ML20203L653
| ML20203L653 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 02/27/1998 |
| From: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-06, GL-96-6, NLS970231, NUDOCS 9803060161 | |
| Download: ML20203L653 (4) | |
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COOPER NUCLE AR STATON P.O. 90X 98, BROWNVILLE. NEBRASKA 68321 Nebraska Public Power District
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G. R. Horn g
Senior Vice-President, linergy Supply (402) 563 5518 NLS970231 February 27,1998 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:
Subject:
Response to NRC Generic Letter 96-06 C*oer Nuclear Station, NRC Dociet 50-298, DPR-46
Reference:
1 NRC Generic Letter 96-06 dated September 30,1996," Assurance of Equipment Operability and Containment Integrity during Design-Basis l
Accident Conditions" l
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" Response to NRC Generic Letter 96-06"
- 3. NRC Generic Letter 96-06, Supplement 1, November 13,1997, " Assurance of Equipment Operability and Containment Integrity during Design-Basis Accident Conditions"
- 4. NEI Letter, dated December 11,1997,"NEl/NRC Workshop on GL96-06" This submittal provides a follow up response to NRC Generic Letter (GL) 96-06 (Reference 1).
In Reference 2, the Nebraska Public Power District (District) stated its position on Penetration X8, the Main Steam Drainline, Penetration X12, Residual Heat Removal (RHR) Shutdown Cooling Suction, and Penetration X14, Reactor Water Clean-up Suction. The following is a I
revised position on these penetrations with regard to their overpressurization potential:
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7 Penetration X8:
Size 3", Main Steam Drainline in Reference 2 the District stated that this line is normally hot during power operation since the primary isolation valves are maintained in the open position. Thus, there is no overpressurization effects on this line post accident 980306d 61 980227 PDR ADOCK 05000298 jigg!\\g}{\\]}Q){
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' f February 27,1998 Page'2 4
since the initial temperature in the line is about the same as the drywell L
temperature post accident, and therefore no modification is necessary. While the valves are normally open, they could be closed during power operation. A calculation has been performed, NEDC 96-058', which demonstrated that, in the event of a LOCA and with the valves closed, over-pressurization of the penetration piping would net ocetr. Therefore, there is no over-pressurization concern with this penetration.
Penetration X12: Size 20", Residual 11 eat Removal Shutdown Cooling Suction In Reference 2 the Di: trict stated that it wot Id investigate various alternatives, including use of administrative co Arols or a modification which would have installed a relief valve, to ensure long term protection against possible overpressurization -f his penetration following a L.OCA.
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The analysis in calculation NEDC 96-058 determined that, incorporating the latest Pressure Isolation Valve (PlV) measurements for valve leakage, the post LOCA stress levels would only slightly exceed code allowables as defined by B31,1-1967, and would remain well within ASME III, Section NB limits for Service Level C (Emergency Condition). Assur. :ng no PlV leakage, the post LOCA stress levels would exceed the Senice Level C limits but would remain within the limits of the ASME code,Section III, Appendix F. As discussed in References 3 and 4, meeting ASME code, Section Ill, Appendix F is acceptable as a temporary resolution of the thermal overpressurization
'ssue.
i Since the st ess levels in the present penetration X12 piping configuration will remain within the limits of ASME,Section III, Appendix F, the District has placed on hold the modification scheduled for this penetration until the NRC makes a detennination on the use of Appendix F criteria as a permanent resolution of the overpressurization issue.
Penetration X14: Size 6", Reactor Water Clean-up Suction Reference 2 indicated that there would only be.
erpressurization concern if the LOCA were to occur after the penetration had been isolated, during power operation, for greater than 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />. To address this issue, the District has revised the RWCU operating procedures to ensure administrative controls are in place to prevent post-LOCA overpressurization concerns in the event that the penetration becomes isolated during power operation. Calculation
' Calculation NEDC 96-058 " Evaluation of the Overpressurization Potential foi isolated Penetrations in Accordance with GL96-06"
,NLS970231 February 27,1998
' Page 3 NEDC 96-058 determined that, even with the isolation valves closed for longer than 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />, the stress levels on the penetration piping as a result of a i
LOCA co not exceed the limits of ASME,Section III, Appendix F. liowever, District will maintain the present procedural testrictions whenever the isolation valves for penetration X14 are closed for greater than 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />, until the NRC makes a final determination on the use of Appendix F criteria.
The Disirict would welcome any comments the NRC staff have on the above District positions.
Should you nave any questions concerning this raatter, please contact me.
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'l Se ' - Vice President - Energy Supply
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cc: Regional Administrator USNRC - Region IV J
~ Senior Project Manager USNRC - NRR Project Directorate IV-1 Senior Resident Inspector USNRC NPG Distribution
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ATTACHMENT 3 LIST OF NRC COMMITMENTS l
Correspondence No:.NLS970231 The following table identifies those actions com.T.itted to by the District in this document. Any other actions discussed in the submittal represent intended or planned actions by the District.
They are described to the NRC for the NRC's information and are not regulatory commitments.
Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
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COMMITTED DATE COMMITMENT OR OUTAGE NONE 1
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PROCEDURE NUMBER 0.42 l
REVISION NUMBER 4 l
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