ML20005C035
| ML20005C035 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 11/13/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| TASK-15-03, TASK-15-3, TASK-RR LSO5-81-11-018, LSO5-81-11-18, NUDOCS 8111180302 | |
| Download: ML20005C035 (9) | |
Text
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November 13, 1981 Docket No. 50-409 Ashf ~
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s ll Mr. Fra n Linder
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Dairyland Power Cooperative 2615 East Avenue South
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%(UTD c,6 Lacrosse, Wisconsin 54601 s
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Dear Mr. Linder:
SUBJECT:
LACROSSE - SEP TOPIC XV-3, LOSS OF EXTERNAL LOAD, TURBINE TRIP, LOSS OF CONDENSER VACUUM, CLOSURE OF MAIN STEAM ISOLATION VALVE, STEAM PRESSURE REGULATOR FAILURE (CLOSED)
In your letter dated June 26,1931 (LAC-7632) you submitted a safety assersment 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
(
'ater 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 HRC criteria relating to this topic are modified before the integrated assessment is conpleted.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licer. sing 56c>Y l
Enclosure:
As stated
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Mr. Frank Linder cc Fritz Schubert, Esquire U. S. Environmental Protection Staff Attorney Agency Dairyland Power Cooperative Federal Activities Branch 2615 East Avenue South Region V Office La Crosse, Wisconsin 54601 ATTN: Regional Radiation Representative 230 South Dearbdtn Street O. 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 Licensing Appeal Board U. S. Nuclear Regulatory Commission 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 Regulatory Commission Resident Inspectors Office Mr. Ralph S. Decker P. ural Route #1, Box 276 Route 4, Box 190D Genoa, Wisconsin 54532 Cambridge, Maryland 21613 Town Chairman Thomas S. Moore Town of Genoa Atomic Safety and Licensing Appeal Board Route 1 U. S. Nuclear Regulatory Connission Genoa, Wisconsin 54632 Washington, D. C.
20555 Chairman, Public Service Commission of Wisconsin Hill Farms State Office Building Madison, Wisconsin 53702 Alan S. Rosenthal, Esq., Chairman 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, Wirconsin 54601
LACROSSE BOILING WMER REACTOR (LACBWR)
TOPIC:
XV-3, LOSS OF EXTERNAL LOAD, TURBINE TRIP, LCSS OF CONDENSER VACUUM, CLOSURE OF MAIN STEAM 19-ATION VALVE (BWR), AND STEAM PRESSURE REGULATORY FAILURE (CLO! J) 1.
INTRODUCTION
~ The events considered in this topic involve a decrease in secondary heat ramoval.
This decrease can cause a sudden increase in reactor pressure.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a con-struction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, including determina-tion of the margins of safety during normal operations and transient conditions anticipated during the life of the fccility.
The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water-cooled reactors.
GCC 10 "Reactc; Design" requires that the core and associated coolant, control and protection 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 " 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 occurrence.
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 anticipated operational occurrences, and with appropriate margin for mal-function.; such as stuck rods, specified acceptable fuel design limits are not exceeded.
III.
RELATED SAFETY TOPICS Various other SEP topics evaluate such items as the reactor protection system.
The effects of single failures on safe shutdown capability are considered under Topic VII-3.
~ - - ~
. IV.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.2.1, 15.2.2, 15.2.3, 15.2.4, (BWR only) and 15.2.5.
V.
INDIVIDUAL EVENT EVALUATIONS Loss of External Load A.
Introduction A loss of external load can result from the opening of electrical circuit breakers or other external electrical malfunctions.
This loss of load to the generator results in the turbine accelerating toward the overspeed trip point.
TFe turbine inlet valves close down rapidly due to the speed governor, attempting to control turbine speed at approximately 105% of rated speed, causing reactor pressure to increase.
The sudden increase in pressure causes a high flux scram to occur if the initial power level is greater than 60% power.
Tne overspeed trip on the turbine occurs a short time later (within a few seconds) and causes the turbine stop valve to close, which causes a scram signal to some of the control rods.
The increase in reactor pressure causes the main steam bypass valve to automatically actuate to maintain pressure.
Following the scram, the main steam bypass valve continues to operate at intermittent in-tervals until the pressure begins to decay slowly due to heat losses from the primary purification system, steam to the gland steam genera-tor and air ejectors.
The Dairyland Pot - Cooperative (DPC) presented an analycis of the loss of external load w nsient, dated 2/28/74, in Volume 3 of the Applica-tion far an Operating License (Ref.1).
The results of a reanalysi.
of this transient are given in a report to NRC dated 2/25/77 (Ref. 2).
G.
Evaluation The loss of load transient is bounded by the turbine trip event.
Reactor scram is initiated directly by signals from closure of the turbine control valves.
During the loss of load transient the steam flow to the turbine is interrupted by closure of the turbine control valves, while for the turbine trip transient the flow is interrupted by closure of the stop valves which is more rapid than control valve closure.
Thus the transient during loss of load is less severe than the turbine trip transient.
I
. C.
Conclusions As part of the SEP review for Lacrosse we have evaluated the licensee's analysis of loss of external load (Ref. 2) against the criteria for SRP Section 15.2.1.
Ba ad on this evaluation we have concluded that this transient is bounded by the turbine trip event which has been evaluated and found in conformance with the criteria of SRP Section 15.2.1.
Turbine Trip A.
Introduction A turbine trip is actuated by fast closure of the turbine stop valves which abruptly interrupt steam flow to the turbine.
Independent of the cause, a turbine trip f s followed by a reactor scram initiated directly by turbine stop valve position switches.
~he effect of turbine trip is rapid increase in pressure in the steam
'es ar.d reactor vessel.
presented an analysis of the turbine trip transient, dated 2/28/74, Volume 3 of the Application for an Operating License (Ref.1).
,nis analysis goes beyond the requirements of SRP Section 15.2.1 in assuming the control rods fail to in3ert upon the receipt of the scram signal, i.e., an ATWS. Since NRC's criteria are different for ATWS events, the turbine trip transient was reanalyzed and the results are given in a report to NRC dated 2/18/77 (Ref. 3).
B.
Evaluation In the 2/28/74 analysis (Ref.1), which was done as an Anticipated Transient Without Scram, it was assumed that the reactor was at 100%
power at the time of th? turbine trip and the control rods did not move in during the transient.
The results of ;his analysis, assuming operation of the turbine bypass and relief systems, indicated that the pressure peaks at about 1315 y sia which is below the design pressure of 1400 psig.
In the 2/18/77 analysis (Ref. 3) it was assumed that the reactor was at 102% power at the time of the turbine trip and that the reactor was operating at the end of the fuel cycle at which time the delayed neutron fraction was calculated to be.0055.
A value of 20 microseconds was used for the neutron lifetime.
The results of this analysis showed that the pressure increase was less than 40 psi and that minimum critical power ratio (MCPR) stayed above 1.32, which is the established CPR criter'on based on the Exxon XN-2 critical heat flux correlation which was approved by the NRC on 6/23/76 (Ref. 4).
. Furthermore, the results of analyses which assume that the turbine hypass is unavailable indicated that pressure peaks at about 1365 psia and that the MCPR stayed above 1.32.
This peak pressure of 1365 psia is still below the design pressure of 1400 psig.
C.
Conclusions As part of the SEP review for Lacrosse we have evaluated the licensee's analysis of the turbine trip event (Ref. 3) against the criteria of SRP Section 15.2.1.
Based on this evaluation we have concluded that the analyses performed adequately bound the turbine trip analysis as required by SRP Section 15.2.1.
We therefore, find the results of the turbine trip analyses acceptable.
Introduction the extreme case of sudden loss of concensyr vacuum the transient ald be identical to the turbine trip transient with failure of bypass.
.e most limiting single failure dhring the transient would be a relief valve failure to open.
The licensee has not presented an ana?ysis of loss of condenser vacuum, but has referenced the results of turbine trip transients LRef, 3),
B.
Evaluation The worst case loss of condenser vacuum transient is identical to the turbine trip transient with failure to bypass.
However, since loss of condenser vacuum results in a loss of bypass, an additional single fail-ure should be assumed to satisfy the SRP 15.2.1, section II acceptance criterion 2d.
The most limiting single failure that could produce the highest pea!.
ressure is a relief valve failure to open.
However, this event is unded by the turbine trip analysis performed assuming the turbine ass and relief valves are not available.
A relief valve failure to a would not liifiuence the minimum MCHFR because this minimum is at-(
.ned before any of the relief valves opens.
L.
Conclusions As part of the SEP review for Lacrosse we have evaluated the licensee's analysis of loss of condenser vacuum (Ref. 5) against the criteria of SRP Section 15.2.1.
Based on this evaluation we have concluded that this transient is bounded by the turbine trip event which has been evaluated and found in conformance with the criteria of SRP Section 15.2.1.
c
v Closure of Main Steam Isolation Valve A.
Introduction Inadvertent closure of the main steam isolation valves results in loss of the ste.am removal path from the reactor to the turbine and may cause vessel overpressurization. A full scram signal is initiated when the main steam isolation valve leaves the open position.
The licensee has analyzed (Ref. 5) closure of the main steam isolation valve with the following initial conditions and assumptions:
- 1) The reactor is initially operating at 102% of rated power.
- 2) The Main Steam Line Isolation Valve closes in 6.5 seconds.
- 3) No credit is taken for the reactor scram and shutdown condenser operation caused by the MSIV closure.
- 4) The reactor is operating at the end of a fuel cycle.
- 5) When the reactor scrams due to 102% overpower recirculation flow is cut back to 80% of full power.
- 6) Operation of the shutdown condenser is initiated when reactor pres-sure is greater than 1 3 psig.
Evaluation s.
The closure of the main steam isolation valve in 6.5 seconds results in a rise in reactor pressure which collapses voids in the core and causes a sharp increase in reactor power.
Six and one-half seconds is a minimum value which causes the maximum pressure rise.
Ten seconds is the time interval allowed by the Technical Specificat %ns. At approximately 1.5 seconds, the reactor reaches 120% of full power and i
This causes the recirculation pumps to cut back to 80% of full i
flow which in turn reduces the reactor power.
Reactor power continues to decay as the control rods are inserted. At approximately 6 seconds i
operation of the shutdown condenser is initiated by a reactor pressure of 1325 psia.
The reactor pressure continues to increase to about 1365 psia.
This is well below the limit (110%) of 1540 psig.
The critical power ratio stays above the 1.12 limit.
l C.
Conclusion The analysis on main steam isolation valve closure has been esaluated against the criteria of SRP 15.2.1 and we have concluded that it is in conformance with the criteria.
_, _. ~. _. - _ _ _..
. Steam Pressure Reculator Failure A.
Introduction In case of a steam pressure regulator failure in the direction of decreasing flow +he turbine control valve starts to close.
This causes an incre % in reactor pressure to the setpoint of the mairi steam bypass va..
which opens to limit the increase in pressure.
As the bypass valve opens the generator output goes down until the reverse power relay actuates to open the output breaker and close the turbine stop valve.
This results in a partial scram.
B.
Evaluation The event induces a very mild transient on the plant (Ref. 5).
In the case of the most limiting single failure the transient is bounded by the turbine trip analyses.
C.
Conclusions Steam pressure regulator failure is not as limiting as the turbine trip transient and a quantitative analysis of its consequences is not needed.
VI.
TOPIC CONCLUSIONS For each of the events included in this topic, the staff has determined either that the event is bounded by another event, or that the analysis provided is in compliance with the criteria.
Therefore, this topic is complete.
REFERENCES
~
1.
Gul? Nuclear Project 3431 for DPC; Anticipated Transients Without Scram :t the Lacrosse Boiling Water Reactor; February 28, 1974; page 3-3 2.
Letter for J. P. Madgett of DPC to R. W. Re.id of NRC entitled DAIRYLAND POWER COOPERATIVE LACROSSE BOILING WATER REACTOR PROVISIONAL OPERATING LICENSE NO. DPR-45 APPLICATION F0ii ~
AMENDMENT TO LICENSE dated Februsry 25, 1977.
3.
Nurlear Energy Services, Inc. report for DPC; JLesponses to Qsestion 4 Transient Analysis for LACBWR Re?oad Fuel; February 18, 1977.
1 4.
Letter from G. Lear of NRC to W. Nechodon of Exxon Nuclear Power Company, on Topical Report Evaluation, June 23, 1976.
5.
Letter from Frank Linder of 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 0F EXTERNAL LOAD, TURBINE TRIP LOSS 0F CONDENSER VACULN, CLOSURE OF MSIV & STEAM PRESSURE REGULATING FAILURF, AND SEP TOPIC XV LOSS OF NON-EMERGENCY AC POWER TO THE STATION AUXILIARIES; June 26, 1981.
1.
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