ML101680595
| ML101680595 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/17/2010 |
| From: | Peter Bamford Plant Licensing Branch 1 |
| To: | Chernoff H Plant Licensing Branch 1 |
| Bamford, Peter J., NRR/DORL 415-2833 | |
| References | |
| TAC ME3587 | |
| Download: ML101680595 (5) | |
Text
June 17, 2010 MEMORANDUM TO: Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation FROM:
Peter Bamford, Project Manager /ra/
Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
SUBJECT:
THREE MILE ISLAND, UNIT NO. 1 - ELECTRONIC TRANSMISSION, DRAFT REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO ADOPT TSTF-425, RELOCATION OF SURVEILLANCE FREQUENCIES TO A LICENSEE CONTROLLED PROGRAM (TAC NO. ME3587)
The attached draft request for additional information (RAI) was transmitted by electronic transmission on June 17, 2010, to Mr. Glenn Stewart, at Exelon Generation Company, LLC (Exelon, the licensee). This draft RAI was transmitted to facilitate the technical review being conducted by the Nuclear Regulatory Commission (NRC) staff and to support a conference call (if needed) with Exelon in order to clarify the licensees amendment request regarding the proposed relocation of certain technical specification surveillance frequencies to a licensee-controlled program. The draft RAI is related to the licensees submittal dated March 24, 2010. The draft questions were sent to ensure that they were understandable, the regulatory basis was clear, and to determine if the information was previously docketed.
Additionally, review of the draft RAI would allow Exelon to evaluate and agree upon a schedule to respond to the RAI. This memorandum and the attachment do not represent an NRC staff position.
Docket Nos. 50-289
Enclosure:
As stated
June 17, 2010 MEMORANDUM TO: Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation FROM:
Peter Bamford, Project Manager /ra/
Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
SUBJECT:
THREE MILE ISLAND, UNIT NO. 1 - ELECTRONIC TRANSMISSION, DRAFT REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO ADOPT TSTF-425, RELOCATION OF SURVEILLANCE FREQUENCIES TO A LICENSEE CONTROLLED PROGRAM (TAC NO. ME3587)
The attached draft request for additional information (RAI) was transmitted by electronic transmission on June 17, 2010, to Mr. Glenn Stewart, at Exelon Generation Company, LLC (Exelon, the licensee). This draft RAI was transmitted to facilitate the technical review being conducted by the Nuclear Regulatory Commission (NRC) staff and to support a conference call (if needed) with Exelon in order to clarify the licensees amendment request regarding the proposed relocation of certain technical specification surveillance frequencies to a licensee-controlled program. The draft RAI is related to the licensees submittal dated March 24, 2010. The draft questions were sent to ensure that they were understandable, the regulatory basis was clear, and to determine if the information was previously docketed.
Additionally, review of the draft RAI would allow Exelon to evaluate and agree upon a schedule to respond to the RAI. This memorandum and the attachment do not represent an NRC staff position.
Docket Nos. 50-289
Enclosure:
As stated DISTRIBUTION:
Public RidsNrrPMThreeMileIsland Resource LPL1-2 R/F Accession No.: ML101680595
- via memo OFFICE LPL1-2/PM APLA/BC NAME PBamford DHarrison*
DATE 6/17/10 05/02/2010 OFFICIAL RECORD COPY
DRAFT Enclosure REQUEST FOR ADDITIONAL INFORMATION THREE MILE ISLAND NUCLEAR STATION, UNIT 1 PROPOSED RELOCATION OF SURVEILLANCE FREQUENCIES TO A LICENSEE CONTROLLED PROGRAM DOCKET NO. 50-289 By letter dated March 24, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100840205), Exelon Generation Company, LLC (the licensee), submitted a license amendment request (LAR) for Three Mile Island Nuclear Station, Unit 1 (TMI-1). The licensee proposes to modify the TMI-1 Technical Specifications (TSs) by relocating certain surveillance frequencies to a licensee-controlled program through the implementation of Nuclear Energy Institute (NEI) 04-10, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies. According to the licensee, the changes are consistent with U.S. Nuclear Regulatory Commission (NRC)-approved Technical Specifications Task Force (TSTF) Standard Technical Specifications (STSs) change TSTF-425, Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specifications Task Force Initiative 5b, Revision 3. In order for the NRC staff to complete its review of the LAR, a response to the following request for additional information (RAI) is requested.
RAI-1
The LAR states that the changes presented are consistent with TSTF-425 and also includes a discussion of the differences in the application that result primarily from the custom TMI-1 TSs as compared to the STSs presented in TSTF-425 and NUREG-1430., TSTF-425 (NUREG-1430) vs. TMI Unit 1 Cross-Reference, is provided to aid in the determination of consistency of the surveillances proposed for relocation as compared to TSTF-425. In order to verify that the surveillances proposed for relocation are consistent with TSTF-425 as the LAR asserts, the NRC staff requests that the licensee provide corresponding TSTF-425 cross references for the following surveillances proposed for relocation: Table 4.1-1, Instrument Surveillance Requirements, Channel Description Nos. 11, 15, 17, 19e, 19f, 45, and 46.
RAI-2
With reference to Attachment 2, Table 2-1, each of the findings in the following table identified an issue or gap that, while individually might not significantly impact the results from an surveillance test interval (STI) risk evaluation performed via the NEI 04-10 methodology, when taken cumulatively could prove significant. The NRC staffs concern associated with each is highlighted in italics. Please address whether, when taken cumulatively, their effects could prove significant to the risk evaluation for an STI TS change and, if not, why not.
DRAFT Finding Issue/Gap IE-A4a-01 The potential for common cause failures [CCFs] was included in examination of potential initiating events resulting from the systematic evaluation for potential initiating events. As recommended per
[Regulatory Guide] RG 1.200, Rev. 2, for Supporting Requirement (SR) IE-A6 (Capability Category [(CC)]-II), this examination should also include CCFs from routine system alignments that could result from preventive and corrective maintenance.
IE-A5-01 No documentation was found of incorporating: (a) events that have occurred at conditions other than at-power operation (i.e., during low-power or shutdown conditions), and for which it is determined that the event could also occur during at-power operation; (b) events resulting in a controlled shutdown that includes a scram prior to reaching low-power conditions, unless it is determined that an event is not applicable to at-power operation. SR IE-A7 requires that, even if not documented, these events have to be incorporated.
IE-A6-01 and IE-A7-01 No documentation was found of interviews with plant personnel (e.g.,
operations, maintenance, engineering, safety analysis) to determine if potential initiating events have been overlooked... No documentation of the review of plant-specific operating experience for initiating event precursors was found in the [probabilistic risk assessment] PRA notebooks. Even if not documented, CC-II for both of these SRs requires that the interviews (SR IE-A8 [CC-II], with finding IE-A6-01) and reviews (SR IE-A9 [CC-II], with finding IE-A7-01) have been conducted.
SC-C2-01 SR SC-C2 requires that, even if not documented (or else still in the process of being documented), computer code limitations or potential conservatisms have to be addressed.
QU-D5-01 Some SSCs that are significant contributors to initiating events, but not to mitigation, are not explicitly identified in the documentation of significant contributors. SR QU-D5 requires that at least a sampling of non-significant contributors (accident sequences and cutsets) have been reviewed, implying that the identification of these potentially significant contributors is necessary.
QU-F5-01
[O]ther than the [large early release frequency] LERF truncation limitation, no evaluations of limitations were presented..., [including]
limitations of the model as they may apply to applications. As implied by SR QU-F5, these limitations need to have been addressed.
LE-C8a-01 The Reactor Building fan coolers are undersized at TMI and have a little to no impact on containment [CNMT] pressure and temperature with respect to early containment failure. SR LE-C9 (CC-II) requires justification for any credit taken for equipment survivability under adverse environmental conditions such that, even if the fan coolers were assumed to be failed, there would be "little to no impact" on CNMT pressure and temperature with respect to early CNMT failure.
DRAFT Finding Issue/Gap LE-E4-01 The level 2 results with the flag file are expected to be conservative.
When the cutsets were reviewed, it was determined that there appears to be non-minimal cutsets in the level 2 model as quantified without the flag file... Some sensitivities have been performed, although a conclusive determination has not been made regarding the current method for quantifying LERF... ([T]he TMI model uses Forte 3.0c as the quantifier). SR LE-E4 requires that LERF be quantified consistently as with core damage frequency. This implies that the LERF quantification be conclusively determined as conservative, e.g.,
by quantifying LERF using Forte 3.0c at a greater truncation value just to assess whether the use of the flag file produces conservative results.
RAI-3
With reference to Attachment 2, Table 2-2, Finding DA-B2-01 states: There is no evidence that the intent of this SR was met. Although the component failure rates are grouped by system and component type, that does not guarantee that outliers are not included in a group. SR DA-B2 (CC-II) requires exclusion of outliers in the definition of system/component failure groups. Were outliers appropriately excluded from group definitions? If not, will their exclusion be part of the sensitivity analysis for an STI evaluation?
RAI-4
With reference to Attachment 2, Table 2-2, Finding IFEV-A5-01 states: Several requirements in establishing flood initiating event frequencies are not met. Specifically cited are SRs IFEV-A5 through IFEV-A7, which require inclusion of plant-specific information and consideration of human-induced floods during maintenance (CC-II). Are any of the valves that may be assigned new STIs potential flooding sources, such that increasing the STI could increase the frequency of a flood due to miscalibration, etc., of one of these valves?