ML093490797

E-mail Request for Additional Information, License Amendment Request to Modify Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for One-time Extension of Integrated Leak Rate Test Interval

ML093490797
Person / Time
Site:	Callaway
Issue date:	12/15/2009
From:	Thadani M Plant Licensing Branch IV
To:	Elwood T, Maglio S Union Electric Co
Thadani, M C, NRR/DORL/LPL4, 415-1476
References
TAC ME0986
Download: ML093490797 (4)

Text

From: Thadani, Mohan Sent: Tuesday, December 15, 2009 10:25 AM To: Maglio, Scott A; Elwood, Thomas B Cc: Burkhardt, Janet; Lent, Susan

Subject:

DRAFT RAI FOR PROPOSED REVISION TO TS 5.5,16, "CONTAINMENT LEAKAGE RATE TESTING PROGRAM" me0986 Scott/Tom:

The NRC staff has reviewed the subject request, and has identified a need for additional information in order for the NRC staff to complete the detailed technical review of your application. Please review the request below and advise if you would like to discuss this request for additional information with the NRC staff.

Thanks.

Mohan C Thadani Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation (301) 415-1476 Mohan.Thadani@nrc.gov REQUEST FOR ADDITIONAL INFORMATION ON THE LICENSE AMENDMENT REQUESTS FOR REVISION TO TECHNICAL SPECIFICATION 5.5.16, "CONTAINMENT LEAKAGE RATE TESTING PROGRAM," FOR CALLAWAY PLANT, UNIT 1 (TAC NO. ME0986)

The following RAIs are related to Table 2.1.3 of Attachment 5 of the subject application.

1. Item 3 identifies credit for repair of hardware faults for certain initiator models without sufficient analysis and data, including repairs of CCFs. To support credit for repairs in the PRA model, the licensee must have identified repair rates and times for the specific components and failures for which the repair is credited. Please provide the non-recovery/non-repair probabilities applied in the PRA model and their bases. Also, please provide a sensitivity calculation for this application that takes no credit for the repair of hardware faults, including associated CCFs.

2. Items 6, 7, 8, 11, 12, 20, and 21 all identify apparent fundamental logic errors in the fault tree/event tree structure of the PRA model, including failure to properly treat dependencies, invalid placement of human error events in the logic, credit for systems which would not be available given the sequence (i.e., station blackout crediting main feedwater, loss of service water crediting instrument air). The dispositons state that correction of these items has been determined to increase CDF by about 1%. Please provide the basis for this conclusion, including exactly how the 1% increase was

determined. Describe how the extent of condition of these logic errors was investigated to ensure other instances do not exist in other places within the PRA. In addition, please revise the PRA to address these F&Os and provide revised results.

3. Item 9 addresses the use of an inaccurate reactor coolant pump seal LOCA model. The disposition states that core uncovery probabilities were increased by 25% resulting in a 1.5% increase in CDF. Typically, the seal LOCA model is used to determine the time to core uncovery, which is then used to estimate the offsite power recovery probability, and higher leak rate scenarios. Although low probability, these seal LOCAs tend to dominate the risk. Please provide the basis for selection of a 25% increase used in the sensitivity study, and how this is known to bound the seal LOCA nonconservatism. In addition, please revise the PRA to reflect the WOG 2000 model if Callaway has high temperature seals installed in all its pumps such that the WOG 2000 model is applicable; otherwise, use the conservative seal LOCA model accepted by the NRC (i.e., the Rhodes model) and provide revised results.

4. Item 10 F&O questions the validity of MAAP 3 for addressing the SGTR sequence with failure to isolate. Please provide the basis for the validity of MAAP 3 for addressing this sequence.

5. Item 10 states it is conservative to assume the SGTR sequence automatically goes to LERF. However for this application, conservatively assuming events result in LERF is non-conservative, since it masks the intact containment frequency, and reduces the delta LERF. Address this item for this application accounting for the non-conservative impact.

6. Item 14 identifies an improper treatment of data. The response indicates a recent update using the correct method per the standard was performed. It is not clear why this item is not therefore resolved if the data has been updated. Clarify this apparent inconsistency.

7. Items 19 and 1 (from Table 2.2.1 of Attachment 5 to ULNRC-05598) identify the failure to consider the "state of knowledge correlation". The disposition states that this only impacts the uncertainty analysis. This is fundamentally not true. The standard requires that quantification of CDF and LERF consider correlated data. This is especially significant for evaluation of ISLOCA, where the primary failure mode leading to overpressurization of low pressure piping involves coincident failure of two or more redundant identical isolation valves. Neglecting the data correlation has the potential to significantly underestimate the overall frequency of the event. Identifying that these events are not significant (when quantified with the non conservative error) does not justify that they would not become significant once the error is corrected. Please provide the basis for why this error is known to be insignificant, especially with regards to the interfacing LOCA contribution. In addition, please revise the PRA to specifically address these F&Os by including the state of knowledge correlation and provide revised results.

8. Item 23 indicates that key assumptions and key sources of uncertainty that influence the current quantification is not addressed in a coherent manner. The disposition indicates that this is solely a documentation issue with any basis for how the licensee determined that there were no assumptions or uncertainties that could impact this application.

Please provide a discussion of the key assumptions and key sources of uncertainty that could impact this application and how the licensee has addressed these key

assumptions and key sources of uncertainty (e.g., by conducting additional sensitivity studies) and as necessary, please provide any additional sensitivity study results.

9. Item 24 indicates that the licensee did not use the ASME definition of significant and the licensee dispositions this item as being solely a documentation issue. The staff disagrees that not including upwards of 7% of the results is a documentation issue.

Please provide revised results that meet the ASME definition of significant.

10. There are numerous B F&Os (significant and should be resolved by next update of PRA) and one A F&O (highly significant and should be resolved immediately) that remain open many years after the peer review and gap analysis. This is not consistent with the expectations of the peer review process and the staff. Please provide a schedule and commitment for the resolution of all remaining open F&Os, including any open C and D F&Os.

E-mail Properties Mail Envelope Properties (0A64B42AAA8FD4418CE1EB5240A6FED10818DA8BCA)

Subject:

DRAFT RAI FOR PROPOSED REVISION TO TS 5.5,16, "CONTAINMENT LEAKAGE RATE TESTING PROGRAM" me0986 Sent Date: 12/15/2009 10:24:56 AM Received Date: 12/15/2009 10:24:56 AM From: Thadani, Mohan Created By: Mohan.Thadani@nrc.gov Recipients:

SMaglio@ameren.com (Maglio, Scott A)

Tracking Status: None TElwood@ameren.com (Elwood, Thomas B)

Tracking Status: None Janet.Burkhardt@nrc.gov (Burkhardt, Janet)

Tracking Status: None Susan.Lent@nrc.gov (Lent, Susan)

Tracking Status: None Post Office:

HQCLSTR02.nrc.gov Files Size Date & Time MESSAGE 16243 12/15/2009 Options Expiration Date:

Priority: olImportanceNormal

ReplyRequested: False Return Notification: False Sensitivity: olNormal Recipients received: