examinations had a negative impact on the operators because the tests were not always perfomance based (i.e., not related to what operators had to do in the plant).
NRC decided to look at l
new methods of administering the requalification exams.
These new methods were developed and tests conducted in five pilot programs.
Dr. Remick said that his understanding was that the pilot programs have been very success-ful.
Mr. Ken Perkins, Chief, Operator Licensing Branch, discussed the requalifi =
cation program for operator licensing. He said that the staff developed and tested a new methodology for assessing the effectiveness of facility requal-ification training programs and, at the same time, for assessing the profi-ciency of operators in maintaining the goal of enhancing plant safety.
The new requalification methodology utilized existing industry training program standards to develop and administer the examinations.
By adminis-tering requalification examinations that are consistent with the existing facility developed training programs, NRC reduces the impact on the facili-ties and their operators while improving the program assessments.
Each NRC requalification examination included an operating test and a-written examination. Each of these was comprised of two distinct parts.
i 344TH ACRS MEETING MINUTES 44-The first part of the operating test was conducted in a simulator facility.
This allowed the examiners to observe selected control room crews during simulated transients and accident scenarios.
The focus of this portion of the examination was on crew perfomance rather than on individual perfor-mance.
The second part of the operating test was conducted during a plant walk-through.
During the plant walk-through, individual operators were evaluated on their ability to correctly perfom plant tasks that are important to safety. The emphasis of this mode of testing was to ensure that the operators have maintained their understanding of and proficiency in performing selected system tasks. The walk-through was conducted by facility-appointed evalva-tors. The NRC examiners evaluated the examination process, asked questions of the operators as necessary to ensure adeouate system knowledge and job perfomance measurement coverage, and made independent assessments of the operator's performance and the evaluator's examination as they,were adminis-tering it.
The written examination is administered in two parts.
The written exam is an open reference examination administered to assess the operator's knowl-edge of plant systems, trocedures, and operating limits.
The plant opers-tions section is administered in a control room environment und was designed to evaluate the operator's knowledge of plant systems, integrated plant operations, and instruments and controls.
This section was also used to evaluate the operator's ability to diagnose postulated events and to recog-nize limiting conditions of operation as defined in technical specifica-tions.
The procedures section of the examination has an open reference fomat and is administered in a classroom setting.
It was designed to evaluate the operator's ability to analyze a given set of conditions and determine the proper procedural steps and administrative practices to follow.
The opera-tors were given access to the same abnomal emergency and administrative procedures that would be available to them in dealing with similar real-world situations in the control room.
The written examination was developed from the examination question' bank proposed and provided by the licensee.
NRC reviewed and modified the proposed items, as necessary, to ensure - accuracy, clarity, importance to safety and appropriateness for an open reference fomat.
NRC and the facility licensees worked together in developing, administering, and grading the examination.
Mr. Perkins said it takes about nine months of training to become an examin-er.
Examiners take a number of courses at the NRC Training Center and.get on-the-job training working with certified examiners.
Upon successful completion of this process they become certified examiners.
344TH ACRS MEETING MINUTES Mr. Michelson asked what the passing grade was for this type of examination.
Mr. Perkins replied the minimum was 80% for-the two written sectiors com-bined.
For the walk-through, there are 10 job performance tests. Eight out of the 10 have to be perfonned satisfactorily for a passing grade. Critical tasks (about five) have been defined for the simulator test.
If an indi-vidual cannot perform one of those tasks he may fail and will fail if he cannot perform two of the tasks. Mr. Perkins pointed out that, although the i
simulator. testing involves teams, it evaluates individual perfomance, X.
Nuclear Power Plant Operations - Actions Taken in Response to Core Power Oscillation Event at LaSalle Unit 2 (0 pen)
I
[ Note:
Mr. Paul Boehnert was the Designated Federal Official for this portionofthemeeting.]
Dr. Kerr, as Chairman of the Core Performance Subcommittee, introduced this topic to the Committee.
He noted that the oscillation event at LaSalle was unexpected, given the stability analyses supplied by GE.
He said the Comittee would hear presentations from the BWR Owners Group (BWROG),
NRC/NRR, and NRC/RES, reporting on the status of actions taken and planned to address the implications of this event.
Mr. Tom Rausch, Commonwealth Edison, spoke on behalf of the BWROG.
Points made by Mr. Rausch as background included:
BWROG initiated generic studies following the March 1988 LaSalle event and BWROG plans to resolve this issue were discussed with the NRC on June 24, 1988, s
Preliminary results from the BWROG analyses are now available and interim corrective actions have been implemented by all BWR utilities.
These findings and interim actions were discussed with the NRC on November 9, 1988.
BWROG stability program now includes investigations of viable long-tem solutions to the stability issue.
The BWROG has two main programs underway. These are:
Phase 1A - Perform an assessment of postulated large amplitude oscil-lations without scram (ATWS) to 1) determine the response of average core power and 2) confirm that current ATWS analyses are not affected.
Phase 1 - Perform an assessment of plant response to postulated region-al thermal-hydraulic instability.
Identify existing mitigation capa-bility and develop appropriate guidance for operators.
L The 'nalyses supporting the above work were performed with the TRAC-G code which includes full three-dimensional capability.
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344TH ACRS MEETING MlHUTES l The Phase 1A analyses did not reveal any concerns.
For Phase I the analyses showed that the safety limit (SL) minimum critical pNer ratio (MCPR) could be exceeded in some cases.
GE, with the support of the BWROG, has issued interim corrective actions which ban operation in regions of the power / flow i
map where instabilities are known to occur (Figure 2),
in addition, 11cen-sees were instructed to scram if oscillations are seen or if the more restrictive operating regimes are entered. In response to Dr. Kerr, Mr.
4 Rausch said all BWR plants have taken actions to assure that the operator 1
l always knows where the plant is operating on the power / flow map.
In response to Dr. Shewmon, Mr. Rausch said the BWROG has the " authority" to require all BWR utilities to adhere to their directives even if a given utility is not a fomal member of the BWROG.
This authority exists when a given issue being addressed is designated " generic."
Future plans of the BWROG include:
EPRI has been requested to provide critical review of the BWROG analy-sis to assure no important issues have been overlooked.
The BWROG Executive Oversight Comittee has authorized a study of intemediate and long-term solutions to this issue.
The BWROG' Stability Committee met December 13-15, 1988, to fomulate plans and initiate studies.
The BWROG plans to review progress with the NRC in 6 months, and to identify viable long-term solution within 12 months.
Mr. L. Phillips. NRR, discussed the status of the NRR review of the BWR stability issue.
In response to Mr. iiichelson, Mr. Phillips said NRC believes the power oscillations can range up to 3-500% of nominal power
- level, Mr. Phillips indicated that NRR is preparing a commission paper that will provide a status of the actions taken on this issue. NRR sees the following safety concerns:
Forregon-wise /out-of-phaseoscillationsfueldesignlimits(MCPR)may be exceeo0d prior to detection.
The maximus, amplitude of oscillations during ATWS is not known for core-wide /in-phase oscillations.
The effects of continuing large amplitude oscillations on core themal power and the effects on pres-sure during ATWS need to be evaluated.
NRR noted that BNL has been perfoming analyses of oscillation phenomena using the RAMONA 3-B code.
Mr. Phillips discussed the main contributors to instability and the actions taken by Aapanese and European BWR operators to prevent / mitigate the phenomenon.
l 26a INDUSTRY ACTIONS RECOMMENDED INTERIN CORRECTIVE ACTIONS-190 =
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REGION C - MINIMIZE OPERATION 0
SCRAM IF OSCILLATIONS ENCOUNTERED FIGURE 2 i
i 344TH ACRS MEETING MINUTES NRR is calling for more calculations to evaluate the effects of large amplitude oscillations on core thennal power.
In response to Dr. Kerr, Mr.
Phillips indicated that the codes available to the Staff cannot at this time adequately model osciliation behaviort however, he expects the codes will be improved in order to do so.
NRR plans to issue a supplement to NRC Bulletin 88-07, " Power Oscillations in Boiling Water Reactors."
The supplement will approve the_ use of the BWROG interim corrective actions with modifications.
The principal addi-tional requirement is that operators manually scram a plant if a dual recir'tulation pump trip occurs.
NRR expects to receive the BWROG's pro-posals for long-term corrective actions within one year.
Mr. D. Bessette, NRC/RES, discussed a proposed program and schedule for i
research on BWR instability.
He said the issues that RES sees for this program are:
What is the minimum critical power ratio a plant may experience during an instability event?
What effect do control systems and operator actions have?
What is plant rcsponse during ATWS, including effect of ATWS proce-dures?
I Dr. Kerr said he believes that the key safety issue here 15-the impact of i
core power oscillations given an ATWS.
He asked if RES agrees with this point.
Mr. Bessette replied in the affirmative.
Dr. Kerr asked RES to focus the remaining presentation on this point.
RES is conducting an assessment of the relevant codes of interest by use of the Swedish FRIGG test facility data and plant data obtained during the LaSalle event.
RES will apply the CSAU methodology to this program.
In response to Dr. Kerr, Mr. Bessette said he believes the codes will eventually be able to adequately model oscillation events.
RES also indi-cated that it should take about one year to complete this effort.
Dr. Kerr asked if the Comittee had any coments on this issue or if they believed some additional action is required.
After some discussion, it was conluded that the current activities to address this concern appear reason-able and the Comittee would follow the resolution effort.
l XI. Executive Sessions (0 pen)
A.
Subcommittee Reports There were no subcomittee reports given during this meeting.
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i 344TH ACRS MEETING MINUTES )
B.-
Reports, Letteg and Memoranda (0 pen)-
1.
Equipment Qualification - Risk Scoping Study (Report to Chairman Zech dated December 20, 1988)
The i Comittee recomended that the requirements of Regulatory-Guide 1.89,
" Environmental Qualification of Certain Electric-Equipment-Important to Safety for Nuclear Power-Plants " be reevaluated-in light' of the conclusions of the study as to the t
overemphasis - cr the importance of: radiation dose in equipment--
l qualification.
It also noted that the failure rates used in PRAs may not be appropriate for accident environments and-recomended j
that the: implications of this observation be studied further. The-Comittee comented favorably on the use of a peer-review process to review the research work and the: results and conclusions of.
risk scoping studies.
4 i
C.
Other Comittee Conclusions (0 pen) 1.
Vogtle Electric Generating Plant a
The Comittee had, in its report dated August 13,1985, on the Vogtle' Electric Generating Plant, Units 1 and 2, stated that the ACRS would consider the need for further review of Unit 2 if there was a significant delay in the schedule for start-up.
The pro-jected start-up date for Unit 2-has slipped about eight months _
(from June 1988 to February 1989). The Comittee decided that it was not necessary to conduct' additional review of Unit' 2 because' l
of this delay in the projected start-up date.-
2.
Implementation'of Severe Accident Policy for Future LWRs The Comittee decided that time should be scheduled at t. future ACRS meeting for an NRC staff briefing on the implementation of the Severe Accident Policy for future LWRs.
Mr. Fraley subse quently scheduled discussions for a future (tentatively Februar 1989) ACRS meeting.
(Dr.El-ZeftawyhastheactiononthisLitem.
3.
Application of leak-Before-Break Technology The Comittee decided to review the - NRC staff's proposal for a Comission policy statement on additional applications of leak-before-break technology and provide comments on this matter.
The Subcomittee on Thermal Hydraulic Phenomena (D. Ward /P. Boehnert) was given this assignment. Mr. Fraley infonned the Comission' and the Office of the Secretary of the Commission of the Comittee's decision on Thursday, December 15, 1988.
A subcommittee meeting will be scheduled during February 1989 for discussion of this matter.
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L 344TH ACRS MEETING MINUTES.
4.
The Comittee decided that 'it wished to continue - to be kept in-formed of the NRC/ industry plans for conducting. experimental research on B&W once-through steam generator thermal-hydraulic performance.
The Committee. requested a briefing - from the' NRC staff at a future ACRS meeting. (See letter from~ R. Fraley to V.
Stello, dated December 20,1988.)
D.
FutureActivities(0 pen) 1.
Future Agenda The Comittee agreed to the tentative future agenda as shown in.
Appendix II.
2.
Future Subcommittee Activities A schedule of future-subcommittee-activities was distributed to _
members (AppendixIII).
t The 344th ACRS meeting was adjourned at 6:15 p.m., - Friday, ' December 16, 1988.
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'REYisE02: m i 6 1988 344TH ACRS MEETING MINUTES APPENDIX III i
ACRS/ACNW COMMITTEE 8' SUBCOMMITTEE MEETINGS t
5th ACNW Meeting, December 21, 1988, Bethesda, MD, Room P-422, REINSTATED.
Regional ~ Programs, January 5-6, 1989, Region lY Office, 611 Ryan Piara Drive.
Arlington, TX (Boehnert), 8:30 a.m.
The Subcommittee will review the activi-ties under the control of the Region lY Office.
Attendance by the following is anticipated, and reservations have been made at the Hawthorn Suites (tele-phone: 817/640-1188),2401 Brookhollow Plaza Drive, Arlington, TX for i
the nights of January 4 and 5:
Dr.Remick(5thonly)
Mr. Michelson
-Mr. Carroll Mr. Ward Dr. Kerr Improved Light Water Reactors, January 10, 1989, 7920 Norfolk Avenue,.
Eethesda, MD (Alderman), 8:30 a.m., Room P-114.
The Subcommittee will review-the proposed final version of 10 CFR Part 52. Early Site Permits. Standard Design Certification.
Lodging will' be. announced later.
Attendance by the following is anticipated:-
Mr. Wylie (tent.)
Dr. Siess Mr. Michelson Mr. Ward Auxiliary and Secondary Systems, January 11, 1989, c 7920 Norfolk Avenue, Bethesda. MD (Duraiswanty), 8:30 a.m. - 12:00 noon, Room P-422. The Subcom-mittee will discuss Control Air System Design and Operating Experience, and the proposed resolution of Generic Issue 43, " Air ' Systems Reliability."
Lodging will be announced later. Attendance by the following is anticipated:
Mr. Michelson Dr. Siess Mr. Carroll l
Mechanical Components, January 11, 1989, 1920 Norfolk Avenue, Bethesda, MD (Igne), 1:00 p.m., Room P-422.
The Subcomittee will discuss Air Operated ValveTestingandOperatingExperience(includingSolenoidAirControlValves)
L and other related matters.
Lodging will be announced later.
Attendance by l
the following is anticipated:
L Mr. Michelson Mr. Wylie Mr. Carroll Mr. Wohld Dr. Siess 345th ACRS Meeting, January 12-14, 1989, Bethesda, MD, Room P-114.
6th ACNW Meeting, January 23-24, 1989, Bethesda, MD, Room P-114.
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l 344TH ACRS liECTitlG filflVTES APPENDIX II (Rev.)
TENTATIVE ACRS AGENDA-ITEMS -
ilanuary 12-14, 1989 i
Sodium Advanced Fast Reactor (SAFR) (0 pen) (DAW /MME) Estimated Time:
2 hrs.
Complete ACR5 discussion and preparation of. ACRS report on the preapplication
. review of this standardized plant.
3 Fitness for Duty (0 pen)
(FJR/HA)EstimatedTime:
I hr. - Review and report u
on proposed NRC rule regarding fitness for duty of nuclear power plant opera tors.
L Standard Design Certification and Combined Licenses for Nuclear Power Plants (0 pen) (CJW/HA) Estimated Time:
2-3/4 hrs. - ACR5 review and report regarding proposed final version of 10 CFR Part 52 regarding Early Site Permits, Standard Design Certifications, and Combined Licenses for nuclear power plants.
Meeting with NRC Comissioner James E. Curtiss (0 pen) (WK/RFF) Estimated Time:
1 hr. - Discuss items of mutual int. rest.regarding ACRS/NRC activities.
ECCS(0 pen)(DAW /PAB)EstimatedTime:
1-3/4 hr. - Prepare ACRS report on proyosed NRC Code Scaling Applicability and Uncertainty Evaluation MetTodology.
Containment Systems (0 pen) (DAW /MDH) Estimated Time: _ 24 hrs. - Complete ACRS discussion and preparation of report to NRC regarding proposed recomendations for BWR Mark I containment performance and improvements.
NRC Quantitative Safety Goals (0 pen) (DAW /MDH) Estimated Time: 2 hrs. -
1 Complete ACRS discussion and preparation of ACRS report to NRC on the proposed plan for implementation of the NRC's Safety Goal Policy.
Generic Issue 43, " Air Systems Reliability" (0 pen) (CM/SD) Estimated Time:
2 hrs. ' Review and coment on proposed resolution of Generic-Issue 43, " Air Sy; isms Reliability."
Nuclear Safety Research Program (0 pen) (CPS /SD) Estimated Time: i hr, '-
Discuss proposed ACRS annual report to the U.S. Congress on the NRC safety l
research program.
AnticipatedACRSActivities(0 pen)'(WK/RFF/MWL)EstimatedTime:
i hr. -
Discuss topics proposed for consideration by the Comittee.
ACRSSubcommitteeReports(0 pen)(FJR/RFF)EstimatedTime:
li hr. - Discuss anticipated ACRS subcommittee activities and hear and discuss the status of' assigned subcomittee and designated members activities.
NewMembers(Closed)(FJR/NSL)EstimatedTime: i=hr. - Discuss the qual fica-tions of candidates proposed for consideration as nominees for' appointment to the ACRS.
Accident Management (0 pen) (WK/MDH) Estimated Time:
I hr. - Briefing regard-'
ing NRC staff development of a program plan on severe accident management.
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APPENDIX 1 MINUTES OF THE 344TH ACRS MEETING DECEMBER 15-16,-1988 THURSDAY, DECEMBER IS,- 1988
-Public Attendees NRC Attendees Steph'anie>Shorron, SERCH-Bechtel R. Landry, RES
- R.
T.; Lancet.Rockwell International L. Soffer, RES-C. L. Allen, SAIC D. Persinko, NRR-G. J. Van Tuyle Brookhaven Natl.-Lab, w.,Beckner, RES N. Suttora, NUS R. W -Houston, RES G. Sherwood. DOE Gil Brown, NUMARC LL.-Gifford, GE W. P. McCaughey, BGSE i
F..T. Stetsom,'SAID J. Russell, DOE FRIDAY DECEMBER 16, 1988 Public Attendees NRC Staff-Wolfgang Wulff -Brookhaven Natl. Lab.
E'. Beckjord, RES
-Ali Tabatabai, PNL C. Bartlett, RES M. E. Waterman, INEL/EG&G Idaho
-F. Coffman, RES' G. S. Lellouche, SLI D. Persinko, NRR F. C. Phifer SERCH/BECHTEL E. Shomaker, 0GC Art Bivens, NUMARC J. Shea,'GPA Ton Tausch, Commonwealth Edison & BWROG Dave Lange.R I DRS L. Wiens, NRR K.'Perkins, NRRL
.H.. Scott, RES I
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345 339 340 341 342 343
_346 347 ACRS MEETING' DATE,DEC. 15-16, 1988-ATTENMES Thursday-Friday
~ Saturday l
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Dr. William Kerr, Chairman
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Dr. Forrest J. Remick, Vice Chairman
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V' I
Mr. James C. Carroll-
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V' Dr. Harold W. Lewis V
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I Mr. Carlyle.Michelson V'
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'l Dr. Paul G. Shewmon Y
V' f
Dr. Chester P. Siess V
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Mr. David A. Ward I/
V'
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APPENDICES MINUTES OF THE' 344TH ACRS MEETING DECEMBER 15-16, 1989 1.
Attendees I
I I.,
Future Agenda
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111.
Future Subcomittee Activitits.
IV.-
Other Documents Received r
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. Human Factors, January 26, 1989, 7920 NorfolkAvenue,Bethesda,MD(Alderman),
8:30 a.m.,
Room P-422.
The Subcommittee will - review the. Human Factors Research Program Plan.
Lodging will be announced later.
Attendance by the following is anticipated:
Dr. Renick Mr. Michelson Mr. Carroll Mr. Ward Dr. Kerr Mr. Wylie Auxiliary and Secondary Systems, January 27, 1989, 7920 Norfolk Avenue,=
. Bethesda,11D (Duraiswanty), 8:30 a.m. - 1:00 p.m., Room P-422.. The Subcommit '
tee will review the adequacy of the proposed Staff's plans to implement the recommendations resulting from the Fire Risk Scoping Study.
Lodging will be announced later.
Attendance by the following is anticipated:
Mr. Michelson Dr. Siess Mr. Carroll Mr. Wylie i
Mechanical Components, January 27, 1989, 7920 Norfolk Avenue, Bethesda, MD (Igne), 2:00 p.m., Room P-422.
The Subcommittee will review the proposed resolution of Generic Issues 70, "PORV Reliability," and 94' " Low Temperature Over-Pressure Protection," and other related matters.
Lodging will be an-nounced later. Attendance by the following is anticipated.
Mr. Michelson Dr. Siess Mr. Carroll Mr.Wylie(tent.)
Babcock & Wilcox Reactor Plants, February 1 & 2,1989, Sacramento, CA (Igne),
8:30 a.m.
.The Subcommittee will discuss-the lessons learned from the a)proxi-mately 2-year shutdown of Rancho Seco that occurred following the:Decem)er 16, 1985, overcooling event.
Topics for discussion include monitoring extended 1
start-up program as well as plant and organization changes as a result of the restart effort.
Lodging will be announced later. Attendance"by the follwing is anticipated:
i Mr. Wylie Mr. Michelson Mr. Carroll Mr. Ward Dr. Kerr Safety Research Program, February 8, 1989, 7920 Norfolk Avenue, Bethesda, MD l
00uraiswamy), 8:30 a.m.,
Room P-114.
The Subcommittee will discuss the l
vngoing and proposed NRC Safety Research program. and budget.
Lodging;will be
.I i
announced later, Attendance by the following is anticipated:
j Dr. Siess Dr. Shewmon L
Dr.-Kerr Mr. Ward Mr. Michelson Mr. Wylie Dr. Remick 346th ACRS Meeting, February 9-11, 1989, Bethesda, MD, Room P-114.
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7th ACNW Meeting, February 22-23 1989, Bethesda, MD, Room P-114.
1 Occupational and Environmental Protection Systems, March 1-2, 1989,. 7920 Horfolk Avenue, Bethesda, MD (Igne), 8:30 a.m., Room P-114 The Subcommittee.
will di&cus; the general status of emergency planning for nuclear. power i
plants.
Lodging will be announced later.
Attendance by the following is' j
ancicipated:
{
-1 Dr. Renick Mr. Kathren
-i tir. Wylie Dr. Shapiro et al.
347th ACRS Meeting, March 9-11, 1989, Bethesda, MD, Room P-114.
1 Materials and Metallurgy, flarch 15-16, 1989, Columbus, OH (Igne), 8:30~ a.m.
The Subcommittee will review the degraded piping program, including NDE and aging of centrifuga11y cast stainless steel piping material.= Lodging will.be announced later. Attendance by the following is. anticipated:
Dr. Shewmon Mr. Etherington Dr. Lewis Dr. Hutchinson Mr. Michelson Dr. Thompson
-i Mr. Ward 8th ACNW Meeting, Ma"ch 22-23, 1989, Bethesda, MD, Room P-114.
Limerick 2, March 28, 1989, Philadelphia,PA(Quittschreiber),.8:30.a.m.
The Subcommittee will review Limerick 2 for a low power operating license.
Lodging will be announced later. Attendance by.the'following is anticipated:
Dr. Kerr Dr. Siess Dr. Lewis Maintenance Practices and Procedures, March 30, 1989, 7920 Norfolk Avenue, Bethesda, MD (Alderman), 8:30 a.m., Room P-114.
The Subcommittee will review the proposed maintenance rule.
Lodging will be announced later.
Attendance i
by the following is anticipated:
Mr. Michelson Mr. Wylie Mr. Carroll i
Materials and Metallurgy, April 27, 1989, Palo Alto, CA (Igne).
The Subcom-mittee will discuss the status of the following matters:
erosion / corrosion of pipes, hydrogen water chemistry, zinc addition to primary coolant loop and its effects on materials, decontamination effects on materials, and other related 4
matters.
Lodging will be announced later.
Attendance by the following-is anticipated:
Dr. Shewmon Mr. Ward Dr. Lewis Mr. Etherington Mr. Michelson l
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International Conference on Quality. May 14-18.-1989, San Diego, CA (Igne). -
Attendance by the following is anticipated *
-J Dr. Remick M r.1 Ward Dr. Siess et al.
(q Advanced' Pressurized Water Reactors', Date to be determined (January / February),
Bethesda, MD (El-Zeitawy).
The Subcommittee will review the. licensing review; bases document being developed by the Staff for Combustion Engineering's Standard -Safety Analysis Report-Design Certification'.(CESSAR-DC).
Attendance-t by the following is anticipated:
~'
Mr. Carroll Dr. Remick i
Dr. Kerr Dr. Shewmon_
3 Mr. Michelson-Mr. Wylie
~
- Systems, Date to.be determined (January / February),
i Bethesda, MD -(Doehnert).
The Subcomnlittee will continue its review of the proposed resolution of Generic ~ lssue 23, "RCP Seal failures.
Attendance by the following is anticipated:
Mr. Ward Dr. Catton Dr. Kerr Mr. Davis-Mr. Wylie General Elsrfric Reactor Plants (Peach Bottom Restart), Date to be. determined (January / February), Bethesda, MD ( Alderman).
The Subcommittee will review the proposed restart plan for the Peach Bottom Plant.
Attendance:by the following is anticipated:
1 Dr. Kerr Mr. Michelson.
Dr. Lewis Dr. Siess 1
l Joint Core Performance / Thermal Hydraulic Phenomena, Date to be - determined (January /Februa ry), Bethesda, MD (Bochnert/ Houston).
The Subcommittee will review the implications of-the core power oscillation event at LaSalle, Unit 2.
Attendance by the following is anticipated:
Dr. Kerr Dr. Lee
'l
-Mr. Ward Dr. Lipinski' Mr. Michelson Dr. Plesset Dr. Shewmon Mr. Schrock Mr. Wylie Dr. Sullivan Dr. Catton Dr. Tien 1
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AC/DC Power Systems Reliability, Date-to' be determined (February), Bethesda,.
MD (El-Zeftawy).
The Subcommittee will review the-proposed resolution of teneric Issue 128, " Electrical Power Reliability." Attendance by the follow-ing is anticipatei:
Mr. Wylie Dr. Lewis Mr. Carroll Mr. Davis Dr. Kerr Dr. Lee Instrumentation and Control Systems, Date to be determined (February / March)',
Bethesda, MD (El-Zeftawy).
The Subcommittee'will review the proposed resolu-tion of Generic Issue 101, " Break Plus.. Single failure in-BWR Water Level Instrumentation." Attendance by the following is anticipated:
s Dr. Kerr Mr. Wy. lie Mr.. Carroll Mr. Davis Dr. Lewis Dr. Lipinski.
Mr. Michelson Extreme External Phenomena, Date to be determined '(February / March), Bethesda, MD (Igne)..
The Subcommittee will review planning documentr on external..
1 Events. Attendance by the following.is anticipated:-
Dr. Siess Mr. Michelson Dr. Kerr Mr. Wylie
]
Dr. Lewis Instrumentation and Control Systems, Date to be 'etermined (March), Bethesda,
{
d MD (El-Zef tawy).
The Subcommittee will-review the ATWS rule implementation status. Attendance by the following is anticipated:
l
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Dr. Kerr Mr. Wylie l
Mr. Carroll Mr. Davis l
Dr. Lewis Dr. Lipinski.
l Mr. Michelson Advanced Pressurized Water Reactors, Date to be determined (April), Bethesda, MD (El-Zef tawy). The Subcommittee will discuss the comparison of WAPWR (RESAR L
IP/90) design with other modern plants (in U.S. and abroad).
Aftendance by L
the following is anticipated:
Mr. Carroll Dr.'Remick Dr. Kerr Dr. Shewmon Mr. Michelson Mr. Wylie 4
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Plant Operating Procedures, Date. to be ! determined '(spring), Bethesda, MD.
(Igne).- The Subcommittee will review the status of the NRC program on Techni-cal Specifications update. Also, it will review an anonymous letter to Ms. E'.
Weiss (Union of Concerned Scientist), dated. Sept. 27, 1988, on Technical Specifications inadequacies. Attendance by the following is anticipated:
Mr. Michelson Mr. Ward
~
Mr. Carroll Mr. Wylie Mr. Remick-Materials and Metallurgy, Date to be determined (2nd or 4th week of May),
Bethesda. MD (Igne).
The Subcommittee will - review low upper ' shelf fracture energy concerns of reactor pressure vessels.
Attendance by 'the following is anticipated:
Dr. Shewmon Mr. Ward' Dr. Lewis Mr. Etherington Mr. flichelson Decay Heat Removal Systems, Date to be determined, ' Bethesda. MD (Boehnert).
The Subcommittee will explore the issue of the use of feed and bleed for decay heat removal in PWRs. Attendance-by the following.is anticipated:
Mr. Ward Mr. Wylie Dr. Kerr Dr. Catton Mr. Michelson Mr. Davis Thermal Hydraulic Phenomena, Date to be determined, Bethesda, MD (Boehnert).
The Subcommittee will discuss the status of Industry best-estimate ECCS model submittals for use with the revised ECCS Rule. Attendance by-the following is anticipated:
fir. Ward Dr. Catton Dr. Kerr Dr. Plesset Mr. Michelson Mr. Schrock Mr. Wylie Dr. Sullivan Dr. Tien Auxiliary and Secondary Systems, Date to. be determined,
- Bethesda, MD (Duraiswamy).
The Subcommittee will discuss the:
(1)criteriabeingused by utilities to design Chilled Water Sy(3) criteria being used by the NRCstems, (2) for Chilled Water Systems design, and staff to review the Chilled Water Systems design. Attendance by the following is anticipated:
Mr. Michelson Mr. Wylie 1
Mr. Carroll l
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APPENDIX-IV-MINUTES OF THE 344TH ACRS MEETING DECEMBER 15-16, 1989 l
l MEETING NOTEBOOK Tab 2
SAFR NUCLEAR POWER PLANT -
Slides used by the speaker during the presentation Table of Contents, Tentative Agenda Status Report with Attachments Att. I:
Letter to V. Stello, NRC-from Mr. D. F. Bunch, DOE, dated August 15, 1988 re selection of SAFR.and PRISM for continued l
Department'[ofEngergy] support Att. II:
Letter to.T. J.-Garrish, DOE from V. Stello, NRC,.
dated August 17, 1988, re two issues developed during NRC review-of three cdvanced reactor conceptual designs, MHTGR, PRISM,1 and SAFR Att. III:
Memo for NRC Commissioners from V. Stello, Subj.:-
COMMISSION ACTION ON THE KEY-LICENSING AND' STANDARDIZATION ISSUES ASSOCIATED WITH THE DOE ADVANCED REACTOR CONCEPTS (SECY-88-202'and SECY-88-203),. dated August 18, 1988 3
CONTAINMENT.. SYSTEMS-- Table of Contents, Tentative Agenda l
Slides used by the-speaker during the presentation' Status Report with Attachments l
Att. I: 'ACRS report of December 17, 1986 l
Att. II:
draf t copy of SECY on Mark I-undated (INTERNAL COMMITTEE l
USE) i Att. Ill:
Draft Generic Letter undated Att. IV:
Selected Slides Used by RES at Dec. 6, 1988 Meeting of Containment Systems Subcommittee (Mark I issues) l Att. V:
Selected Slides Used by NRR at December 6,1988 Meeting of Containment S/ stems Subcomittee. (Overview of EPG Revision 4 Details on Venting) 4 EQUIPMENT QUALIFICATION-RISK SCOPING STUDY - Table of Contents, Tentative Agenda Status Report: Memo to ACRS Members and ACRS Technical Staff from-S. Duraiswamy, ACRS Staff, Subj.:
STATUSLREPORT - EQUIPMENT QUALIFICATION-RISK SCOPING STUDY - 344TH ACRS MEETING',
DECEMBER 15-17, 1988,.BETHESDA MARYLAND, dated November 29 1988 Executive Summary - Equipment Qualificaton-Risk Scoping Study General Conclusions /Recomendations - Equipment Qualification Risk-Scoping Study-Discussion of Peer Review Comments - Equipment Qualificaton-Risk l
Scoping Study 5
QUANTITATIVE SAFETY GOALS - Table of Contents, Tentative Agenda-Status Report Slides used by the speaker during-the presentation l
Certified Minutes of Safety Philosophy, Technology, and Criteria y
(SPT&C) Subcommittee Meeting on September 1, 1988 i
y D_
344th ACRS Meeting IV-2 ACRS Report of 4/12/88 Draft Plan, " Implementation of Safety Goal Policy," received December.7, 1988-7
. REVIEW OF NRC RES CODE SCALING ~ APPLICABILITY AND UNCERTAINTY'~
(CSAU) METHODOLOGY --
Slides used by the speaker during his presentation Table of Contents, Presentation Schedule, Status Report-ACRS ltr.12/16/87 ACRS ltr, Prop Rev._ECCS Rule, May-10, 1988 Memo to Paul Boehnert from Ivan Catton, ACRS Consultant, Subj.:
TECHNICAL PROGRAM GROUP ~(TPG)-MEETING, DEVELOPMENT OF CODE
. SCALING, APPLICABILITY,ANDUNCERTAINTY-(CSAU) METHODOLOGY, NICHOLSON LANE BUILDING, SEPTEMBER 27-28,;1988-(INTERNAL COMMITTEE USE) 8.1 FUTURE ACRS ACTIVITIES - Memo for ACRS Members from~R.- Fraley, i
Subj:
FUTURE ACRS ACTIVITIES - 345TH ACRS MEETINGJ > JANUARY.
12-14, 1989, dated December 14, 1989 w/ attachment (FutureAgenda) 8.2 empty 8.3 Memorandum for ACRS Staff and ACRS Fell;-
from R. Fraley..Subj.:
ASSIGNMENT OF ACNW/ACRS RESPONSIBILITIES, dated November 23 1988 with Attachment (Chart " Distribution of Responsibilities,"
Revision 2:
November 23, 1988 10 OPERATOR REQUALIFICATION Slides used by the speaker'during-his-presentation Table of Contents, Tentative Agenda ES-601, " Administration of NRC Requalification Evaluation" 11 BRIEFING ON ACTIONS TAKEN IN RESPONSE-T0 LASALLE CORE' POWER OSCILLATION EVENT Slides used by the speaker during the presentation Table of Contents, Agenda, Status-Report Memorandum for W. Kerr from P. Boehnert,
Subject:
NRC Augmented Inspection Team (AIT) Report:
LaSalle Unit 2: Core Power Fluctuations Event of March 9,.'1988, dated May 25, 1988 AE0D Special Report:
"AEOD Concerns Regarding the Power Oscillation Event at LaSalle 2,"
NRC Bulletin No. 88-07: Power Oscillations in Boiling Water Reactors (BWRs), dated June 15, 1088 Memorandum for V. Stello from Chairman Zech,
Subject:
' Power Oscillations Event of March 0,1988 at LaSalle 2, dated July 5, 1988 Memorandum for W. Kerr from P. Boehnert,
Subject:
NRC Meeting with BWR Owners Group Representatives:
LaSalle Core Power Oscillation Event - June 24, 1988, Rockville, MD, dated July 15, 1988 Memo from P. Boehnert to W. Kerr/D. Ward-Report on NRC/BWROG Meeting of November 14, 1988 on T/H Stability (PROPRIETARY)
............q eTI 344th ACRS Meeting Minutes 1V :
q I4.1 REPORT ON PLANNING SESSION FOR INTERNATIONAL CONFERENCE ON QUALITY l
IN THE NUCLEAR POWER INDUSTRY Table of Contents Hemorandum for C. P. Siess from-H. Alderman,
Subject:
Planning Session-for the International Conference on Quality in the Nuclear
- Power Industry 14.2 REPORT OF-VISIT'TO FERMI-2 PLANT-
-Table of Contents, Status Report 4
15.1 ACRS Activities l
Memorandum for ACRS Members from M. Libarkin, Subject
- Information on Topics-Discussed and Time Spent by ACRS, h
dated December 8, 1988
]
1 MEETING HANDOUTS Agenda M.
Item 1
6.1 Schedule i
Memorandum for ACRS Members and Staff from S. Duraiswamy, 1
1 dated December 1, 1988, Status Report - Meeting with the Director of the Office of Nuclear Regulatory Researach (RES)
- 344th-ACRS Meeting, December 15-17,-1988, Bethesda,-
a Maryland with the
Attachment:
. NRC Announcement No.-118, dated July 6, 1988,
SUBJECT:
REORGANIZATION OF:THE OFFICE OF q
NUCLEAR REGULATORY.RESEARCH 2
December 13, 1988 Trip Report from Dr..Kerr on^ Visit to Korea-3 7.0
- 1. Catton Report on' December-7, 1988 T/H Phenomena Subcomittee Meeting (INTERNAL COMMITTEE USE)
Working Minutes of December 7, 1988 T/H' Phenomena Subcomittee Meeting -(INTERNAL COMMITTEE. USE) 4 3
- 1. Catton, Consultant's Coments on Mark i Subcomittee.
Meeting of December 6, 1988, dated December 11, 1988 (INTERNAL COMMITTEE USE)
Latest version of Staff's proposed generic letter on BWR Mark Is, obtained on December 14,1988(INTERNALCOMMITTEEUSE) 5 15.1 Memorandum for Comissioners from-V. Stella, EDO,
Subject:
Implementation of Severe Accident Policy for Evolutionary LWR Designs, dated December l', 1988
en-
,.- g; is t n
'344th ACRS Meeting. Minutes!
IY-4L 6
2
. Memorandum for D. Ward from Stewart W. Long, ACRS Fellow
Subject:
Comments on SAFR and PRISM Design Features in Support of Upcoming Subcommittee Meeting, dated 9 December 1988 7
7.0 Memorandum for D. Ward,'from P. Boehnert,
Subject:
_ Response of T/H Phenomena Consultants to your Question-Concerning
~j Usefulness of T/H Codes, dated December. 14, 1988-l 8
8.2 Memorandum for ACRS Members from R. Fraley, dated December 14, 1988,
Subject:
- Future ACRS Activities - 345th ACRS Meeting - January 12-14, 1989 i
a i
1
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