ML20054A045
| ML20054A045 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 04/13/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| TASK-03-07.D, TASK-3-7.D, TASK-RR LSO5-82-04-036, LSO5-82-4-36, NUDOCS 8204150242 | |
| Download: ML20054A045 (6) | |
Text
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April 13,1982 Docket No. 50-409 LS05 04-036 (o
Mr. Frank Linder d
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General Manager 2
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Dairyland Power Cooperative a
6 2615 East Avenue South
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Lacrosse, Wisconsin 54601 1>
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Dear Mr. Linder:
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SUBJECT:
SEP TOPIC III-7.0, CONTAlflMENT STRUCTURAL IrlTEGRITY TEST LACROSSE BOILIt:G WATER REACTOR Enclosed is a copy of our final evaluation of SEP Topic III-7.D.
This evaluation compares your f acility as described in the Safety Analysis Report you supplied on June 29, 1981 and other information on Docket No. 50-409 with criteria used by the staff for licensing new facilities.
The evaluation concludes that the structural integrity test performed corpares favorably with current criteria.
~ This eval"' tion will be a basic input to the integrated assesse.ent of your facility and may be changed in the future if your facility design is changed or if NRC criteria relating to this topic is modified before the integrated assessnent is corpleted.
Sincerely, original signed by:
Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
As stated nh:
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Lacrosse Docket No. 50-409 Revised 3/30/82 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 233 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 Licensing Appeal Board U. S. Nuclear Regulatory Commission
' Mr. R. E. Shimshak Washington, D. C.
20555 La Crosse Boiling Water Reactor Dairyland Power Cooperative Mr. Ralph S. Decker P. O. Box 135 Route 4, Box 190D Genca, 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 Commission 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 l'niversity of Washington Seattle, Washington 98195 U. S. Nuclear Regulatory Commission Resident Inspectors Office James 3. Keppler, Regional Administrator Rural Route.#1, Box 276 Nuclear Regulatory Commission, Region III Genoa, Wisconsin 54632 799 Roosevelt Road G1en Ellyn, Illinois 60137 Town Chairman Thomas S. Moore Town of Genoa Atomic Safety and Licensing Appeal Board Route 1 U. S. Nuclear Regulatory Commission 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
Ef1 CLOSURE 1 LACROSSE BOILIf1G WATER REACTOR SEP TOPIC III-7.D ASSESSMErlT OF C0flTAlf1MEllT STRUCTURAL IllTEGRITY TEST I.
IfiTR00VCTI0fl The structural integrity test procedure and test results for the Lacrosse containment were evaluated against current criteria for such i
tests. The purpose of the evaluation was to verify that the contain-ment structural integrity test was in compliance with the current requirements of 10 CFR 50 and thus provided assurance that the structure I
would perform its intended safety function.
II.
REhlEWCRITERIA
References:
A.
IIRC Standard Review Plan Section 3.8.2 C.
ASME Boiler and Pressure Vessel Code,Section III, Division 1, Article flE-6000, 1980 edition D.
Lacrosse SAR for SEP Topic III-7.D (Transmitted to f1RC staff by a letter dated June 29,1981, LAC-7643)
References A, B, and C outline current criteria for conducting and evaluating containment structural integrity tests. Reference D describes the containment tests actually conducted at the Lacrosse plant.
III.
RELATED TOPICS AtlD IfiTERFACES SEP Topic VI-3 " Containment Pressure and Heat Removal Capability" will provide an assessment of the adequacy of the original design pressure for the containment. The evaluation described herein is based on the originial design and test pressure loading of the containment as presented in reference D.
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e IV.
REVIEW GUIDELIfiES The test procedure and results were compared with current f1RC criteria for such tests in order to determine if any significant deviations existed.
V.
EVALVATI0f1 The Lacrosse containment building is a right circular cylinder with a hemispherical dome and semi-ellipsoidal bottom.
It has an overall internal height of 144 feet and an inside diameter of 60 feet, and it extends 26 feet 6 inches below grade level. The steel shell thick-ness is 1.16 inches, except for the upper hemispherical dome, which is 0.60 inch thick.
Thg inter,ior of the shell is lined with a 9 inch thick layer of con-crete up to an elevation of 727 feet 10 inches, to limit direct radiationdoses in the event of fission-product release within the containment building.
The shell includes two airlocks. The principal access to the shell is through the personnel airlock that connects the containment building to the turbine building. The airlock is 21 feet 6 inches long be-tweenitstwodoors,whicharefihefeet6inchesby7feetandare large enough to permit passage of a spent fuel element shipping cask.
The emergency airlock is 7 feet long and 5 feet in diameter, with two circular doors of 324 inches diameter (with a 30-inch opening). Both airlocks are at elevation 642 feet 9 inches. The airlocks and doors are designed to reamain gastight under a pressure of 52 psig from inside the containment shell and 1 psig from outside of the shell.
The airlock doors are manually operated.
An 8 feet by 10 feet freight door opening in the containment building 1
accommodates large pieces of equipment.
It is used only when the reactor is shut dotm and only if large pieces of equipment must be removed. During operation, 9-inch thick concrete blocks are placed on the outside of the door for shielding. The door is bolted internal-ly to the door frame in the shell. Two rubber gaskets in parallel between the door and door frame ensure a pressure-tight seal.
Approximately 300 til cables and 75 bulkhead conductors penetrate the containment shell.
i Thedesignandconstructionofthecontainmenthesselconformstothe applicable requirements of the 1962 edition of the ASf4E Boiler and Pressure Vessel Code,Section VIII, Unfired Vessels, and applicable code cases 1270ti, 1271tf and 1272tl. The containment vessel has the ASf4E code stamp.
The design conditions for the code calculations were as follows:
(1) maximum internal pressure.............. 52 psig (2) maximum negative pressure............. 0.5 psig (3) maximumtemperature......................28bF (4) minimumtemperature.......................-2bUF (5) we lded j oint ef ficiency................... 100%
2 (6) Basic wind pressure.................. 20 lb/ft Requirements for the initial testing of the containment vessel included the following:
(1) visual inspection of welding.
(2) pressurizing the containment Yessel to 5.0 psig and performing soap bubble tests on all welds.
(3) pressurizing the shell to 59.80 (1.15 x Design) psig, then reducing the prassure for final soap bubble tests on all welds at a vessel pressure of 52 psig. The 52 psig pressure was maintained for a 3-day period, with leakage measurements taken hourly. The inner chamber method was used to measure leakage.
The overload test procedure fulfilled the requirements of the ASME Code,Section VIII, as modified by Code Case 1272 t1.
The criteria that was applied during the overload test on the Lacrosse Boiling Water Reactor Containment Vessel conformed to all the require-ments of reference C.
Current requirements are for a test to 1.10 x
=
Design (57.2 psig) and therefore the test pressure used is conservative by current standards.
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No evidence of unusual response of the containment building was observed during the test.
VI CONCLUSION Based on the review of the original structural integrity test in comparison to current test requirements, it is considered that the test was satisfactory and thus demonstrated that the structure is I
capable of performing its intended safety function.
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