ML071220191
| ML071220191 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 05/07/2007 |
| From: | Veronica Rodriguez NRC/NRR/ADRO/DLR/RLRB |
| To: | Wolf Creek |
| rodriguez v m, ADRO/DLR/RLRB, 415-3703 | |
| References | |
| Download: ML071220191 (10) | |
Text
May 7, 2007LICENSEE: Wolf Creek Nuclear Operating CorporationFACILITY: Wolf Creek Generating Station, Unit 1
SUBJECT:
SUMMARY
OF TELEPHONE CONFERENCE CALL HELD ON MARCH 7, 2007,BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND WOLF CREEK NUCLEAR OPERATING CORPORATION, CONCERNING DRAFT REQUESTS FOR ADDITIONAL INFORMATION PERTAINING TO THE WOLF CREEK GENERATING STATION, UNIT 1, LICENSE RENEWAL APPLICATIONThe U.S. Nuclear Regulatory Commission (NRC or the staff) and representatives of Wolf CreekNuclear Operating Corporation held a telephone conference call on March 7, 2007, to discuss and clarify the staff's draft requests for additional information (D-RAIs) concerning the Wolf Creek Generating Station, Unit 1, license renewal application. The telephone conference call was useful in clarifying the intent of the staff's D-RAIs.Enclosure 1 provides a listing of the participants and Enclosure 2 contains a listing of theD-RAIs discussed with the applicant, including a brief description on the status of the items.The applicant had an opportunity to comment on this summary./RA/ RA/ Verónica M. Rodríguez, Project Manager License Renewal Branch B Division of License Renewal Office of Nuclear Reactor RegulationDocket No. 50-482
Enclosures:
- 1. List of Participants
- 2. List of Draft Requests for Additional Informationcc w/encls: See next page
ML071220191OFFICEPM:RLRB:DLRLA:DLRBC:RLRB:DLRNAMEVRodríguezSFigueroaRAuluck DATE05/04/0705/04/0705/07/07 TELEPHONE CONFERENCE CALLWOLF CREEK GENERATING STATION, UNIT 1LICENSE RENEWAL APPLICATIONLIST OF PARTICIPANTSMARCH 7, 2007PARTICIPANTSAFFILIATIONSVerónica M. RodríguezU.S. Nuclear Regulatory Commission (NRC)Devender ReddyNRC Billy RogersNRC Lorrie BellWolf Creek Nuclear Operating Corporation (WCNOC)
Charlie MedenciyWCNOC Rick FoustWCNOC Ron TraudtWCNOC Eric BlocherStrategic Teaming and Resource Sharing Alliance (STARS)
Paul CrawleySTARS Arden AldridgeSTARS DRAFT REQUESTS FOR ADDITIONAL INFORMATIONWOLF CREEK GENERATING STATION, UNIT 1LICENSE RENEWAL APPLICATIONMARCH 7, 2007The U.S. Nuclear Regulatory Commission (NRC or the staff) and representatives of Wolf CreekNuclear Operating Corporation held a telephone conference call on March 7, 2007, to discuss and clarify the following draft requests for additional information (D-RAIs) concerning the Wolf Creek Generating Station (WCGS), Unit 1, license renewal application (LRA). Scoping and Screening MethodologyD-RAI 2.1-1LRA Section 2.1.2.1 states that safety-related classifications for systems and structures atWCGS, 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 structure, system, and components (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 address the impact, if any, of the use of differing definitionof 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 facility's current licensing basis. Discussion: The applicant indicated that the question is clear. This D-RAI will be sent as aformal RAI. D-RAI 2.1-2NRC Regulatory Guide (RG) 1.188, "Standard Format and Content for Applications to RenewNuclear 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 directlyconnected 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 within 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 ofthe 10 CFR 54.4(a)(2) criteria: (1)Explain which option (mitigative or preventive) was used for nonsafety-related SSCs notdirectly 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 thepurposes 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 aninsignificant 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, theapplicant 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. Discussion: The applicant indicated that the question is clear. This D-RAI will be sent as aformal RAI. D-RAI 2.1-3During 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 insulationsupports 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 indicatewhether 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 toprevent 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 towithstand a seismic event and that are located in areas containing safety-related equipment were reviewed for inclusion within the scope of license renewal.Discussion: The applicant indicated that the question is clear. This D-RAI will be sent as aformal RAI. Quality Assurance ProgramD-RAI 3.0.4-1The 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 theirassociated 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 ofthe 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 ofcorrective 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 ADiscussion: The staff requested that the applicant evaluate whether this is an exception to theprogram, as stated in the LRA, or an enhancement. The staff stated that if this is an exception,an alternative to the 10 CFR Part 50, Appendix B, quality assurance program, must be provided. The applicant indicated that the question is clear. This D-RAI will be sent as a formal RAI.
Note to Wolf Creak Nuclear Operating Corp. from V. Rodriguez dated May 7, 2007DISTRIBUTION:
SUBJECT:
SUMMARY
OF TELEPHONE CONFERENCE CALL HELD ON MARCH 7,2007, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND WOLF CREEK NUCLEAR OPERATING CORPORATION, CONCERNING DRAFT REQUESTS FOR ADDITIONAL INFORMATION PERTAINING TO THE WOLF CREEK GENERATING STATION, UNIT 1, LICENSE RENEWAL APPLICATIONHARD COPY DLR RF E-MAIL:PUBLICRWeisman SSmith (srs3)
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CJacobs JDonohew DReddy BRogers KGreen GPick, RIV SCochrum, RIV CLong, RIV Wolf Creek Generating Station cc:Jay Silberg, Esq.Pillsbury Winthrop Shaw Pittman, LLP 2300 N Street, NW Washington, DC 20037Regional Administrator, Region IVU.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-7005Senior Resident InspectorU.S. Nuclear Regulatory Commission P.O. Box 311 Burlington, KS 66839Chief Engineer, Utilities DivisionKansas Corporation Commission 1500 SW Arrowhead Road Topeka, KS 66604-4027Office of the GovernorState of Kansas Topeka, KS 66612Attorney General120 S.W. 10 th Avenue, 2 nd FloorTopeka, KS 66612-1597County ClerkCoffey County Courthouse 110 South 6 th StreetBurlington, KS 66839Thomas A. Conley, Section ChiefRadiation and Asbestos Control Kansas Department of Health and Environment Bureau of Air and Radiation 1000 SW Jackson, Suite 310 Topeka, KS 66612-1366Vice President Operations/Plant ManagerWolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839Supervisor LicensingWolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839U.S. Nuclear Regulatory CommissionResident Inspectors Office/Callaway Plant
8201 NRC Road Steedman, MO 65077-1032Kevin J. Moles, ManagerRegulatory Affairs Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839Lorrie I. Bell, Project ManagerWolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839Mr. James RossNuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708