ML20054H114
| ML20054H114 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/18/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Vandewalle D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TASK-15-03, TASK-15-3, TASK-RR LSO5-82-06-062, LSO5-82-6-62, NUDOCS 8206220586 | |
| Download: ML20054H114 (9) | |
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June 18, 1982 Docket flo. 50-155 LS05-82-06-062 Mr. David J. VandeWalle Nuclear Licensing Administrator Consumers Power Company 1945 W Parnall Road Jackson, Michigan 49201
Dear Mr. VandeWalle:
SUBJECT:
BIG ROCr, POINT - SEP TOPIC XV-3, LOSS OF EXTERNAL LOAD, TURBINE TRIP, LOSS OF CONDENSER VACUUM, CLOSE OF MAIN STEAM ISOLATION VALVE, AND STEAM PRESSUFC REGULATOR FAILURE By letter dated October 16, 1981, the staff issued its safety evaluation on SEP Topic XV-3.
Your letter of June 3,1982, identified a correction to that evaluation. The staff also modified the evaluation to be more consistent with the SEP topic format.
Enclosed is our final topic evaluation which concludes that with respect to this topic, your facility meets current licensing criteria.
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 future if your facility design is changed or if NRC criteria relating to this subject are modified before the integrated assessnent is completed.
Sincerely, l
Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
As stated 3404 QSugSSX/f/f) cc u/ enclosure:
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NRC FORM 318 (10-80) NRCM O24o OFFIClAL RECORD COPY usam ini-m.co
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.Mr. David J. VandeWalle cc Mr. Paul A. Pe'rry, Secretary U. S. Environmental Protection Consumers Power Company Agency 212 West Michigan Avenue Federal Activities Branch Jackson, Michigan 49201 Region Y Office ATTN:
Regional Radiation Representative Judd L. Bacon, Esquire 230 South Dearborn Street
-Consumers Power Company Chicago, Illinois 60604 212 West Michigan Avenue Jackson, Michigan 49201 Peter B. Bloch, Chairman Atomic' Safety and Licensing Board Joseph Gallo, Esquire U. S. Nuclear Regulatory Commission Isham, Lincoln & Beale Washington, D. C.
~20555 1120 Connecticut Avenue Room 325 Dr. Oscar H. Paris Washington, D. C.
20036 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Peter W. Steketee, Esquire Washington, D. C.
20555 505 Peoples Building Grand Rapids, Michigan 49503 Mr. Frederick J. Shon Atomic Safety and Licensing Board Alan S. Rosenthal, Esq., Chairman U. S. Nuclear Regulatory Commission Atomic Safety & Licensing Appeal Board Washington, D. C.
20555 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Big Rock Point Nuclear Power Plant ATTN: Mr. C. J. Hartman Mr. John O'Neill, II Plant Superintendent Route 2, Box 44' Charlevoix, Michigan 49720 Maple City, Michigan 49664 Christa-Maria Mr. Jim E. Mills Route 2, Box 108C Route 2, Box 108C Charlevoix, Michigan 49720 Charlevoix, Michigan 49720 William J. Scanlon, Esquire Chairman 2034 Pauline Bouisvard County Board of Supervisors Ann Arbor, Michigan 48103 Charlevoix County Charlevoix, Michigan 49720 Resident Inspector Big Rock Point Plant Office of the Governor (2) c/o U.S. NRC Room 1 - Capitol Building RR #3, Box 600 Lansing, Michigan 48913 Charlevoix, Michigan 49720 Herbert Semmel Counsel for Christa Maria, et al.
Urban Law Institute Antioch School of Law 263316th Street, NW Washington, D. C.
20460
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r Mr. David J. VandeWalle cc Dr. John H. Buck Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission 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 Commission Washington, D. C.
20555 James G. Keppler, Regional Administrator Nuclear Regulatory Commission, Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 a
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SYSTEMATIC EVALUATION PROGRAM TOPIC XV-3 BIG ROCK POINT Loss of External Load I.
INTRODUCTION The plant is designed to accomodate loss of external load without any significant transients felt by the reactor (Ref.1).
The turbine control valves close only partially and the plant transfers to house load.
During the transfer steam is rejected to the condenser through the turbine bypass valve.
If the turbine bypass valve fails to open the reactor is tripped hy high neutron flux or by high pressure.
The licensee has not analyzed the loss of external load, but has identified the turbine trip without hypass as a bounding event (Ref. 2).
II.
EVALUATION In the extreme case of complete load rejection and failure of the bypass valve to open, the transient is identical to the turbine trip without
___m bypass. Otherwise loss of external load is a less severe event.
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III.
CONCLUSION J-
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~ Loss of external load transient is bounded by the turbine trip event which has been evaluated and found in conformance with the criteria of SRP Section 15.2.1.
Turbine Trip I.
INTRODUCTION A turbine trip is actuated by fast closure of the turbine stop valve which abruptly interrupts the steam flow to the turbine.
The stop valve closes in 0.5 seconds.
The plant is designed to accomodate a 138KV transmission line trip at reactor power up to 160MWt without j
a reactor scram by automatic opening of the turbine bypass valve.
If the turbine bypass fails'to open or if the trip occurs from a higher load, the pressure starts to increase and the reactor is scrammed from,-
high neutron flux.
Pressure increase is buffered by a steam drum which is located on the steam flow path between the reactor vessel and the turbine.
High pressure in the steam drum actuates the emergency conden-ser by opening a valve in each of two redundant lines which pass the steam through the emergency condenser.
If the emergency condenser fails, the steam is relieved by six safety valves.
These are large enough to pass all steam generated in the reactor even if the reactor is not tripped.
. The licensee has recently analyzed the turbine trip event, assuming a failure of the bypass valve to open, and has presented the results in Ref. 2.
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 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 determina-tion of the margins of safety during normal operations and, transients 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 i,ntegrity of the physical barriers which guard against the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to 10 CFR Part 50) set forth the criteria for the design of water-cooled reactors.
GDC 10 " Reactor Design" requires that the core and associated cooling, 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 operational occurrences.
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 26 " Reactivity Control System Redundance and Capability" requires that the reactivity control system be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for mal-functions such as stuck rods, specified acceptable fuel design limits are not exceeded.
III.
RELATED SAFETY TOPICS Various other SEP topics evaluate features of the reactor protection system.
The effects of single failures on safe shutdown capability are considered under Topic VII-3,
IV.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.2.1.
The evaluation included review of the analysis for the event and identifica-tion 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.
Devia-tions from the criteria specified in the Standard Review Plan are identified.
V.
EVALUATION The turbine trip analysis presented in Ref. 2, assumes that the bypass valve fails to open.
Thus the only protective action taken into account is the reactor trip, actuated by high neutron flux.
The codes used in the analysis are RETRAN and a modified version of COBRA-IV-I. These have not been formally reviewed and accepted by NRC.
The plant parameters used as code input are conservative and the assumptions made in the analysis are consistent with the acceptance criteria of SRP Section 15.2.1.
The calculation is continued 14 seconds into the transient.
The reactor is tripped in 0.95 seconds and the pressure reaches its maximum value in 8.25 seconds.
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The normal operating pressure of the plant is 1350 psia and the reactor vessel design pressure is 1715 psia.
The calculated maximum pressure in the reactor vessel is 1445 psia and 1432 psia in the steam drum.
The initial pressure transient is terminated by the cold feedwater added to the steam drum.
It is uncertain whether the cold feedwater would condense the steam and.
reduce the pressure as predicted by the calculational model.
- However, we have estimated that the pressure transient would be rapidly turned down by proper operation of the emergency condenser set to actuate at 1450 psia.
In case of the emergency condenser failure the safety valves would open at 1550 psia limiting the pressure well below the design pressure.
The initial steady state CPR used in the calculation (1.68) is lower than the minimum CPR calculated for the current fuel cycle, The lowest CPR during the transient is 1.40 while the acceptance criteria for XN-2 correlation is 1.32.
A possible inaccuracy in the modelling of the feed-water mixing in the steam drum does not significantly influence the minimum CPR which occurs early in the transient.
Thus the design limits are not exceeded.
. l Taking into account the margin betweer the calculated results and the acceptance criteria, and based on our experience on expected BWR behavior, we conclude that the turbine trip event in the Big Rock. Point Plant is in conformance with the criteria of SRP Section 15.2.1.
VI.
CONCLUSION As part of the SEP reviews for Big Rock Point, we have evaluated the licensee's analysis of the turbine trip transient (Ref. 2), against the criteria for SRP Section 15.2.1.
Based on this evaluation we have concluded that the analysis is in conformance with the present SRP criteria.
INTRODUCTION If the condenser vacuum starts to decrease the reactor is tripped at a preset condenser pressure.
The setting for turbine trip is somewhat higher and thus the loss of condenser vacuum is expected to be a milder transient than the turbine trip without bypass.
The licensee has not analyzed the loss of condenser vacuum, but has identified the turbine trip without bypass as a bounding event (Ref. 2).
II.
EVALUATION In the extreme case of sudden loss of condenser vacuum or in the case of failed reactor trip at increased condenser pressure the transient is identical to the turbine trip without bypass.
Otherwise loss of condenser vacuum is a less severe event.
III.
CONCLUSION The loss of condenser vacuum transient is bounded by the turbine trip which has been evaluated and found in conformance with the criteria of SRP Section 15.2.1.
Closure of Main Steam Isolation Valve I.
INTRODUCTION Big Rock Point has only one steam line, and there is a single isolation j
l valve in that line.
The itolation valve closes very slowly, with a closure time of 40 seconds while the turbine stop valve closes in 0.5 seconds.
The reactor is tripped directly by isolation valve position switches when the valve has closed 50%.
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. i The licensee has not analyzed the closure of the main steam isolation valve, but has identified the turbine trip without bypass as a bounding event (Ref. 2).
II.
EVALUATION Closure of the main steam isolation valve transient is terminated hy reactor trip from isolation valve position before the reactor poser or steam pressure has increased.
If the direct trip signal fails the reactor is tripped from the higher power or the high pressure.
In this case the transient is much slower and thus less severe than the turbine trip without bypass.
III.
CONCLUSION Closure of main steam isolation valve is bounded hy the turbine trip event which has been evaluated and found in conformance with the criteria of SRP Section 15.2.1.
Steam Pressure Regulator Failure I.
INTRODUCTION The steam line pressure is normally maintained at a constant value by the pressure regulator which positions the turbine control valves without regard to the generator load.
If the pressure regulator fails and starts to inadvertently close the control valves, the increased pressure is sensed by two independent pressure sensors which start to open the turbine bypass valve.
The event induces only a mild transient in the core.
The licensee has not analyzed the steam pressure regulator failure, but has identified the turbine trip without bypass as a bounding event (Ref. 2).,.
II.
EVALUATION In the most limiting case, full closure of the control valves at maximum speed and failure of the bypass to open, the event would be similar to the turbine trip without bypass.
Otherwise steam pressure regulator failure is a.less severe event.
III.
CONCLUSION The steam pressure regulator failure is bounded hy the turbine trip event which has been evaluated and found in conformance with the criteria of SRP Section 15.2.1.
REFERENCES 1.
Final Hazards Summary Report for Big Rock Point Plant, November 14, 1961.
2.
Letter from Robert A. Vincent, Consumers Power Company, to Dennis M.
Crutchfield, NRC,
Subject:
Docket 50-155-License DPR-6-Big Rock Point Plant, SEP Design Basis Event Topics XV-1, XV-3, XV-4, XV-5, XV-7, and XV-9, dated July 15, 1981.
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