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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION.,3

( ')SUPPORTING AMENDMENT NO.14 TO FACILITY LICENSE NO. DPR-6"

CONSUMERS POWER COMPANY

BIG ROCK POINT PLANT

DOCKET N0. 50-155

Introduction

Consumers Power Company (CPCo) proposed the following changes tothe Technical Specifications for the Big Rock Point Plant:

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1. Revision of reactor vessel pressure temperature operatinglimits to comply with Appendix G of 10 CFR Part 50,(CPCo application dated May 30, 1975, as supplementedby letter dated June 30,1975),

2. Incorporation of a provision to allow automatic bypassingof tne high condenser pressure reactor trip anytimesteam drum pressures are less than 500 psig insteadof the current 350 psig reactor coolant pressure limit(CPCo application dated September 10, 1975, as supplementedby letter dated May 25,1977),

3. Inclusion of a definition of administrative control requirementsassociated with the air ejector off-gas monitoring system(CPCo application dated May 26, 1976, formerly included inproposed draft Technical Specifications forwarded withCPCo letter dated June 7, 1974),

4. Correction of an error in Table 4.1.2(b) on chloride ionconcentration limits in the primary coolant (CPCo applicationdated April 21,1977) , and

5. Deletion of the 100-inch per minute limitation on winchspeed during refueling operations (CPCo application datedMay 18, 1977).

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Item 1 above contained other proposed changes to the TechnicalSpecifications. Our review of the remaining portions of item1 will be the subject of a later action.

Evaluation

1. Reactor Vessel Pressure Temperature Limits

The current pressure-temperature limits for operation of Big 'Rock Point are based on rgactor vessel-nil ductility transition(NDT) temperature plus 60 F and do not comply with all therequirements of Appendix G,10 CFR Part 50, " FractureToughness Requirements." The proposed pressure-temperatureoperating limits are based on the requirements of AppendixG, 10 CFR Part 50. These limits were developed from thetest results of the Big Rock Point reactor vessel materialsurveillance program.

Three withdrawals of specimens have been made to date. Thesespygimens provjge radiation damage data at fluences of 1.5 x10 9, 2.3 x 10 and 1.08 x 1020 n/cm2 The proposedoperating limits are calculated for a fluence of 2.8 x1019 n/cm . This ' fluence will be reached in about three2

Jaars. The proposed Technical Specifications state thattheir pressure-temperature limits will be recalculatedprior to exceeding 2.8 x 1019 NYT.

We conclude that the CPCo proposed pressure-temperature limitsfor operation of Big Rock Point comply with the requirementsof Appendix G,10 CFR Part 50 and are acceptable.

2. Automatic High Condenser Pressure Reactor Tr,ip Bypass

The low condenser vacuum (high pressuie) trip prevents operationwhenever condenser absolute pressure is greater than 8 inchesof mercury by causing reactor scram. The purpose of the lowvacuum trip is to assure that the main condenser is availableto condense the steam exhausted from the main turbine-generatorduring normal power operation or from the turbine and turbine by-pass during abnormal transient conditions.

Low condenser vacuum during operation can cause overheatingof the low pressure turbine casing. Loss of condenservacuum while at rated power . level can cause unacceptablereactor coolant pressure transients unless an anticipatory lowvacuum signal causes the reactor to scram. The requirement forreactor scram with loss of condenser vacuum diminishes as

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reactor power level decreases because the resultant reactorcoolant pressure transients are smaller. At very low powerlevel, loss of condenser vacuum is equivalent to a smallsteam line break with respect to radioactivity escapingto the atmosphere and automatic reactor scram is unnecessary.

Loss of vacuum can result from loss of main condenser circulating(cooling) water, excessive air leakage, insufficient steamto the condenser air ejector or faulty operation ofthe_ air ejector. Condenser vacuum is maintained by a singletwin element, two-stage steam-jet ejector which removesnon-condensibles and de-aerates the condensate.

During plant start-up, the low condenser vacuum scram switchmust be bypassed until the condenser vacuum can be pulleddown to levels (25 inches of mercury) lower than the normaloperational reactor scram set point (21 inches of mercury).The existing technical specification limits reactor pressureto 350 psig for automatic condenser vacuum reactor scramswitch by-pass.

CPCo has found that to produce a vacuum of 25 inches of Hgthe reactor coolant pressure must be very close to the 350 psigtechnical specification limit. Instrument drift has frequentlyresulted in violation of the technical specification 350 psigreactor coolant system pressure limitations for automaticar,tuation of the low condenser vacuum reactor scram switch.

(PCu has therefore proposed that the 350 psig reactor coolantpre.ssure be increased to 500 psig steam drum pressure toprovide a larger margin to the point where the reactor scramtrip function might be required to operate thereby allowinggreater flexibility during warm up of the main lines andthe main condenser.

He agree with the CPCo statement that at steam drum pressuresless than 500 psig, power level will be very low. Even withthe turbine stop valve open to wam up the turbine powerlevel should not exceed 63. Thennal practice, however, willbe to pull the condenser vacuum to the required level(approximately 15" Hg) at some pressure less than theprooosed 500 psig steam drum pressure technical specificationlimit before automatic resetting of the condenser vacuumscram switch. After the 25" Hg vacuum has been reached,the turbine stop valve will open and the turbine will bewarmed up. Until turbine warm up begins, the steam demandis very low. During turbine warm up and prior to electivepower generation, reactor power level will generally notexceed 6*.' according to CPCo. As noted above, we agree thatthe power will be acceptably low. Normally, turbine wam updoes not begin until steam drum pressure is greater than *

500 psig.

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On the basis of this infomation, we have concluded thatthe 350 psig primary coolant technical specification limitfor automatic bypass of the low condenser vacuum reactorscram switch is unnecessarily restrictive. We havedetennined that loss of condenser vacuum with the turbinestop valve closed could result in relatively small amounts*

of steam escaping into the turbine building. The releaseof radioactivity, however, is only a small fraction of theradioactivity released from a steam line break outsidecontainment while at rated power level. The large steamline break accident has been analyzed and found to beacceptable. Therefore, the small steam line break, equivalentto loss of condenser during plant heat up, is also acceptable.If loss of condenser vacuum is postulated to occur duringturbine warm up before steam drum pressure reaches 500 psig(contrary to normal operating practices) and before thelow condenser vacuum scram trip is automatically activated,the turbine stop valve will close automatically because ofthe trip which occurs et condenser absolute pressures inexcess of 10 inches of mercury and the radiological conse-quences will be equivalent to the small steam line breakthat we have considered and which requires timely actionof the operator to terminate. We have therefore concludedthat the CPCo proposed change to allow automatic bypassof the high condenser pressure reactor trip when steamdrum pressure is no greater than 500 psig should be approved.

3. Air Ejector Off-Gas Monitoring System

The air ejector off-gas monitoring system for the Big RockPoint Plant audits continuously the level of radioactivityin the gases released from the main turbine condenser to theoff-gas hold up pipe and 240 ft. stack. During normal operation,these levels are very low compared with the allowabletechnical specification limits and the time delay betweenthe off-gas system radiation monitor at the air ejectorexhaust and the stack is at least one half hour. This delay

permits off-gas system isolation and reactor shutdown beforehigh radioactivity levels can reach the stack and environment.In addition to prov'it;ng a continuous measure of theradiation in the centrolled release of gases from the mainturbine condenser to the off-gas piping and stack,the off-gasmonitoring system is also a fuel failui detection system.The system therefore provides an alann at the control roomto alert the operator if there is a significant increase inoff-gas radioactivity. The alann set point can be no higherthan allowed in technical specification 6.4.3. Before theradiation levels reach the limits stated in this sametechnical specification, the isolation valve in the off-gas

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system must close automatically..

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The proposed chanqe lowers the present technical specifi-cation alarm set point limit from 5 curies per second to-

about.5curiespgrgecondassumingEoftheproposedalarmset point limit, -, is 0.845 MeV per disintearation.-

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The value 0.845 MeV/ disintegration i535accordinq to CPCo., theaverage gamma ray energy for a pure U fission spectrum. Thisvalug3gesults in a lower alarm limit set point than allowing forthe ;Pu fission contribution and is therefore conservative,i.e. lower than it would be using more realistic conditions.

We have conclud'ed that the proposed change to lower theoff-gas system alarm set point will provide an earlier

, warning of an unlikely sudden increase in radioactivitythat could be indicative of fuel failure. The reducedsetting should not interfere with normal plant operationsince releases during normal plant operation have beenreduced in recent years by a factor of 10 to 100 whilethe alarm set point limit is hereby reduced by only afactor of about 10. The lowered alarm set pointallows more time for corrective actions by the operatorand is, therefore, acceptable. The technical specificationsshould.therefore be changed as proposed. CPCo has alsoproposed to reduce power level immediately if the off-gassystem alarm set points are reached. According to theproposal, the operator would reduce power level untilthe off-gas radiation levels are less than the off-gasalarm set point lic:its. "

CPCo reported that a new off-gas isolation valve war,installed at Big Rock Point and that extensive maintenancewas performed to improve the off-gas system leak tight-

However, tests performed later during stable powerness.conditions (January 31,1976) showed that when the isolationvalve was closed the stack gas release rate decreased' initially but later as the off-gas hold up p; essure increasedsignificantly the stack gas release rate increased to avalue slightly in excess of the initial value. During thetest, condenser vacuum was not noticeably reduced, andtherefore there was no i Tactor scram resulting from lowreactor vacuum. CPCo has concluded that it is rot feasibleto reduce stack gas releases by automatic isolation valveclosure followed by reactor scram due to loss of condenservacuum and has proposed administrative actions to require thatthe operator reduce reactor power level whenever specifiedoff-gas radiation alarm limits are exceeded.

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We note that:

- Testing at Big Rock Point has shown that closureof the off-gas system isolation valve causes onlya temporary reduction of the stack gas releaserate, i.e. isolation valve closure does notprevent continued reactor operation with radio-active gaseous release rates in excess of thetechnical specification limit as CPCo and webelieve was originally intended.

- During the 15 year operating history of the BigRock Point Plant, there has never been a need forautomatic closure of the off-gas system isolationvalve to remain within technical specificationlimits. In fact, the highest measured stack gasrelease rate,78,000 micro-curies per second, occurredin 1966. This value is nearly a factor of 10lower than the proposed alarm set point limit andmore than a factor of 100 below the present tech-nical specification requirement for closure of the off-gas system isolation valve.

- Operating experience has shown that off-gas radio-activity increases between normal levels and 100,000micro-wries per second do not oce ur rapidly.Increase!, when and if they occur, are easilycontrolled by reducing reactor poter level.

- The off-gas radioactivity release rate during thelast year of Big Rock Point operation has been thelowest in the history of opera +,fonLI). The off-gasrelease rate stabilized following start-up forfuel cycle No. 14 at approximately 750p Ci/s. Therelease rate near the end of fuel cycle No.14 is

j averaging 700-800p Ci/s. Power level over theperiod has ranged betwaan 206 and 216 MWt most ofthe time. This historically low radiation levelis attributed by the licensee, to the gradual removalof copper based crud from the primary system by wayof deposits on spent fuel assemblies removed fromthe core. As a result of improved fuel rod designand fabrication and reduced crud deposition on the

(I) Big Rock Point Technical Specification Change Request, G-3,April 15, 1977

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fuel rod. surfaces, the number of leaking fuelrods is noticeably reduced. Consequently: 'cheamount of gaseous radioactivity escaping from thefuel rods to the atmosphere through the primarycooling system, the main-turbine-condenser, theoff-gas exhaust system and the exhaust stack is -significantly reduced in contrast to earlier yearsof operation when stainless and inconel fuelcladding, developmental fuel assemblies with>

centermelt fuel and many components containingcopper were used. These efforts to improve fuelrod integrity have resulted in a significantreduction in radioactive off-gas released fromthe stack during normal plant operation.

We have concluded that operator action to reduce power levelif radioactive off-gas alarm set point limits are reached isacceptable based on the operating experience to date at BigRock Point. The normal gaseous release levels are relativelylow compared with the alarm set point and fuel rod claddingfailure experience indicates that the operator can react inample time to prevent release levels in excess of limits. Wehave also concluded that the CPCo proposal to administrativelyrequire reactor power level reductions whenever a high radiationstack gas release alarm is activated provides added assurancethat the 10 Ci/sec gaseous release limit will never beattained. In this respect, the safety margin is increased.Both technical specification changes further enhance thehealth and safety of the public and should therefore be made.

4. Chloridc lon Concentration

The Big Rock Point chloride ion concentration limit in theprimary coolant has been one (1) part per million since theplant start up in 1962. The Technical Specification for thechloride ion concentration was unintentionally changed from1 ppm to a less conservative limit of 140 ppm at the timeother approved changes were made. We agree with CPCo thatan error has been made in changing the technical specificationlimit from 1 to 140 ppa and that the mistake should becorrected by once again specifying a chloride concentrationlimit of 1 ppm as originally specified.

5. Winch Speed During Refueling

A 1/4 ton winch mounted on the fuel transfer cask has been usedover the past 15 years to raise and lower the 450 pound fuelassemblies used in the Big Rock Point core. Because the winchmanufacturer could not guarantee the winch performance above

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the rated 1/4 ton capacity and to provide added safety marginagainst .thepossibility of exceeding winch capacity limits,Consumers Power Company (CPCo) plans to replace the 1/4 tonwinch that has been used for Big Rock Point fuel assemblymovement into and out of the core with a new one ton winch.The administrative procedures that have been used to testand limit operation of the 1/4' ton winch will continue in

The 750 pound load limit cut off switch used oneffect.the 1/4 ton crane will also be retained (tested prior to use) inthe one ton winch during normal conditions for fuel withdrawalfrom the core.

CPCo reports, however, that the nbw winch speed is 112inches per minute compared to 96 inches per minute for the1/4 ton winch. The technical specification limits themaximum hoist speed to 100 inches per minute. The new

winch speed is, therefore, about 17% faster than the old1/4 ton winch and about 11% above the current technicalspecification limit.We have concluded that the difference in speed is notsignificant in considering the potential for (1) fuelassembly or core damage while lowering a fuel assemblytowards the top of the core, or (2) unintentional criticalitywhile lowering a new fuel assembly into a specified core

By letter dated March 21, 1977, CPC0 submittedposition.an evaluation to NRC that included fuel bundle drop onto

We have not completed our evaluation ofthe reactor core.this CPCo submittal but we agree with CPCo that the fuelbundle drop accident will not cause radioactive releasesto the environment in excess of 10 CFR 100. The accidentanalysis assumes that the bundle drops from the top of thevessel to the core (free fall). The free fall drop velocityis therefore faster than any winch speed and resulting

Ondamages can be equated with the insertion speed values.these bases,no winch speed limi,ts are required.

With respect to the potential for unintentional reactorcriticality while inserting a fuel bundle: The fastestinsertion time possible with the new one ton winch is 39seconds for the 72" Big Rock Point fuel bundle compared withthe 43 second time limit imposed by the technical specifications.In the unlikely situation of fuel bundle insertion at thefastest speed using either winch there is adequate timefor the winch operator to reverse direction and withdraw the

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bundle if there is any indication of an approach to anunplanned criticality. The neutron papulation is monitoredcontinuously during refueling operations to guard againstunplanned criticality, floreover, technical specificationsection 5.2.5 imposes reactor shutdown requirements thatprovide assurance that criticality will not be attainedwhile loading new fuel bundles into the core.

lDn the basis that fuel bundle or core damage is bounded by thefuel bundle drop analysis that shows radioactivity releaselimits are not exceeded even without containment isolation,that the new one ton winch speed charige from the 1/4 tonwinch is insignificant relative to unplanned criticalityconsideration, and that the standard technical specificationfor BWRs have no refueling hoist speed limits, we have~

concluded that the current technical specificationWeupper limit of 100 inches per minute is unjustified.

have concluded therefore, that the technical specification 2

transfer cask winch speed limits should be eliminatedas proposed by CPCo.

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Environmental Considerations

i d that this amendment does not authorize a change.We have determ nein effluent types or total anounts nor an increase ir, power level andwill not result in any significant environmental impact. Having

made this determination, we have further concluded that theamendment involves an action which is insignificant from thestandpoint of environmental impact and pursuant to 10 CFR 51.5(d)(4) that an environmental impact statement, or negative,

declaration and environmental appraisal need not be prepared '

in connection with the issuance of this amendment.

Conclusion

We have concluded, based on the considerations discussed above, that:4

(1) because the amendment does not involve a significant increasein the probability or consequences of accidents previously consideredand does not involve a significant decrease in a safety margin, theamendment does not involve a significant hazards consideration, (2)there is reasonable assurance that the health and safety of the publicwill not be endangered by operation in the proposed manner, and(3) such activities will be conducted in compliance with theCommission's regulations and the issuance of this amendment willnot be inimical to the common defense and security or to thehealth and safety of the public.

Date: June 24,1977

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