ML053120186
| ML053120186 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 01/25/2006 |
| From: | Jaffe D H Plant Licensing Branch III-2 |
| To: | Cobb D K Detroit Edison |
| Jaffe D H, NRR/DLPM, 415-1439 | |
| Shared Package | |
| ML | List: |
| References | |
| TAC MC6468 | |
| Download: ML053120186 (14) | |
Text
January 25, 2006Mr. Donald K. CobbAssistant Vice President - Nuclear Generation Detroit Edison Company 6400 North Dixie Highway Newport, MI 48166
SUBJECT:
FERMI 2 - ISSUANCE OF AMENDMENT RE: REACTOR COOLANT SYSTEMPRESSURE AND TEMPERATURE CURVES (TAC NO. MC6468)
Dear Mr. Cobb:
The Commission has issued the enclosed Amendment No. 168 to Facility Operating LicenseNo. NPF-43 for the Fermi 2 facility. The amendment revises the Technical Specifications (TSs)in response to your application dated March 17, 2005, as supplemented by letter dated April 15, 2005. The amendment revises TS 3.4.10, "RCS Pressure and Temperature (P/T) Limits." Specifically, the amendment revises the P/T curves for the hydrostatic pressure test, non-nuclear heatup and cooldown, and nuclear (core critical) limits illustrated in TS Figure 3.4.10-1 with six recalculated separate curves for 24 and 32 effective full power years of reactor operation. In addition, the amendment also revises associated surveillancerequirements.A copy of our safety evaluation is also enclosed. The Notice of Issuance will be included in theCommission's biweekly Federal Register notice.Sincerely,/RA/David H. Jaffe, Sr. Project ManagerPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-341
Enclosures:
- 1. Amendment No. 168 to NPF-43 2. Safety Evaluationcc w/encls: See next page January 25, 2006Mr. Donald K. CobbAssistant Vice President - Nuclear Generation Detroit Edison Company 6400 North Dixie Highway Newport, MI 48166
SUBJECT:
FERMI 2 - ISSUANCE OF AMENDMENT RE: REACTOR COOLANT SYSTEMPRESSURE AND TEMPERATURE CURVES (TAC NO. MC6468)
Dear Mr. Cobb:
The Commission has issued the enclosed Amendment No. 168 to Facility Operating LicenseNo. NPF-43 for the Fermi 2 facility. The amendment revises the Technical Specifications (TSs)in response to your application dated March 17, 2005, as supplemented by letter dated April 15, 2005. The amendment revises TS 3.4.10, "RCS Pressure and Temperature (P/T) Limits." Specifically, the amendment revises the P/T curves for the hydrostatic pressure test, non-nuclear heatup and cooldown, and nuclear (core critical) limits illustrated in TS Figure 3.4.10-1 with six recalculated separate curves for 24 and 32 effective full power years of reactor operation. In addition, the amendment also revises associated surveillancerequirements.A copy of our safety evaluation is also enclosed. The Notice of Issuance will be included in theCommission's biweekly Federal Register notice.Sincerely,/RA/David H. Jaffe, Sr. Project ManagerPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-341
Enclosures:
- 1. Amendment No. 168 to NPF-43 2. Safety Evaluationcc w/encls: See next page DISTRIBUTION
- PUBLICRidsNrrPMDJaffeRidsAcrsAcnwMailCenterRidsNrrDciCvib LPLIII-1 R/FRidsNrrLADClarkeRidsOgcRp GHill (2)DORL DPR RidsNrrDorlLpleRidsNrrDirsItsbRidsNrrDssSbwb NRayLLoisRidsRgn3MailCenterADAMS Accession Number: ML053120191 (Package)ADAMS Accession Number: ML053120186 (Letter)
ADAMS Accession Number: ML060260552 (Tech Spec)*SE input by memo w/ minor changes OFFICELPLIII-1/PMLPLIII-1/LAITSB/BCCVIB/BC*SBWB/BC(A)*OGCLPLIII-1/BC (A)NAMEDJaffe:RSADClarkeTBoyceMMitchellGHsiiMSpencerTKobetzDATE1/20/061/20/061/4/068/4/055/2/051/17/061/24/06OFFICIAL RECORD COPY Fermi 2 cc:
David G. PettinariLegal Department 688 WCB Detroit Edison Company 2000 2nd Avenue Detroit, MI 48226-1279Michigan Department of EnvironmentalQuality Waste and Hazardous Materials Division Radiological Protection and Medical Waste Section Nuclear Facilities Unit Constitution Hall, Lower-Level North 525 West Allegan Street P.O. Box 30241 Lansing, MI 48909-7741U.S. Nuclear Regulatory CommissionResident Inspector's Office 6400 N Dixie Highway Newport, MI 48166Mr. M. V. Yudasz, Jr., DirectorMonroe County Emergency Management Division 965 South Raisinville RoadMonroe, MI 48161Regional Administrator, Region IIIU.S. Nuclear Regulatory Commission 2443 Warrenville Road, Suite 210Lisle, IL 60532-4352Ronald W. GastonManager, Nuclear Licensing Detroit Edison Company Fermi 2 - 200 TAC 6400 North Dixie Highway Newport, MI 48166Supervisor - Electric OperatorsMichigan Public Service Commission 6545 Merchantile Way P.O. Box 30221 Lansing, MI 48909Wayne County Emergency Management Division 10250 Middlebelt Road Detroit, MI 48242 DETROIT EDISON COMPANYDOCKET NO. 50-341FERMI 2AMENDMENT TO FACILITY OPERATING LICENSEAmendment No. 168License No. NPF-431.The U.S. Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by the Detroit Edison Company (DECo) datedMarch 17, 2005, as supplemented by letter dated April 15, 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 FacilityOperating License No. NPF-43 is hereby amended to read as follows:Technical Specifications and Environmental Protection PlanThe Technical Specifications contained in Appendix A, as revised throughAmendment No. 168, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. DECo shall operate the facility in accordance with the Technical Specifications and the EnvironmentalProtection Plan.3.This license amendment is effective as of its date of issuance and shall be implementedwithin 60 days of the date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION/RA/ Timothy Kobetz, Acting Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: January 25, 2006 ATTACHMENT TO LICENSE AMENDMENT NO. 168FACILITY OPERATING LICENSE NO. NPF-43DOCKET NO. 50-341Replace 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.REMOVEINSERT3.4-243.4-243.4-253.4-25 3.4-273.4-27 3.4-283.4-28
- - -3.4-28a
- - -3.4-28b
- - -3.4-28c
- - -3.4-28d
- - -3.4-28e SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATED TO AMENDMENT NO. 168 FACILITY OPERATING LICENSE NO. NPF-43DETROIT EDISON COMPANYFERMI 2DOCKET NO. 50-34
11.0 INTRODUCTION
By application to the U.S. Nuclear Regulatory Commission (NRC, Commission) dated March 17, 2005 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML050830480), as supplemented by letter dated April 15, 2005 (ADAMS Accession No. ML051230442), the Detroit Edison Company (the licensee) requested changes to its Technical Specifications (TSs) for Fermi 2. The supplement dated April 15, 2005, providedadditional information that clarified the application, did not expand the scope of the applicationas originally noticed, and did not change the NRC staff's original proposed no significant hazardsdetermination as published in the Federal Register on April 26, 2005 (70 FR 21453).The proposed amendment would revise TS 3.4.10, "RCS [Reactor Coolant System] Pressureand Temperature (P/T) Limits." Specifically, the proposed amendment would revise the P/T curves for the hydrostatic pressure test, non-nuclear heatup and cooldown, and nuclear (core critical) limits illustrated in TS Figure 3.4.10-1 with six recalculated separate curves for 24 and 32 effective full power years (EFPY) of reactor operation. In addition, the amendment would revise associated surveillance requirements (SRs).
2.0 REGULATORY EVALUATION
The NRC staff evaluated the acceptability of the licensee's proposed P/T limit curves based onthe following regulations and guidance:Title 10 of the Code of Federal Regulations (10 CFR), Section 50.60(a) states:Except as provided in paragraph (b) of this section, all light-water nuclear powerreactors, other than reactor facilities for which the certifications required under§50.82(a)(1) have been submitted, must meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in appendices G and H to this part.Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements," requires that facility P/Tlimit curves for the reactor pressure vessel (RPV) be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). The most recent version of Appendix G to Section XI of the ASME Code, which has been endorsed in 10 CFR 50.55a, and,therefore, by reference in Appendix G to 10 CFR Part 50, is the 2001 Edition through the 2003 Addenda of the ASME Code. This edition of Appendix G to Section XI of the ASME Code incorporates the provisions of ASME Code Cases N-588, "Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels,Section XI, Division 1,"
and N-640, "Alternative Reference Fracture Toughness for Development of P-T Curves for ASME Section XI, Division 1." In addition, Appendix G to 10 CFR Part 50 imposes minimum head flange temperatures when system pressure is above 20 percent of the preservicehydrostatic test pressure.Appendix H to 10 CFR Part 50, "Reactor Vessel Material Surveillance Program Requirements,"establishes requirements related to facility RPV material surveillance programs. NRCRegulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," describes methods and assumptions acceptable to the NRC staff fordetermining the pressure vessel neutron fluence. RG 1.99, Revision 2, "Radiation Embrittlementof Reactor Vessel Materials," contains methodologies for determining the increase in transition temperature resulting from neutron radiation.Generic Letter (GL) 92-01, Revision 1, "Reactor Vessel Structural Integrity," requested thatlicensees submit their plant's RPV data for NRC staff review. GL 92-01, Revision 1, Supplement1, requested that licensees provide and assess data from other licensees that could affect theirRPV integrity evaluations. NUREG-0800, "Standard Review Plan," Section 5.3.2, "Pressure Temperature Limits andPressurized Thermal Shock," provides guidance on using these regulations and documents in the NRC staff's review. In addition, Section 5.3.2 provides guidance to the NRC staff inperforming check calculations of the licensee's submittal.Appendix A to 10 CFR Part 50, "General Design Criteria [GDC] for Nuclear Power Plants,"Criterion 30, "Quality of reactor coolant pressure boundary," and Criterion 31, "Fracture prevention of reactor coolant pressure boundary," establish requirements for pressure vessel fluence calculations. The NRC staff has approved vessel fluence calculation methodologieswhich satisfy the requirements of GDC 30 and 31 and adhere to RG 1.190 guidance. Fluence calculations are acceptable if they are done with approved methodologies or with methods which are shown to conform to RG 1.190 guidance.
3.0 TECHNICAL EVALUATION
The NRC staff has reviewed the licensee's technical and regulatory analyses in support of itsproposed license amendment.3.1 Assessment of Neutron Fluence LevelsThe NRC staff reviewed the licensee's submittal to establish that the pressure vessel fluenceused in the calculation of the vessel material properties for the estimation of the pressure vessel P/T limit curves is acceptable. The current Fermi 2 P/T limit curves were approved in Amendment No. 87 and their estimated applicability is to December 31, 2005. The licenseeprovided new vessel fluence calculations for both the 24 and 32 EFPY. The licensee submitted General Electric Company (GE) Document No. GE-NE-0000-0031-6254-R1, "DTE Energy Fermi 2 Energy Center Neutron Flux Evaluation,"
dated February 2005, which describes in detail the methodology, and justifies the input to thecalculations in view of RG 1.190 guidance. The methodology is based on GE Document No.
NEDC-32983-P-A, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation," that the NRC approved in a letter to GE dated September 14, 2001 (ADAMSAccession No. ML012400381). The method is based on a discrete ordinates method and cross sections derived from the ENDF/B-VI cross section file as recommended in RG 1.190. The transport approximations are P 3 for the inelastic scattering approximation and an acceptablevalue for the angular quadrature approximation.Fermi 2 is currently in fuel cycle 11 with the next refueling outage scheduled for March 2006. Fermi 2 implemented a 3293 to 3430 megawatts thermal (MWt) power uprate at the end of the third fuel cycle. The Fermi 2 RPV flux for the first 10 cycles of operation was determined utilizingcycle 6 as a representative cycle with a thermal power limit of 3430 MWt, which is conservativedue to the fact that the reactor operated at 3293 MWt for the first three cycles. An additionalconservative factor of 10 percent was added to the cycle 6 RPV flux. The total energygenerated by the Fermi 2 reactor for cycles 1 through 10 is equivalent to 12 EFPY at a 3430MWt licensed thermal power. Regarding future plant operations, the fluence calculations conservatively assumed the reactor would operate at 3952 MWt (the planned extended power uprate (EPU)) for the remaining cycles until the end of the 40-year license in 2025. The fluencecalculations assumed that the total energy at that time would be equivalent to 32 EFPY. This isconservative because the current licensed power level is limited to 3430 MWt. The NRC staffdetermined that the source computation is conservative and, therefore, it is acceptable. The core operating conditions were retrieved from plant engineering data, including materialcompositions and core void fractions. The core, shroud, vessel, jet pump and surveillancecapsule geometry and location were derived from design data. The fuel location and fuel type reflect actual design core loadings for the pre-EPU and EPU phase of the operation. The geometric representation of the core and its surroundings is in excellent agreement with the actual plant geometry and, therefore, it is acceptable.The neutron source distribution was based on integrated energy production in conjunction withisotopic density distribution and corresponding fission energy and fission yield data. The calculated peak, inside vessel surface fluence for 24 EFPY, is 7.13x10 17 neutrons/centimetersquared (n/cm 2), and for 32 EFPY is 9.68x10 17 n/cm 2. Because these values were calculated inaccordance with RG 1.190 guidance as discussed above, they are acceptable.
3.2 P/T Limit Curve AssessmentThe NRC staff evaluated the acceptability of the licensee's proposed Fermi 2 P/T limit curvesthat provide new limits that are valid for 24 and 32 EFPY of operation.The methodology in Appendix G to Section XI of the ASME Code postulates the existence of asharp surface flaw normal to the direction of the maximum applied stress for axial welds, plates, and forgings. A sharp surface flaw parallel to the weld is postulated for the evaluation of circumferential welds. For materials in the beltline, upper head, and lower head regions of the RPV, the maximum flaw size is postulated to have a depth that is equal to one-fourth of the thickness and a length equal to 11/2 times the wall thickness. Thus, the critical locations in the RPV beltline and head regions are the 1/4-thickness (1/4T) and 3/4-thickness (3/4T) locations,which correspond to the crack tips of the postulated inside and outside surface flaws of the vessel. For RPV nozzles, the surface flaw is postulated to be parallel to the axis of the nozzle's corner radius.The basic parameter, in Appendix G to Section XI of the ASME Code, for calculating P/T limitcurves is the stress intensity factor, K I, which is a function of the stress state and flawconfiguration. Beginning with the 1999 Addenda to the 1998 Edition, the static crack initiationfracture toughness curve, K IC, has been provided in Figure G-2210-1 in Appendix G to Section XI of the ASME Code. The axes in Figure G-2210-1 are K IC and T-RTNDT, where T istemperature and RTNDT is the reference temperature of the material. For beltline materials, the RTNDT is increased due to neutron radiation embrittlement. This value is described as anadjusted reference temperature (ART).The methodology in Appendix G to Section XI of the ASME Code requires that licenseesdetermine the ART value at the maximum postulated flaw depth of beltline materials. The ARTvalue is defined as the sum of the initial (unirradiated) reference temperature (initial RTNDT), themean value of the adjustment in reference temperature caused by irradiation (RTNDT), and amargin (M) term. The RTNDT is the product of a chemistry factor (CF) and a fluence factor. TheCF is dependent upon the amount of copper and nickel in the material and may be determined from tables in RG 1.99, Revision 2 or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The M term is dependent uponwhether the initial RTNDT is a plant-specific or a generic value and whether the CF wasdetermined using the tables in RG 1.99, Revision 2 or from surveillance data. The M term is used to account for uncertainties in the values of initial RTNDT, copper and nickelcontents, fluence, and calculational procedures. RG 1.99, Revision 2 describes the methodology used in calculating the M term. The methodology in Appendix G to Section XI of the ASME Code requires that P/T limit curvesmust satisfy a factor of 2.0 on stress intensity factors arising from primary membrane and bending stresses during normal plant operation (including heatups, cooldowns, and anticipated transients), and a factor of 1.5 on stress intensity factors arising from primary membrane and bending stresses when leak rate or hydrostatic pressure tests are performed on the RCS.
Table 1 of Appendix G to 10 CFR Part 50 provides the NRC staff's criteria for meeting the P/Tlimit requirements and the minimum temperature requirements for bolting up the vessel during normal and pressure testing operations. Table 1 of Appendix G to 10 CFR Part 50 also identifies P/T limit requirements based on the RTNDT of the materials in the closure flange regionwhich is highly stressed by the bolt preload.The proposed amendment would replace the following TS figures with P/T limit curves valid to24 and 32 EFPY of operation:Applicable to 24 EFPY Figure 3.4.10-1System Pressure Test P/T Curve (Curve A)Figure 3.4.10-2Core Not Critical P/T Curve (Curve B)
Figure 3.4.10-3Core Critical P/T Curve (Curve C) Applicable to 32 EFPYFigure 3.4.10-4System Pressure Test P/T Curve (Curve A)Figure 3.4.10-5Core Not Critical P/T Curve (Curve B)
Figure 3.4.10-6Core Critical P/T Curve (Curve C)Composite curves were generated by enveloping the most restrictive P/T limits from theseparate bottom head, beltline, upper vessel and closure assembly P/T limits.The licensee's ART calculation is in accordance with RG 1.99, Revision 2. For P/T calculations,the licensee used the 1998 Edition of the ASME Code, including 2000 Addenda, which is equivalent to the 2001 Edition through 2003 Addenda of the ASME Code. This version of the ASME Code has been endorsed in 10 CFR 50.55a. The P/T limit curve methodology used K ICfrom Figure A-4200-1 of Section XI in Appendix A to determine the relationship between fracture toughness and T-RTNDT. P/T limit curves were developed using the geometry of the RPV shellsand discontinuities, the initial RTNDT of the RPV materials, and the ART for the beltline materials. As a conservative simplification, the thermal gradient stress at the 1/4T location is assumed tobe tensile for both heatup and cooldown, thereby resulting in the approach of applying themaximum tensile stress at the 1/4T location. The licensee stated that this approach isconservative because (1) the stresses at the 1/4T location during heatup are assumed to be tensile, and (2) radiation effects cause the allowable toughness at 1/4T to be less than that at3/4T for a given metal temperature.In addition to the beltline considerations, limits related to non-beltline discontinuities such asnozzles, penetrations, and flanges influence the construction of P/T limit curves. The non-beltline limits were included to allow monitoring of the vessel bottom head and upper vessel regions, separate from the beltline region, to help minimize heating requirements prior to pressure testing.3.2.1 Bottom Head Curves Bottom head curves are utilized because the water in the vessel lower head is separated fromthe water in contact with the vessel beltline and upper head regions by the reactor baffle plates. The water in the regions above the baffle plate is heated by decay heat from the reactor core,while the water in the lower head is at a lower temperature due to the injection of control rod drive (CRD) water for vessel pressurization. With little or no circulation through the recirculationpump loops, these regions are therefore maintained at different temperatures during inservice leak and hydrostatic testing and non-nuclear heatup/cooldown conditions.The bottom head thermal transient and pressure stresses were developed using the limitingnormal and upset conditions and generic boiling water reactor (BWR)/6 dimensions. GenericP/Ts for the bottom head curves were determined using the applied stress intensity factor (K I)based on a revised finite element analysis that is described in Appendix H of the licensee'sMarch 17, 2005, submittal.
The generic bottom head curves have been adjusted to reflect thehighest RTNDT value of 30 F for the bottom head plates and welds.Evaluations were also performed to determine whether the CRD nozzle analyzed in the earlierfinite element analysis was the limiting location. The licensee concluded that these analyses showed that the generic BWR/6 P/T limit curve, indexed to the RTNDT of the limiting bottom headmaterial, is conservative when applied to the Fermi 2 bottom head.The NRC staff concluded that the licensee's evaluation for the bottom head curves is acceptablebecause the licensee showed that all discontinuities in the bottom head region were bounded by the CRD discontinuity, and that stresses in the generic BWR/6 vessel were higher than those inthe Fermi 2 vessel. 3.2.2 Upper Vessel, Flange, and Beltline Region Curves The P/T limit curves for inservice leak and hydrostatic testing and non-nuclear heatup/cooldownoperations were developed from curves based on the material properties for the upper vessel (including feedwater nozzle), vessel flange, and vessel beltline regions. The P/T limit curve for core-critical operations, however, did not consider feedwater nozzles. Since the bottom head curves are less limiting than the upper vessel, vessel flange, and beltline region curves, the bottom head curves are not utilized for developing the core-critical operations curve. Using thehighest RTNDT for the materials in the beltline, upper vessel, and closure flange regions, thelicensee developed P/T limit curves to meet the criteria in 10 CFR Part 50, Appendix G and ASME Code,Section XI, Appendix G.The upper vessel region P/T limits were based on analysis of the feedwater nozzle and beltlineregions. The K I for the feedwater nozzle during pressure test conditions was computed usingthe methods from Weld Research Council (WRC) Bulletin 175 together with the geometry from a feedwater nozzle. Since Appendix G to Section XI of the ASME Code indicates that the methods used in WRC Bulletin 175 are acceptable for analyzing the inside corner of a nozzle and cylindrical shell for elastic stresses due to internal pressure stress, the method of analysis proposed by the licensee for the upper vessel and feedwater nozzle will satisfy the requirementsof Appendix G to 10 CFR Part 50. The K I for the upper vessel curve during normal operationwas determined using the primary and secondary stresses from a feedwater nozzle finite element analysis and a membrane stress intensity factor, M m, based on the values identified inAppendix G to Section XI of the ASME Code for a postulated surface flaw normal to the direction of maximum stress. The P/Ts for the upper vessel curve included the limiting feedwater transient for normal and upset conditions.The beltline region P/T limits were based on the ART for the limiting materials in the beltline ofthe Fermi 2 RPV. The limiting beltline material is the lower shell axial weld (heat 13253 and heat 12008). Heat CE-2(WM), which has been determined to be heat 13253,12008 is the heat for the surveillance weld material as defined by the BWR integrated surveillance program (ISP). For this material, the CF was adjusted by multiplying the ISP least-squares fit CF developed in accordance with RG 1.99, Revision 2, as reported in BWRVIP-102, by the ratio of the CF for thevessel weld chemistry to the CF for the ISP surveillance chemistry. This results in an adjusted CF of 354.5 [326.96*(224/206.6)=354.5]. The key parameters for the licensee's ARTdetermination for the 1/4T location are shown as follows: Plate/HeatEFPYAdjusted CF Init. RTNDT1/4T FluenceRTNDTMarginARTCE-2(WM)(13253, 12008)24354.5-44 F3.04x10 17 n/cm 2 78 F28 F62 FCE-2(WM)(13253, 12008)32354.5 -44 F4.06x10 17 n/cm 2 93 F28 F77 FThe fluence values were calculated using a computer code that was based on an NRC-approved methodology. The fluence calculation was approved by the NRC staff in amemorandum dated May 2, 2005 (ADAMS Accession No. ML005123002).The P/T limit curves apply to both heatup and cooldown and for both 1/4T and 3/4T locationsbecause the maximum tensile stress of either heatup or cooldown is applied at the 1/4T location. The NRC staff's assessment also included an independent calculation of the ART value for the1/4T location of Fermi 2 RPV beltline regions. For the evaluation of the limiting beltline materials, the NRC staff confirmed that the ARTs were based on the methodology of RG 1.99,Revision 2. The P/T limit curve methodology used the formula in Appendix G to Section XI to determine the K I values. The licensee stated that the feedwater nozzle, which experiences feedwater flow that is colderrelative to the vessel coolant, was selected to represent all nozzles for determining fracture toughness in the upper shell region. The highest non-beltline RTNDT for the feedwater nozzle atFermi 2 is 12 F. However, the licensee has increased the RTNDT to 25 F to consider thestresses in the recirculation inlet nozzle. The generic curve applied to the Fermi 2 upper vessel is modified for upper vessel by shifting the pressure versus (T-RTNDT) values to reflect the RTNDTvalue of 25 F.In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes a minimumtemperature at the closure head flange based on the reference temperature for the flangematerial.Section IV.A.2 of Appendix G states that when the pressure exceeds 20 percent of the preservice system hydrostatic test pressure, the temperature of the closure flange regions(highly stressed by the bolt preload) must exceed the reference temperature of the material in those regions by at least 120 F for normal operation and by 90 F for hydrostatic pressure testsand leak tests. Based on the flange RTNDT of 12 F for Fermi 2, the NRC staff has determinedthat the proposed P/T limits have satisfied the requirement for the closure flange region duringnormal operation and inservice leak and hydrostatic testing.3.2.3 Changes to Associated Surveillance RequirementsIn addition to the introduction of six P/T curves, the license amendment request proposed P/Tchanges to surveillance requirements (SR) 3.4.10.1 and SR 3.4.10.2 for RCS limits. Therequested changes to SR 3.4.10.1 and SR 3.4.10.2 would reflect the replacement of one P/Tcurve with six curves by changing the figure numbers to be used in performance of the surveillances. These changes would make the SRs reference the correct figures and are therefore acceptable. The licensee requested changes to the reactor vessel flange and head flange temperaturesstated in SR 3.4.10.7, SR 3.4.10.8, and SR 3.4.10.9. The changes would reflect the newminimum temperatures indicated on new Figures 3.4.10.1 through 6. As stated earlier in thisSE, the methodology used to create the figures is acceptable to the NRC, therefore, it isacceptable to use the information in the figures as a new minimum value in the SRs.3.3 Technical Evaluation ConclusionThe NRC staff has reviewed the licensee's submittal, and based on the review discussed above,the NRC staff finds the proposed changes to be acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Michigan State official was notified of theproposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
This amendment changes a requirement with respect to the installation or use of a facilitycomponent located within the restricted area as defined in 10 CFR Part 20 and with respect to surveillance requirements. The NRC staff has determined that this amendment involves nosignificant increase in the amounts, and no significant change in the types, of any effluent 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 this amendment involves no significant hazards consideration and there has been no public comment on such finding published April 26, 2005 (70 FR 21453). Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to10 CFR 51.22(b), no environmental impact statement or environmental assessment need beprepared in connection with the issuance of this amendment.
6.0 CONCLUSION
The NRC staff 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 this amendment will not be inimical to thecommon defense and security or to the health and safety of the public.Principal Contributors: N. Ray L. LoisDate: January 25, 2006