ML070710050
| ML070710050 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 04/04/2007 |
| From: | Veronica Rodriguez NRC/NRR/ADRO/DLR/RLRB |
| To: | Garrett T Wolf Creek |
| rodriguez v m, ADRO/DLR/RLRB, 415-3703 | |
| References | |
| Download: ML070710050 (7) | |
Text
April 4, 2007 Mr. Terry J. Garrett Vice President Engineering Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE WOLF CREEK GENERATING STATION, UNIT 1, LICENSE RENEWAL APPLICATION
Dear Mr. Garrett:
By letter dated September 27, 2006, Wolf Creek Nuclear Operating Corporation submitted an application pursuant to 10 CFR Part 54, to renew the operating license for Wolf Creek Generating Station, Unit 1, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review.
These requests for additional information were discussed with Lorrie Bell, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-3703 or e-mail VMR1@nrc.gov.
Sincerely,
/RA/
Verónica M. Rodríguez, Project Manager License Renewal Branch B Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-482
Enclosure:
Requests for Additional Information cc w/encl: See next page
ML070710050 OFFICE LA:DLR PM:RLRB:DLR BC:RLRB:DLR NAME IKing VRodríguez RAuluck DATE 03/15/07 04/4/07 04/4/07
Enclosure WOLF CREEK GENERATING STATION, UNIT 1 LICENSE RENEWAL APPLICATION (LRA)
REQUEST FOR ADDITIONAL INFORMATION (RAI)
RAI 2.1-1 LRA Section 2.1.2.1 states that safety-related classifications for systems and structures at Wolf Creek Generating Station (WCGS), are reported in the updated safety analysis report (USAR) or in design basis documents such as engineering drawings, evaluations, or calculations.
Safety-related classifications for components are documented on engineering drawings and in the WCGS Q-List. The safety-related classification as reported in these source documents has been relied upon to identify SSCs satisfying one or more of the criteria of 10 CFR 54.4(a)(1).
These SSCs have been identified as within the scope of license renewal.
However, during the audit the staff noted that source documents, such as USAR Section 3.2, and procedures AP 05-007, Section 6.1.4, and AP 23M-001, Section 4.17.1, have differing definitions for the term safety-related. In addition, these documents currently cite superseded regulatory text for establishing the scoping criteria to be used in identifying SSCs in accordance with the requirements of 10 CFR 54.4(a)(1).
The staff requests that the applicant addresses the impact, if any, of the use of differing definition of safety-related. In addition, the applicant is requested to address the impact of not having considered these different definitions in its scoping methodology for those SSCs that are relied upon to ensure the capability to prevent or mitigate the consequences of accidents that could result in potential off-site exposures comparable to the guidelines in 10 CFR Sections 50.34(a)(1), 50.67(b)(2), or 100.11 of this chapter, as applicable, consistent with the facilitys current licensing basis.
RAI 2.1-2 NRC Regulatory Guide (RG) 1.188, Standard Format and Content for Applications to Renew Nuclear Power Plant Operating Licenses, Revision 1, dated September 2005, provides endorsement on the use of Nuclear Energy Institute (NEI) 95-10, Industry Guidelines for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule,Revision 6, dated June 2005. RG 1.188 indicates that NEI 95 -10, Revision 6, provides methods that the staff considers acceptable for complying with the requirements of 10 CFR Part 54 for preparing an LRA.
NEI 95-10, Appendix F, states in part, that nonsafety-related SSCs that are not directly connected to safety-related SSCs, or that are connected downstream of the first equivalent anchor, may be within the scope of license renewal if their failure could prevent the performance of the system safety function for which the safety-related SSC is required.
NEI 95-10, Appendix F, describes two options for determining which nonsafety-related SSCs may be withing the scope of license renewal. The applicant's methodology for scoping of nonsafety-related components affecting safety-related components is briefly described in LRA Section 2.1.2.2.
The staff requests that the applicant provide the following information related to its evaluation of the 10 CFR 54.4(a)(2) criteria:
(1)
Explain which option (mitigative or preventive) was used for nonsafety-related SSCs not directly connected to safety-related SSCs. In addition, describe the process for scoping portions of nonsafety-related systems in rooms or building, level, or areas (BLA) that contain safety-related components.
(2)
Define a room or BLA as used to determine the location of safety-related equipment. In addition, explain how portions of nonsafety-related systems were scoped for spatial interaction.
(3)
Discuss how interactions between adjacent rooms and/or BLAs were evaluated for the purposes of 10 CFR 54.4(a)(2). For example, describe how the effects of a pipe break in a room that may not contain a safety-related component were evaluated for interaction with an adjacent room that may contain a safety-related component.
(4)
During the audit, the applicant indicated that portions of piping systems containing an insignificant amount of liquid that would not typically be replenished, such as small isolated drain lines, were not considered to be fluid filled, and as such, were not included within the scope of 10 CFR 54.4(a)(2).
Provide the technical justification and extent of condition for the exclusion of such portions of systems from the scope of license renewal.
(5)
Similarly, roof drain piping was also considered not to be filled with fluid.
However, the applicant has identified that some roof drain piping passes through portions of the auxiliary, control, and diesel generator buildings where safety-related equipment may be located. During the audit, the applicant stated that the design and installation of roof drain piping precludes spatial interaction concerns. Provide the technical justification and extent of condition for the exclusion of the roof drain piping that passes through the portions of the above buildings where safety-related equipment is located.
RAI 2.1-3 During the audit, the applicant stated that it applied the same justification provided in NEI 95-10, Appendix F, Section 5.2.2.3 to exclude the insulation section from the scope of 10 CFR 54.4(a)(2) due to the physical impact hazard. The insulation is supported by insulation supports that are designed to withstand a seismic event for piping and equipment that are seismic designed, or that are classified as seismic II/I. However, the applicant did not indicate whether the insulation supports are within the scope of license renewal. As stated in NEI 95-10, Appendix F, piping supports for seismic II/I piping need to be intact in order to prevent physical impacts on safety-related equipment during a seismic event and as a result must be included within the scope of license renewal.
The staff requests that the applicant explain how the insulation supports that are designed to withstand a seismic event and that are located in areas containing safety-related equipment were reviewed for inclusion within the scope of license renewal.
RAI 3.0.4-1 The staff reviewed the applicant's aging management programs (AMPs) as described in LRA Sections A and B. In addition, the staff reviewed each individual AMP basis document to ensure consistency in the use of the quality assurance attributes for each program. The purpose of this review was to assure that the aging management activities were consistent with the staff's guidance described in the Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (SRP-LR), Section A.2, Quality Assurance for Aging Management Programs (Branch Technical Position IQMB-1).
Based on its evaluation of the descriptions and applicability of the plant-specific AMPs and their associated quality attributes provided in LRA Section B.1.3, the staff finds that the quality assurance attributes are generally consistent with the NRC position regarding quality assurance for aging management. However, the applicant has not sufficiently described the use of the quality assurance program and its associated attributes (corrective action, confirmation process, and administrative controls) in LRA Section A. In addition, AMPs discussed in LRA Sections B.2.1.1 and B.2.1.3 address exceptions taken to the corrective actions program element. However, there is no indication or description of the use of an alternative method to the WCGS 10 CFR Part 50, Appendix B, quality assurance program being applied to the area of corrective action.
The staff requests that the applicant provide the following information to address these issues:
(1)
A supplement to the description in LRA Section A.1 to clearly indicate the application of the WCGS 10 CFR Part 50, Appendix B, quality assurance program, or an alternative, for the corrective action, confirmation process, and administrative control attributes in each program, applicable to nonsafety-related and safety-related structures and components (SCs) during the period of extended operation. If any alternative approaches are identified, provide a detailed description such that the staff can determine if the quality attributes for the AMPs are consistent with the review acceptance criteria contained in SRP-LR, Section A.2.
(2)
As described in LRA Section B, for each AMP that take exceptions in the area of corrective action, confirmation process, and administrative controls, indicate whether the exceptions include an alternative to the application of the WCGS 10 CFR Part 50, Appendix B, quality assurance program as described in LRA Section B.1.3. If alternative approaches are identified, provide a detailed description such that the staff can determine if the quality attributes for the AMPs are consistent with the review acceptance criteria contained in SRP-LR, Section A.2.
Letter to T. Garrett from V. Rodriguez dated April 4, 2007 DISTRIBUTION:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE WOLF CREEK GENERATING STATION, UNIT 1, LICENSE RENEWAL APPLICATION HARD COPY:
DLR RF E-MAIL:
PUBLIC RWeisman GGalletti SSmith (srs3)
SDuraiswamy RidsNrrDlr RidsNrrDlrRlra RidsNrrDlrRlrb RidsNrrDlrRlrc RidsNrrDlrReba RidsNrrDlrRebb RidsNrrDci RidsNrrDra RidsNrrDe RidsNrrDeEemb RidsNrrDeEeeb RidsNrrDeEqvb RidsNrrDss RidsOgcMailCenter RidsNrrAdes VRodriguez CJacobs JDonohew CLong, RIV GPick, RIV SCochrum, RIV DReddy KGreen BRogers GGalletti
Wolf Creek Generating Station cc:
Jay Silberg, Esq.
Pillsbury Winthrop Shaw Pittman, LLP 2300 N Street, NW Washington, DC 20037 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-7005 Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 311 Burlington, KS 66839 Chief Engineer, Utilities Division Kansas Corporation Commission 1500 SW Arrowhead Road Topeka, KS 66604-4027 Office of the Governor State of Kansas Topeka, KS 66612 Attorney General 120 S.W. 10th Avenue, 2nd Floor Topeka, KS 66612-1597 County Clerk Coffey County Courthouse 110 South 6th Street Burlington, KS 66839 Thomas A. Conley, Section Chief Radiation and Asbestos Control Kansas Department of Health and Environment Bureau of Air and Radiation 1000 SW Jackson, Suite 310 Topeka, KS 66612-1366 Vice President Operations/Plant Manager Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839 Supervisor Licensing Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839 U.S. Nuclear Regulatory Commission Resident Inspectors Office/Callaway Plant 8201 NRC Road Steedman, MO 65077-1032 Kevin J. Moles, Manager Regulatory Affairs Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839 Lorrie I. Bell, Project Manager Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839 Mr. James Ross Nuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708