ML20062B362

Discusses Operation W/One Loop Out of Svc,Per 900301 Feedwater Pump Trip & Manual Recirculation Pump Trip

ML20062B362
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	10/15/1990
From:	Beckman W CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9010240289
Download: ML20062B362 (3)

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1%nt Manager MM AIEMEAF5 Peasarm

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. Nuclear Regulatory Commission Document Control. Desk Washington, DC 20555-DOCKET 50-155 - LICENSE'DPR BIG ROCK point PLANT.-

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.0PERATION WITH ONE LOOP OUT OP. SERVICE During. a'.recent feedwater pump trip event and subsequent manual recirculation i

' pump' trip.at Big Rock Point on March 1, 1990, some questions,and concerns were raised dealing with the. isolation requirements contained ^in our'pla'nt operating procedures.^- 'The current procedures state that extended operation.

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-(greater.than'one hour):with one loop idle requires:

'l) The' suction, discharge, and discharge.-bypass valves of the inactive loop to be! closed and caution tagged.

12). The. inactive pump motor breaker to be opened and lowered.

3) 'A determination of the maximum allowable reactor power to that-permitt'ed by: Technical-Specifications, for one. loop ~ operation.

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'These' guidelines were established during resolution'of Systematic Evaluation-J i

Program (SEP)' Topic IV-l'.A:- Operation With Less-Than AlloLoops in Operation..

As discussed in NRC. letter dated-August 9, 1979-(Attachment 1),.the' analysis used'to. justify > operating in the N-1 mode assumes that-the inactive loop is isolated. allowing'no bypass flow through the inactive. loop. The'second j

restriction dealt'with concerns associated with a cold water accident.. This

'l led to the reconunendation that ;the power to 'the inoperative pump be locked.out

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-during N-liloop operationL to provide additional' assurance that an inadvertent E

. cold water injection accident will'not occur. This' evaluation'also noted that j

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thistrestriction may be-removed bycperforming a reanalysis of the cold water 3

accident'and by obtaining the= acceptance of-that analysis from the NRC staff.

In a'recent review-of these operating restrictions,-a concern regarding

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l-b reliability of the recirculation pump seals has'been raised.

If a pump is

-removed from serviceifor other than a seal failure, completely isolating the

. pump can' lead.to-sea 15 degradation during restart.

If the plant operators are v

-unable to start the inactive pump within the one hour limit, closing the suction ~ valve in addition to the discharge and discharge bypass valves will cause.the pump shaft seals to begin to depressurize. When pump restart is h

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Big Rock Point Plant.

4 Op.With One Loop Out of Service October 15; 1990.

F conducted,,the pressure transient on the pump seals may cause'them to seal improperly initiating shaft seal leakage leading to-pump isolation and unit derate. Leaving the suction valve open maintains system pressure on the seals

during shutdown, minimizing the transient affects during restart.

LIn order to reduce the potential-for seal degradation, the operating

. procedures for one pump operation have been revised tot

1) Close and caution tag the discharge and discharge bypass valves of the inactive loop.

2) A determination of the maximum allowable reactor power permitted by Technical Specifications for one loop' operation.

g Justification for these changes is contained in the attached 10CFR50.59 W

evaluation (Attachment 2); however, it can be'susunarized as follows:-

1) Closing the discharge and' discharge bypass valves alone prevents

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reverse flow through the idled-pump. The suction valve,would only-be redundant to the discharge;and discharge bypass valves.

'2) The powe'r lockout requirement of the idled 11oop was based upon'not having an accepted analysis for single Icop operation with restarting.

the idled pump-(cold water injection). The analysis was performed for SEP' Topics XV-1, XV-3, XV-4, XV-5, XV-7 and XV-9 and submitted on July 15, 1981.- In a' letter dated April 7, 1982 (Attachment 3), the staff reviewed-and accepted the analysis.

Although. changes made to plant procedures pursuant to 10CFR50.59 do not

require NRC approval,fConsumers. Power Company felt it was appropriate to-u 1 notify the staff of,these changes. Although none of the initial procedural controls-discussed-_ earlier appear in the plant Technical. Specifications, they do appear in the staff's Safety Evaluation for Amendment No. 48 dated

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. September 21,-1981 (Attachment 4).

%l MdrA W L Neckman-Plant Manager y

.CC. Administrator,-Region III, USNRC y

HRC Resident. Inspector - Big Rock Point.

Attachments 1

OC0990-0074-NLO4 l

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.i ATTACHMENT-1

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Big' Rock Point Plant-

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Docket 50-155 I

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. i, NRC LETTER DATED AUGUST 9, 1979 l

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NUCL M LIC!N!

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August 9,1979 d

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- Docket'.No. 50-155 ~

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U Mr. David Bixel

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l Nuclear Licensing Administrator l

Consumers Power, Company -

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-212 West' Michigan ~ Avenue H

' Jackson, Michigan 49201 4

Dear Mr.- Bixel:

RE: -TOPIC IV-1.A' bPERATION WITH LESS THAN ALL LOOPS IN OPERATION -

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4 Enclosed is a copy of our: revised: safety-assessment of Topic IV-1.A, 10peration With Less Than All Loops In Operation.

This revision supersedes r

y the evaluation l issued by our letter dated August 17, 1978.

1 This revision completes ourl assessment of Topic IV-1.A which _will.

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'be used as' input to the integrated-review of the Big Rock Point Plant; 1

y However,11t'should be: noted that;the acceptability of this topic.

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,eyaluation is-contingent:upon your agreement to (1): include in a

procedure for N-lDloop operationia statement that'the bypass: and-isolation-valves'intthe inactive loop be closed during N-1 operation,

(2)Jphysically lock-out power to the inactive pump, 'and (3)Jincorporate

!the MAPLHGR. limits!for_ N-1 loop operation in. the Technical Specifications.

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If there are'any errors in:the facts of this revised assessment, please j

supply corrected infonnation-and your response with respect to: items (1)-

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ithrough:(3)_ above within-30 days of the date you receive this letter. '

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l Sincerely, 1

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Y Dennis L. Ziemann lef Operating Reactors: Branch #2 Q

Division of Operating Reactors a

Enclosure:

Revised Assessment for N'

Topic IV-1.A!

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"See next page

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2-August 9;-1979 y={, LMr' David-Bixel m

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.A Mr. Pauli A. Perry, Secretary UEh l Consumers.PowerCompany.

J 212 West' Michigan Avenue.

' Jackson.. Michigan 49201 y

~'Judd L. Bacon, Esquire 3

Consumers Power Company 212: West Michigan Avenue:

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? Jackson,- Michigan. 49201 b

i Hunton & Williams George C. Freeman,LJr., Esquire

' P. O. Box 1535 1

Richmond, Virginia : 23212 Peter;W. Steketee, Esquire 505 Peoples Building -

L Grand Rapids : Michigan'.,.49503 a:

a l Charlevoix Public Library..

107.Clinton Street

.:Charlevoix,: Michigan 49720 L K'M C 'Inc.

LATTN:L'Mr. Richard E. Schaffsta11 11747 Pennsylvanial Avenue, N. W.-

Sulte 1050 3 Washington,L 0 l C. J20006 '

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-SYSTEMATIC EVALUATION PROGRAM:

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Topic IV-1.A; Operation With less Than All Loops In Service Plant: _ Sig.R_ock Point'

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Discussion a

Th'e majority of the; prisently operating BWRs. and 'PWRs are designed-a;

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tolpermit operation with'less than full reactor coolant flow. That' 3

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is, if; a' PWR reactor coolan't pump'or a BWR recirculation pump becomes-inoperative, the' flowIprovided by the remaining loop-or <1 oops is,

7 sufficient for steady state operation at.some definable power level,.

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.usually less than' full power.

Plants authorized for long' ternt operation with one reactor coolant:

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pump' ou't of. service have submitted, and the staff has: approved,.

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.the necessary ECCS, steady. state, and transient analysis. The L

~ remaining 'PWR and BWR? licensees!have -Technical Specifications which

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require: reactor shutdown within/ 24. hours if one of.' the operating

,loopsTbecomes; inoperable.and cannot'be returned to operation within'-

E he-time' period.

t In a letter dated August 17, 1978, Consumers Power Company (the licensee) was 'sent draf t evaluations of eight essentially complet'ed 1'

-Systematic Evaluation Program (SEP) Topics. We requested that the licensee review and verify that the information was factual-and

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that all documentation cited was: current. Topic IV-1. A, O' eration p

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With less Than All loops In Service, was one of the eight essentially completed for Big Rock Point. The assessment stated that at.nori-zation' for N-1 loop operation lis'provided'(Technical Specification 4.l.2(b)); however..there is no supporting ECCS analysis to -justify 4

this' mode of operation.

By'1etter ' dated.-October 24, 1978, the licensee responded to our -

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. August 17,1978 request and provided comments concerning the

. correctness of-our assessments. With regard to Topic IV-1,A, the' licensee took exception to our assessment and coricluded that opera-tion with.less ;than all loops in service at Big Rock Point was justified. This conclusion was based on an analysis performed by.

.GeneralElectricCompany-(GE)inearly1977. The licensee derived

,y and implemented operating limits for Big Rock Poin.t based on this

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The October 24, 1978 letter from the-licensee-contained thr.ee

-1 enclosures:

(1) Consumers Power Company statement concerning N-1 j

loop operation, (2) the GE Single Loop L'OCA Analysis, and (3) a 9

document entitled Operation, of.the Big Rock Point Reactor With One

Loop Out of Service - Impact on MCHFR Limits.- Enclosure 2-contained 3 attachments:.'(1) the General Electric Analysis, (2)' Addendum A i

GeneralElectric'sAnswersthConsumersPower'squestions,and(3)

' Addendum B~ - fMPLHGR-Limits for Exxon F0els.

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' operation:

(1)theimpacto'nnormaloper'ation'(i.e.,'arethere adequate thermal margins,when one considers the effect of antici-pated transients), (2) the potential effect on accidents which are y

analyzed (principally the 1.0CA and locked rotor accident), and_

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(3) the' potential for~ a new accident (in this case, a coldwater accident caused by. the startup of the inactive pump).

A One factor that can affect all three of these considerations is the

,effect of one;1oop operation on reactor coolant flow distribution.

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. Big Rock Point is a 2 ' loop, General Electric design, non-jet pump boiling water reac' tor. The coolant flows through two inlet. nozzles L(one~ per loop)' which lie 72' degrees apart on-the vessel lower head.

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Thefflow-entering lthrough e'ach nozzle impinges on a diffuset plate f

L(one plate per nozzle). A flow diffuser baffle; connected to'the.

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! core support-plate surrounds the fuel channel-support tubes and '

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causes the: pressure at the inlet to the core sepport tubes to be

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'relatively uniform. The fact that the vessel entrance re'gion acts,

-as s plenum has been-supported by test- (" Core' Performance and -

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Transient Flow Testing - Big Rock Point Boiling' Water Reactor",

The test GEAp-4496,' July 19'65, USAEC contract AT (04-3)-361).

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showed that the frictional pressure drop between the vessel nozzles 1

and the support tube inlets'to be nearly 5 times the velocity head

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'in the support tubes.

Instrumented fuel assembly measurements during j

forced-circulationtests(Figure 4-7oftheabovereference)have shown relative assembly' power to be insensitive to the number o<

loops in operation, further indicating that the relative flow to the assemblies is not substantially effected by the number of loops in operation. When considering the high losses due to flow resis4 i

tance caused by the orifices in the assemblies, a small pressure difference in the lower plenum at the support tube entrance elevatiun

,i should have a negligible effec,t on the core flow distribution.

Since the physical arrangement of the forced circulation systems at Big-Rock Point, flow through the core, and supportive testing indi-cate that flow perturbations will not be introduced to the system, it is expected that the reactor will not discern the difference between one pump and two pump operation. The staff, therefore, concludes that uneven or asymetric flow conditions will have a negligible avfect on Big Rock Point during N-1 loop operation.

With regard to the effects of anticipated transients: The licensee has provided (Enclosure 3 to the October 24, 1978 submittal)a

. discussion on the effects of transients on the minimum critical heat flux ratio /MCHFR) when operating in the reduced flow con -

figuration. The licensee stated that the 3.0 MCHFR limit derived 4

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I for full flow steady state operation is valid for single loop gg operation.

Oca rev W of the codes used to predict the MCHFR limit i

r support this c. % 1usion. Although the critical heat flux correlation

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H used by the licensee (synthesized ENC Hench Levy) does not, per se, have a flow terni which would directly support the licensee's state-ment that MCHFR is insensitive to flow changes; it does have a L

fluid quality factor. Since we have indicated above that the reactor does not see the difference between one or two pump operation, j

the quality of the fluid does not change; therefore, the computed MCHFRs for two loop operation are bounding for the one loop operation.

Furthermore, as discussed in the Cycle 15 reload analysis, a transient MCHFR pf 2.15 was established; to this limit was added

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additional margin to account for the worst case transients. The i

s'taff further added conservatism to the limit which yielded the 4

steady state MCHFR of 3.0.

The total steady state.MCHFR provides

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assurance that u Jer the worst case transient the resulting;MCHFR will not be below the safety limit of 2.15. This margin of safety precludes operation of the reactor in a region conducive to fuel failure.

In addition to the operating restrictions of MCHFR, high neutron flux and high reactor pressure trips are maintained within the same

.L proximity for single loop operation as for two loop operation.

That is, the reactor protective system is realigned to cause trips dthin the sane tolerance at the reduced power level as they would e

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6-l at the. full power level (e.g., high flux trip - 12C% of maximum allowed operating power level, at reduced power, say 50%, in the I

event of a transient the reactor would trip if power reached or exceededthe60%powerlevel). The same reduction in setpoint would relate to the overpressure trip.

i General Electric (GE) has performed for Big Rock Point an:ECCS-LOCA caleviation at 102% of rated power, with one loop out-of-service, withtheout-of-serviceloopisolated(pumpsuction, discharge,and bypass valve closed), and has compared the results to calculations for the two pump in service conditions. The calculations and comparisons performed demonstrate all effects of one loop operation.

i that might significantly affect peak clad temperature (PCT).

The analysis of the break spectrum revealed that the worst case break would result from a break in the recirculation discharge line of 2

the inactive loop (0.500 ft breaksize). This analysis predicts the PCT for this break size to be 2192 degrees F and a peak local

' oxidation fraction of 0.072. The Appendi'x K to 10 CFR 50.46 PCT limit is 2200 degrees F and the Big Rock Point oxidation, upper limit is 3.17..The calculated PCTs and oxidation fractions for all breaks analyzed for single loop operation are reported in Table 3 of to the October 24, 1978 submittal by the licensee.

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The analysis has demonstrated that a correction factor must be-applied to the all-loop 3-in-service maximum average planar linear heat generation rate (MAPLHGR) 11ait for conservative one-loop-out.

of-service operation. The MAPLHGR limits for double loop operation were calculated for the' design basis accident (DBA) for the cycle 15 reload. New MAPLHGRs were calculated by the licensee considering' 2

the worst case break (0.5 f t recirculationdischargeline)and compared them with those for tae two loop worst case break. The maximum offset'in MAPLHGRs wt.s 0.5 kW/ft. The licensee proposed to operate' the plant in the i-1 loop configuration with MAPLHGR t

limits calculated by subtractii10.5 kW/ft from the double loop MAPLHGR numbers. The analysis yielding approximately a 5% reduction in MAPLHGR applies to the GE fuel.

Big Rock Point also employs fuel fabricated by Exxon; however, an analysis for the behavior of the Exxon fuel in single loop operation has not been performed.

The licensee states that since MAPLHGR is insensitive to the number -

y of loops in service, since they changed only by at most 5% in single loop for the GE fuel, MAPLHGRs can be conservatively derived by reducing the exis' ting two-loop Exxon MAPLHGR limits by 10%.

Based on our review of the methodology employed to calculate the MAPLHGR limits associated with Cycle 15 and our review of the analysis performed for the single loop mode of operation, we conclude that the new limits are conservatively derived and are therefore acceptable. However, since these limits are subject to cnange with each reload review, we recommend that the single loop

i-y v iMPLHGR limits be made part of the Technical Specifications and j

subsequent char.ges to them be evaluated by the staff in the same -

manner as any other change to safety limits.

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Regarding the locked roter accident, Big Rock Point has not provided i

an analysis of the effects that this ecident might have when operating 1

in the N-1' loop configuration. However, d nce operation of the i

e facility with less than all loops in service is a relatively low likelihood event (based on operating experience with several reactors of this design) 'we conclude that an event such as a locked rotor while in this mode is even more remote. Furthermore, the Systematic i

Evaluation Program in the course of reviewing design basis events will review the locked rotor accident in both the N loop and N-1 loop conditions. Thus, we conclude that in the interim this deficiency is acceptable for Big Rock Point.

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With regard to the potential for'a new accident the staff considered the potential for a cold water injection, accident caused by the startup of the inactive loop. Staff criteria requires that an ar.alysis 'of this event be performed to determine the pot *ential con sequences. Technical Specifications prohibiting startup of an inactive loop are not considered by the staff to be an acceptable alternative to analyzing the event. However, in lieu of an analysis, 9

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reducing the credibility of the event is an acceptable alternative.

Methods such as the use of temperature differential interlocks, which prevent opening a valve or, starting a pump unless a predetermined minimum temperature differential exists between the active and inactive loop, or requiring the' isolation valves to remain open when the pump is inactive, thereby maintaining the idle loop in thermal equilibrilsm with the operating loop, are examples of effective measures to reduce the likelihood of the event which the staff would review for acceptability.

The analysis performed by the licensee to justify the operating MCFR and the limiting MAPLHGR in the N-1 loop configuration assumes that the inactive loop is completely isolated allowing no bypass flow through the inactive loop. Although operating with the isolation valves open will establish a thermal equilibrium between the loops and resolve the cold water injection accident,.the diversion of bypass flow (backflow through the loop) may impact the previously.

discussed limits in such a way that they can no longer be considered q

conservative without additional analysis'. On this basis the staff cannot permit Big Rock Poin.t to operate with an idle loop (non-isolated) when in the N-1 configuration. Based on the above, we' recommend that the licensee establish a procedure that administrative 1y dictates the closing of the isolation and bypass valves' in the inactive l

loop if operation is to continue. Section 12.8 of the Big Rock' l

point Final Hazards Summary Report presents the cold water accident tr.nlysis performed by the licensee. A comparison of this analysis

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e to the methods used in current criteria (Standard Rehiew Plan 15.4.4) rehealsomedehiations.'TheSRPmethodismoredetailedthanthatof k

• the licensee in that it' discusses the effects of the cold water accident on operating limits (MCFR), linear heat generation rate (LHGR), and the potential 'for oherpressurization.

However, re-analysis of'the cold' water accident need not be performed if the potential for occurrence is substantially reduced or removed. Therefore, we recommend that the power to the inoperatihe pump be physically re-mohed (locked out) during N-1 loop operation. This action will prohide additional assurance that an inadhertent cold water injection accident will not occur.

It should further be noted that this admin-

.istrative operating restriction may be removed by performing a re analysis of the cold water accident in accordance with SRP 15.4.4 and the acceptability of that analysis determined by the NRC staff.

We therefore conclude, based on our review of the docketed and submitted material, that operation'with less than all loops in

,m f.ervice is acceptably resolved. However, i.t should be noted that the_acceptabilityofthis' topic (IV-1.-A)iscontingentuponthe licensee's agreement to (1)' include in a procedure for N-1 loop operation a statement that the bypass and isolation valves in the inactive loop ie closed during N-1 operation, (2) physically lock-out power to the inadive pmp, and (3) incorporate the l'APLHGR limits,

for N-1 loop operation in the Technical Specifi:ations.

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ATTACHMENT'2 r

Consumers Power Company i

Big Rock Point Plant Docket 50-155 I

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'10CFR50.59 EVALUATION DATED JUNE 15, 1990 I

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5 Pages OC0990-0074-NLO4~

l-NUCLEAR OPERATIONS DEPARTMENT Q

SAFETY EVAL.UATION FORM Item identification 4tsm To be E.atuated Pro (cdue t b M d.

NDCUWR % S/,4ty SymoA ko.ruman{1 &r c>nt. h4.bp gd No. Sof-2.cl asv lW t.10 CFR 60.59 DETERMINATION 1

Is ths itsm saicty.releted or can it af feet another safety related item?

kiYes RNo L"t affected item (s). htq) yo_,_ld Q Is the item changed from its description in the FSAR/FHSR7 List affected section(s). *1tdion 4S Ontar% u4klt46TL Ati train A,%utt (.a Aac/us)

MYes No 2.

Yes

@No 3.

Does the item involve a test or experiment not previously described in FSARIFHSR7 4 Does the item require a change to Technical Specifications?

MYes

%No List affected section(s).

11.10 CFR 50.59 EVALUATION Is the probability of occurrence or the consequences of an accident or malfunction of equipment 1.

important to safety previously evaluated in the Safety Analysis Report increased?

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is the possibility of an accident or malfuncthn cf a different type th:n any evaluated

' tweviously in the Safety Analysb Report created?

' Basis:

h A0lA GM0lke b Yes NNo l

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3. " ls the margin of safety as defined in the basis for any Technical Specification reduced?

h b1Q dk Yes

% No Basis:

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Ill. NRC NOTIFICATIOtt i

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Should this be included in FSAR/FHSR update?

k Yes Il No

2. ' Should item be included in Annual Report to NRC7 Yes [] No 3.

is prior NFIC approval and an application for amendment to License required?

Yes No IV. APPROVALS Date Reviewed by Date i

Prepared by.

DCSIM 3//4/90 6b

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4.5 OPERATION WITE LESS THAN ALL LOOPS Topic IV-1. A of the Systematic Evaluation Program deals with operating the' reactor at power with one of the recirculation loops out of l

s e rvice. NRC letter of October 9,1979 to David Bixel from DLZiemann presents the safety assessment of this topic. The acceptability of operating with one loop out of service was contingent upon satisfying certain conditions.

The bypass and isolation valves in the inactive loop must be closed.

This requirement was to be controlled by procedure. The power to the recirculating pump in the inactive loop must be physically locked out.

MAPLRGR limits for N-1 loop operations i

must be incorporated into the Technical Specifications.

The proposed Technical Specifications to incorporate NAPLHGR limits -

for N-1 loop operation was included in Consumers Pcwer Company letter of February 25, 1980.

NRC letter dated June. 9.1981 included a safety evaluation of the revised MAPLHGR limits for Exxon fuel for a one loop operation. At l

that time evaluation of the one loop MAPLHGR limits for Exxon fuel was not complete.

A follow-up letter in response to NRC questions was issued.by Consumers Power Company to DMCrutchfield on June,19, 1981. A new MAPLHGR limit for Exxon fuel with a one loop operation was proposed. This change is further documented by letter to DMCrutchfield from GCWithrow on July 22, 1981.

The incorporation of these contingencies pending approval of the Technical Specification change was reported to the NRC by letter dated September 3,1981 from TCBordine to DHCrutchfield.

Topic IV-1A was acceptably resolved and documented by letter to DPHoffman from DMCrutchfield on October 8,1981. With the conditions previously mentioned, it is permissible to operate the Big Rock Point reactor with only one recirculation loop in service.

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4.5-1 M11287-1835A-BX01

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y Response to Safety Evaluation Report - Part 2 s

Is the probability of occurrence or the consequences of an accident or. malfunction of equipment important to safety previously evaluated in the Safety Analysis Report increased?

No, the probability of occurrence or the consequences of an accident or. malfunction of equipment important to safety has not been increased. Section 4.5 of the Final nazards Summary Report includes the requirements for plant operations with less than all loops in service.

These requirements include:

1) closing the bypass and isolation valves for the idled loop and,

2) the power to the inactive pump must be physically locked out.

These requirements have different bases.

The requirement to isolate the inactive loop is based upon not having an acceptable analysis-for single loop operation with backflow assuming ECCS conditions.

In

.1 lieu of performing the analysis KAPLHGR limits were reduced and the idled loop isolated.

The power lockout requirement of the idled loop is based upon not having an acceptable analysis for single loop 1

l operation with restarting the inactive loop (cold water addition).

Complete details are contained in the 10/7/79 safety assessment written by the NRC on SEP Topic IV-4. A " Operation With Less Than All Loops In Service".

This assessment was also used in the basis for i

Technical Specification Amendment No.48 which placed the MAPLHGR limits for single loop operation into the Tech Specs.

The NRC stated in their assessment that the administrative limits regarding single loop operation could be removed provided the cold water accident analysis was re-performed in accordance with Standard Review Plan 15.4.4.

The analysis was performed for SEP Topics XV-1, XV-3, XV-4, XV-5, XV-7 and XV-9 and submitted on 7/15/81.

Startup of an inactive loop was acceptable and SOP-29 was changed to allow restart providing power had been reduced to a level specified by the reactor engineer.

Therefore, with the new analysis, the administrative requirement to lockout power for an idled recire pump can be removed from SOP-29.

With respect to isolating the idled pump, SOP-29 allows the operator one hour from the time of pump trip until the valves must be closed.

Closing of the suction valve to the idled pump is not a prudent action from a reliability standpoint.

If the operators are unable to start the inactive pump within the one hour limit, closing the suc-tion valve will cause the pump shaft seals to begin to depressurize.

If at some point after the one hour expires and pump restart is per-formed the pressure transient on the pump seals may cause them to i

seal improperly causing potential shaft seal leakage which may lead to pump isolation and unit derate.

By allowing the operators to keep the suction valve open, pressure is maintained on the seals, thus when the pump is restarted less of a transient is experienced by the seals.

Reverse flow is prevented by having the bypass and discharge valves closed.

..g Therefore, by eliminating the requirements to 1) lockout power to the b

idled recirc pump and 2) closing the suction valve of the idled pump

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during one loop operation, the probability of an' accident or mal-function previously evaluated in the Safety Analysis Report are not increased.

Question 2 Is the possibility of an accident or malfunction of a different type than evaluated previously in the Safety Analysis Report created?

No, changing the requirements on the system configuration during single loop-operation does not create a new possibility of accident or malfunction _of a different type than previously evaluated-in the Safety Analysis Report.

The changes (not locking power out and suc-tion valve open) do not place the plant in an un-analyzed configura-tion.' Previous single loop analysis assumed an isolated inactive loop, no change is being made to this requirement, the idled recirc pump discharge valves still must be closed within one hour.

Power lockout was based solely on not having an analysis which has sub-sequently been performed showing the acceptability of restarting an inactive loop..In addition power level requirements are currently in place in SOP-29 prior to restarting an inactive loop.

Question 3 Is the margin of safety as defined in any Technical Specification reduced?

No, the basis for Tech Spec change No.48 refers to the administrative requirements imposed in the SEP Topic IV-4.A safety assessment (refer to the 10/9/79 NRC to CPCo letter).

This assessment allowed for changing the administrative limits should the proper re-analysis be performed.

The analysis was subsequently performed and restart of an

-inactive loop was found acceptable provided the power level was reduced.

Administrative limits currently in SOP-29 require the Reactor engineer to provide an acceptable powar level prior to pump' i

restart.

In addition the requirement to isolate an inactive loop within'one hour is maintained as the requirement-to close the dis-charge valves is still imposed, only the suction valve is to allowed to remain open. -

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l' ATTACHMENT 3 Consumers Power Company i

Big Rock Point Plant Docket 50-155 l

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NRC. LETTER DATED APRIL 7,=1982 l

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6 Pages OC0990-0074-NLO4

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_.. _ _ _ _ _ _ _. _. _ i Docket No. 50-15$.

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• Rsvised 3/30/82

.Mr.' David J. VandeWalle cc Mr. Paul A. Perry, Secretary U. S. Enviromental Protection Consumers Power Company Agency

. 212 West Michigan Avenue Federal Activities Branch Jackson, Michigan 49201 Region V Office j

ATTW: Regional Radiation Representative j

Judd L. Bacon, Esquire 230 South Dearborn Street 1

1

• Consumers Power Company Chicago, Illinois 60604 212 West Michigan Avenue e

Jackson, Michigan 49201 Peter B. Bloch, Chaiman Atomic Safety and Licensing Board Joseph Gallo, Esquire U. S. Nuclear Regul.atory Commission Isham, Lincoln & Beale Washington, D. C.

20555 1120 Connecticut Avenue Room 325 Dr. Oscar H. Paris Washington, D. C.

20036 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Peter W. Steketee, Esquire Washington, D. C.

20555 505 Peoples Building Grand Rapids, Michigan 49503 Hr. Frederick J. Shon Atomic Safety and Licensing Board Alan S. Rosenthal, Esq., Chairman U. S. Nuclear Regulatory Commission i

Atomic Safety & Licensing Appeal Board Washington, D. C.

20555 U. S. Nuclear Regulatory Co'mmission Washington, D. C.

20555 Big Rock Point Nuclear Power Plant ATH: Mr. C. J. Hartman Mr. John O'Neill, II Plant Superintendent Route 2 Box 4(

Charlevcix, Michigan 49720 Maple City, Michigan 49664 Christa-Maria Mr. Jim E. Mills Route 2. Box 108C Route 2, Box 108C Charlevoix, Michigan 49720 Charlevoix, Michigan 49720 William J. Scanlon, Esquire Chairman 2034 Pauline Boulevard County Board of Supervisors Ann Arbor, Michigan 48103 Charlevoix County Charlevoix, Michigan 49720 Resident Inspector Big Rock Point Plant Office of the Governor (2) c/o U.S. WRC Room 1 - Capitol Building RR #3, Box 600 Lansing,- Mich'igan 48913 Charlevoix, Michigan 49720 Herbert Semmel Counsel for Christa Maria, et al.

Urban Law Institute Antioch School of Law 26S316th Street, NW Washington,.D. C.

20460 e

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-Dr. John H. Buck -

Atomic Sa ety and Licensing Appts1 Board f

U. 5. Nuclear Regulatory Commissio.7 20555-Washington, D.~ C.

Ms. JoAnn Bier-204 Clinton Street Charlevoix. Michigan 49720

' Thomas 5'. Moore Atomic Safety and Licensing Appeal,8oard J. 5.. Nuclear Regulatory Commission i

Washington, D. C.

20555 James G. Keppler, Regional Administrator Nuclear Regulatory Commission, Region !!!

79g Roosevelt Road Glen Ellyn,1111nois' 60137 J.

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SYSTDIATIC EVALUATION PROGRAM

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TOPIC XV-9 BIG ROCK POINT _

TOPIC:

,XV-9. Startup of an Inactive Loop I.

INTRODUCTION The startup of an inactive or idle recirculation loop at an incorrect temperature is examined to assure that t he c onsequences are acceptable.

The guidanc e for the review of this topic is provided by SRP Sec t ions 15.4.4 and 15.4.5.

This t ransient is

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evaluated because it reduces voids in the corer which causes an increase in power and reduces thermal margins.

The calculated Minimum Critical Power Ratio (MCPR). is c ompared t o t he MCPR safety limit to demonstrate that fuel f ailures will not occur.

II.

REVIEW CRITERIA Sect ion 50.34 uf 10 CFR Part 50 requires that each applicant for a const ruct ion permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and component s of t he f ac it i.t y wit h t he objective of assessing the risk to public health and saf ety resulting f rom ope r at ion of t he i

facility, inc tyding determination of the margins of safety during normal oper'at ins and t ransient s condit ions anticipat ed during t he lif e of t he f ac ilit y.

Section 50.36 of 10 CFR Fart 50 requir es the Technic al Specific ations to include -saf et y limit s whic h protect the integrity of the physical ba r r ier s whic h gu a rd aga in s,t the uncontrolled release of radioactivity.

1 The General Design Criteria (Appendix A t o 10. C FR Pa rt 50) set forth the criteria f or the design of water-cooled reactces.

GDC 10 " Reactor Design" requires that the core and associated coolings control and protection system be designed with appropriate margin to assure that specified acceptable fuel design limit s are i ot exceeded

'during normal operation, including the ef fects of anticipated opera-tional occurrences.

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.,,. y i-GDC 15 "Reac t or Coot ant Syst em Design" requires that the reactor coolant and ' a s soc iat ed pr ot ec t ion syst ems be designed with suf ficient

. margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operations including j

the effects of anticipated operational occurrences.

GDC 20 " Prot ec t ion Syst em Functions" requires that the protection

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syst en be designed to init (at e automat ically t he operat ion of reac t ivit y c ont rol syst em's t o. a s sure t hat specified acceptable fuel i

1 d e s ig'n limit s are not exceeded as a result of anticipated operational occurrences.

J GDC 26 "Reac t ivity Cont rol Syst em Redundancy and C apabilit y" requires that the reac tivity cont rol syst em be capable of reliably controlling reactivity changes to assure that under condit ions of normal opera-tions including anticipated operat ional oc currences, and with appropriate margin for malfunctions such as stuck rods, specified accept abt'e fuel des.ign limit s are not exceeded.

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GDC 28 " React ivit y Limit s" requires that the reac t ivit y cont rol sys-t ems be de signed with appropriate limit's on the potential amount and rate of react ivity increase to ensure t' hat th'e' effects of postulated react ivit y a c c ident s c ar, ne it he r (1) result in dan. age t o t he react or coolant pres sure boundary great er t han limited local yielding nor I

(2) suf ficiently disturb the cores i ts support structures or other reactor pressure vessel internals t o impair significantly the capa-7 i

bilit y to cool t he core.

III. RELATED SAFETY TOPICS Various other SEP topics evaluate such itses as the reactor protection systas. The I

ef fects of single failure on saf e simitdown capability are considered under Topic VII-3.

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IV.

REVIEW GUIDELINES The review is conducted in accordance with SRP 15.4.4 and 15.4.5.

i The evaluation includes review of the analysis for the event and identification i

of the features in the plant that mitigate the consequences of the event as well as the ability of these systems to function as required. The extent to which operator action is required is also evaluated. Deviations from the cri-teria specified in the Standard Review Plan are identified.

i V.

EVALUATION _

The st'attup of an inactive loop was analyzed for the Big Rock Point plant in the reference below.

The more important assumptions weret i

1.

Initial power is 102% of singt e loop operating

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power (188.7 MWt) 2.

Water in isolat ed loop is at 100*F 3.

Scram initiat ed on high flux (120% powe r) 4.

C on se r vat ive re a'c t iv it y c oe f fic ient s and sc ram characteristic s are used 5.

Pump in idle loop reaches full speed in i second, bypa ss and discharge valves open sequentially at design speed i

These and other assumpt ions described in the ref erence are in accordance with the SRP.

T he c ompu t e r c od e u's ed in the analysis is COBR A-IV-I modified f or t he Big Roc k Point plant.

The result s of t he ' analysis' show that t he peak power of 127 % initial value is reac hed 8.25 seconds af ter 'initiat ion of t he event. The reactor pressure rises 19 psi to 1349 psia, which is below design pressure.

The minimum c rit ical powe r rat io wa s 1.77, whic h is above the specified acceptable fuel design limit for N PR (1.32).

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VI.

CONCLUSION-i The staf f has reviewed the Big Rock Point submittal on SEP Topic XV-9, Startup

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of an Inactive Loop. The assumptions used in the analysis are in confonnance.

with' SRP Sections 15.4.4 and 15.4.5 and the results satisfy the SRP acceptance i

criteria and are therefore, acceptable.

Letter to D. Crutchfield, NRC from R. A. Vincent, Consumars Power REFERENCEt Company, " Big Rock Point Plant - SEP Design Basis Event Topics",

dated July 15, 1981.

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l ATTACHMENT 4 1

Consumers Power Company Big Rock Point Plant Docket 50-155 a

AMENDMENT NO.48 TO FACILITY OPERATINC LICENSE NO. DPR-6 FOR BIG ROCK POINT DATED SEPTEMBER 21, 1981 l

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14 Pages

' OC0990-0074 NLO4 l

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gggiVED 4,W UNITED STATER J

' NUCLEAR REGULATORY COMMISSION e

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SEP 26 $g;

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September 21, 1981 NUCttAR llCfNstNn e...

Docket No. 50-155

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.L505 81-09-059 7 S.,4 P C 5,. //

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4 6 b g". u*..".."o Mr. David P. Hoffman Nuclear Licensing Administrator

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Consumers Power Company

'1945 W. Parnall Road Jackson, Michigan 49201

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SUBJECT:

OPERATION WITH ONE RECIRCULATION LOOP.0UT OF SERVICE

-Re:.

Big Rock Point Plant The Comission has issued the enclosed Amendment No. 48 to Facility Operating

. License No. DPR-6 for the Big Rock Point Plant.

This amendment consists of changes to the Technical' Specifications in response to your appitcation dated

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February 25, 1980 and supplements thereto dated June 19,1981, July 22,1981, and SFptember 3,1981.

The amendment authorizes operation of the reactor with one recirculation loop out of service.

Co' pies o'f our related Safety Evaluation 'and the Notice of issuance are also enclosed.

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Sincerely.

Dennis M.. Crutchfield dief Operating Reactors Branch f5 Division of Licensing

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Enclosures:

1.. Amendment No. 48 to License No. DPR-6 "m-2.

Safety Evaluation.

4 3.

Notice of Issuance cc w/ enclosures:

See next page-l

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Mr. David P. Hoffman September 21, 1981 I.

lcc Mr. Paul A. Perry, Secretary U. S. Environmental-Protection - -

i Consumers Power Cogany Agency 212 West Michigan Avenue Federal Activities Branch o

Jackson, Michigan 49201 Region V Office ATTW: Regional Radiation Representative Judd L. Bacon Esquire 230 South Dearborn Street Consumers Power Cogany Chicago, Illinois 60604 212 West Michigan Avenue Jackson, Michigan 49201 Herbert Grossman, Esq., Chairman Atomic Safety and Licensing Board Joseph Gallo, Esquire U. S. Nuclear Regulatory Comission Isham, Lincoln & Beale Washington, D. C.

20555 1120 Connecticut Avenue Room 325 Dr. Oscar H. Paris Washington, D. C.

20036 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comission Peter W. Steketee. Esquire j

Washington, D. C.

20555 505 Peoples Building l:

Grand Rapids, Michigan 49503 Mr. Frederick J. Shon Atomic Safety and Licensing Board Alan 5. Rosenthal, Esq., Chairman U. S. Nuclear Regulatory Comission Atomic Safety & Licensing Appeal Board Washington, D. C.

20555 1

U. S. Nuclear Regulatory Comission Washington, D. C.

20555 Bia Rock Point Nuclear Power Plant ATTN: Mr. C. J. Hartman i

Mr. John O'Neill,11 Plant Superintendent Route 2. Box 44 Charlevoix, Michigan 49720 Maple City, Michigan 49664 Christa-Maria Charlevoix Public Library Route 2 Box 108C 107 Clinton Street Charlevoix, Michigan 49720 Charlevoix, Michigan William J. Scanlon, Esquire Chairman 2034 Pauline Boulevard County Board of Supervisors Ann Arbor, Michigan 48103 Charlevoix County Charlevoix, Michigan 49720 Resident Inspector Big Rock Point Plant Office of the Governor (2) c/o U.S. NRC" Room 1 - Capitol Building RR #3, Box 600 Lansing, Michigan 48913 Charlevoix, Michigan 49720 r

Herbert Semel Mr. Jim E. Mills Council for Christa Maria, et al.

Route 2. Box 108C Urban Law Institute -

Charlevoix, Michigan 49720 Antioch School of Law-263316th Street, NW Washington, D. C.

20460-i B

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3-September 21, 1981 Mr. David P. Hoffman cc Dr. John H. Buck Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Ms. JoAnn Bier 204 C1tnton Street Charlevoix, Michigan 49720 Thomas S. Moore Atonde Safety and Licensing Appeal Board l

U. S. Nuclear Regulatory Commission Washington, D. C.

20555 F

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UNITED STATES NUCLEAR REGULATORY COMMISSION o

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CONSUMERS POWER COMPANY 00CKET NO. 50-155 BIG ROCK POINT NUCLEAR PLANT AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 48 License No. DPR-6

'1.

The Nuclear Regulatory Comission (the Comission) has found that:

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A.

The application for amendnent by Consumers Power Company (the 1

licensee) dated. February 25, 1980, as supplemented June 19, 1

July 22 and September 3,1981,' complies with the standards and requirements of the Atomic Energy Act of 1954, as amended 1

(the Act), and the Comission's rules and regulations set

'j forth in,10 CFR Chapter'I; B.

The' facility will ' operate in conformity with. the application -

the provisions of the Act, and the: rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized -

.by this amendment can-be conducted without endangering the health l

and safety of the public, and (ii) that such activities will be

. conducted in cemn11ance with the Comission's regulations, j

D.

The issuance of this ' amendment will not be inimical to the common 4

defense.and security or to the health and safeth of the pubitc; and E.

The. issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements -

have been satisfied.

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2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C(2) of Facility Operating License No.

DPR-S hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 48, are hereby incorporated in the licenso. The licensee shall operate the facility 1

in accordance with.the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION l

Dennis M. Crutchfield.dhief Operating Reactors Branch #5 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: September 21, 1981 i

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ATTACHMENT TO LICENSE AMENDMENT NO. 48 FACILITY OPERATING LICENSE NO. DPR-6 DOCKET NO. 50-155 i

Revise Appendix A Technical Specifications by removing the following page J

and inserting the enclosed page.

This revised page includes the captioned amendment number and contains a vertical line indicating the area of change.

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Reloads:

Reload Reload T ti Modified F Reload 0 G-1U G-3/G-4 Minimus Critical Neat Flux Ratio ct Normal Operating Conditions

• 3.0 3.0.

3.0 3.0 Minimum Bundle Dry Out Time **

Tigure 1 Maximus Heat Flux at Overpower, Stu/h-Tg 500,000 395,000 407,000 392,900 Maximus Steady State Rest Flux, 5Stu/h-Ft8 410,000 324,000 333,600 _

322,100.

4..

Maximum Average Planar Linear Heat Generation Rate, Steady (tate, kW/Tt ***

Table 2 Table 2 Table 2 Table 2 Max}.aus Bundle Power, W Figure 2 x 0.95 Tigure 2 Tigure 2 'Tigure 2 c

e stability Criterion: Maximum c

8 Measured Zero-to-Peak T1ux Amplitude, Percent of Average

Operating Flux 20

. 20 20 20

. Maximum Steady State Power Level, W 240 240 240 240 g

a Maximus Value of Average Core-Power Density 9 240 W,,

kW/L 46 46 a6 46.

'.\\*emina ' ?.eacts: Pressure During Stead */ State - ? ver Operation, Psig 1,335

,335 1,335 1,333 Minimum Recirculation Tiow Rate th/h' 6's 106 6 x 106 6 x 106 6 x 106 Rate of-Change-of-Reactor-e Power During Pcwer Operation:

Control rod withdrawal during power operation shall be such that the average rata-of-change-of reactor power is less than 50 We per :cinute when power is less than.

and 200 Wg, and 120 W.,. less than 20 W per minute when power is between 120 We g

and 240 W per minute when power is between 200 Wg JO Wg 2

g

.. The bundle Mir.laus Critical Heat Flux Ration (McHFR), brsed on the Exxon Nuclear

' Corporation Synthesized Hench-Levy Correlation, must be above this value.

• • The actual dry out time for GE 9x9 fuel (based on the General Electric Dry Out

. Correlation for pon-jet Pump 3siling Water Reactors, NEDE-20566) should be above the-dry out time shown in Figure 1.

• • • For operation with onik one recirculation loop in service these limits shall be reduced by 5 percent for Reload F and Modified F, and reduced by 20 percent for other fuel types.

Amendme'nt No, M g, Jh,48 5-9a-4.

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SAFETY EVALUATION BY-THE OFFICE OF NUCLEAR REACTOR REGULATION e

SUPPORTING AMENDMENT-NO.:48 TO FACILITY OPERATING LICENSE NO. DPR-6 CONSUMERS POWER COMPANY t

BIG ROCK POINT PLANT i

DOCKET N0; 50-155 ~

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.1.02 INTRODUCTION By' letter dated February 25,-1980 and supplements dated June 19, 1981,

' July 22,1981, and September 3,1981, Consumers Power Company (the s-

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. licensee) requested an amendment to_ Facility Operating License No.. DPR-6

~ for the Big Rock Poing Plant. -This amendment would add reactor operating i

limits. for operation with one recirculation -loop out of service.-.In a i

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f related action, the' Commission issued Amendment No. 44 on June 9,1981-

'which authorized a change in the reactor operating limits for operation L

with both recirculation loops-in service.

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2.0 -BACKGROUND.'

1:

25, 1980 submittal requested (1) revised reactor l,

,The licensee's February operating limits for operation'with both recirculation loops:in' service, e

and (2) operating limits <for one recirculation loop out of service for the

. Big Rock Point Plant. - Amendment No. 44 dated June 9,1981 approved new p

f, ECCS operating limits for two loop operation.

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J By letters dated June 19 1981' and July 22,1981, Consumers Power Company

.i provided additional information with regard to single loop operation. This f

safety evaluation addresses operation with one recirculation loop in service.

t3.0.4 DISCUSSION AND EVALUATION, s The NRC staff.has completed a. review of the March 31,19N, General Electric lGE)LBig Rock ' Point ECCS analysis (reference 1) for single loop operation.

Mthough single loop operation (SLO)-is still;an outstanding issue in jet pump BWRs Big-Rock Point is a non-jet pump BWR 'and has sufficiently demon-strated that this mode of operation is safe and acceptable, i

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.2 The GE' analysis is performed with the same codes as used in the two loop:

4 case. Certain ing.t parameters are modified to simulate' single loop operation.

In the analysis, the location of the worst break was determined i

to be in the 20-inch recirculation discharge of the isolated _ loop.

Further analyses of a break spectrum was made for the worst break location. A 0.5 ft2 break was determined to result in the highest calculated peak cladding -

e temperature of 2192*F with a local oxidation of 7.2%.

A detailed description of the codes.used can be found in NEDE-20566, " General Electric Company Analytical Model for loss-of-Coolant Analysis in Accordance with 10 CFR 50 Appendix K."

Although this_ would appear to satisfy requirements, Big Rock Po' int is in a unique situation with respect to the simultaneous or exclusive use of i

EXXON or GE fuel during a particular. cycle. The use of a GE two loop LOCA analysis to_ describe the behavior of an all EXXON core in a 'ransient-1 situation was a previous staff concern and was reviewed and found acceptable in reference 2.

A further complication arises by the fact that there was.

a no SLO ECCS performed for EXXON fuel.

Rather than performing EXXON SLO ECCS, the licensee proposed to place a 20% MAPLHGR reduction (reference 3) on the EXXON fuel whereas GE fuel has only a 5% MAPLHGR reduction. The-5% MAPLHGR-1 reduction for GE fuel is a result of faster core uncovery in SLO.. The staff has concluded that the 20%-reduction for EXXON fuel is an adequate margin 1

such that in the' event-of a LOCA, the fuel would not be expected to violate 10 CFR 50.46 limits. In reference 4, the licensee stated that an Exxon analysis would probably-predict a similar response in the most limiting' break situnW during single loop operation. This can be based on the similarity oei. ween EXXON and GE two loop analysis response times for the-0.5 ft2 most limiting break size.- A comparison of the analysis response can be found in Table 1 (copy attached).

1 Supporting analysis for safe op(eration in the SLO mode can be found in the Systematic Evaluation Program-SEP) review (reference 5). Although this review did not address the acceptability of the GE ECCSt analysis, which prompted.this review, several other areas of concern were addressed. ' A L

H, major point was that SLO will not affect the reactor coolant flow distri-I bution. The coolant flows through two inlet nozzles (one per loop) which are 72 degrees apart on the vessel lower head.1The; flow entering through n

each nozzle' impinges on a diffuser plate (one plant per nozzle). A flow diffuser baffle connected to the core support plate surrounds the fuel channel support tubes and causes the pressure at the inlet to the core support

. tubes to be relatively uniform. The fact that'the vessel entrance region acts as a' plenum has been supported by test (" Core Perfomance and Transient Flow Testing - Big-Rock Point Boiling Water Reactor," GEAP-4496, July 1965. USAEC-q contract AT (04-3)-361). The test showed that the frictional pressure drop-between the vessel nozzles and the support tube inlets to be nearly 5 times the velocity head in the support tubes.

Instrumented fuel assembly measurements j

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-during forced-circulation tests (Figure 4-7 of the above reference) have shown -

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relative assembly power to be: insensitive to the number of loops in operation. -

l further indicating that the relative flow to.the assemblics is.not substantially a,

laffected by-the number of loops in operation. When considering the high losses-idue to flow resistance caused _by-the orifices in the assemblies, a small pressure difference in the lower plenum at the support tube entrance elevation should

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e have a negligible effect on the core flow distribution.

'The SEP review also concluded that the acceptability of the SLO is contingent upon the. licensee's agreement' to (1) include _in a procedure for N-1 loop opera-t tion. a statement that the bypass and. isolation valves intthe inactive loop be' 4

closed during N-1 operation, (2) physically lock-out power to the inactive pump, 3

and'(3) incorporate the MAPLHGR limits for N-l~ loop operation in the Technical J

Speci fica tions. By letter dated September 3,1981, the licensee stated that items (1)_ and (2), above,have been included in the plant procedures. This L

amendment would incorporate MAPLHGR limits for SLO into the Technical Specifications.

Based on the references presented, the NRC staff agrees with the licensee that--

the GE~ SLO ECCS analysis will adequately represent the behavior of EXXON fuel elements;in a LOCA situation. Based on the supportive analyses, the NRC staff concludes that SLO is a safe means of reactor oparation.

i<

j

4.0 ENVIRONMENTAL CONSIDERATION

We have. determined that the amendnent does not authorize a change in effluent

. types or total. amounts nor an increase-in power level and will not result in-1 any significant environmental impact., Having made this determination, we have J

further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement or negative declaration and environmental

. impact appraisal need not be prepared in. connection with the issuance of this.

1 amendment.

'5'0 CONCLUSIONSL We have. concluded, based on the considerations discussed above, that: (1) 1

-because the amendment does not involve a significant increase in the pro--

~

bability or consequences of accidents previously considered and does not

. involve.a. significant decrease in a safety margin, the amendment does not

-involve a' significant hazards. consideration, (2)' there is reasonable

_ assurance that the health and safety of the

^ by operation in the proposed manner, and (3)public will not be endangered such activities will be con-ducted in compliance with the Commission's regulations and the issuance of this amendment will-not be inimical to the common defense and security or to the health and safety of the public.

1

+

5

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~ ' . ~..J. p+ ~'s je : ..a, .' 4 1 t. K,1 ,;s.'.,, y.. ,...s ' 3 7 ' j!e 6;0' REFERENCES-4 / (1): " Big Rock-' Point Single-Loop Operation Loss-of-Coolant AccidNt Analysis-g f 3 for General Electric Fuel in Conformance with 10 CFR 50 Appendix K (Non-q Jet Pump _ Boiling Water Reactor),"' March 31. 1977. (2) ~ Safety Evaluation Report from D. Zieman (NRC) to Consumers Power, June 4, 1976. ' [ (3) Letter from G.. C. Winthrow (Consumers Power)- to D. Crutchfield,; July 22', 1981, i ,c.t (4)' Letter from G. C.- Winthrow (Consumers Power) to D. Crutchfield, June 19.-1981. ?(5)' Letter from D. Ziemann (NRC) to D. Bixel- (Consumers Power), August 9,1979. l Dated:- September 21,.1981 " Attached: Table 1 ,u j ' -J ' 7~. ,1

p y ld ~

t 3 '.g'. j ,i-I o q* .c '( 't n e .I 9

Q '.f."i "'..[.,^ ,.9,. . }[g " ~ 3-y x vg., - S g; ' 7 "[ -N. 4'. 5D4 ~ es TABLE 1 C M q LS911 pf,,,M y_gy1 H M_R[$tOf!SC P M i[R}. A AS Pftf otCIED BY ExgGIO AIID GE APPLIIDIM IL LOCA PIOOELS I p ~ Time (Sect l Cold Lo9 Roted Core gerev Flow Core Mldelene L ___.~i Breek Low Reactor Pressu_r_9_L100 Psiel_ CE ' Exxon GE CE Exxon CE-CE Exxon- - CE. 'Sl2gI IFt. -t2 Loosel f2 Loess) I1 t. coal i2 Loess) '(2 Leonel i1 Leos) f2 Loomst f2 Leenal. 11 Leeml - 7.1 - 20.4 5. 3.926 ~- I s' -e 26.9 i: 3.53 11.9 4.9 8-27.7 -l 12. 1.6 17.2 6.1 1.0 '20.4 21.5. 20. 35.4

36.9_

35.4 29.6 10.8 ( .50 36.3 40. sec. 5,1. 3, 55.1 6al. - ~ 13.6 " 71 ~23.(1) 61.5. 54. .375 ,46.5 103. 29.9. 19.4' .25 67.9 75. 75. - e5.6 l98.2 .10 161.k2) 140. 140. 189. 156. ~ 15s. 62.2 ' 48.9 1 i 4 ~ Implies caso not snelyzed.

• teplies enrormation not reedily evelleblo.

9 (1) Time et which einellty at hot node (coec midplenel 90es to 1.0. ft2,,,,gglg,,,3 ,,g,,gg,,, g,,g blowdown le prolonged In the Exxon (2)Sreek smel ter then about' O.1 snelyses versus the CE enelyses bedeuse Exxon esstamed only thryo operable blowdown pethe wheroes.. l - S CE assumed four. j' . l 1 p I e e I 4 f* l 4 [ nuG681-03156 ; p 9 - O e S 9 -,~ ,...-..h.w.: _,_2. %,...n m .1 ,-,.~m

L t ..! <3.r.v. o _,. 1 @ j,3L' &lw% o + .n. b -UNITED STATES NUCLEAR R GULATORY COMMISSION E -.o Y DOCKET-N0. 50-155 b CONSUMERS POWER COMPANY m NOTICE OF 1SSUANCE OF AMENDMENT TO FACILITY L OPERATING LICENSE c, ? The U.. S. Nuclear Regulatory Commission (the Commission) has issued m, t Dm Amendment No. 48 to Facility Operating Licensing No'. OPR-6, issued to y . the Consumers-Power Company (the licensee), which revised the Technical Specifications for operation of the Big Rock Point Plant (the facility) -located in Charlevoix County.. Michigan. The amendment is effective as of + its date of issuance. The amendment authorizes operation of the. reactor with one recirculation 1 l d o loop out-of service. E! .i ' The application for the amendment complies' with the standardsf and j t.>. 1 requirements of the Atomic Energy Act'of 1954, as' amended (the Act), and' ] i kr-f::': 13 r t rs;, r.':n. ' " i % ~ ' : ', ': 1:

2 6

r:pria a 5

findings as' required by the Act and the Commission's rules and regulations If in 10 CFR Chapter I, which are set-forth in the license amendment. Prior q ,public' notice of 'this amendment was not required since the amendment does 1 r m n:t involve a significant hazards consideration. The-Commission has determined that the. issuance of this amendment m M P willL not result -in any significant environmental impact and that' pursuant l; 1 i to 10 -CFR 51.5(d)(4);an environmental impact statement or negative 1a g* H o declaration and environmental impact appraisal need not be prepared in i hi connection with issuance of this amendment. 4 o e t .A \\ p \\ o f-ed I + 4. n

[fE[ 4 [ ~"~~#*~~,r"~~'"~- ~ ~ " ' ' ' ~ ~ ~ ~ ~ ~ ~ ~ ' ~ ~ " - ~ ~ ~~~~ g n. ~. ...1... , e: - + s a a i;',qF- - 2. g a, ~ For further details' with respect to this action,: see-(1) the i application for amendment dated February 25, 1980 and 'its: supplements dateddune 19,1981 July 22,1981, L and September 3. 1981, (2) Amendment No. 48 to License No. DPR-6,' and-(3) the Commission's related Safety: Evaluation. ' All of these items are available for public inspection at the Commission's Public Document Room,1717-H. Street, N. W., Wdshington, D.- C. and at-the Charlevoix Public Library,107 Clinton Street, i Charlevoix, Michigan 49720. A' copy of items (2) and (3) may be obtained upon request addressed m- - to the U. S. Nuclear Regulatory Commission, Washington, D. ' C.

20555,

't 4 Attent, ion Director, Division of.Licen' sing.- . Dated at Bethesda, Maryland, this 21st day of September,1981. FOR-THE NUCLEAR-REGULATORY COMMISSION. ]-

• M h N/ '

..:n-t a,'s y.;:.' A c Dennis-M. Crutchfield, Chief a Operating Reactors Branch #5 Division.of Licensing .3 b v f l ' i I' A /. e f .}}