ML20050C783
| ML20050C783 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 04/05/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| TASK-15-08, TASK-15-11, TASK-15-13, TASK-RR LSO5-82-04-012, NUDOCS 8204090352 | |
| Download: ML20050C783 (12) | |
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April 5,1982 Gi Ev, Docket No. 50-409 gf!C.L' LS05 04-012 A9p p B WB2" Z_
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Mr. Frank Linder General Manager Dairyland Power Cooperative b
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Lacrosse, Wisconsin 54601
Dear Mr. Linder:
SUBJECT:
SAFETY EVALUATION REPORTS FOR SEP TOPICS XV-8, XV-11, XV-13 (SYSTEMS) - LACROSSE BOILING WATER REACTOR In your letter dated February 26, 1982, you submitted safety assessment reports on the above topics. The staff has reviewed your assessments and our conclusions are presented in the enclosed safety evaluation reports which completes these topic evaluations for the Lacrosse Boiling Water Reactor (LACBWR).
The potential radiological consequence of a rod drop accident (Topic XV-13) will be addressed in a separate evaluation.
The enclosed safety evaluations will be basic input to the integrated safety assesstaent for your facility. The assessments may be revised in the future if your facility design is changed or if NRC criteria relating to these topics are modified before the integrated assessment is completed.
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j/s Sincerely, j
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Original SICned byt AOD'.
Dennis M. Crutchfield, Chief G,Sh6[
Operating Reactors Branch No. 5 g, ytaP'$
Divisden of Licensing Enclosures :
As stated cc w/ enclosures:
See next page 8204090352 820405 PDR ADOCK 05000409
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Lacrosse Docket No. 50-409 Revised 3/30/82 Mr. Frank Linder cc Fritz Schubert, Esquire U. S. Entironmental Protection Staff Attorney Agency Dairyland Power Cooperative Federal Activities Branch 2615 East Avenue South Region V Office La Crosse, Wisconsin 54601 ATTN:
Regional Radiation Representativt 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 135 Route 4, Box 190D Genoa, Wisconsin 54632 C:mb, idge, Maryland 21613 Mr. George R. Nygaard Charles Bechheefer, 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
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Kendal at Longwood, Apt. 51 Department of Oceanography Kenneth Square, Pennsylvania. 19348 University of Washington Seattle, Washington 98195 U. S. Nuclear Regulatory Commission Resident Inspectors Office James G. Keppler, Regional Administrator Rural Route.#1, Box 276 Nuclear Regulatory Commission, Region III Genoa, Wisconsin 54632 799 Roosevelt Road Glen Ellyn, Illinois 60137 Town Chairman Thomas S. Moore Town of Genoa Atomic Safety and Licensing Appeal Boced 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 Conmission Washington, D. C.
20555
SYSTEMATIC EVALUATION PROGRAM TOPIC XV-8 LACROSSE TOPIC:
XV-8, Control Rod Misoperation (System Malfunction or Operator Error)
I.
INTRODUCTION By letter dated February 26, 1982 the Dairyland Power Cooperative sub-mitted information concerning the control rod misoperation event in the Lacrosse Boiling Water Reactor (LACBWR).
Such an event can cause an increase in core power and reduction in margin to fuel thermal limits.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a con-struction pennit or operating license provide an analysis and evaluation of the design 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 deter-mination 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 inclufe safety limits which protect the integrity of the physical bar-riers which guard a gainst the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 10 "Reacter Design" requires that the core and associated coolant, control and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during normal operation, including the effects of anticipated oper-ational occurrences.
GDC 20 " Protection System Functions" requires that the protection system be designed to initiate aut.omatically the operation of reactivity con-trol systems to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences.
GDC 25 " Protection System Ree:uirements for Reactivity Control Malfunc-tions" requires tFat specificd acceptable fuel design limits not be exceeded for any single malfunction of the reactivity control systems such as accidental withdrawal of control rods.
III.
RELATED SAFETY TOPICS Topic IV-2 describes the reactivity control system and any failure modes that could led to control rod misoperation.
Other SEP topics address l
such items as the reactor protection system.
. IV.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.4.1 and 15.4.2.
The evaluation includes review of the analysis for the event and iden-tification 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.
EVALUATION In present day boiling water reactors control rods are n'oved one at a time in predetermined sequences. During startup and low power operation the sequences are designed to limit the worth of potential dropped rods.
At higher powers the sequences are designed to keep the power distri-butions within limiting conditions for operation or to achieve a par-ticular power shape. A rod misoperation event occurs when out of se-quence withdrawals occur.
Low power (startup) and at-power events are considered.
The LACBWR reactor does not have equipment which enforces a particular withdrawal sequer.ce during starthp.
Procedures are employed to ensure compliance. Rods are kept within one-quarter inch of each other until criticality is achieved. After this rods are withdrawn in 3 banks (i.e.
divided into three groups with all rods within each group kept within a small distance of each other). This tends to limit potential rod worths during rise to power.
Rod Misoperation During Startup In spite of the low probability of occurrence of an out-of-sequence rod withdrawal the consequences of withdrawing a high worth (s 2.5% react-ivity change) rod at maximum speed (20 inches per minute) have been calculated. The results show no fuel damage. This procedure is similar to that used in current analyses for boiling water reactors and is acceptable.
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Rod Misoperation At Power The consequences of the complete withdrawal of the highest worth rod l
from the controlling rod pattern are calculated for each cycle. A reactor simulator code is used and it is required that fuel ther.nal limits not be violated. This procedure is similar to that used in pre-l sent day analyses for boiling water reactors 'and is acceptable.
Operator Action No operator action is required for this event.
I Deviations from SRP 15.4.1 and 15.4.2 No deviations from the requirements of the Standard Review Plan, Sec-l tions 15.4.1 and 15.4.2 are present.
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3-VI Conclusion Based on the evaluation presented above we conclude that LACBWR meets present day requirements for the rod misoperation events.
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SYSTEMATIC EVALUATION PROGRAM TOPIC XV-ll LACROSSE TOPIC:
7,V-11, Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position I.
INTRODUCTION Two misloading events are analyzed for boiling water reactors - misorientation of a bundle in its proper position, and mislocation of a bundle.
Fuel misload-ing events could result in operation with power distributions that provide re-duced margin to fuel thermal limits.
The licensee provided a topic assessment for this event in their letter of February 26,1982 (LAC-8117).
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 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.
2-The General Design Criteria ( Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 13 " Instrumentation and Control" requires that instrumentation and controls be provided to monitor variables over anticipated ranges for normal operations, anticipated operational iccurrences and for accident conditions as appropriate to assure adequate safety.
10 CFR Part 100.11 provides dose guidelines for reactor siting against which calculated accident dose consequences may be compared.
III.
RELATED SAFETY TOPICS None IV.
REVIEW G'JIDELINES This review is conducted in accordance with SRP 15.4.7.
The evaluation includes review of the analysis for the event and identification 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.
T+r extent to which operator action is required is also evaluated. Deviations, if any, from the criteria specified in the Standard Review Plan are identified.
. V.
EVALUATION The consequences of misloading events are considered a part of the reload design analysis for LACBWR. The improper orientation (rotation) of a bundle in its proper location has a negligible impact. The enrichment distribution is symmetric with respect to rotation and symmetric rod patterns are used during operation.
Careful procedures are followed during refueling in order to reduce the proba-bility of mislocating an assembly and a visual verification is performed on the reloaded core. LACBWR does not have incore po,ver instrumentation so this method of loading verification is not available. No credit is taken for detection of a misloaded assembly.
The analysis of the consequences of misloadings is performed by use of a core simulator code as a function of cycle burnup. Such analyses are performed for each reload. This is consistent with current practice for boiling water reactors and is acceptable. These analyses have shown that fuel thermal limits are not violated when LACBWR is operated at full power with the potentially limiting fuel loading error.
This meets the criterion for this event (Standard Review l
Plan, Section 1.5.4.7) which states that any misloading that cannot be detected by use of incore instrumentation shall not result in offsite consequences greater i
than a small part of 10 CFR Part 100 guidelines.
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. VI. CONCLUSION We conclude that the r nalysis of this event for LACBWR meets present day criteria e
and is acceptable.
SYSTEMATIC EVALUATION PROGRAM TOPIC XV-13 LACROSSE TOPIC:
XV-13, Spectrum of Rod Drop Accidents I.
INTRODUCTION Rod drop accidents in a boiling water reactor cause a rapid increase in core reactivity. The licensee provided an assessment of this event in their letter of February 26,1982 (LAC-8117).
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a con-struction permit or operating license provide an analysis and evaluation of the design and perfornance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operating of the facility, including deter-mination of the margins of safety during normal operations and tran-sients 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 b,arriers which guard against the uncontrolled release of radioactivity.
The General Design criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal criteria for water-cooled re-actors. GDC 28 " Reactivity Limits: requires that the reactivity acci-dents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently dis-turb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core.
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III.
RELATED SAFETY TOPICS SEP Topic XV-19 considers the effects of rupture of the reactor coolant pressure boundary of the ejected rod.
IV.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.4.9.
The evalu~ation 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 re-quired. The extent to which operator action is required is also eval-uated.
Deviations if any, from the criteria specified in the Standard Review Plan are identified.
The potential radiological consequences are assessed in a separate eval-uation.
V.
EVALUATION The rod drop accident in a boiling water reactor occurs when a rod blade becomes disconnected from its drive and sticks in the cere.
The drive is then withdrawn leaving the rod behind. At some later time, the rod blade becomes unstuck and falls rapidly out of the core.
If the control rod has sufficient reactivity worth the potential for localized fuel damage exists.
The control drop accident for LACBWR has been analyzed and the results presented in a docunent, NES-81 A0033, entitled "LACBWR Rod Drop Prob-ability Study". This report was reviewed and the staff concluded that backfitting a rod worth minimizer or rod sequence control system was not necessary for LACBWR. This conclusion was based on analysis which i
showed that, with conservative estimates of tte various component prob-abilities including that of having a rod of sufficient worth, the nx-imum probability of a rod drop accident exceeding a maximum fuel enthalpy of 280 calories per gram is 4.6 X 10-9 This is sufficiently low to permit the conclusion that additional protection against the event is not warranted.
- The analysis of the rod drop accident that was performed for LACBWR did not include the effects of thermal hydraulic feedback on the event.
Analysis performed by Brookhaven National Laboratories and reported in BNL-NUREG-28109, " Thermal-Hydraulic Effects on Center Rod Drop Acci-dents in a Boiling Water Reactor", dated July,1980, show that inclusion of such effects greatly reduces the consequences of such events. While the analysis was not performed specifically for LACBWR the results are applicable to it. These results permit the conclusion that if these effects were included in the rod drop analysis, peak enthalpies in excess of 280 calories per gram would not be calculated for LACBWR.
On the basis of the discussion presented above which concludes that a rod drop accident which would result in rapid dispersal of fuel into the coolant (i.e. fuel enthalpy greater than 280 calories per gram) does not occur in LACBWR, we conclude that a significant pressure pulse, does not becur. The power increase occurs chiefly in the area of the dropped rod and in LACBWR can only occur when significant voids are present. This would tend to mitigate the effects of any pressure rise.
VI.
CONCLUSION On the basis of the evaluation presented above we conclude that the core performance analysis of the rod drop accident for LACBWR meets present day criteria and is acceptable.
The potential radiological consequences will be a,ddressed in a separate evaluation.
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