ML20033D135
| ML20033D135 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 11/17/1981 |
| From: | Wuller G ILLINOIS POWER CO. |
| To: | John Miller Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.44, TASK-TM GL-81-32, U-0331, U-331, NUDOCS 8112070275 | |
| Download: ML20033D135 (3) | |
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U-0331 ILLINDIS POWER 00MPANY fp L39_gt(11_tyy_5 500 SOUTH 27TH STREET, DECATUR, ILLINOIS 62525 November 17, 1981 Mr. James R. Miller, Chief Standardization & Special Projects Branch i
Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Miller:
Reference.
NRC letter 8/7/81, D. G. Eisenhut to all Applicants /
Licensees Referencing BWR Owners Group Response to NUREG-0737, ITEM II.K.3.44 Clinton Power Station Unit 1 Docket No. 50-461 This is in reply to the referenced letter on the subject of NUREG-0737, ITEM II.K.3.44 - Evaluation of kricipated Transients Combined with Single Failure (Generic Letter No. 81-32).
Illinois Power's response to this item for the Clinton Power Station was provided in the CPS-FSAR Appendix D, page D-107 in Amendment 7, September 1981 (copy attached).
The BWR/6 generic analysis is representative for the Clint.on Power Station.
Sincerely,.
?
G. E. Wuller Supervisor - Licensing Nuclear Station Engineering GEW/mak cc:
Mr..J. H. Williams, NRC Clinton Project Manager Mr. H. H. Livermore, NRC Resident Inspector
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CPS-FSAR AMENDMENT 7 SEPTEMBER 1981 NRC ACTION PLAN (NUREG-0660 as clarified by NUREG-0737)
II.K.3.44 Evaluation of Anticipated Transients with Sinole Failure to Verify No Fuel Failure r
NRC Position For anticipated transients combined with the worst single failure and assuming proper operator actions, licensees should demonstrate that the core remains covered or provide analysis to show that no significant fuel damage results from core uncovery.
Transients which result in a stuck-open relief valve should be included in this category.
CPS Response In response to this requirement, Illinois Power Company, jointly sponsored through the BWR Owners Group, conducted an evaluation of the worst anticipated transient with the worst single failure.
These results were submitted to the NRC via a letter from D. B.
Waters, Chairman BWR Owners Group, to D. G. Eisenhut, Director
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NRC, dated December 29, 1980.
The letter from D. G. Eisenhut (NRC)- to D. B. Waters (BWR Owner's Group) transmitted the NRC evaluation of this item.
The report was found to be acceptable for referencing by individual licensee / applicants.
A summary of the results of the analysis follows.
The anticipated transients in NRC Regulatory Guide 1.70, Revision 3 were reviewed for all BWR product line BWR/2 through BWR/6 from a core cooling viewpoint.
The loss of feedwater event was identified to be the most limiting transient which would challenge core cooling.
The BWR/6 is designed so that the HPCS or ADS with subsequent low pressure makeup is independently capable of maintaining the water level above the top of the active fuel given a loss of feedwater.
The detailed analysis shows that even with the worst single failure in combination with i
the worst transient the core remains covered.
Furthermore, even with degraded conditions involving one SORV in addition to the worst transient with the worst single failure, studies show that the core remains covered during the whole course of the transient either due to RCIC operation or due to automatic depressurization via the ADS or manual depressurization by the oeprator so that low pressure inventory makeup can be used.
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It is concluded that for anticipated transients combined with the worst single failure, the core remains covered.
Additionally, it is concluded that for severely degraded transients beyond the D-107 j
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CPS-FSAR AMENDMENT 7 SEPTEMBER 1981 h
design basis where it is assumed that SRV sticks open and an additional failure occurs, the core remains covered with proper operator action.
The generic analysis for the BWR/6 product line was performed using the Kuo Sheng Station design.
The key analysis parameters of core thermal power, reactor coolant inventory and RCIC system flow at Clinton Power Station are identitical to the Kuo Sheng Station parameters.
Based on this fact, it is concluded that the BWR/6 generic analysis is representative for the Clinton Power Station.
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D-108
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