ML071590342

Email from Christian Jacobs (NRC) to Charlie Medenciy (Wolf Creek Nuclear Operating Corporation) Regarding SAMA Questions on Wolf Creek LRA

ML071590342
Person / Time
Site:	Wolf Creek
Issue date:	05/23/2007
From:	Christian Jacobs NRC/NRR/ADRO/DLR/REBB
To:	Bell L, Medenciy C Wolf Creek
References
Download: ML071590342 (4)

Text

1Chrilstian f466bs' -- WO'SA-MAQuesio' Pag 1 From: Christian Jacobs To: Charlie Medenciy; Lorrie Bell Date: 5/23/2007 2:55:04 PM

Subject:

WC SAMA Questions Charlie, Lorrie, Attached are some additional questions that our SAMA reviewers would like to discuss. I'm checking availability for all parties to set up a teleconference call. I suggest some time later next week (Thursday, May 31) if that works for everyone else.

Chris Christian Jacobs Environmental Project Manager Division of License Renewal U.S. Nuclear Regulatory Commission 301-415-3874 O-12D11 cjj @nrc.gov CC: Alicia Williamson; bmrowca@islinc.com; Robert Palla

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Mail Envelope Properties (46548E08.D98: 2 : 35062)

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WC SAMA Questions Creation Date 5/23/2007 2:55:04 PM From: Christian Jacobs Created By: CJJ@nrc.gov Recipients Action Date & Time islinc.com Transferred 5/23/2007 2:55:16 PM bmrowca CC (Bruce Mrowca) nrc.gov TWGWPO03.HQGWDO01 Delivered 5/23/2007 2:55:10 PM RLP3 CC (Robert Palla) nrc.gov TWGWPO04.HQGWDOO1 Delivered 5/23/2007 2:55:10 PM ARWI CC (Alicia Williamson)

WCNOC.com Transferred 5/23/2007 2:55:16 PM chmeden (Charlie Medenciy) loyokum (Lorrie Bell)

Post Office Delivered Route islinc.com TWGWPO03.HQGWDOO1 5/23/2007 2:55:10 PIM nrc.gov TWGWPO04.HQGWDO01 5/23/2007 2:55:10 PIV nrc.gov WCNOC.com Files Size Date & Time MESSAGE 965 5/23/2007 2:55:04 P M WCGS followup RAIs 5.22.07 Rev 1.wpd 8072 5/22/2007 12:34:52 PM Options Auto Delete: No Expiration Date: None Notify Recipients: Yes Priority: Standard

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Subject:

No Security: Standard To Be Delivered: Immediate Status Tracking: Delivered & Opened

5/22/2007 Wolf Creek Followup Questions for Discussion

1. In response to Question 1b (QU-9) it is stated that "internal flooding scenarios have not been included in PSA updates and that new internal flooding criteria may identify new human-induced floods through consideration of errors of commission." State how this could impact the SAMA analysis.

2. The answer provided to Question lb (TH-6) talks about the definition used in the I PE while the F&O is focused on the MAAP calculation and states that "the 30-'minute criteria was not likely a factor in determining success or failure in many (if any) of the analyses performed for the IPE." State the issue associated with the core damage criteria used in the MAAP calculations and its impact on the SAMA evaluation.

3. In response to Question 2b, it is stated that NUREG-1 570 asserts that only 2 percent of the high pressure melt scenarios with dry steam generators would result in an induced SGTR and that the applicability of percentage is predicated on the conditions that the secondary side is not depressurized and that the RCPs are not operated. A review of NUREG-1 570 found that for cases where a RCP Seal LOCA occurs that the TI-SGTR failure probability for 3 steam generators (4 SG scenario is not available) is 1.0. State how this increased failure likelihood on RCP seal failure was accounted for in the response to Question 2b.

4. In response to Question 3c, it is stated that "WCGS has high temperature seal material installed in only one RCP." State the seal LOCA model used. Discuss plans for installing qualified seals in the remaining RCPs. Provide a cost-benefit assessment for installing qualified seals in the remaining RCPs, if this is not currently planned.

5. In response to Question 7a, it is stated that procedures exist at WCGS to operate the turbine driven AFW pump after battery depletion. Describe these procedures, including the instrumentation used to effectively control SG level. State the human error probability or probabilities used for this action.

6. In response to Question 6.b.ii it is stated that extending the EDG completion time actions 24/7 would potentially place Technical Specification Amendment #163 in jeopardy. This appears to be due to the presence of a reduced base risk profile as a result of implementing the proposed alternative making the removal of an EDG more significant (Its not that an EDG outage is more significant, it is that the plant's risk profile is lower making the delta risk greater). Please clarify this.