ML20002C940
| ML20002C940 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/24/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20002C936 | List: |
| References | |
| NUDOCS 8101120484 | |
| Download: ML20002C940 (9) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
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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 to the Technical Specifications for the Big Rock Point Plant:
1.
Revision of reactor vessel pressure temperature operating limits to comply with Appendix G of 10 CFR Part 50, (CPCo application dated May 30, 1975, as supplemented by letter dated June 30,1975),
2.
Incorporation of a provision to allow automatic bypassing of tne high condenser pressure reactor trip anytime steam drum pressures are less than 500 psig instead of the current 350 psig reactor coolant pressure limit (CPCo application dated September 10, 1975, as supplemented by letter dated May 25,1977),
3.
Inclusion of a definition of administrative control requirements associated with the air ejector off-gas monitoring system (CPCo application dated May 26, 1976, formerly included in proposed draft Technical Specifications forwarded with CPCo letter dated June 7, 1974),
4.
Correction of an error in Table 4.1.2(b) on chloride ion concentration limits in the primary coolant (CPCo application dated April 21,1977), and 5.
Deletion of the 100-inch per minute limitation on winch speed during refueling operations (CPCo application dated May 18, 1977).
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- Item 1 above contained other proposed changes to the Technical Specifications. Our review of the remaining portions of item 1 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 the requirements of Appendix G,10 CFR Part 50, " Fracture Toughness Requirements." The proposed pressure-temperature operating limits are based on the requirements of Appendix G,
10 CFR Part 50. These limits were developed from the test results of the Big Rock Point reactor vessel material surveillance program.
Three withdrawals of specimens have been made to date. These spygimens provjge radiation damage data at fluences of 1.5 x 10 9, 2.3 x 10 and 1.08 x 1020 n/cm2 The proposed operating limits are calculated for a fluence of 2.8 x 1019 n/cm. This ' fluence will be reached in about three 2
Jaars. The proposed Technical Specifications state that their pressure-temperature limits will be recalculated prior to exceeding 2.8 x 1019 NYT.
We conclude that the CPCo proposed pressure-temperature limits for operation of Big Rock Point comply with the requirements of 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 operation whenever condenser absolute pressure is greater than 8 inches of mercury by causing reactor scram. The purpose of the low vacuum trip is to assure that the main condenser is available to condense the steam exhausted from the main turbine-generator during normal power operation or from the turbine and turbine by-pass during abnormal transient conditions.
Low condenser vacuum during operation can cause overheating of the low pressure turbine casing.
Loss of condenser vacuum while at rated power. level can cause unacceptable reactor coolant pressure transients unless an anticipatory low vacuum signal causes the reactor to scram. The requirement for reactor scram with loss of condenser vacuum diminishes as 9
~ reactor power level decreases because the resultant reactor coolant pressure transients are smaller. At very low power level, loss of condenser vacuum is equivalent to a small steam line break with respect to radioactivity escaping to 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 steam to the condenser air ejector or faulty operation of the_ air ejector.
Condenser vacuum is maintained by a single twin element, two-stage steam-jet ejector which removes non-condensibles and de-aerates the condensate.
During plant start-up, the low condenser vacuum scram switch must be bypassed until the condenser vacuum can be pulled down to levels (25 inches of mercury) lower than the normal operational reactor scram set point (21 inches of mercury).
The existing technical specification limits reactor pressure to 350 psig for automatic condenser vacuum reactor scram switch by-pass.
CPCo has found that to produce a vacuum of 25 inches of Hg the reactor coolant pressure must be very close to the 350 psig technical specification limit.
Instrument drift has frequently resulted in violation of the technical specification 350 psig reactor coolant system pressure limitations for automatic ar,tuation of the low condenser vacuum reactor scram switch.
(PCu has therefore proposed that the 350 psig reactor coolant pre.ssure be increased to 500 psig steam drum pressure to provide a larger margin to the point where the reactor scram trip function might be required to operate thereby allowing greater flexibility during warm up of the main lines and the main condenser.
He agree with the CPCo statement that at steam drum pressures less than 500 psig, power level will be very low.
Even with the turbine stop valve open to wam up the turbine power level should not exceed 63.
Thennal practice, however, will be to pull the condenser vacuum to the required level (approximately 15" Hg) at some pressure less than the prooosed 500 psig steam drum pressure technical specification limit before automatic resetting of the condenser vacuum scram switch. After the 25" Hg vacuum has been reached, the turbine stop valve will open and the turbine will be warmed up.
Until turbine warm up begins, the steam demand is very low.
During turbine warm up and prior to elective power generation, reactor power level will generally not exceed 6*.' according to CPCo. As noted above, we agree that the power will be acceptably low.
Normally, turbine wam up does not begin until steam drum pressure is greater than 500 psig.
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, On the basis of this infomation, we have concluded that the 350 psig primary coolant technical specification limit for automatic bypass of the low condenser vacuum reactor scram switch is unnecessarily restrictive.
We have detennined that loss of condenser vacuum with the turbine stop valve closed could result in relatively small amounts of steam escaping into the turbine building. The release of radioactivity, however, is only a small fraction of the radioactivity released from a steam line break outside containment while at rated power level. The large steam line break accident has been analyzed and found to be acceptable. Therefore, the small steam line break, equivalent to loss of condenser during plant heat up, is also acceptable.
If loss of condenser vacuum is postulated to occur during turbine warm up before steam drum pressure reaches 500 psig (contrary to normal operating practices) and before the low condenser vacuum scram trip is automatically activated, the turbine stop valve will close automatically because of the trip which occurs et condenser absolute pressures in excess of 10 inches of mercury and the radiological conse-quences will be equivalent to the small steam line break that we have considered and which requires timely action of the operator to terminate. We have therefore concluded that the CPCo proposed change to allow automatic bypass of the high condenser pressure reactor trip when steam drum 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 Rock Point Plant audits continuously the level of radioactivity in the gases released from the main turbine condenser to the off-gas hold up pipe and 240 ft. stack.
During normal operation, these levels are very low compared with the allowable technical specification limits and the time delay between the off-gas system radiation monitor at the air ejector exhaust and the stack is at least one half hour. This delay permits off-gas system isolation and reactor shutdown before high radioactivity levels can reach the stack and environment.
In addition to prov'it;ng a continuous measure of the radiation in the centrolled release of gases from the main turbine condenser to the off-gas piping and stack,the off-gas monitoring system is also a fuel failui detection system.
The system therefore provides an alann at the control room to alert the operator if there is a significant increase in off-gas radioactivity. The alann set point can be no higher than allowed in technical specification 6.4.3.
Before the radiation levels reach the limits stated in this same technical specification, the isolation valve in the off-gas system must close automatically.
( i The proposed chanqe lowers the present technical specifi-cation alarm set point limit from 5 curies per second to about.5curiespgrgecondassumingEoftheproposedalarm set point limit,
-, is 0.845 MeV per disintearation.
E The value 0.845 MeV/ disintegration i535accordinq to CPCo., the average gamma ray energy for a pure U fission spectrum. This valug3gesults in a lower alarm limit set point than allowing for the
- 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 the off-gas system alarm set point will provide an earlier warning of an unlikely sudden increase in radioactivity that could be indicative of fuel failure.
The reduced setting should not interfere with normal plant operation since releases during normal plant operation have been reduced in recent years by a factor of 10 to 100 while the alarm set point limit is hereby reduced by only a factor of about 10.
The lowered alarm set point allows more time for corrective actions by the operator and is, therefore, acceptable. The technical specifications should.therefore be changed as proposed.
CPCo has also proposed to reduce power level immediately if the off-gas system alarm set points are reached. According to the proposal, the operator would reduce power level until the off-gas radiation levels are less than the off-gas alarm set point lic:its.
CPCo reported that a new off-gas isolation valve war, installed at Big Rock Point and that extensive maintenance was performed to improve the off-gas system leak tight-However, tests performed later during stable power ness.
conditions (January 31,1976) showed that when the isolation valve was closed the stack gas release rate decreased
' initially but later as the off-gas hold up p; essure increased significantly the stack gas release rate increased to a value slightly in excess of the initial value.
During the test, condenser vacuum was not noticeably reduced, and therefore there was no i Tactor scram resulting from low reactor vacuum.
CPCo has concluded that it is rot feasible to reduce stack gas releases by automatic isolation valve closure followed by reactor scram due to loss of condenser vacuum and has proposed administrative actions to require that the operator reduce reactor power level whenever specified off-gas radiation alarm limits are exceeded.
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.. We note that:
- Testing at Big Rock Point has shown that closure of the off-gas system isolation valve causes only a temporary reduction of the stack gas release rate, i.e. isolation valve closure does not prevent continued reactor operation with radio-active gaseous release rates in excess of the technical specification limit as CPCo and we believe was originally intended.
- During the 15 year operating history of the Big Rock Point Plant, there has never been a need for automatic closure of the off-gas system isolation valve to remain within technical specification limits.
In fact, the highest measured stack gas release rate,78,000 micro-curies per second, occurred in 1966. This value is nearly a factor of 10 lower than the proposed alarm set point limit and more 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 100,000 micro-wries per second do not oce ur rapidly.
Increase!, when and if they occur, are easily controlled by reducing reactor poter level.
- The off-gas radioactivity release rate during the last year of Big Rock Point operation has been the lowest in the history of opera +,fonLI). The off-gas release rate stabilized following start-up for fuel cycle No. 14 at approximately 750p Ci/s. The release rate near the end of fuel cycle No.14 is j
averaging 700-800p Ci/s.
Power level over the period has ranged betwaan 206 and 216 MWt most of the time. This historically low radiation level is attributed by the licensee, to the gradual removal of copper based crud from the primary system by way of deposits on spent fuel assemblies removed from the core. As a result of improved fuel rod design and fabrication and reduced crud deposition on the (I) Big Rock Point Technical Specification Change Request, G-3, April 15, 1977 l
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. fuel rod. surfaces, the number of leaking fuel rods is noticeably reduced. Consequently: 'che amount of gaseous radioactivity escaping from the fuel rods to the atmosphere through the primary cooling system, the main-turbine-condenser, the off-gas exhaust system and the exhaust stack is -
significantly reduced in contrast to earlier years of operation when stainless and inconel fuel cladding, developmental fuel assemblies with centermelt fuel and many components containing copper were used. These efforts to improve fuel rod integrity have resulted in a significant reduction in radioactive off-gas released from the stack during normal plant operation.
We have concluded that operator action to reduce power level if radioactive off-gas alarm set point limits are reached is acceptable based on the operating experience to date at Big Rock Point. The normal gaseous release levels are relatively low compared with the alarm set point and fuel rod cladding failure experience indicates that the operator can react in ample time to prevent release levels in excess of limits. We have also concluded that the CPCo proposal to administratively require reactor power level reductions whenever a high radiation stack gas release alarm is activated provides added assurance that the 10 Ci/sec gaseous release limit will never be attained.
In this respect, the safety margin is increased.
Both technical specification changes further enhance the health and safety of the public and should therefore be made.
4.
Chloridc lon Concentration The Big Rock Point chloride ion concentration limit in the primary coolant has been one (1) part per million since the plant start up in 1962. The Technical Specification for the chloride ion concentration was unintentionally changed from 1 ppm to a less conservative limit of 140 ppm at the time other approved changes were made. We agree with CPCo that an error has been made in changing the technical specification limit from 1 to 140 ppa and that the mistake should be corrected by once again specifying a chloride concentration limit of 1 ppm as originally specified.
5.
Winch Speed During Refueling A 1/4 ton winch mounted on the fuel transfer cask has been used over the past 15 years to raise and lower the 450 pound fuel assemblies used in the Big Rock Point core.
Because the winch manufacturer could not guarantee the winch performance above
' the rated 1/4 ton capacity and to provide added safety margin against.thepossibility of exceeding winch capacity limits, Consumers Power Company (CPCo) plans to replace the 1/4 ton winch that has been used for Big Rock Point fuel assembly movement into and out of the core with a new one ton winch.
The administrative procedures that have been used to test and limit operation of the 1/4' ton winch will continue in The 750 pound load limit cut off switch used on effect.
the 1/4 ton crane will also be retained (tested prior to use) in the one ton winch during normal conditions for fuel withdrawal from the core.
CPCo reports, however, that the nbw winch speed is 112 inches per minute compared to 96 inches per minute for the 1/4 ton winch. The technical specification limits the maximum hoist speed to 100 inches per minute.
The new winch speed is, therefore, about 17% faster than the old 1/4 ton winch and about 11% above the current technical specification limit.
We have concluded that the difference in speed is not significant in considering the potential for (1) fuel assembly or core damage while lowering a fuel assembly towards the top of the core, or (2) unintentional criticality while lowering a new fuel assembly into a specified core By letter dated March 21, 1977, CPC0 submitted position.
an evaluation to NRC that included fuel bundle drop onto We have not completed our evaluation of the reactor core.
this CPCo submittal but we agree with CPCo that the fuel bundle drop accident will not cause radioactive releases to the environment in excess of 10 CFR 100. The accident analysis assumes that the bundle drops from the top of the vessel to the core (free fall). The free fall drop velocity is therefore faster than any winch speed and resulting On damages can be equated with the insertion speed values.
these bases,no winch speed limi,ts are required.
With respect to the potential for unintentional reactor criticality while inserting a fuel bundle: The fastest insertion time possible with the new one ton winch is 39 seconds for the 72" Big Rock Point fuel bundle compared with the 43 second time limit imposed by the technical specifications.
In the unlikely situation of fuel bundle insertion at the fastest speed using either winch there is adequate time for the winch operator to reverse direction and withdraw the L
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' bundle if there is any indication of an approach to an unplanned criticality. The neutron papulation is monitored continuously during refueling operations to guard against unplanned criticality, floreover, technical specification section 5.2.5 imposes reactor shutdown requirements that provide assurance that criticality will not be attained while loading new fuel bundles into the core.
lDn the basis that fuel bundle or core damage is bounded by the fuel bundle drop analysis that shows radioactivity release limits are not exceeded even without containment isolation, that the new one ton winch speed charige from the 1/4 ton winch is insignificant relative to unplanned criticality consideration, and that the standard technical specification for BWRs have no refueling hoist speed limits, we have
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concluded that the current technical specification We upper limit of 100 inches per minute is unjustified.
have concluded therefore, that the technical specification 2
transfer cask winch speed limits should be eliminated as proposed by CPCo.
Environmental Considerations i d that this amendment does not authorize a change
.We have determ ne in effluent types or total anounts nor an increase ir, power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint 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 increase in the probability 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 public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted 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.
Date:
June 24,1977
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