ML20005B662
| ML20005B662 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 06/30/1981 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| TASK-15-15, TASK-RR LAC-7637, NUDOCS 8107080512 | |
| Download: ML20005B662 (6) | |
Text
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e DlDA/RYLAND
[k COOPERAT/W
a (608) 788-4000 June 30, 1981 In reply, please refer to LAC-7637 DOCKET NO. 50-409 U. S. Nuclear Regulatory Commission
,gIgj ATTN:
Mr. Darrell G. Eisenhut, Director s
eg Division of Licensing 4
'f Office of Nuclear Reactor Regulation g'
Division of Operating Reactors
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Washington, D. C.
20555 9_
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SUBJECT:
DAIRYLAND POWER COOPERATIVE g
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/Cj LA CROSSE BOILING WATER REACTOR (LACBWR)
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PROVISIONAL OPERATING LICENSE NO. DPR-45 g jTy,Y SEP TOPIC XV INADVERTENT OPENING OF A LACBWR SAFETY VALVE
Reference:
(1)
DPC Letter, LAC-7387, Linder to Eisenhut, dated February 27, 1981.
Gentlemen:
Encl.osed find Safety Evaluation Report (SER) for Inadvertent Opening of a LACBWR Safety Valve (SEP-XV-15) which we have pre-pared for the La Crosse Boiling Water Reactor.
Our letter, Reference 1, identified topics for DPC to submit for NRC evaluation.
The subject topics were listed in the schedule submitted with Reference 1.
If there are any questions regarding this report, please contact us.
Very truly yours, DAIRYLAND POWER, COOPERATIVE 0D
--- 2 q u i.-
Frank Linder, General Manager FL: CWA:af cc:
J. G. Keppler, Dir., NRC-DRO III NRC Resident Inspectors l
0107080512 810630 t PDR ADOCK 05000409 P
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._...___.~._,--_,_ _._ _ _ _.._ __..___..._._ _..__.,
LA CROSSE BOILING WATER REACTOR SYSTEMATIC EVALUATION PROGRAM SAFETY EVALUATION REPORT TOPIC XV-15 INADVERTENT OPENING OF A i.ACBWR SAFETY VALVE INTRODUCTION:
The inadvertent opening of a LACBWR safety valve or failure to close results in a reactor coolant inventory decrease and corres-ponding decrease in reactor coolant system pressure and water level.
In LACBWR (a BWR of Allis-Chalmers Company manufacture),
three spring-loaded safety valves mounted in a branch 10" steam line approximately 13 feet above the reactor vessel (RV) operati.g water level protect the reactor vessel and primary system piping from overpressurization.
The valves vent steam directly to the containment building which automatically isolates on high containment building radiation, high containment building pressure (5 psig), high reactor pressure and low RV water level.
The three valve setpoints are a combination of 1390 and 1426 psig.
Since the relief valves are located in a branch of the 20" steam line supply to the shutdown condenser, a safety valve that inadvertently opens or remains stuck-open, would be equivalent to a steam line break or an above the core LOCA.
The probability of the requirement for a steam safety valve to operate is extremely low.
The transients with the highest pressure, the Reactor Building Main Steam Isolation Valve (MSIV) closure, and Turbine Trip Without Bypass, were analyzed in Reference 6 (DPC Letter LAC-4654, Answer to Question 14 dated April 27, 1977) and DPC Letter LAC-6846, Turbine Trip Without Bypass Pressurization Transient.
In order to obtain a safety valve opening, conservative assumptions of 102% operating power were made.
No credit was taken for specified trip signals or operation of the shutdown condenser (Figure 1).
Figure 2 includes operation of the shutdown condenser.
System Response to Stuck-Open Valve The effects due to inadvertent pressure relief and a stuck-open safety valve (two simultaneous active failures) at an operating pressure of 1300 psig would result in a loss of approximately 280,000 lbm/hr of saturated steam from the reactor vessel coolant inventory.
The turbine inlet governor valves would begin to close to maintain reactor pressure, feedwater would increase to maintain RV water level and a high radiation signals in the containment building would alarm causing automatic isolation of containment building.
A mismatch in steam flow and feedwater flow would immediately occur and the specially installed safety valve position indicators and redundant safety valve exhaust temperature indicators would indicate that a safety valve was open.
Operator action could be taken to trip the reactor... --
m i
Assuming no operator action is taken, in approximately 6-8 minutes, the hot well condensate inventory would be depleted and the feedwater pump would trip on low suction pressure.
Reactor pressure would decrease and water level would decrease to -12 inches within one minute, at which point a reactor trip would occur.
The single failure proof HPCS pumps would automatically start and core cooling would be maintained throughout the transient.
Summary:
The effects of steam line breaks, stuck-open safety valves, emerge'Ay core cooling methods and the results of the analyses were used to demonstrate the adequacy of the LACBWR emergency core cooling system (ECCS) in Reference (1).
The effects on fuel integrity and system response was further addressed in References (2) through (14).
The consequences of a stuck-open relief valvc are less severe than cther types of feedwater transients, extensively addressed in Reference (6), " Transient Analyses for LACBWR Reload Fuel", and Reference (2), " Technical Evaluation of the Adequacy of LACBWR Emergency Core Cooling System".
In these analyses, the trarsient conditions of complete loss of feedwater, uncontrolled feedwater increase to raximum two-pump capacity and loss-of-coolant from steam-line breaks were conservatively addressed.
In all cases, 0
the peak cladding temperatures were maintained below 2300 F, the allowable limit for stainless steel clad fuel.
==
Conclusions:==
The review of this SEP topic concludes that the La Crosse Boiling Water Reactor complies with the guidelines and acceptance criteria given in Standard Review Plan, " Inadvertent Opening of a BWR Safety /
Relief Valve", Section 15.6.1.
The consequences of the reviewed transient are of extremely low frequency and are less severe than other transients that result in decrease of reactor coolant inventory, such as steam line break.
The plant and operating systems will respond satisfactorily to safety valve opening transients to prevent excessive fuel damage.
System pressurization will be maintained below 110% of the design pressure (1540 psia).
j This concludes the review of this SEP Topic XV-15.
' i r
Technical
References:
(1)
DPC Letter LAC-6705, Linder to Director of Nuclear Regulation, D. Ziemann,
Subject:
Information on Small Break Analysis, dated December 20, 1979.
(2)
Gulf United Report, " Technical Evaluation of the Adequacy of LACBWR Emergency Core Cooling System", SS-942, May 31, 1972.
(3)
Gulf United Report, " Response to Questions by AEC with regard to Gulf United Report SS-942, Technical Evaluation of the Adequacy of LACBWR ECCS", SS-1075, Revision 1, November 15, 1973.
(4)
Gulf United Report, " Supplemental Information on the LACBWR Emergency Core Cooling System", SS-ll26, October 10, 1973.
(5)
Letter to Director, Nuclear Reactor Regulaton, from Dairyland Power Cooperative, "Dairyland Power Cooperative, La Crosse Boiling Water Reactor, LACBWR Emergency Core Cooling System",
LAC-4082, Enclosure TR-7, "The Effect of LOCA Environment and Subsecuent Containment Flooding on LACBWR Safety Systems",
July 21, 1976.
(6)
NES Report, " Transient Analyses for IACBWR Reload Fuel",
NES-81A0025, February 18, 1977.
(7)
G h2 Nuclear Fuels Company Report, " Anticipated Transients Without Scram at LACBWR", SS-ll78, February 28, 1974.
(8)
"La Crosse Boiling Water Reactor Safeguards Report",
ACNP-65544, August 1967.
(9)
NES 81A0244, " Comparison of LACBWR ECCS Results to AEC Final Acceptance Criteria", December 9, 1974.
(10)
Gulf United Report, " Evaluation of Reactor Vessel Water Level Indication at LACBWR", SS-1182.
(11)
Gulf United Report, " Technical Evaluatior. of Water Level Instrument Behavior During Main Steam Bypass Valve Malfunction Incident at LACBWR", SS-676, May 1971.
(12)
NES 81A0019, " Single Failure Analysis of the LACBWR Emergency Core Cooling Systems" November 1975.
(13)
Application for Amendment to License, DPC Letter LAC-4440, dated January 17-1977.
(14)
" Description of Post-Accident Safeguards Provisions for the LACBWR", ACNP-66564, Amendment No. 29, September 1966.
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