ML20062E454
| ML20062E454 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 08/06/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| LSO5-82-08-006, LSO5-82-8-6, NUDOCS 8208100109 | |
| Download: ML20062E454 (6) | |
Text
f I
August 6,1982 Docket No. 50-409 LS05-82-08-006 Nr. Frank Linder General Manager Dairyland Power Cooperative 2615 East Avenue South Lacrosse, Wisconsin 54601
Dear Mr. Linder:
SUBJECT:
SEP TOPIC XV-19, LOSS OF COOLANT ACCIDENTS RESULTING FROM SPECTRUM 0F POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY - LACROSSE BOIL 1NG WATER REACTOR Byletter(LAC-8119)datedFebruary 26,1982 (received May 28, 1982),
you submitted a safety assessment on this topic. The staff has reviewed your assessment and our conclusions are presented in the entlosed final topic evaluation. With respect to the systens aspects of this topic, the staff concludes that the Lacrosse Boiling Water Reactor meets current licensing criteria.
Potential radiological consequences will be addres-l sed in a separate evaluation.
This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility.
This assessment may be revised in the i
future if your facility design is changed or if NRC criteria relating to this subject are/ modified before the integrated assessment is completed.
Sincerely, g60Y uif Original signed byi DSM Dennis M. Crutchfield, Chief
- 6. SIm/6/
Operating Reactors Branch No. 5 j
Division of Licensing
/
f T. Me t.ec, Q
Enclosure:
l As stated l
cc w/ enclosure:
See next page 8208100109 820806 PDR ADOCK 05000409 P
PDR w.
.SEEBS12L/.M.SERS..[/.h...S EP.B...(k,.
,..S.EPB... @I 0RB,#.5.,,,,
0RB#5 <
.AD.Yg)fod,
- S/;DL omce>
.C..G.j.i.me.g.g......
W. R..u.sse..l l.. RDudley D27uTENfield T
'..T.M.i c h.g.el.s...
..gg,g.a..:..b..l..
EMcKenn gr..j s
a suoruwe)
...g
,.782" -
"8/4r/82"-
gy q g
.....a......
mn>
' sac rORM 318 00-80) NRCM ONO OFFICIAL RECORD COPY usom,u-mm
LaCrassb c--
Docket No. 50-409 Revised May 12, 1982 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 Dearborn 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 Mr. Ralph S. Decker P. O. Box 275 Route 4, Box 190D Genoa, Wisconsin 54632 Cambridge, Maryland 21613 Mr. George R. Nygaard Charles Bechhoefer, Esq., Chairman Atomic Safety and Licensing Board Coulee Region Energy Coalition U. S. ' Nuclear Regulatory Commission 2307 East Avenue La Crosse, Wisconsin 54601 Washington, D. C.
20555 Dr. Lawrence R. Quarles Dr. George C. Anderson Kendal at Longwood, Apt. 51 Department of Oceanography Kennpth Square, Pennsylvania 19348 University of Washington Seattle, Washington 98195 l
U. S. Nuclear Regulatory Commission Resident Inspectors Office James G. Keppler, Regional Administrator I
Rural Route #1, Box 276 Nuclear Regulatory Commission, Region III l
Genoa, Wisconsin 54632 799 Roosevelt Road Glen Ellyn, Illinois 60137 Town Chairman Thomas S. Moore Atomic Safety and Licensing Appeal Board Town of Genoa Route 1 U. S. Nuclear Regulatory Commission Genoa, Wisconsin 54632 Washington, D. C.
20555 Chairman, Public Service Commission l
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 l
l l
l f
A-IV.
REVIEW GUIDELINES The review of ECCS performance during a LOCA is conducted in accordance with Standard Review Plan Sections 15.6.5 and 6.3.
A plant using stainless steel
~
clad fuel is considered to be adequately designed against a LOCA if the Interim Acceptance Criteria are met.
The radiological consequences are addressed in a separate evaluation.
V.
EVALUATION Assuming a conservative combination of circumstances which could lead to core uncovery and excessive heatup following a loss-of-coolant accident, fuel clad-ding integrity is maintained by successful operation of the Emergency Core Coolant System (ECCS).
The following systems in the Lacrosse Boiling Water Reactor provides the necessary protection to mitigate the consequences of a 3-loss-of-coolant accident:
~
(1) The High Pressure Core Spray (HPCS) system which is put into operation manually or automatically on reactor low water level or high containment building p.ressure.
(2) The Alternate Core Spray (ACS) system which is also put into operation manually or automatically on coincident low reactor water level and high containment building pressure.
The combined operation of the HPCS and ACS provides long-term cooling of the core.
A manual depressurization system is provided to equalize reactor vessel and containment pressure following a LOCA.
The LOCA analysis for LACBWR has t
b'een performed with no credit taken for blowdown to the shutdown condenser.
Manual operation is permissible since at least 20 minutes is available to the operator to make a decision.
The adequacy of the LACBWR ECCS evaluation model was discussed by the licensee in Reference 2.
This evaluation was made with respect to the requirements for analysis of blowdown phenomena, as prescribed in the Interim Acceptance Criteria.
Small break results in the LOCA analysis have been extrapolated from large and 3
&e4
,7
-.s.
- ~".
TOPIC XV-19 (SYSTEMS)
LOSS OF COOLANT ACCIDENTS RESULTING FROM SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY LACROSSE BOILING QATER REACTOR
~~
I.
IkTRODUCTION
]-
The objective of this review is to assure that the consequences of a Loss of Coolant Accident (LOCA) are acceptable, i.e., that the requirements of the AEC Interim Policy Statement and Appendix K to 10 CFR 50 are met.
Loss-of-coolant accidents are postulated accidents that would result from the loss of reactor coolant, at a r' ate in excess of the capability of the reactor coolant make up
~
~
system, from piping breaks in the reactor coolant pressure boundary'.
The review consists of evaluating the licensee's analysis of the spectrum of loss-of-coolant accidents including break location, break size, initial conditions assume'd, the evaluation model used, failure modes and the acceptability.of auxiliary systems used.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of systems provided for the prevention of accidents and the
- mitigation of the consequences of accidents.
.1 1
Th2 AEC Int:ric Policy St;te:.:nt r:quir:s thit all light w:;ter rnet;rs shall be provided with an emergency core cooling system designed so that its perfor-manc'e following a LOCA satisfies the criteria set forth in the Interim Accept-ance Criteria.
Performance is calculated with an evaluation model satisfying the applicable requirements of Appendix K to 10 CFR 50.
The General Design Criteria (Appendix A to 10 CFR Part 50) set forth the cri-teria for the design of water-ccoled reactors.
GDC 35 " Emergency Core Cooling" requires that a system be provided to provide abundant emergency core cooling whose function is to transfer heat from the core following a loss of coolant such that (1) fuel and clad damage that could interfere with continued effec-tivecorecoolingisprevented,and(2)ciadmetalwaterreactionislimitedto negligible amounts.
The system should have suitable redundancy and inter-connections such that function can be maintained assuming a single failure and t:'
assuming availabil_ity of only onsite or only offsite power supplies.-
-M III.
RELATED' SAFETY TOPICS Topic III-5.A, " Effects of Pipe Breaks on Structures, Systems and. Components.
Inside Containment" ensures that the ability to achieve safe shutdown or miti-
~
gate the consequences of an accident are maintained.
The adequacy of the features provided for Switchover from Injection to Recir-culation modes is addressed in Topic VI-7.B.
Other SEP topics consider the emergency power supplies (VIII-2), effects of
~
flooding of safety-related equipment (VI-7.D), prevention of boron.pfecipita4f I
tion (IX-4) as well as failure modes of the ECCS (VI-7.C).
In addition, such areas as containment integrity and isolation, postaccident chemistry and Engineere,d Safety Feature systems are considered as part of SEP topics.
Topics VI-2.D and VI-3 address the capability of the containment heat removal systems to alleviate the pressure / temperature transient so that the contain-l ment is not overpressurized.
L '.
l 2
.-,.. =.
2:~,
intermediate size breaks as discussed in Reference 1.
The LOCA results have shown that the HPCS is adequate, even with a single failure, to maintain core parameters within Interim Acceptance Criteria limits.
The limiting single failure is the failure of one HPCS pump.
The break spectrum analysis performed with the LAC 8WR ECCS evaluation model identified the most limiting break as an intermediate size break (0.072 sq. ft.) in the 20-inch recirculation line.
The highest peak clad temperature (2296 F) is calculated for this break, with no more than 0.15% of the cladding reacting chemically.
These values are within the IAC limits of 2300 F peak clad temperature and 1 percent cladding steam reaction.
The ECCS performance has been found acceptable by the staff (Refer-ence 4) based on information provided by the licensee in Reference 3.
VI.
CONCLUSIONS As part of the SEP review of Lacrosse Boiling Water Reactor, the loss-of-coolant analysis was reviewed against the Interim Acceptance Criteria, and the acceptance criteria of SRP Sections 15.6.5 and 6.3.
The initial conditions relative to single failure, break size and location, power level and operating conditions have been reviewed and found to conform to the requirements of the SRP.
The analysis was performed with an approved evaluation model and the results were found to be acceptable.
VII.
REFERENCES 1.
DPC Letter LAC-5705, Linder to D. Ziemann,
SUBJECT:
Information on Small Break Analysis, dated December 20, 1979.
2.
NES-81A0244, "Compa'rison of LACBWR ECCS Results to AEC Final Acceptance Cri,teria," December 9,1974.
3.
Gulf United Report, " Technical Evaluation of the Adequacy of the LACBWR Emergency Core Cooling System," 55-942, May 31, 1972.
4.
Safety Evaluation Supporting Amendment No. 11 to Provisional Operating License DPR-45, Lacrosse Boiling Water Reactor, March 3, 1978.
~~ ~
4
_- -.