ML20032B375

Forwards Evaluation of SEP Topic XV-4 Re Loss of Nonemergency Ac Power to Station Auxiliaries.Review Will Be Basic Input to Integrated Safety Assessment

ML20032B375
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	11/03/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Linder F DAIRYLAND POWER COOPERATIVE
References
TASK-15-05, TASK-15-5, TASK-RR LSO5-81-058, LSO5-81-58, NUDOCS 8111050483
Download: ML20032B375 (8)

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Docket No. 50-409 1 y

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Mr. Frank Linder c

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Dairyland Power Cooperative 2615 East Avenue South Lacrosse, Wisconsin 54601

SUBJECT:

LACROSSE - SEP TOPIC XV-4, LOSS OF NON-EMERGENCY A-C POWER TO THE STATION AUXILIARIES In your letter dated June 26,1981 (LAC-7632) you submitted a safety assessment report on the above topic. The staff has reviewed your assessment and our conclusions are presented in the enclosed safety evaluation report, which completes this topic for the Lacrosse Boiling Water Reactor (LACBWR).

The enclosed safety evaluation will be a basic input to the integrated safety assessment for your facility. The assessment may be revised in the future if your facility design is changed or if NRC criteria relati..J to this topic are modified before the integrated assessment is completed.

Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing y

Enclosure:

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use, m m - u G OFFICIAL RECORD COPY hAC FORM 318 (10-60) NRCM 024a

Mr. Frank Linder c'c Fritz Schubert, Esqu' ire U. S. Environmental Protection Staff Attorney Agency Dairyland Power Cooperative Federal Activ?;ies Branch 2615 East Avenue South Region V Office La Crosse, Wisconsin 54601 ATTN:

EIS C0ORDINATOR 230 South Dearborn Street

0. S. Heistand, Jr., Esquire Chicago, Illinois 60604 Morgan, Lewis & Bockius 1800 M Street, N. W.

Mr. John H. Buck Washington, D. C.

20036 Atomic Safety and Licens,ing Appeal Board U. S. Nuclear Regulatory Conimissiion'

' Mr. R. E. Shimshak Washington, D. C.

20555 La Crosse Boiling Water Reactor Dairyland Power Cooperative Dr. Lawrence R. Quarles P. O. Box 135 Kendal at Longwood, Apt. 51 Genoa,. Wisconsin 54632 Kenneth Square, Pennsylvania 19348 Ms. Anne K. Morse Charles Bechhoefer, Esq., Chairman Coulee Region Energy Coalition Atomic Safety and Licensing Board P. O. Box 1583 U. S. Nuclear Regulatory Commission La Crosse, Wisconsin 54601 Washington, D. C.

20555 La Crosse Public Library Dr. George C. Anderson 800. Main Street Department of Oceanography La Crosse, Wisconsin 54601 University of Washington Seattle, Washington 98195 U. S. Nuclear Regulatofy Commission Resident Inspectors Office Mr. Ralph S. Decker Rural Route #1, Box 276 Route 4, Box 1930

- Genoa, Wisconsin 54632 Cambridge, Maryland 21613 Town Chairman Thomas S. Moore Town of Genoa Atomic Sefe+y sad Licensing Appeal Board Route 1 U. S. Nuclear Regulatory Commission Genoa, Wisconsin 54632 Washington, D. C.

20555 Chairman,' Public Service Commission of Wisconsin Hill Farns State Office Building Madison, Wisconsin 53702 Alan S. Rosenthal, Esq., Chairnen Atomic Safety and Licensing-Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Mr. Frederick Milton Olsen, III 609 North lith Street Lacrosse, Wisconsin ~ 54601

r LACROSSE BOILING WATER REACTOR (LACBWR)

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SEP TOPIC:

XV-4 Loss of Non-Emergency A-C Power to the Station Auxiliaries I.

INTRODUCTION The loss of ac power to station auxiliaries while the plant is at power result.s in loss of power to:

(1) the forced recirculation pumps (loss of reactor coolant flow), (2) the circulating water pumps (loss of main condenser cooling), and (3) the main feedwater pumps (loss of feedwater)'.

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II.

REVIEW CRITERIA 4

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Section 50.34 of 10 CFR Part~ 50 requires that each applicant for a construction permit or operating license. provide an analysis and 4

e'aluation of the design and performance of structures, systems, v

and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facilit'y, including determination of the margins of safety during normal operations and transient conditions anticipated durina the life of the facility.

The General Design Criteria (dppendix A to 10 CFR Part 50) establish

' minimum requirements for the principal design criteria for water-cooled reactors.

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GDC 10 " Reactor Design" requires that the core and associated coolant, control and prot'ection systems be designed with appropriate margin to assure that specified acceptable fuel _ design limits are not exceeded during normal operation, including the effects of anticipated operational occurrence.

'GDC 15 "Reac. tor 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 norm'al operation, including the effects of anticipated operational occurrences.

GDC 26 " Reactivity Control System Redundance and Capability" requires that the reactivity control systems be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipa'ted operational occurrences, and with appropriate margin for malfuntions such as stuck rods, specified acceptable fuel design limits are not exceeded.

II'.

RELATED SAFETY TOPICS

'Various other SEP topics evaluate such items as the reactor protection system and the-adequacy of onsite and offsite power sources.

The effects of single failures on safe shutdown capability are considered under Topic VII-3.

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IV. ~ REVIEW GUIDELINES l

The' review is conducted in accordance with SRP 15.2.6.

The evaluation includes review of the analysis for the event and.

identificatien of ~the features in the plant'that mitigate,the consequences -

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.of the event as well as the. ability.of these systems:to functioti as re-

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quired.

The extent to which operator action is. required is'also evaluated.

Deviations from the c'riteria specified in the Standard j

' Review Plan are identified.

V.

EVALUATION The Lacrosse "lant protection system is designed to trip the turbine on the loss of non-emergency A-C power and to partially scram'the reactor.

1 if 2400 volts is lost to either of two buses for longer than the time required for automatic switching to reserve power. The plant protection system is designed to scram all of the control rods if voltage is lost to both 2400 volt busses or to both Essential Busses IA and 1B. When the gen'erator trips, 'the power to the forced circulation pumps and to

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the circulating water pumps is interrupted.

The subsequent coasting down of the recirculation flow'and sudden loss of feedwater make the-l loss of non-emergency A-C power transient different from transients analyzed under topic XV-3, which cauce a decrease, in heat removal by the secondary system.

The Dairyland Power Cooperative (DPC) presented calculations (Ref.1) which show that the Critical Power Ratio (CPR) stays above the minimum value of i.32 (Ref. 2) throughout the transient.

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. In the survey of the analyses made for Reference 1 it was found (Ref. 3) that the inost severe transient resulting from a decrease in coolant flow through. the core is the loss of power to both recirculation pumps. -Thus the CPR in a. loss of A-C power transient

.is bounded by the CPR following the trip of both recirculation pumps.

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The initial conditions and assumptions used in the analysis ~of the" loss of recirculation flow transient (Ref. 4) are as follows:

1.

The reactor is initially operating at 102% of rated power. -

2.

Both recirculation pumps are lost.

3.

No credit is taken for a scram resulting from the event that caused the loss of both pumps.

4.

The reactor scrams when the recirculation flow is 30% of full fl ow.

5.

The reactor is operating at the.begirning of. a fuel cycle.

This analysis she,s that (1) the reactor power rapidly decreases due to th: increase in voids in the core as the recirculation flow decreases, (2) the loss of feedwater causes a decrease in the core wat;r level which closes the main steam isolation valves and starts the shutdown condenser and (3) the CPR stays above 1.32 at al'1 times during the transient.

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VI.

CONCLUSION As part of the SEP review of Lacrosse, the analysis for 1.oss of non-

emerger.cy A-C power has been evaluated and we have concluded that the consequences of this event are bounded by a sudden loss of recirculation

. flow which will' be evaluated under SEP Topic XV-7.

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REFERENCES 1.

Nuclear Energy Services, Inc. report for DPC: 81A0025; Response to Question 4, Transient Analysis for LACBWR Reload Fuel; February 18, 1977.

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2.

Letter from G. Lear of NRC to W. Nechodon of Exxon Nuclear Power Company, on Topical Report Evaluation, June 23, 1976.

3.

Nuclear Energy Services, Inc. report for DPC: 81A0025; _Re_sponse to Questien 4, Traresient Analysis for LACBWR Reload Fuel; February 18, 1977; page 8.

4.

Letter from Frank Linder ci DPC to D. G. Eisenhut of NRC; Dairyland Power Cooperative Lacrosse Boiling Water Reactor (LACBWR) Provisional Operating License No. DPR-45 SEP Topic XV Loss of External Load, Turbine Trip less of Condenser Vacuum, Closure of MSIV & Steam Pressure Regulating. Failure,

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and SEP Topic XV Loss of Non-Emergency AC Power to the Station Auxiliaries; June 26 1981.

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