ML20126E564

Summary of 760318-19 Meeting W/Util & NUS Corp in Charlevoix,Mi Re Review of ECCS Performance.Attendee List & Generic Info Request Re ECCS Review in Electrical, Instrumentation & Control Areas Encl

ML20126E564
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	04/08/1976
From:	Reeves E Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8102050310
Download: ML20126E564 (12)

Text

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APR 0 e 1976 Docket No.. 50-155 LICENSEC: Consumers Power Conq2any FACI LITY:

dig Rock Point Sul4%RY OF MEETING HELD ON WdtOi 18-19, 1976, REGARDING EMERGENCY Coil COOLINO SYSTini On March 18-19, 1976, representatives of Consumers Power Company (CPCo) and their consultant, NUS Corporation, met with a task force of the Nuclear Regulatory Commission (NRC) staff in Charlevoix, Michigan. A list of attendees is attached. NRC requested the meeting to review the Emergency Coro Cooling System (ECCS).

Information required to complete our LCCS review in the electrical, instrumentation and control areas are shown in Enclosure 1.

Other concerns relating to the existing Dig Rock Point ECCS are discussed below.

Following is a status of the discussion hold and the CPCo responses

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(note.

the first 13 responses are numbered to coincide with the questions in Enclosure 1):

1.

Paragraph 3.2 of the February 27. 1976 CPCo report lists ten short-term single failure recommendations.

Items 5, 6, 9 and 10 fail to moet the criteria of IEEL Std 273-1971. CPCo agreed to respond but stated the requirement to meet the standard appears unnecessary at Big Rock Point.

2.

Item 7, paragraph 3.2 of the February 27, 1976 CPCo report discusses clistination of the emergencyddiesel generator's dependency on the station battery. CPCo will bring this into coy 11ance with IEEE Std 279-1971.

5.

LOCA environnent, s.eismic e information and potential high energy lines breaks were discussed. Although these items were previously addressed by CPCo they agreed to provide additional information on these subjects.

4.

Submerged LOCA equipment was previously reviewod and documented in a fmy 15,197S CPCo report. Other submerged equipment which might affect LCCS power busses will be reviewed by CPCo.

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Previously identified valves MO-70W (spray crossconnect) and'!iG-7072 6

(long-term cooling heat exchanger bypass) do not conform to branch 6.

Technical Position 131 CSS 18. ' CPCo agreed to provide an evaluation of this subject. Mechanical blocks for each valve will be considered.

The station transfomer fire systen. deluge valve failure probability.

is considered extrencly low by CPCo. The NRC staff task force advised snat an ECCS actuation overrido should be considered.

The task force made an onsite inspection of the entire plant fire protection syster,i.

7.

Previously identified need for an electrical interlock for enclosure spray valves M0-7064 and MD-7068 was discussed. 'Ihc NRC staff noted that corrective action would be relatively easy and should be considered.

CPCo agreed to consider this matter and advise NP.C of its conclusions.

8.

During a tour of the facility it was observed that the physical separation criteria. may not be met for some cabic and wiring. CPro noted that the separation criteria for FCCS is as varstated in their 1071 submittal for the redundant core spray systen. All wiring for the backup core spray syste u is routed in metal conduit, physically and electrically separate from the ring core spray systen. CPCo agreed to review this matter and provide additional infomation.

9.

CPCo advised that these questions were answered in the natrix of L

Table 4.1 and Table 5.1 respectively of Attachment 2 to the February 27, 10.

1976 CPCo report. NRC agreed.

11.

'Inis question was iaodified by HRC to include not only the dicsol l

generator but also the diesel fire pump. CPCo agreed to res ond but c

stated they considered noncoupliance with i&C Branch Position EICS317 to be one of the current hRC guidelines not applicable to liig Rock i

Point.

12.

It was mutually agreed that Bi;; Rock Point does not cowply with !mC i

Uranch Position LICSb 27.

NRC explained currently acceptanle alterna-tives for the five valves in question.

CPCo agreed to evaluato and provide c response to this concern.

13.

CPCo's risk analysis, Attacluncut 4 to the Pobruary 27, 1076 CPCo report i

was discussed in considerabic detail.

CPCo was advised that overn]1 reliability of the system must includo considerctions of the test and maintenance unavailabilit'res. connon modo failure considerations and

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human error considerations. Secondly, the defenso in depth concept including possible use of the feedwater systen for emergoney core cooling should be addressed. CPCo agreed to update their previous submittal by letter.

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14.

Recent General Electric information 'relati ' g to effectiveness of the nozzle core spray was discussed.

CPCo was advised of the current status and requested to study the effect of this infor-mation on their nozzle core spray effectiveness, CPCo agreed to evaluate this subject and provide the results of their review to the NRC.

15.

An inconsistency was noted by the'NRC task force between the current Big Rock Point technical specification relating to the LOCA analysis p. essure of 1350 psig and the specification which would allow operation to 1485 psig.

CPCo agreed to evaluate this f

inconsistency and advise us of their conclusions.

16.

CPCo's November 26, 1975 submittal relating to the core spray coefficients correlation between the 7 x 7 and 11 x 11 assemblies was discussed.- CPCo advised that the information'was included in an EXXON report submitted to NRC. CPCo agreed to advise us of the date of the report.

17. NRC requested information relating to the fire protection system.

The above ground and below ground piping system may be subject. to passive failures..CPCo agreed to provide piping specifications, pipe schedules and test requirements.

i 18.

NRC requested additional information regarding the' ring core spray system which was installed in 1961. Specifically,'the pipe schedules, and the detailed valve specification or design' class for No-7051 and the adjacent check valve was requested.

CPCo agreed to provide the in formation.

During telephone conversations. following the meeting, NRC asked and CPCo agreed to respond to the following:

1.

Since the deluge valve is near the station transformer, would a i

break in this valve cause a' loss of all offsite power?

I 2.

Clarify that the break sizes analyzed included the correct calculated break size for the 24 inch reactor recirculation pump suction line.

3.

Consider diesel generator single failure in relation to ECCS l'

acceptability. Although this matter has been addressed by NRC previously, NRC considers it appropriate to reconsider the matter at this time, l

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4-r CPCo agreed to respond to the outstanding NRC concerns described above by March 26, 1976, if at all possible.

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Edward A. Reeves, Project Mannger Operating Reactors Branch #2 Division of Operating Reactors Enclosures-1.

List of Attendees 2.

Generic Infromation Request for Review of ECCS in the Llectrical, Instrumentation and Control Areas

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h. b. Sewell nuclear Licensing Administrator Consurprs Power Company 212 West Michigan Avenue Jackson, Michigan 49201

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GL:RIC Ilif 0PJ'\\T10N REQU::ST FOP. RE'JIEW 01-LCCS IN lilE ELECTRICAL, 1

IN5'idU.E TAT 10N AND CONTRutS ALLAS l

'lhe Acceptance Criteria for Emergency Core Cooling Systems for Light Kater Nuclear Power Reactors,10 CFR Part 50.46, requires that an analysis of possible failure modes of ECCS equipment and of their effects-on ECCS perfort:.nce be performd.

This analysis should deronstrate that i

i your ECCS and supporting subsystems ncet the. single failure criterion.

i We require that documentation of this analysis be provided in sufficient detail to cnnble the staff to fli verify that the analysis de=nstr

t es t h at the ECCS and supportin;; subsyster.s-nt. the single fnilure c:iterion l

l as def]ned in 1LI:E Std 279-1971, and (2) date:::,ine the acecptabili ty

.r. c ve ri f;, the itzpletx nta t ion of..ny p roc os t ! d.... nodi fi cat ien re, oi r. _

l a result of your analysis.

Therefore, we vn.ui re that the followin; infer-cat i on 1"> sts::i t ted t o s uyort the sin;.le f'.rilure analysis of the ECCS md support ina, subsyste: -

1.

J a a ti fy any 'noneenforr.ance o f the ail;,.. of tav LCCS actuation h stu 01

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Descr:be hi tn the s' ingle f ailure requi rc.cn any chan~cs prc?os:' for n2etir:,

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Id. :tify any nonconfo::mnce of the J.t :

cf the onsite eter;;cncy fa lure requirc=;nts of pou r system, a-c c.:w d-c, with the s.1 s

IEEE Std 279-1971.

Describe any changes preposed for meeting thesc requirements.

3.

Identify all the electrical equipment required for. the ECCS and i

supporting subsystems to enable periorr.ance of the ECCS safety' function.

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1 Define the qualification status (ability to withstand the design basis seismic and environmental conditions) of this equipt.ent, and the basis for such qualification, to provide.reasonahic -

assurance that the equipment will be capabic of perforning its safety function.

Describe any proposed design nodifications, analyses, or test programs for meeting the environmental and sei smic qua li fi ca tion requirc::xnts.

4.

Identify all electrical equipnent, both sa fety and r.on-sa fety, that ray beco:2 sub cr;ed as a result of a LOCA.

For all such equipment that is not cu.ilified f 9r t crvice in such t.n envircn-tent, provide an r.nalysis to determine the follouing:

(1) the safet, 3.icnificc.nce of the failure of the cquipter.: (o.g., spurican a

operatica, W s of function, loss of accident / pest-accident i onit oring,

I et c.) as a result of flooding, (2) the effects of Class IE elect rier.1 power sources sc rving this equipment as a result of such failurer, and (3) the prcpered de/ipt changes resulting fren your analyrir.

Your respc-nse to i tm (2) rhould specifically cadress breater and fuse coordinaticn ana tne isolation capabilitics of this aspect c:

your design.

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5.

Identify any sin;;1e electrically opcrated fluid system component, including uanually-controlled electrically-operated valves, whoso failure could result :in loss of capability of the hCCS to periore its safety function.

Failure in both the " fail to function" sense and in i

the " undesirable function" sense should be considered, and this should i

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apply even though the component may not be required to function in a given safety operational sequence.

6.

With regard to the equipnent identified in item (5), provide a i

detailed description of any proposed design changes deemed necessary by your analysis for meeting the single failure criterion.

Your response should specifically address but should not be limited to chan;;es inade to : et the single failure criterion by confor..nce to Branch Teenad eal Position LICSE IS, " Application of the Single l aj 1ure Crite: ' ' to.'imual]y-cont ro11c, L lec t ri ca lly-Ope ra t e d Ya l ve.s", o f.ip:. ' ', 7A of t',e l'.ej.ul a t or Standa rd f.evi c.

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'lhis position est:.blishes the necept:.bility of discennectinp. p; s r to the electrier:! ccaponents of a fluid systen as one rean.s of 1.2ctini the sint e fai]ure cri tc rj on.

l 7.

Idtntify ar,y electrical interlocl:s between redundant portions of the ECCS and suprortin a subsystems.

D tine the consequence of failu:e of h e c:,,J.,ili ty of ths : CC5 to perferm its safety any interloc:. :

functica, 1: scribe ra; preposed desi;n :.i:Jific::tions rer.ultin: ' '

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8.

Provide the electrical and physical separation criteria for your i

design of redundat sa fety equipe.cn't and functions.

Include the features in your design that minimize the vulnerability of the ECCS and supporting subsyste..:s to col =on failure uodes.

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9.

In reference to the NUS Corporation report (5069-P-01, Rev. C) entitled " Big Rock Point Nuclear Power Plant ECCS and Associated Systems Evaluation" dated February 23, 1976, the Failure !: ode and Effects Anclysis (Fi.:EA) identified 37 electrical sipgic failures which result in unacceptab]c ECCS or associated systems operating conditions.

The irpact of these failures on the ECCS i

and.i t s associ r t -

rystc :s focused only en 10 explicit systen features which require design todifientica due to unacccptable sinr,1c failurcs.

Three of these 10 SUS desir,n todification L

re co:n.c n d at i on; are bein:; irpic;..cnted durin; the present plant shutdown while the rc~ninin:, 7 reccc ended nodifications Imy or may not be innlemented dependin:/ on Plant Life i!xenpt ion Action.

1 In sum.ary, E ef 37 identified electrical single failures whi ch result in unacceptable ECCS or associated systems operating conditicr.c are bei ng i:rgic::..nt ed.

Provide justification thy the remaining 20 e]cetrical' sin.de failures id2ntificd in the II.uA.thould not be corrected aaJ the r;ss.Iting impact cn the risk analysis as n result of not implencnting corrective action.

10.

As shown in Figure 5 (Attachment 9), of your February 27, J976 Report on Evaluation of.\\deque.cy of ECCS, the station battery system loads include, but are not licited to, the plant annunciators, ECCS valves 110- 7051/F10- 7061, ESS valve F10-7061, and the emergency diesel generator 1

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-S-and its associated breaker,(I A-2B and 2A-2B). controi circuits.

Identify any nonconformance of the station battery system design and any changes proposed for meeting the requircr:.cnts of GDC 17 and ]EEE Std 50E-3971 for the ECCS and its associated systems with regard to sufficient capacity, capability and reliability to supply the required distribution system loads in, the event of.a 1.0CA.

Dis-cuss schether the statica hnttery syster has been si:ed to acco:modate l

the added non-ra:'ety-relatcJ loads during crerrency conditier.; and if not, t.hether automatic disconnectien of those non-safcty-related lo:.ds upon d:

.c.in, of the emer.7ency coni.i t ion is provided.

11.

Pro' vide a descriplic n of the dicsc) generator protection syster. and any nonconfor: mee of this design with regard t o Iiroviding prot ection r

a,,ainr t st ucj ou. trips as the posit ion described in Appendix 7A of the negulatot Staadard P.eview P]cn (EICE 17).

12.

Provid> a description of ycur desir,n criteria for therc.a1 overload protectien for

,3 tors of '.otor operated valves in the ECCS cad cc this to the Positie (El(5!; 27) found in Appendix 7A of the Repuletcry

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13.

The risk analysis presented in Attachment 4 of the February 27, 1976 ECCS report failed to consider common raode failures, 1l hen considering, 3

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the ' risk of failure to achieve ECCS perform::nce within FAC limits due to single failurcs, common node and dependency con-siderations must be considered.

Modify your rish analysis

- accordingly to include the. effects of cc:non modo failu);cs.

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MEETING Sit'J1ARY DISTitIntJI10N:

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