ML20027C372
| ML20027C372 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 10/12/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| TASK-15-09, TASK-15-9, TASK-RR LSO5-82-10-028, LSO5-82-10-28, NUDOCS 8210150375 | |
| Download: ML20027C372 (7) | |
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.i October 12, 1982 Docket No. 50-409 LS05-82-10-028 Mr. Frank Linder General Manager Dairyland Power Cooperative 2615 East Avenue South Lacrosse, Wisconsin 54601
Dear Mr. Linder:
SUBJECT:
SEP TOPIC XV-9, STARTUP OF AN INACTIVE LOOP OR RECIRCULATION LOOP AT AN INCORRECT TEMPERATURE - LACROSSE BOILING WATER REACTOR (LACBWR)
By letter dated August 25, 1981 (LAC-7756), you submitted a safety assessment report for the above topic. The staff has reviewed this assessment and our conclusions are presented in the enclosed safety evaluation report, which completes this topic for the Lacrosse Boiling Water Reactor.
This evaluation will be a basic input to the integrated assessment for your facility. The evaluation may be revised in the future if your facility design is changed or if NRC criteria relating to this topic is modified before the integrated assessment is conpleted, j
Sincerely,
$,04 f
usE Dennis M. Crutchfield, Chief pta Operating Reactors Branch #5 Division of Licensing
Enclosure:
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- NRC FORM 318 (10 80) NRCM 0240 OFF1ClAL RECORD COPY usam mmeso
Mr. Frank Linder CC Fritz Schubert, Esquire U. S. Environmental Protection Staff Attorney Agency Dairylar.d Power Cooperative Federal Activities Branch.
2615 East Avenue South Region V Office La Crosse, Wisconsin 54601 ATTN: Regional Radiation Representative 230 South Dearborn Street O. S. Heistand, Jr., Esquire Chicago, Illinois 60604 Morgan, Lewis & Bockius 1800 M Street, N. W.
Janes G. Keppler, Regional Administrator Nuclear Regulatory Commission, Region III Washington, D. C.
20036 799 Roosevelt Road Mr. John Parkyn Glen Ellyn, Illinois 60137 La Crosse Boiling Water Reactor Dairyland Power Cooperative Mr. Ralph S. Decker P. O. Box 275 Route 4, Box 190D Genoa, Wisconsin 54632 Cambridge, Maryland 21613 Mr. George R. Nygaard Charles Bechhoefer, Esq., Chairman Coulee Region Energy Coalition Atomic Safety and Licensing Board 2307 East Avenue U. S. Nuclear Regulatory Conmission La Crosse, Wisconsin 54601 Washington, D. C.
20555 Dr. Lawrence R. Quarles Dr. George C. Anderson Kendal at Longwood, Apt. 51 Department of Oceanography Kenneth Square, Pennsylvania 19348 University of Washington Seattle, Washington 98195 U. S. Nuclear Regulatory Commission Resident Inspectors Office Rural Route #1, Box 276 Genoa, Wisconsin 54632 Town Chairman Town of Genoa Route 1 Genoa, Wisconsin 54632 Chairman, Public Service Commission of Wisconsin Hill Farms State Office Building Madison, Wisconsin 53702 e
0 9
LACROSSE SEP TOPIC XV-9 STARTUP OF AN INACTIVE LOOP I.
INTRODUCTION The startup of an inactive or idle loop at an incorrect temperature and the malfunction of a recirculation flow controller are examined to assure that the consequences are acceptable.
The guidance for the review of this topic is provided by SRP Sections 15.4.4 and 15.4.5.
This transient is evaluated because the addition of cooler water to the core reduces the core void fraction, which causes an increase in power and reduces thermal margins.
The calculated Minimum Critical Power Ratio (MCPR) is compared to the MCPR safety limit to demonstrate that fuel failures will not occur.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structueesi systemsr and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, including determination of the margins of safety during normal operations and transients conditions anticipated during the life of the facility.
Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radioactivity.
4
_2 The General Design Criteria (Appendix A to 10 CFR Part 50) set forth the criteria for the design of water-cooled reactors.
GDC 10 " Reactor Design" requires that the core and associated cooling, control and protection system be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during normal operationi including the effects of anticipated operational occurrences.
GDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated operational occurrences.
GDC 20 " Protection System Functions requires that the protection i
system be designed to initiate automatically the operation of i
reactivity control systems to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences.
GDC 26 " Reactivity Cc~t J System Redundancy and Capability" -
requires that t h+
r.. e '.ity control system be capable of reliably controlling reactivity ch)nges to assure that under conditions of normal operations includin: anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded.
e
. GD C 28 " Reactivity Limits" requires that the reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core.
III.
R EL AT ED SAFETY TOPIC The review is conducted in accordance wi t h SR P 15.4.4 and 15.4.5.
The Evaluation includes review of the analysis for the event and identification of the features in tf 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.
D eviations f rom the criteria speci-fied in the Standard Review Plan are identified.
IV.
EVALUATION Two types of events were analyzed as part of this review:
Inactive Loop Startup from hot conditions and the recirculation flow control failure.
The Licensee states that the Inactive Loop Startup event was the most severe of the two events and presents only the results of the Inactive Loop Startup.
The startup of an inactive loop from a cold condition, by opening the bypass valves, was found to be bounded by the increase in feedwater flow transient which has been reviewed and found accept-able as part of SEP Topic XV-1 9
_4-The Inactive Loop Startup Event described in the reference assumes the opening of a loop discharge valve while the reactor is at power.
The main assumptions include:
1.
Initial power at 52%.
2.
The isolated loop is 25 F Lower than the operating loop.
3.
The isolated loop flowrate reaches the active loop flowrate in 20 seconds.
4.
No credit is taken for reactor scram.
The idle loop discharge valve is assumed to open in spite of the interlock which is set to prevent valve opening when the Loop temperatures differ by more than 10 F.
The valve opening time used in the analysis is 20 seconds rather than the 4 minutes actually required.
End of cycle core physics parameters are used for conservatism.
The results of the analyses show that the peak reactor power of 130% is reached 18 seconds after startup of the inactive loop and that a steady power Level of 96% occurs 50 seconds after loop startup.
The licensee states that t h e CPR stays above the minimumCPR limit of 1.32 so no fuel damage occurs.
The peak pressure for att transients analyzed is 1540 psia, which is less than the ASME emergency condition design overpressure limit of 1690 psia.
No credit was taken for reactor scram or other protection systems.
The licensee states that their analysis of the recirculation flow controtter failure, which is a more rapid transient, had a
.~
. l lower peak heat flux than the startup of an inactive loop, and is therefore bounded by the startup of an inactive loop event.
V.
CONCLUSION The staff has reviewed the Lacrosse Boiling Water Reactor analysis of SEP Topic XV-9."Startup of an Inactive Loop and Recirculation Flow Controller Malfunction."
The results of the analysis indicate that the Lacrosse plant is in conformance with SRP sections 15.4.4 and 15.4.5 requirements and is acceptable,
Reference:
Letter to R. W. Reid, NRC, from J. P. Madgett, Dairyland Power Coop, dated February 25, 1977.
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