NRC-97-0003, Responds to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions
| ML20134F249 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 01/28/1997 |
| From: | Gipson D DETROIT EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-97-0003, CON-NRC-97-3 GL-96-06, GL-96-6, NUDOCS 9702070327 | |
| Download: ML20134F249 (7) | |
Text
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Douglas R. Gipson
, Sen,or Vee President j
Nuclear Generation OR r.rmi 6400 North Dix 6e Highway Edison
~(313)SE5249
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l January 28,1997 NRC-97-0003 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555
References:
1)
Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43 2)
NRC Generic Letter 96-06: Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions, dated September 30,1996 3)
Detroit Edison Letter to NRC," Detroit Edison 30-Day Respone *o NRC Generic Letter 96-06", NRC-96-0118, dated October 30,1996
Subject:
Detroit Edison 120-Day Response to NRC Generic Letter 96-06 Reference 2 requested a 120 day response describing actions taken and conclusions reached relative to the issues communicated by Generic Letter 96-06. Reference 3 committed Detroit Edison to the 120 day response. Detroit Edison was requested to determine: 1) if containment air cooling water systems are susceptible to either waterhammer or two-phase flow conditions during postulated accident conditions; 2) if piping systems that penetrate the containment are susceptible to thermal expansion of fluid so that overpressurization of piping could occur.
lj Based on engineering reviews that were performed it was determined that the containment air cooling water systems continue to be operable. Detroit Edison has i
j determined containment penetrations are operable based on engineering judgment, taking credit for expected leakage of boundary valves to prevent over pressurization l
of piping. Please refer to the Attachment for Detroit Edison's complete response to the Generic Letter.
9702070327 970128 PDR ADOCK 05000341 p
j-USNRC NRC-97-0003 -
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The following commitment is being made in this letter-i l.
Detroit Edison will monitor the ongoing discussion of these generic issues between the Nuclear Energy Institute and NRC and design and install modifications as necessary, to provide adequate overpressure protection of j
affected drywell piping penetrations. If necessary to maintain the Fermi 2 j
containment integrity licensing basis, these modifications will be installed by i
the completion of the next refueling outage.
i If you have any questions, please contact Joseph M. Pendergast, Licensing Engineer at (313) 586-1682.
1 Sincerely, 5
i Attachment cc: A. B. Beach I
M. J. Jordan A. J. Kugler A. Vegel j
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USNRC NRC-97-0003 Page 3 1, DOUGLAS R. GIPSON, do hereby affirm that the foregoing statements are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.
I DOUGLAS R.DIPSON Senior Vice President-On this 8
day of
'44/46[1997 before me personally appeared Douglas R. Gipson, being fir 8 duly sword and says that he executed the foregoing as his free act and deed.
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Nbtary Public ROSAUE A. ARMETTA NOTARYPU3UC MONROECOUNTYMl MYCOMMISSION EXPlRES 10/11/99
, to NRC-97-0003 Page1 RESPONSE to GENERIC LETTER %-06 Reauested Action Number 1: Determine if containment air cooler cooling water systems at Fermi 2 are susceptible to either waterhammer or two-phase flow conditions during postulated accident conditions.
Background
The design basis Loss of Coolant Accident (LOCA) concurrent with a Loss of Offsite Power (LOP) is the limiting event with regard to the concerns described in Generic Letter 96-06. At Fermi 2 the Reactor Building Closed Cooling Water (RBCCW) 4 system supplies coolant for the drywell coolers. The RBCCW System and drywell coolers are not required for either safe shutdown or to mitigate postulated accidents.
The Emergency Equipment Cooling Water (EECW) system provides emergency 4
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cooling to the essential equipment necessary for safe shutdown of the plant normally cooled by the RBCCW system.
Actions Taken and Basis for Continued Operability As requested by the Generic Letter, Detroit Edison has performed an evaluation to i
determine if containment air cooler cooling water systems at Fermi 2 are susceptible to either waterhammer or two-phase flow conditions during postulated or Design Basis Accidents (DBA) conditions. The results of this evaluation are discussed below.
In the postulated accident scenario, the RBCCW supply to the drywell coolers is isolated automatically on high drywell pressure. The RBCCW isolation valves would be closing during the EECW pump start and acceleration interval. Although some steam bubble formation in the coolers during the early stages of the accident is possible, there would be virtually no flow through the cooling coils at this time and the EECW piping would not be affected.
Steam bubble collapse is not expected to occur because there is no flow of subcooled water into the system. This flow is precluded initially by automatic closure of the drywell closed cooling water supply line as noted above. Procedural controls prevent reestablishing cooling water flow to the drywell coolers if predetermined limits on drywell pressure and torus temperature have been exceeded thus minimizing the potential for waterhammer during the accident recovery process. Also, the effects of any possible steam formation in the RBCCW/EECW system would be mitigated by relief valves on EECW piping inside the drywell and / or RBCCW/EECW head tanks.
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Attachment I to j
j NRC-97-0003 Page 2 In summary, Fermi 2 does not rely on drywell coolers for post accident containment heat removal or safe shutdown. Waterhammer or two phase flow events in the 4
RBCCW/EECW systems are not a concern based on review of the system design including the time it takes for the flow of RBCCW cooling water to stop during the design basis accident, the time to establish cooling water flow in the EECW system, and operating restrictions on system restart to reestablish flow to the drywell coolers.
Based on this evaluation, it was concluded that the EECW, RBCCW and drywell cooling systems continue to be operable without the need for any modifications to the design or procedures. Therefore, no further corrective active actions are planned.
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i Attachment I to NRC-97-0003 Page 3 Reunested Action Number 2: Determine if piping systems that penetrate the containment at Fermi 2 are susceptible to thermal expansion of fluid so that overpressurization of piping could occur.
Actions Taken Prior to restart from the fifth refuel outage an evaluation of containment piping penetrations was performed. This evaluation determined that some piping penetrations are susceptible to thermal overpressurization. These penetrations are:
- 1. Main Steam Drains [3" Nominal Pipe Size (NPS)]
- 4. Reactor Water Sample (l" NPS)
The review of drywell piping systems inside of the inboard isolation valve also identified that the two drywell drain sump discharge lines between the pump discharge check valves and the inboard isolation valves are susceptible to thermal overpressurization.
As noted in Generic Letter 96-06 the potential for system failure as a result of thermally induced overpressurization is dependent on many factors. The Generic Letter further notes that these factors include leak tightness of valve seats, bonnets, packing glands, etc. Taking such expected leakage into account the determination was made that for these piping penetrations isolation valve leakage will limit the pressure rise to less than the failure pressure.
The Basis for Continued Operability The review concluded that operability of containment penetrations listed above was not adversely affected by this phenomenon. The review was based on engineering judgment and included considerations such as those described in the following paragraphs.
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Attachment I to NRC-97-0003 Page 4 For Reactor Water Sample, and the RRS Pump Seal Purge (2 penetrations), the containment isolation valves are globe valves and the expanding volume of water is small. The valves would be expected to leak to limit thermal overpressurization before the 3/4" and I" schedule (sch.) 80 or 160 piping would fail.
l The main steam drains drywell penetration is isolated hot after drain flow has reduced so the penetration piping is not likely to be solid. Also, the piping should be above ambient temperature due to interfacing high temperature systems on each side of the j
isolation valves. Lastly, although the isolation valves are gate valves, valve leakage to l
limit thermal overpressurization would be expected before the 3" sch.160 piping would fail.
The piping between the drywell sump pump discharge check valves and the inboard j
isolation valves cannot significantly pressurize since the two check valves are hard seated and would leak to limit thermal overpressure. This piping is not safety related.
Also, since the inboard isolation valves are normally open, check valve leakage would drain the penetration piping at the high point of the piping system so the penetration piping is not likely to be solid. Although the isolation valves are gate valves, the j
valve will leak to limit thermal overpressurization before the 3" sch. 40 piping would j
fail.
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The evaluation was based on the assumption that leakage would exist as discussed above. Local Leak Rate Test data was reviewed to confirm that all of the penetrations exhibited at least some leakage when last tested in the fifth refuel outage.
The evaluation was based primarily on engineering judgment.
Corrective Actions Planned Detroit Edison will monitor the ongoing discussion of these generic issues between the Nuclear Energy Institute and NRC and design and install modifications as necessary, to provide adequate overpressure protection of affected drywell piping penetrations. If necessary to maintain the Fermi 2 containment integrity licensing basis, these modifications will be installed by the completion of the next refueling outage.