ML20039B663
| ML20039B663 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 11/17/1981 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 LAC-7983, NUDOCS 8112230360 | |
| Download: ML20039B663 (5) | |
Text
.
D DA/RYLAND h
[k COOPERAT/VE PO. BOX 817 2615 EAST AV SOUTH LA CROSSE. WISCONSIN 54601 (608) 788 4 000 December 17, 198,1 In reply, please refer to LAC-7983 DOCKET NO. 50-409 l
N Director of Nuclear Reactor Regulation g
ATTN:
Mr. Dennis M. Crutchfield, Chief p
/
Operating Reactors Branch #5 SpCD)g D g
Division of Operating Reactors U.
S. Nuclear Regulatory Commission D, c gIC
~~
Washington, D. C.
20555
/
SUBJECT:
DAIRYLAND POWER COOPERATIVE
A
% {q~
j d
LA CROSSE BOIL NG WATER REACTOR (LAC ' 'R)
/
PROVISIONAL OPERATING LICENSE NO. DPR ESTS g)-
,c ; : S 1 TMI LESSONE LEARNED - SHORT TERM REQUIR NUREG-0578, SECTION 2.1.2
References:
(1).
NUREG-0578, Section 2.1.2, dated July 1979.
(2)
DPC Letter, Linder to Denton, LAC-6680, dated December 6, 1979.
i (3)
NUREG-0737, dated October 31, 1980.
(4)
Evaluation of Fluid Conditions at the Main Steam Safety Valve Inlets for Expected Transients, DPC/NES Report, dated June, 1981.
(5)
DPC Letter, Linder to Denton, LAC-7633, dated June 29, 1981.
(6)
Wyle Lab Test Report 45711-01, dated July 15, 1981.
Gentlemen:
This letter is to present final information regarding the status of the LACBWR program to satisfy the intent of the recommendation provisions of NUREG-0578, Section 2.1.2, entitled " Performance Testing for BWR and PWR Relief and Safety Valves", (Reference 1).
The purpose of the NUREG recommendations was to require qualifi-cation of relief and safety valves under off-normal reactor over-pressure conditions.
At LACBWR there are three Crosby HCU-Spec. 3"xMx6" spring-loaded safety valves installed on a horizontal run of shutdown condenser steam line.
They are designed to meet the reactor primary system
[Oib 3
pah22ggggogahajjj
- /g O P
Mr. Dennis M. Crutchfield, Chief December 17, 1981 Operating Reactors Branch #5 LAC-7983 overpressurization requirements of ASME Nuclear and Boiler Pressure Vessel Codes.
The safety valves are located approximately four feet above the reactor vessel steam line outlet and approximately 13 feet above the normal operating unvoided water level.
The three valves, set at 1390 psig and 1426 psig relief pressure (Note:
At least one of three must be set at 1390 psig and at least one of three must be set at 1426 psig) are designed to free dis-charge directly to the containment building and are sized to accommodate full design steam flow of 611,500 lbm/hr at 100%
(165 MWth) reactor power.
The valves are frequently verified to be operable by bench testing and they have never been required to actuate in 14 years of reactor operation.
Power operated relief valves are not used to control over-pressurization transients at LACBWR, In the event of a postulated ove.rpressurization transient coincident with the loss of several safety related equipment functions, the spring-loaded valves would safely relieve steam to the containment building.
A general descrip-tion of the safety valves is contained in Reference (2).
An evaluation which addresses the fluid conditions that could exist at the safety valves (Reference 4) was sent to you in Reference 5.
It concluded that liquid or two-phase flow conditions would not be present in the safety-valve steam line during any postulated trans-ient condition in which high-pressure relief would be required.
In order to assure that the HCU safety valves would operate satis-factorily even under the most improbable postulated transient con-ditions (as discussed in Reference 4), the valves were tested with saturated eteam at approximately 200,000 lbm/hr'., a steam pressure of 1465 psia) at the Wyle Laboratory testing facility in Huntsville, Alabama, in conjunction with Consumers Power Company, which has generically equivalent safety valves at the Big Rock plant.
The testing program of other BWR Owners (whose operating reactor are not generically equivalent to LACBWR and BRP) was factored into our review and test plan.
The full-flow test program was designed to monitor and record impor-tant variables such as valve body temperature, relief pressure, back-pressure, accumulation pressure, reset pressure, and valve disc lift distance.
(Live steam tests of the LACBWR valves were also satis-factorily conducted on limited flow test facilities in 1977).
The full-flow tests were completed on June 26, 1981.
The NRC Project Manager for LACBWR was notified of the test.
Mr. Dennis M. Crutchfield, Chief December 17, 1981 Operating Reactors Branch #5 LAC-7983 The tests conducted at Wyle Laboratory were witnessed by technical representatives from Wyle Lab, Dairyland Power Cooperative, Consumers Power Company and Crosby Valve Company.
The results of the full-flow tests indicate the valves operated satisfactorily.
Minor adjustments were made to the valve nozzle ring and guide ring settings to obtain optimum performance.
A final test repart was issued to Dairyland Power Cooperative summarizing the results of the test (Reference 6).
An extension for submittal of this report was granted to Mr. C. Angle by Mr. Ralph Caruso.
If you have any questions regarding this letter, please let us know.
Very truly yours, DAIRYLAND POWER COOPERATIVE
/1 G '
< 2=~
(
w-Frank Linder, General Manager FL: CWA:af Attachment cc:
NRC Resident Inspectors J.
G.
Keppler, Dir., NRC-DRO III D.
Vandewalle-Consumers Power. -
DATA SIIEET FOR NUREG-0737 S/RV TEST A11-5 WYLE LABS, IIUNTSVILLE, ALA.
l LA CROSSE BOILING WATER REACTOR DATE TESTED:
June 25, 1981 RUN/ TEST NO.:
14 S/RV MODEL NO.: 3XM2X6, IICU-65 SPEC.
S/RV MANUFACTURER:
CROSBY S/RV SERIAL NO.:
62-27-009 TYPE OF TEST:
STEAM-SATURATED, 1446 PSIA MAX.
TEST RE'SULTS:
S/RV OPENED ON COMMAND YES S/RV CLOSED ON COMMAND YES STABILIZED FLUID CONDITIONS:
200,000 #/hr.
I FLOW RATE N
INLET RELIEF PRESSURE 1390 PSIG TEMPERATURE SAT.
S/RV OPENING TIME (MAIN DISK 0.05 SEC.
LIFT TIME)
LIFT 0.42*
VALVE INTEGRITY DEMONSTRATED BY:
VISUAL INSPECTION YES POST-TESTXilYDRO*
N/A POST-TEST DISASSEMBLY YES INSPECTION
- AND CLEANED YES VALVE INTEGRITY DEMONSTRATED YES
- ONLY ONCE PER VALVE AT END OF TEST SEQUENCE
s DISTRIBUTION OF LAC-7983 SRC Trommel Shimshak Parkyn Towsley Rybarik Brimer Files A-11, T5C, V-9 TMI-NB-T.L.
Reading File
.