ML20039E135
| ML20039E135 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/31/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TASK-15-08, TASK-15-8, TASK-RR LSO5-81-12-108, NUDOCS 8201060569 | |
| Download: ML20039E135 (6) | |
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o December 31, 1981 Docket No. 50-155 LS05 12-108 g
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[M Mr. David P. Hoffman Nuclear Licensing Administrator J!w 51082>ill Consumers Power Company 1945 W Parnall Road u ^
Jackson, Michigan 49201 8
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Dear Mr. Hoffman:
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SUBJECT:
BIG ROCK POINT - SEP TOPIC XV-8, CONTROL R00 MISOPERATION 5
By letter dated September 23, 1981, you submitted a safety assessment report for the above topic. The staff has reviewed this assessment and our conclu-sions are presented in the enclosed safety evaluation report, which completes the review of this topic for Big Rock Point.
This evaluation will be a basic input to the integrated assessment for your facility. The evaluation may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5
-Division of Licensing
Enclosure:
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OFFICIAL RECORD COPY usom i,ei_ x.o
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i Mr. David P. Hoffman i
cc U. S. Environmental Protection Mr. Paul A. Perry, Secretary Agency Consumers Power Company Federal Activities Branch 212 West Michigan Avenue Region V Office-.
Jackson, Michigan 49201 ATTN: Regional Radiation Representative 230 South Dearborn Street Judd L. Bacon, Esquire Chicago, Illinois 60604 Consumers Power Company 212 West Michigan Avenue Herbert Grossman, Esq., Chairman Jackson, Michigan 49201 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Joseph Gallo, Esquire Washington, D. C.
20555 Isham, Lincoln & Beale 1120 Connecticut Avenue Dr. Oscar H. Paris Room 325 Atomic Safety and Licensing Board Washington,- D. C.
20036 U. S. Nuclear Regulatory Commission Washington, D..C.
20555 Peter W. Steketee, Esquire,
505 Peoples Building Mr. Frederick J..Shon' Grand Rapids, Michigan 49503 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Alan S. Rosenthal, Esq., Chairman Atomic Safety & Licensing Appeal Board Washington, D. C.
20555 U. 's. Nuclear Regulatory Commission Big Rock Point Nuclear Power Plant Washington, D. C.
20555 ATTN: Mr. C. J. Hartman Plant Superintendent Mr. John O'Neill, II Charlevoix, Michigan 49720 Route 2, Box 44 Maple City, Michigan 49664 Christa-Maria Route 2, Box 108C l
Charlevoix Public Library Charlevoix, Michigan 49720 107 Clinton Street Charlevoix, Michigan William J. Scanlon, Esquire 2034 Pauline Boulevard Chairman Ann Arbor, Michigan 481.3 0
County Board of Supervisors l
Charlevoix County Resident Inspector Charlevoix, Michigan 49720 Big Rock Point Plant c/o U.S. NRC Office of the Governor (2)
RR #3, Box 600 f,
Room 1 - Capitol Building Charlevoix, Mic,higan 49720 Lansing, Michigan 48913 Mr. Jim E. Mills Herbert Sennel Route 2, Box 108C Counsel for Christa Maria, et al.
Charlevoix, Michigan 49720 Urban Law Institute 4
Antioch School of Law 263316th Street, NW Washington, D. C.
20460 I
o-Mr. David P. Hoffman f
i cc Dr. John H. Buck Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Comission Washington, D. C.
20555 Ms. JoAnn Bier 204 Clinton Street Charlevoix, Michigan 49720 Thomas S. Moore Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Comission Washington, D. C.
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a SYSTEMATIC EVALUATION PROGRAM TOPIC XV-8 BIG ROCK POINT TOPIC: XV-8, CONTROL R0D MISOPERATION (SYSTEM MALFUNCTION OR OPERATOR ERROR)
I.
INTRODUCTION The control rod misoperation event in boiling water reactors consists
'of moving an out of sequence rod or of mov ng an insequence rod beyond i
its allowable limits.
Such an event leads to an increase in core power.
i II.
REVIEW CRITERIA Section 50.34 of 10 CFR 50 reouires that each applicant for a construction -
permit or operating license provide an-analysis and evaluation of the design 4
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 determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility.
Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to 10 CFR Part 50) estabi~ish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 10," Reactor Design" requires that the core and associated coolant, control and protection systems be designed with appropriate umrgin to assure that specified acceptable fuel design limits are not exceeded during normal operation, includir.7 the effects of anticipated operational occur-rence.
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 occurrences.
GDC 20, " Protection System Functions" requires that the portection system be designed to initiate automatical-ly the operation of reactivity control systems to assure that specified acceptable fuel design limits are'not exceeded as a result of anticipated operational occurrences.
GDC 25, " Protection System Requirements for Reactivity Control Malfunctions" requires that specified acceptable fuel design limits not be exceeded for.
any single malfunction of the reactivity control systems such as accidental withdrawal of control rods.
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III.
RELATED SAFETY TOPICS
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Topic I'V-2 describes the reactivity control system and any failure modes ~
.that could lead to control rod misoperation.
Other SEP topics address such items as the reactor protection system.
IV.
REVIEW GUIDELINES
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The review is conducted in accordance with SRP 15.4.1,15.4.2 and 15.4.3.
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The. evaluation includes review of the analysis for the event and identi-fication of the features in the plant that mitigate the consequences of the event as well as the ability of these systems to function as required.
The extent to which operator action is required is also evaluated.
Deviations from the criteria specified in the Standard Review Plan are identified.
V.
EVAldATION
. This event has been analyzed for Big Rock Point at both startup. and operating power levels. At startup. levels no fuel dashge occurs'for continuous withdrawal of..the rod.of maximum anticipated worth.- This is 4
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consistent with results for other boiling water reactors and is. acceptable.
The Big Rocik Point Reactor does not have a Rod'Blocli Monitor which is present.
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in later boiling water reactors.
As a result the amount of withdrawal of a control rod at power is not limited by equipment. A single operater error or equipment failure can result in the inadvertent withdrawal of a high worth rod. A conservative analysis of this event has been performed..Beginning of cycle conditions which maximizes rod worths and event consequences are assumed.
A maximum worth rod is chosen and scram at 120 percent of full power does not occur.
Instead the power is assumed to increase to 140 percent of full power.
The thermal analysis in performed at this power. The critical pouar ratio
. correlation which is used.in the themal evaluation is the XN-2 correlation of Exxon Nuclear Corporation. This correlation predicts that, with 95 percent confidence,' 95 percent of the rods will not experience departure from nucleate boiling if the critical power ratio of an assembly is greater than 1.225.
For the conservative analysis perfomed here two assemblies have. lower critical power ratios.
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The following discussion. relates to the need to backfit the Big Rock plant to meet current requirements. The plant was designed and con-structed prior to the existence of tne General Design Criteria and was
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not required to meet these criteria.
Such an event has a very low pro-bability of occurrence.
Further, the analysis of the consequences ofs -
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the event ignored three possible reactor scrams which would have reduced the consequences.
These are the high flux trip, a trip on loss of feed-water suction and a trip on loss of condense'r vacuum.' It is likely that the plant cannot sustain ~ a power level of 140 percent of full power though no systems analysis has beer) performed to support this conclusion.
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An additional conservatism exists in the critical power ratio correlation used for-the analysis.
The quoted limit is based on a chopped cosine a.3 al power distribution.
If a flat axial power distribution is assumed i
no rods would fail the critical power ratio criterion. The actual power distribution in Big Rock will be relatively flat over a large portion of the elevation and the thermal conditions will be improved over those existing for a chopped cosine distribution.
VI.
CONCLUSIONS Based on the analysis described above we conclude that the Big Rock Point reactor does not satisfy General Design Criterion 25 which requires that specified acceptable fuel design limits not be exceeded for events that proceed from single failures in the reactivity control systems.
Big Rock
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does not, therefore, meet current requirements for this event.
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