ML20207D468
| ML20207D468 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/24/1999 |
| From: | Alexander J BOSTON EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BECO-2.99.052, GL-96-06, GL-96-6, NUDOCS 9906030346 | |
| Download: ML20207D468 (3) | |
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L 3 Boston Edison N A BEC ENERGY COMPANY J.F. Alexander Nuclear Assessment Group Manager l
May 24,1999 l
BECo Ltr. 2.99.052 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 RESPONSE TO GENERIC LETTER 96-06 VERBAL REQUEST FOR INFORMATION Backaround -
This letter provides additional information to that included in Boston Edison (BECo) letter 98-123 dated October 1,1998.
Letter 98-123 addressed NRC concems described in Generic Letter (GL) 96-06 concerning waterhammer in the reactor building closed cooling water (RBCCW) system.
This submittal j
summarizes the results of a waterhammer analysis calculation performed in response to an NRC l
verbal request. The calculation supports the conclusion provided in letter 98-123 that containment
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penetrations are not damaged by waterhammer in RBCCW piping in the drywell.
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ResDOnse The RBCCW system inside containment at Pilgrim that is subject to heatup during a loss-of-coolant-accident (LOCA) will not experience waterhammer when the system performs its design-basis safety function. Only Loop B of Pilgrim's RBCCW provides cooling water to the drywell coolers. During a loss of AC power (LOOP) assumed with the LOCA, the RBCCW system restarts on emergency AC power within sufficient time to preclude steam formation in the drywell coolers.
However, there are accident scenarios involving active failures that prevent the RBCCW Loop B I
l pumps from restarting at appropriate times, such as failure of the Loop B Emergency Diesel
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Generator. As described in letter 98-123, a delayed restart of the RBCCW Loop B pumps possibly U
could result in a waterhammer event. Under these single failure conditions, it is not necessary for RBCCW Loop B to perform its safety funct!on. However, there must not be any adverse effect on containment isolation or integrity, p
9906030346 990524 '
PDR ADOCK 05000293 P
pyg Pilgrim Nuclear Power Station. Rocky Hill Road. Plymouth. Massachusetts 02360 u
Letter 98-123 postulated that the drywell coolers may be damaged by a waterhammec, but that containment integrity would not be affected. To confirm this postulation, Pilgrim performed a simplified waterhammer analysis (Pilgrim Calculation M-955) using the methods of NUREG/CR 5220.
Based on this calculation, Pilgrim concludes that waterhammer caused by the worst case LOCA/ LOOP will not compromise the integrity or operability of the primary containment system because the peak magnitude of the resulting pressure wave plus the maximum operating pressure in the 6" drywell penetration piping is less than the design pressure of 150 pounds per square inch (psig). This analysis is based on the maximum, steam void geometry (F = 1), a bounding sonic velocity (4900 ft/sec), and only the attenuation due to the fittings and junctions. Since no attenuation factors for elbows are provided in NUREG/CR 5220, these factors were obtained from EPRI NP-6766. The pressure in the 1-1/2" piping to the drywell coolers, in which the waterhammer event occurs, was significantly greater and it is considered the drywell coolers may be damaged as a result.
However, in this scenario it can be assumed that the 1-1/2" piping failure is an acceptable consequence because the waterhammer event occurs only after RBCCW Loop B has failed to
. perform its safety function. Assuming an additional failure of either the supply or return containment isolation valves would be beyond the PNPS design basis requirements. Thus, containment integrity remains intact.
Procedure Update, Letter 98-123 stated that to preclude the potential for waterhammer Procedure 2.2.19.5, "RHR Modes of Operation for Transients," was revised to prevent operators from initiating flow through the coolers when the drywell temperature exceeds 250 F. The change was delayed by other, unrelated changes to procedure 2.2.19.5.
This delay in issuing Procedure 2.2.19.5 was identified, and this procedure and Procedure 2.4.42,
" Loss of RBCCW," were both revised and issued May 6,1999, to address this issue.
Conclusion This response completes Pilgrim's GL 96-06 effort. Should you require additional information on GL 96-06, please contact P.M. Kahler at (508) 830-7939.
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nder PK/nb 299052 cc: see attached I
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i U.S. Nuclear Regulatory Commis-lon cc: Mr. Alan B. Wang, Project Manager -
Project Directorate 14
- Office of Nuclear Reactor Regulation Mail Stop: OWFN 14020 U. S. Nuclear Regule.cory Commission 1 White Flint North ~
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11555 Rockville Pike Rockville, MD 20852 U. S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406 Senior NRC Resident inspector Pilgrim Nuclear Power Station I
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