ML060190605
| ML060190605 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/15/2006 |
| From: | Milano P D Plant Licensing Branch III-2 |
| To: | Korsnick M G Ginna |
| Milano P, NRR/DLPM 415-1457 | |
| References | |
| TAC MC6867 | |
| Download: ML060190605 (15) | |
Text
February 15, 2006Mrs. Mary G. KorsnickVice President R.E. Ginna Nuclear Power Plant R.E. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519
SUBJECT:
R.E. GINNA NUCLEAR POWER PLANT - AMENDMENT RE: RELAXED AXIALOFFSET CONTROL (TAC NO. MC6867)
Dear Mrs. Korsnick:
The Commission has issued the enclosed Amendment No. 94 to Renewed Facility OperatingLicense No. DPR-18 for the R.E. Ginna Nuclear Power Plant. This amendment is in response to your application dated April 29, 2005, as supplemented on September 19, 2005.The amendment revises the Technical Specifications to incorporate the relaxed axial offsetcontrol and heat flux hot channel (F Q) surveillance methodologies. These methodologies areused to reduce operator action required to maintain conformance with power distribution control requirements and to increase the ability to return to power after a plant trip or transient. Thechanges are consistent with Westinghouse Electric Company Report WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control/F Q Surveillance Technical Specification."A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be incl udedin the Commission's biweekly Federal Register notice. Sincerely,/RA/Patrick D. Milano, Sr. Project ManagerPlant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-244
Enclosures:
- 1. Amendment No. 94 to Renewed License No. DPR-18 2. Safety Evaluationcc w/encls: See next page February 15, 2006Mrs. Mary G. KorsnickVice President R. E. Ginna Nuclear Power Plant R.E. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519
SUBJECT:
R.E. GINNA NUCLEAR POWER PLANT - AMENDMENT RE: RELAXED AXIALOFFSET CONTROL (TAC NO. MC6867)
Dear Mrs. Korsnick:
The Commission has issued the enclosed Amendment No. 94 to Renewed Facility OperatingLicense No. DPR-18 for the R.E. Ginna Nuclear Power Plant. This amendment is in response to your application dated April 29, 2005, as supplemented on September 19, 2005.The amendment revises the Technical Specifications to incorporate the relaxed axial offsetcontrol and heat flux hot channel (F Q) surveillance methodologies. These methodologies areused to reduce operator action required to maintain conformance with power distribution control requirements and to increase the ability to return to power after a plant trip or transient. Thechanges are consistent with Westinghouse Electric Company Report WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control/F Q Surveillance Technical Specification."A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be incl udedin the Commission's biweekly Federal Register notice. Sincerely,/RA/Patrick D. Milano, Sr. Project ManagerPlant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-244
Enclosures:
- 1. Amendment No. 94 to Renewed License No. DPR-18 2. Safety Evaluationcc w/encls: See next page Accession Number: ML060190605OFFICELPLI-1\PMLPLI-1\LASPWB\BCOGCLPLI-1\BCNAMEPMilanoSLittleJNakoskiRLauferDATE01/26/0601/26/0601/27/0602/09/0602/14/06Official Record Copy DATED: February 15, 2006AMENDMENT NO. 94 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 R.E. GINNA NUCLEAR POWER PLANTPUBLICLPLI-1 R/F R. LauferRidsNrrDorlLpla S. LittleRidsNrrLASLittle P. MilanoRidsNrrPMPMilano J. NakoskiRidsNrrDssSpwb F. AkstulewiczRidsNrrDssSpnb T. BoyceRidsNrrDirsItsb K. Wood G. Hill (2)
OGC RidsOgcMailCenter ACRSRidsAcrsAcnwMailCenter J. Trapp, RIcc: Plant Service list R.E. Ginna Nuclear Power Plant cc:
Mr. Michael J. WallacePresident R.E. Ginna Nuclear Power Plant, LLC c/o Constellation Energy 750 East Pratt Street Baltimore, MD 21202Mr. John M. HeffleySenior Vice President and Chief Nuclear Officer Constellation Generation Group 1997 Annapolis Exchange Parkway Suite 500 Annapolis, MD 21401Kenneth Kolaczyk, Sr. Resident InspectorR.E. Ginna Nuclear Power Plant U.S. Nuclear Regulatory Commission 1503 Lake Road Ontario, NY 14519Regional Administrator, Region IU.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406Mr. Peter R. Smith, PresidentNew York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399Mr. Carey W. Fleming, EsquireSenior Counsel - Nuclear Generation Constellation Generation Group, LLC 750 East Pratt Street, 17th Floor Baltimore, MD 21202Mr. Charles Donaldson, EsquireAssistant Attorney General New York Department of Law 120 Broadway New York, NY 10271Ms. Thelma Wideman, DirectorWayne County Emergency Management Office Wayne County Emergency Operations Center 7336 Route 31 Lyons, NY 14489Ms. Mary Louise MeisenzahlAdministrator, Monroe County Office of Emergency Preparedness 1190 Scottsville Road, Suite 200Rochester, NY 14624Mr. Paul EddyNew York State Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223 R.E. GINNA NUCLEAR POWER PLANT, LLCDOCKET NO. 50-244R.E. GINNA NUCLEAR POWER PLANTAMENDMENT TO RENEWED FACILITY OPERATING LICENSEAmendment No. 94 Renewed License No. DPR-181.The Nuclear Regulatory Commission (the Commission or the NRC) has f ound that:A.The application for amendment filed by the R.E. Ginna Nuclear Power Plant, LLC (the licensee) dated April 29, 2005, as supplemented on September 19, 2005,complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;C.There is reasonable assurance: (i) that the activities authorized by thisamendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with theCommission's regulations;D.The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public; andE.The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied. 2.Accordingly, the license is amended by changes to the Technical Specifications asindicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-18 is hereby amended to read as follows: (2)Technical SpecificationsThe Technical Specifications contained in Appendix A, as revised throughAmendment No. 94, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.3.This license amendment is effective as of the date of its issuance and shall beimplemented prior to startup from the fall 2006 refueling outage. FOR THE NUCLEAR REGULATORY COMMISSION/RA/Richard J. Laufer, ChiefPlant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical SpecificationsDate of Issuance: February 15, 2006 ATTACHMENT TO LICENSE AMENDMENT NO. 94RENEWED FACILITY OPERATING LICENSE NO. DPR-18DOCKET NO. 50-244Replace the following pages of the Appendix A Technical Specifications with the attachedrevised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.Remove Insert3.2.1-13.2.1-13.2.1-23.2.1-2 3.2.1-33.2.1-3 3.2.1-43.2.1-4 3.2.3-13.2.3-1 3.2.4-13.2.4-1 3.2.4-23.2.4-2 3.2.4-33.2.4-3 3.3.1-13.3.1-1 3.3.1-23.3.1-2 3.3.1-33.3.1-3 3.3.1-43.3.1-4 3.3.1-53.3.1-5 3.3.1-63.3.1-6 3.3.1-73.3.1-7 3.3.1-83.3.1-8 3.3.1-93.3.1-9 3.3.1-103.3.1-10 3.3.1-113.3.1-11 3.3.1-123.3.1-12 3.3.1-133.3.1-13 3.3.1-143.3.1-14 3.3.1-153.3.1-15 3.3.1-163.3.1-16 5.6-15.6-1 5.6-25.6-2 5.6-35.6-3 5.6-45.6-4 5.6-55.6-5 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATED TO AMENDMENT NO. 94 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18R.E. GINNA NUCLEAR POWER PLANT, INC.R.E. GINNA NUCLEAR POWER PLANTDOCKET NO. 50-24
41.0 INTRODUCTION
By letter dated April 29, 2005, as supplemented on September 19, 2005 (References 1 and 2;Agencywide Documents Access and Management System Accession Nos. ML051300330 and ML052720369), R.E. Ginna Nuclear Power Plant, LLC. (Ginna LLC, or the licensee) submitted a request for changes to the R.E. Ginna Nuclear Power Plant (Ginna) Technical Specifications (TSs). The requested changes would incorporate relaxed axial offset control (RAOC) and heat flux hot channel (F Q) surveillance methodologies. These methodologies are used to reduceoperator action required to maintain conformance with power distribution control requirements and to increase the ability to return to power after a plant trip or transient. The RAOCmethodology allows the axial flux difference control bands to be widened by removing conservatism inherent in the constant axial offset control (CAOC) methodology currently usedat Ginna. The proposed changes are in accordance with Westinghouse Electric Company Report WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control/F Q SurveillanceTechnical Specification" (Reference 3). The requested changes would also move requirements for the axial flux difference (AFD) and quadrant power tilt ratio (QPTR) monitor alarms to the Technical Requirements Manual (TRM). The September 19, 2005, letter provided additional information that clarified the application, did not expand the scope of the application asoriginally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on June 7, 2005(70 FR 33220)
2.0 REGULATORY EVALUATION
2.1 Power
Distribution LimitsDuring power operation, the global power distribution is monitored by the AFD and the quadrantpower tilt ratio QPTR, which are directly and continuously measured process variables. Local power peaking is monitored by F Q(z), and nuclear enthalpy rise hot-channel factor, (F NH). Limits on these parameters ensures that the nuclear core limits are not exceeded. AFD is a measure of the balance of power between the top and bottom halves of the core. Limits are established on the values of the AFD in order to limit the axial power distribution skewing to either the top or bottom of the core. By limiting the amount of power distributionskewing, core peaking factors are consistent with the assumptions used in the safety analyses.
Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which is a significant factor in axial power distribution control.QPTR is a measure of the balance of power between the radial quadrants of the core. Limitsare established on the values of the QPTR in order to limit the radial power distribution skewing to a quadrant. By limiting the amount of power distribution skewing, core peaking factors are consistent with the assumptions used in the safety analyses. Limiting the QPTR prevents radial xenon oscillations and will indicate any core asymmetries.
F Q(z) is a measure of the local power. The value of F Q(z) varies along the axial height of thecore (z). F Q(z) is defined as the maximum local fuel rod linear power density divided by theaverage fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions adjusted for uncertainty. Therefore, F Q(z) is a measure of the peak pellet power within thereactor core.
F NH is a measure of the maximum total power produced in a fuel rod. F NH is defined as theratio of the integral of the linear power along the fuel rod with the highest integrated power tothe average integrated fuel rod power. 2.2 Current CAOC MethodologyThe current TSs stipulate the use of the CAOC methodology for analyzing and controlling theAFD parameter. In the power range, CAOC methodology calls for a constant band about areference AFD, without regard to reactor power. In the power range, RAOC methodology generates a power dependent AFD curve. Generally, the RAOC allowed AFD increases with decreasing power. In some plants, the CAOC methodology results in significant margin between the CAOC generated AFD and the maximum permitted AFD in the loss-of-coolant accident (LOCA) analysis. RAOC methodology allows a licensee to transfer some of that margin to operating flexibility.2.3 Regulatory RequirementsThe Nuclear Regulatory Commission (NRC) staff based its acceptance on the followingregulatory requirements:Appendix A, "General Design Criteria (GDC) for Nuclear Power Plants," to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR) establishes the minimum requirements for theprincipal design criteria for water-cooled nuclear power plants. In particular, GDC 10, "Reactor Design," states that "[t]he reactor core and associated coolant, control, and protection systemsshall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipatedoperational occurrences." As noted in the Ginna Updated Final Safety Analysis Report (UFSAR), the design criteria used at Ginna predate Appendix A to 10 CFR Part 50. The Ginna design criteria comprise the proposed Atomic Industrial Forum (AIF) versions of the criteria issued for comment by the Atomic Energy Commission on July 10, 1967. These criteria define or describe safety objectives and approaches incorporated in the design of Ginna.In February 1978, the NRC initiated its Systematic Evaluation Program (SEP) for elevenoperating plants, including Ginna, that had received construction permits between 1956 and 1967. The SEP consisted of a limited review of the designs of these older plants. The purpose of the SEP was to reconfirm and document the design safety because the safety criteria had changed since the plants were originally licensed. As part of the SEP, the original codes and standards used in the design of structures, systems, and components at Ginna were comparedwith later licensing criteria. The results for Ginna were documented in NRC Report NUREG-0821, "Integrated Plant Safety Assessment, Systematic Evaluation Program, R.E. Ginna Nuclear Power Plant, Final Report," December 1982 (ADAMS No. 8309200476). The current UFSAR incorporates the SEP review into the current licensing basis at Ginna. The adequacy of the Ginna design relative to the 1972 version of the GDC is described in Section 3.1.2 of the UFSAR. Specifically, the conformance to GDC 10 at Ginna is described in UFSAR Section 3.1.2.2.1. Therefore, the NRC staff reviewed the application and based its acceptanceon the design criteria in GDC 10 rather than the criteria in the 1967 version of the AIF criteria.The NRC approved Westinghouse Report WCAP-10216-P-A, Revision 1, on November 26,1993, (Reference 4) for referencing in license applications to the extent specified and under the limitations stated in the topical report and NRC evaluation. Since the NRC staff f ound thereport acceptable, the purpose of its review of the licensee's April 29, 2005, application was toensure that WCAP-10216-P-A was applicable to Ginna and that WCAP-10216-P-A was being implemented appropriately.The regulatory requirements related to the content of the TSs are set forth in Section 50.36,"Technical specifications," of 10 CFR Part 50. This regulation requires that the TSs include items in five specific categories. These categories are: (1) safety limits, limiting safety systemsettings and limiting control settings, (2) limiting conditions for operation (LCOs),
(3) surveillance requirements (SRs), (4) design features, and (5) administrative controls.
3.0 TECHNICAL EVALUATION
The NRC staff has reviewed the licensee's regulatory and technical analyses in support of itsproposed license amendment, which are described in the licensee's April 29, 2005, application, as supplemented on September 19, 2005.3.1 Proposed TS Changesa. The licensee proposed to revise TS 3.2.1, "Heat Flux Hot Channel Factor FQ(Z),"TS 3.2.3, "Axial Flux Difference," and TS 5.6.5, "Core Operating Limits Report (COLR)," to adopt the RAOC calculational procedure as described in NUREG-1431, "StandardWestinghouse Technical Specifications Westinghouse Plants."b.The licensee proposed changes to TS 3.2.4, "Quadrant Power Tilt Ratio (QPTR)" toprovide the necessary consistency with the changes made to TS 3.2.1 and TS 3.2.3.c.The licensee proposed changes to TS 3.3.1, "Reactor Trip System (RTS)Instrumentation," to accommodate the change to the RAOC methodologies by revisingthe form of the f(I) penalty (f(I) is a function of the indicated difference between the top and bottom detectors of the power range neutron flux channels) for overtemperaturedelta temperature (T) and for overpower T (initially set to 0).d.The licensee proposed to relocated the requirements for the QPTR alarm monitor from TS 3.2.4 to the TRM.e.The licensee proposed to relocated the requirements for the AFD alarm monitor from TS 3.2.3 to the TRM.f. The licensee proposed to revise the listing of analysis methodologies contained in TS5.6.5, "Core Operating Limits Report (COLR)," to include refences for the RAOC methodology.3.2 RAOC/F Q SurveillencesThe NRC's approval of Westinghouse Report WCAP-10216-P-A was limited to pressurized-water reactors designed by Westinghouse. Ginna meets these applicability criteria. The NRCapproval did not impose any additional limitations beyond those already present in WCAP-10216-P-A. WCAP-10216-P-A is comprised of two parts. Part A covers the relaxation of constant axial offset control. Part B covers the F Q (heat flux hot channel factor) TSsurveillance requirements.Part A describes a method for determining an acceptable AFD profile. AFD is a measure of thepower differential between the top and bottom halves of the core. If power is shifted too muchin either direction it can exacerbate local peaking, resulting in challenges to fuel integrity. The RAOC method generates a range of AFD shapes at different power levels. Those shapes are used in LOCA analysis, loss of flow accident analysis, anticipated operational occurrence analysis. Unacceptable shapes are eliminated by each subsequent step. The remaining AFD shapes constitute a power dependent AFD profile, which is then used to operate the reactor.
As noted previously, the RAOC method identifies margin in the analysis and transfers it to operational flexibility. In its application, as supplemented, the licensee demonstrated andaffirmed its ability to implement the WCAP-10216-P-A RAOC methodology.Part B describes an alternate means for monitoring F Q. The WCAP-10216-P-A methodmeasures F Q directly under equilibrium conditions. The method then increases the measured F Q to account for manufacturing tolerance and measurement uncertainties. A further increaseis made to account for the effect of normal operating transients. This final F Q is compared toTS limits to determine whether this parameter is out of specification. By the application, assupplemented, the licensee demonstrated and affirmed its ability to implement theWCAP-10216-P-A RAOC methodology.In its proposed implementation of the WCAP-10216-P-A, RAOC/F Q Surveillance, the licenseestated the supporting analysis was performed at extended power uprate (EPU) conditions, and that the analysis is bounded at the current power level. While the use of WCAP-10216-P-A isnot predicated on a particular reactor power or plant configuration, changes thereto could alterthe results of analysis when using WCAP-10216-P-A. The licensee submitted an application dated July 7, 2005 (Reference 5), to allow a 16.8% increase in its licensed thermal power level (extended power uprate (EPU)). In order to accommodate the EPU, the licensee is making numerous changes to its facility and analyses. Some of those changes would affect the analyses used to support the RAOC/F Q Surveillance determination. Two such changes arereflected in license amendment applications dated April 29, 2005, regarding revised LOCAmethodologies and changes in main feedwater isolation (References 6 and 7). The synergistic combination of these changes is inherent in the analysis that led to the RAOC/F Q Surveillancedetermination. The NRC staff had a concern that, absent those synergistic effects, the analysisused to support the licensee's request may not bound current conditions. In its response to a request for additional information dated September 19, 2005 (Reference 2), the licensee provided a discussion of several items affected by the EPU and associated changes that could impact the analysis used to support the WCAP-10216-P-A acceptability determination, and apotential course of action should those changes not take place in concert with WCAP-10216-P-A implementation. The licensee affirmed that the reload methodology in use at Ginna (WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"
(Reference 8)) would verify the use of the AFD profile and identify any non-bounding conditions.
Therefore, the NRC staff has reasonable assurance the WCAP-10216-P-A, RAOC/F QSurveillance, methodologies will be implemented appropriately.The changes to the Ginna TSs being made to incorporate WCAP-10216-P-A, RAOC/F QSurveillance, are consistent with NUREG-1431, "Standard Technical SpecificationsWestinghouse Plants," Revision 3.0 (Reference 9), and are acceptable.Based on the above evaluations, the NRC staff has determined that Westinghouse ReportWCAP-10216-P-A is applicable to Ginna and that the licensee's request to implement the report is acceptable.3.3 Relocation of QPTR and AFD Monitor Alarm RequirementsThe QPTR is monitored on an automatic basis using the plant process computer that has aQPTR monitor alarm. The plant process computer determines the ratio of the highest average power in any quadrant to the average power in the 4 quadrants by using the inputs from theexcore detectors. If the calculated QPTR exceeds the 1.02 limit, an alarm message is generated so that the operators can take corrective action.The AFD is monitored on an automatic basis using the plant process computer that has an AFDmonitor alarm. The computer determines the 1 minute average of each of the operable excore detector outputs and provides an alarm message immediately if the AFD for two or more operable excore channels is outside its specified limits.Originally, the requirements of 10 CFR 50.36 established the categories of items for inclusion inthe TSs, but not the particular requirements for the TSs of each individual plant. The NRC provided guidance for the specific contents of the TSs in the "Final Policy Statement onTechnical Specifications Improvements for Nuclear Power Reactors" (Final Policy Statement),
58 FR 39132 (July 22, 1993). In particular, the NRC indicated that certain items could berelocated from the TSs to licensee-controlled documents. The Final Policy Statement established four criteria for determining the items required for inclusion in the TSs:Criterion 1 Installed instrumentation that is used to detect, and indicate in the controlroom, a significant abnormal degradation of the reactor coolant pressure boundary. Criterion 2 A process variable, design feature, or operating restriction that is an initialcondition of a design basis accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of the fission product barrier.Criterion 3 A structure, system, or component that is part of the primary successpath and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of the fission product barrier.Criterion 4 A structure, system, or component which operating experience orprobabilistic risk assessment has shown to be significant to public healthand safety.These criteria have been codified in 10 CFR 50.36(c)(2)(ii). See Final Rule, "TechnicalSpecifications," 60 FR 36953 (July 19, 1995). As a result, TS requirements which fall within or satisfy any of the criteria in the Final Policy Statement must be retained in the TSs. The Final Policy Statement allows the relocation of items not meeting these four specified criteria from theTSs to licensee-controlled documents, such that future changes can be made to these provisions pursuant to 10 CFR 50.59. The NRC also concluded that compliance with the FinalPolicy Statement satisfied Section 182a of the Act, precipitating a revision to 10 CFR 50.36 which superseded the Final Policy Statement. Thus, within this general framework, licensees may remove material from their TSs on two conditions: (1) the material is not required to be in the TSs based on the NRC staff's interpretation of 10 CFR 50.36, including judgments aboutthe level of detail required in the TSs, and (2) there exist suitable alternative regulatory controlsfor the material.The licensee proposed a revision to TS 3.2.3 in order to relocate the portion of the LCO andassociated Action and SRs regarding the AFD alarm monitor that are not required to be contained in the TSs. The licensee proposed a revision to TS 3.2.4 in order to relocate the portion of the LCO and associated SRs regarding the QPTR alarm monitor that are not requiredto be contained in the TSs. These requirements will be relocated to the TRM, which is alicensee-controlled document that is subject to the provisions of 10 CFR 50.59 and provides anappropriate level of review and approval for the revision of requirements that are important to safety, but do not satisfy the criteria of 10 CFR 50.36(c)(2)(ii) for TS requirements. Changes to the TRM requirements are subject to the regulations of 10 CFR 50.59 because the licensee hasincorporated the TRM into the Ginna UFSAR. The proposed revision will maintain anappropriate level of control of the relocated requirements and an improved level of consistency with NUREG-1431, which does not contain requirements for the TSs proposed for relocation. The application of these requirements as part of the TRM will be identical to those applicable tothe TSs.The proposed change will relocate the identified TSs to the licensee-controlled TRM consistentwith the 10 CFR 50.36 requirements. The NRC staff has reviewed the licensee's submittal andfinds that relocation of these requirements to a licensee-controlled document is acceptable inthat the LCO and associated requirements were found not to fall within the scope of the criteriacontained in 10 CFR 50.36(c)(2)(ii), and changes to licensee-controlled documents will beadequately controlled by 10 CFR 50.59, as applicable. Relocating the requirements for the AFD and QPTR monitor alarms to the TRM is consistentwith TSTF-110, "Delete SR Frequencies Based on Inoperable Alarms," (Ref. 10). 3.4 SummaryThe NRC staff has reviewed the license amendment request and has concluded that the safetymargin will remain acceptable and Ginna continues to meet regulatory requirements. SinceGinna is a Westinghouse plant and the RAOC control limits will be determined by the approvedWestinghouse LOCA and non-LOCA methods, the NRC staff concludes the licensee's use ofthe methods documented in WCAP-10216-P-A is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the New York State official was notified of theproposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facilitycomponent located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves nosignificant increase in the amounts, and no significant change in the types, of any effluents thatmay be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been nopublic comment on such finding (70 FR 33220). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR51.22(b) no environmental impact statement or environmental assessment need be prepared inconnection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above that (1) there isreasonable assurance that the health and safety of the public will not be endangered byoperation in the proposed manner, (2) such activities will be conducted in compliance with theCommission's regulations, and (3) the issuance of the amendment will not be inimical to thecommon defense and security or to the health and safety of the public.
7.0 REFERENCES
1.Ginna LLC letter, M. G. Korsnick to NRC, "License Amendment Request RegardingAdoption of Relaxed Axial Offset Control (RAOC), R.E. Ginna Nuclear Power Plant, Docket No. 50-334," dated April 29, 2005. (ML051300330)2.Ginna LLC letter, M. G. Korsnick to NRC, "Response to NRC Generic Request forAdditional Information Regarding Relaxed Axial Offset Control, R. E. Ginna Nuclear Power Plant, (TAC No. MC6867)," dated September 19, 2005. (ML052720369)3.Westinghouse Report WCAP-10216-P-A, "Relaxation of Constant Axial OffsetControl/FQ Surveillance Technical Specification," Revision 1, February 1994.4.NRC letter, Ashok C. Thadani, Director, Division of Systems Safety and Analysis, toNicholas J. Liparulo, Manager, Nuclear Safety and Regulatory Activities, Westinghouse Electric Corporation, "Acceptance for Referencing of Revised Version of Licensing Topical Report WCAP-10216-P-A, Rev. 1, Relaxation of Constant Axial Offset Control -
F Q Surveillance Technical Specification," (TAC No. M88206), dated November 26, 1993.5.Ginna LLC letter, M. G. Korsnick to NRC, "R.E. Ginna Nuclear Power Plant, DocketNo. 50-244, License Amendment Request Regarding Extended Power Uprate," dated July 7, 2005. (ML051950123)6.Ginna LLC letter, M. G. Korsnick to NRC, "License Amendment Request RegardingRevised Loss of Coolant Accident (LOCA) Analyses-Changes to Accumulator, Refueling Water Storage (RWST),and Administrative Control Technical Specifications, R.E. Ginna Nuclear Power Plant, Docket No. 50-334," dated April 29, 2005. (ML051260239)7.Ginna LLC letter, M. G. Korsnick to NRC, "License Amendment Request RegardingMain Feedwater Isolation Valves, R.E. Ginna Nuclear Power Plant, Docket No. 50-334,"
dated April 29, 2005. (ML051260236)8.Westinghouse Report WCAP-9272-P-A, "Westinghouse Reload Safety EvaluationMethodology," July 1985.9.NRC Report NUREG-1431, "Standard Technical Specifications Westinghouse Plants,"Revision 3.0.10.TSTF-110, "Delete SR Frequencies Based on Inoperable Alarms," Revision 2, October2, 1997. (ML040490071) Principal Contributor: K. Wood Date: February 15, 2006