ML051160277

Issuance of AMD.193 Reactor Coolant System Pressure Temeprature Limits (Tac. MC3591)

ML051160277
Person / Time
Site:	Columbia
Issue date:	05/12/2005
From:	Brian Benney NRC/NRR/DLPM/LPD4
To:	Parrish J Energy Northwest
Benney B, NRR/DLPM, 415-3764
Shared Package
ML051360468	List: ML051160277 ML051360214
References
TAC MC3591
Download: ML051160277 (12)

Text

May 12, 2005 Mr. J. V. Parrish Chief Executive Officer Energy Northwest P.O. Box 968 (Mail Drop 1023)

Richland, WA 99352-0968

SUBJECT:

COLUMBIA GENERATING STATION - ISSUANCE OF AMENDMENT RE: REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (TAC NO. MC3591)

Dear Mr. Parrish:

The Commission has issued the enclosed Amendment No. 193 to Facility Operating License No. NPF-21 for the Columbia Generating Station. This amendment is in response to your application dated June 9, 2004, as supplemented on April 1, 2005.

This amendment revises Technical Specification (TS) Section 3.4.11, "RCS Pressure and Temperature (P/T) Limits," to replace the P/T limit curves for Inservice Leak and Hydrostatic Testing, Non-Nuclear Heating and Cooldown, and Nuclear Heating and Cooldown currently illustrated in TS Figures 3.4.11-1, 3.4.11-2, and 3.4.11-3, respectively.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register Notice.

Sincerely,

/RA/

Brian Benney, Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-397

Enclosures:

1. Amendment No. 193 to License No. NPF-21

2. Safety Evaluation cc w/encls: See next page

Columbia Generating Station cc:

Mr. W. Scott Oxenford (Mail Drop PE04)

Vice President, Technical Services Energy Northwest P. O. Box 968 Richland, WA 99352-0968 Mr. Albert E. Mouncer (Mail Drop PE01)

Vice President, Corporate Services/

General Counsel/CFO Energy Northwest P.O. Box 968 Richland, WA 99352-0968 Chairman Energy Facility Site Evaluation Council P.O. Box 43172 Olympia, WA 98504-3172 Mr. Douglas W. Coleman (Mail Drop PE20)

Manager, Regulatory Programs Energy Northwest P.O. Box 968 Richland, WA 99352-0968 Mr. Gregory V. Cullen (Mail Drop PE20)

Supervisor, Licensing Energy Northwest P.O. Box 968 Richland, WA 99352-0968 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Chairman Benton County Board of Commissioners P.O. Box 190 Prosser, WA 99350-0190 Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 69 Richland, WA 99352-0069 Mr. Dale K. Atkinson (Mail Drop PE08)

Vice President, Nuclear Generation Energy Northwest P.O. Box 968 Richland, WA 99352-0968 Mr. William A. Horin, Esq.

Winston & Strawn 1400 L Street, N.W.

Washington, DC 20005-3502 Mr. Matt Steuerwalt Executive Policy Division Office of the Governor P.O. Box 43113 Olympia, WA 98504-3113 Ms. Lynn Albin Washington State Department of Health P.O. Box 7827 Olympia, WA 98504-7827 Technical Services Branch Chief FEMA Region X 130 228th Street S.W.

Bothell, WA 98201-9796

May 12, 2005 Mr. J. V. Parrish Chief Executive Officer Energy Northwest P.O. Box 968 (Mail Drop 1023)

Richland, WA 99352-0968

SUBJECT:

COLUMBIA GENERATING STATION - ISSUANCE OF AMENDMENT RE: REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (TAC NO. MC3591)

Dear Mr. Parrish:

The Commission has issued the enclosed Amendment No. 193 to Facility Operating License No. NPF-21 for the Columbia Generating Station. This amendment is in response to your application dated June 9, 2004, as supplemented on April 1, 2005.

This amendment revises Technical Specification (TS) Section 3.4.11, "RCS Pressure and Temperature (P/T) Limits," to replace the P/T limit curves for Inservice Leak and Hydrostatic Testing, Non-Nuclear Heating and Cooldown, and Nuclear Heating and Cooldown currently illustrated in TS Figures 3.4.11-1, 3.4.11-2, and 3.4.11-3, respectively.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register Notice.

Sincerely,

/RA/

Brian Benney, Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-397 DISTRIBUTION:

PUBLIC GHill (2)

Enclosures:

1. Amendment No. 193 to NPF-21 PDIV-2 Reading TBoyce

2. Safety Evaluation RidsNrrDlpmPdiv(HBerkow)

RidsNrrDlpmPdiv2(RGramm) cc w/encls: See next page RidsNrrPMBBenney RidsNrrLALFeizollahi RidsOGCRp RidsACRSACNWMailCenter RidsRegion4MailCenter (BJones)

TS: ML051360214 NRR-100 PKG.: ML051360468 ACCESSION NO.: ML051160277 NRR-058 OFFICE PDIV-2/PM PDIV-2/LA SRXB/SC EMCB/SC IROB/SC OGC Nlo PDIV-2/SC NAME BBenney:sp LFeizollahi LLois MMitchell TBoyce Uttal RGramm DATE 5/3/05 5/3/05 5/4/05 5/5/05 5/6/05 5/12/05 5/12/05 DOCUMENT NAME: E:\\Filenet\\ML051160277.wpd OFFICIAL RECORD COPY

ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 193 License No. NPF-21

1. The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Energy Northwest (licensee) dated June 9, 2004, as supplemented on April 1, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter 1; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-21 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 193 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Robert A. Gramm, Chief, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: May 12, 2005

ATTACHMENT TO LICENSE AMENDMENT NO. 193 FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.

Remove Insert 3.4.11-7 3.4.11-7 3.4.11-8 3.4.11-8 3.4.11-9 3.4.11-9

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 193 TO FACILITY OPERATING LICENSE NO. NPF-21 ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397

1.0 INTRODUCTION

By letter dated June 9, 2004 (under the Agencywide Documents Access Management System (ADAMS) Accession No. ML041680103), Energy Northwest, licensee for Columbia Generating Station (Columbia), submitted a license amendment to change their Technical Specification (TS) 3.4.11. Additional information was provided in a letter dated April 1, 2005 (ADAMS Accession No. ML051030283). The proposed amendment would replace the Pressure and Temperature (P/T) limit curves for Inservice Leak and Hydrostatic Testing, Non-Nuclear Heatup and Cooldown, and Nuclear Heatup and Cooldown. The licensee revised the P/T limit curves to provide new limits that are valid to 33.1 effective full power years (EFPY) of operation.

2.0 REGULATORY EVALUATION

The NRC staff evaluates the acceptability of a licensees proposed P/T limit curves based on the following regulations and guidance:

Section 50.60(a) of Title 10 of the Code of Federal Regulations (10 CFR) states, Except as provided in Paragraph (b) of this section, all light-water nuclear power reactors, 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 program pressure boundary set forth in Appendices G and H to this Part.

Appendix H to 10 CFR Part 50, Reactor Vessel Material Surveillance Program Requirements, establishes requirements related to facility reactor pressure vessel (RPV) material surveillance programs. Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, describes methods and assumptions acceptable to the Nuclear Regulatory Commission (NRC) staff for determining the pressure vessel neutron fluence. RG 1.99 Revision 2, Radiation Embrittlement of Reactor Vessel Materials, contains methodologies for determining the increase in transition temperature resulting from neutron radiation.

Appendix G to 10 CFR Part 50, Fracture Toughness Requirements, requires that facility P/T limit curves for the 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 (ASME) Code. The most recent version of Appendix G to Section XI of the ASME Code, which has been endorsed in 10 CFR Part 50.55a, and, therefore, by reference in Appendix G to 10 CFR Part 50, is the 1998 Edition through the 2000 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 and N-640. In addition, Appendix G to 10 CFR Part 50 imposes minimum head flange temperatures when system pressure is at or above 20 percent of the pre-service hydrostatic test pressure.

Generic Letter (GL) 1992-01, Revision 1, requested that licensees submit the RPV data for their plants to the NRC staff for review. GL 1992-01, Revision 1, Supplement 1, requested that licensees provide and assess data from their licensees that could affect their RPV integrity evaluations. RG 1.99, Revision 2, describes the general procedures for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels.

NUREG-0800, Standard Review Plan, Section 5.3.2, Pressure Temperature Limits, provides guidance on using these regulations and documents in the NRC staffs review. In addition, Section 5.3.2 provides guidance to the NRC staff in performing check calculations of the licensees submittal.

The regulatory requirements for pressure vessel fluence calculations are specified in General Design Criteria (GDCs) 30 and 31. In March 2001, the staff issued Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. The staff approved vessel fluence calculation methodologies which satisfy the requirements of GDCs 30 and 31 and adhere to the guidance in RG 1.190. Fluence calculations are acceptable if they are done with approved methodologies or with methods which are shown to conform to the guidance in RG 1.190.

3.0 TECHNICAL EVALUATION

3.1 Vessel Fluence Methodology The licensee submitted a General Electric (GE) technical report, Pressure-Temperature Curves for Energy Northwest Columbia, which reports on the methodology and the results for 33.1 effective full power years (EFPYs) of operation. The computer code used to calculate the vessel fluence is in accordance with GE Licensing Topical Report NEDC-32983P, which was approved by the NRC in a letter (MFN 01-050) to GE dated September 14, 2001. The calculation accounted for a power uprate which was implemented in the 11th fuel cycle. Based on the above, the staff finds the proposed fluence values for 33.1 EFPYs acceptable for the calculation of the PT curves.

3.2 Appendix G The methodology of Appendix G to Section XI of the ASME Code postulates the existence of a sharp 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 1.5 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 points of the crack tips if the flaws are initiated and propagated from the inside and outside surfaces of the vessel, respectively. For the case of RPV nozzles, the surface flaw is postulated to propagate parallel to the axis of the nozzles corner radius.

The basic parameter in Appendix G to Section XI of the ASME Code for calculating P/T limit curves is the stress intensity factor, KI, which is a function of the stress state and flaw configuration. Beginning with the 1999 Addenda to the 1998 Edition, KIC (static crack initiation fracture toughness curve) is determined from Figure G-2210-1 in Appendix G to Section XI of the ASME Code. The axis in Figure G-2210-1 are KIC and T-RTNDT, where T is temperature 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 an adjusted reference temperature (ART), which is described later in this section.

The methodology of Appendix G to Section XI of the ASME Code requires that P/T limit curves must satisfy a safety factor of 2.0 on stress intensities arising from primary membrane and bending stresses during normal plant operation (including heatups, cooldowns, and transients),

and a safety factor of 1.5 on stress intensities arising from primary membrane and bending stresses when leak rate or hydrostatic pressure tests are performed on the reactor coolant system. Table 1 of Appendix G to 10 CFR Part 50 provides the NRC staffs criteria for meeting the P/T limit requirements of Appendix G to the ASME Code 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 region, which is highly stressed by the bolt preload.

The methodology in Appendix G to Section XI of the ASME Code requires that licensees determine the ART value at the maximum postulated flaw depth or beltline materials. The ART value is defined as the sum of the initial (unirradiated) reference temperature (initial RTNDT), the mean value of the adjustment in reference temperature caused by irradiation (RTNDT), and a margin (M) term. The RTNDT is the product of a chemistry factor and a fluence factor. The chemistry factor 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 margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the chemistry factor was determined using the tables in RG 1.99, Revision 2 or surveillance data. The margin term is used to account for uncertainties in the values of initial RTNDT, copper and nickel contents, fluence, and calculational procedures. RG 1.99, Revision 2, describes the methodology used in calculating the margin term.

The proposed amendment would replace the following TS figures with P/T limit curves valid to 33.1 EFPYs of operation:

Figure 3.4.11-1 Inservice Leak and Hydrostatic Testing Curve (Curve A)

Figure 3.4.11-2 Non-Nuclear Heating and Cooldown Curve (Curve B)

Figure 3.4.11-3 Nuclear Heating and Cooldown Curve (Curve C)

Composite curves were generated by enveloping the most restrictive P/T limits from the separate bottom head, beltline, upper vessel, and closure assembly P/T limits.

With regard to methodology, the licensee used the 1998 Edition of the ASME Code, including the 2000 Addenda, in the evaluation. The P/T limit curve methodology included the use of KIC from Figure G-2210-1 in Appendix G to Section XI of the ASME Code to determine the relationship between fracture toughness and T-RTNDT, and the use of the Mm calculation in Section XI, Paragraph G-2214.1 of the ASME Code. P/T limit curves were developed using geometry of the RPV shells and 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 to be tensile for both heatup and cooldown, thereby resulting in the approach of applying the maximum tensile stress at the 1/4T location. The licensee stated that this approach is conservative because radiation effects cause the allowable toughness at 1/4T to be less than that at 3/4T for a given metal temperature. The staff agrees with the licensee's methodology because it is conservative.

The licensee stated that part of the analysis involved in developing P/T limit curves is to account for radiation embrittlement effects in the beltline region. The method used to account for radiation embrittlement is described in RG 1.99, Revision 2. In addition to the beltline considerations, limits related to non-beltline discontinuities such as nozzles, penetrations, and flanges influence the construction of P/T limit curves. The non-beltline limits are based on generic analyses that are adjusted to the maximum RTNDT for the applicable vessel components. Curves 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.

The staff noted that the adjusted chemistry factor (CF) for beltline weld heat 5P6756 in Tables 4-5b and 4-6b (Columbia Beltline Weld ART Values) in the June 9, 2004 submittal was calculated to be 157.68EF, which is a significant increase from the RG 1.99 Revision 2 CF of 108EF. The licensee was asked to provide additional information regarding the determination of the adjusted CF. The licensee responded that the adjusted CF was calculated using the fitted CF of 119.72EF that was provided by the Boiling Water Reactor Vessel and Internals Project (BWRVIP) from their evaluation, in accordance with BWRVIP-86-A, of three Integrated Surveillance Program surveillance capsule data sets for weld heat 5P6756. (The BWRVIP is an industry-sponsored program for evaluating the integrity of the reactor vessel and internal components.) The adjusted CF for the Columbia RPV beltline weld was determined by multiplying the fitted CF by the ratio of the CF for the RPV beltline weld, as established by the best estimate chemistry for weld wire heat 5P6756, to the CF for the surveillance weld, as established by the chemistry of the surveillance weld. This is equivalent to the methodology in Section 2.1 of RG 1.99, Revision 2. Therefore, the adjusted CF should be used to determine the ART. The surveillance data was determined by BWRVIP to be credible in accordance with RG 1.99, Revision 2 and verified by the NRC staff.

Bottom Head Curves Bottom head curves are utilized because the water in the vessel lower head is separated from the 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 recirculation pump loops, these regions are therefore maintained at different temperatures during non-nuclear inservice leak and hydrostatic testing and non-nuclear heatup/cooldown conditions.

The applied stress intensity factors, KI, for the bottom head curves were determined using the primary and secondary stresses from a CRD/bottom head finite element analysis that was performed by a boiling water reactor (BWR) vessel vendor in the early 1970's and a membrane stress intensity factor, Mm, based on Paragraph G-2214.1 in Appendix G to Section XI of the ASME Code. The stress analysis used commonly accepted practices and their applications are consistent with analyses performed to demonstrate conformance with Section III of the ASME Code.

The pressures and temperatures for the bottom head curves were determined using the KI methods as described in Paragraph G-2214 in Appendix G to Section XI of the ASME Code, and KIC values were calculated in accordance with Paragraph A-4200 in Appendix A of the ASME Code. In addition, an RTNDT value of 20EF for the limiting material (bottom head dollar plate 13-2-2, heat B5130-2) was used, as was an adjustment in the RTNDT for heatup/cooldown curves based on a revised finite element analysis that is described in Appendix H of the licensees June 9, 2004 submittal.

Evaluations were performed to determine whether the CRD nozzle analyzed in the earlier finite element analysis was the limiting location. The first evaluation, using specific operating conditions, showed that all other discontinuities were bounded by the CRD discontinuity. The second evaluation involved a comparison which showed that stresses in the BWR/6 vessel were higher than those in the Columbia vessel for a given crack depth. The licensee concluded that these analyses showed that the generic BWR/6 P/T limit curve, indexed to the RTNDT of the limiting bottom head material, is conservative when applied to the Columbia bottom head.

The staff concluded that the licensees evaluation for the bottom head curves are acceptable because 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 in the Columbia vessel.

Upper Vessel, Flange, and Beltline Region Curves The P/T limit curves for non-nuclear inservice leak and hydrostatic testing and non-nuclear heatup/cooldown operations 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 was developed from curves based on the material properties for the upper vessel, vessel flange, and vessel beltline regions. 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 the highest RTNDT for the materials in the beltline, upper vessel, and closure flange regions, the licensee developed P/T limit curves to meet the criteria in 10 CFR Part 50, Appendix G and Section XI of the ASME Code, Appendix G.

The upper vessel region P/T limits were based on analysis of the feedwater nozzle and beltline regions. The KI for the feedwater nozzle during pressure test conditions was computed using the methods from Welding 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 from WRC Bulletin 175 provide approximate methods 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 Appendix G to 10 CFR Part 50.

The applied stress intensity factors, KI, for the upper vessel curve during normal operation were determined using the primary and secondary stresses from a feedwater nozzle finite element analysis and a membrane stress intensity factor, Mm, based on the values identified in Appendix G to Section XI of the ASME Code for a postulated surface flaw normal to the direction of maximum stress. The pressures and temperatures for the upper vessel curve were determined using the methods described above, and 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 of the Columbia RPV. The limiting beltline material is the lower shell plate 12-1-1, heat C1272-1.

The licensee calculated the ART at the 1/4T location to be 63EF (33.1 EFPY). The critical parameters for the licensees ART determination for the 1/4T location are shown as follows:

Plate/Heat EFPY

%Cu

%Ni CF Init.

RTNDT 1/4T Fluence RTNDT Margin Shift ART Mk 21-1-1/

C1272-1 33.1 0.15 0.6 0

110 28EF 1.75E17 n/cm2 17EF 17EF 35EF 63EF The P/T limit curves apply to both heatup and cooldown and for both 1/4T and 3/4T locations because the maximum tensile stress of either heatup or cooldown is applied at the 1/4T location. For the beltline curves this approach has added conservatism because radiation effects cause the allowable KIC at 1/4T to be less than at the 3/4T for a given temperature. As a result, the 1/4T location is limiting at all temperatures. The NRC staffs assessment also included an independent calculation of the ART value for the 1/4T location of Columbias 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 beltline region contains a vessel wall thickness transition discontinuity located between the lower and lower-intermediate shells. The licensee performed an evaluation which demonstrates that the discontinuity was bounded by the beltline P/T limit curve. Therefore, the NRC staff agreed that the proposed P/T limit curves do not need any adjustment due to the higher stresses in the transition discontinuity region.

The licensee stated that the feedwater nozzle, which experiences feedwater flow that is colder, relative to the vessel coolant, was selected to represent all nozzles for fracture toughness in the upper shell region. The licensee performed an analysis, based on the methods outlined earlier, for the feedwater nozzle (N4 feedwater nozzle 56-1-3, heat Q2Q55W 786S-3). In determining the Inservice Leak and Hydrostatic Testing Curve (Curve A) and the Non-Nuclear Heating and Cooldown Curve (Curve B), the licensee determined that an addition of 34EF to the initial RTNDT of the feedwater nozzle was necessary to ensure that the feedwater nozzle analysis would be bounding for all upper shell region discontinuities during the pressure test. For the Nuclear Heating and Cooldown Curve (Curve C), the licensee used an analysis which included a feedwater injection temperature of 40EF into the vessel while at an operating temperature of 551EF to determine the stresses.

Table 1 in Appendix G to 10 CFR Part 50 establishes additional requirements for the closure flange region. The requirements are related to the RTNDT for the limiting closure flange material.

The limiting RTNDT material for the closure flange region is for upper shell plate 21-1-2, heat C1307-2 which has an RTNDT of 20EF. The NRC staff confirmed that the proposed P/T limit curves satisfy the closure flange limits in Appendix G to 10 CFR 50.

Based on the NRC staffs review and evaluation of the licensees proposed P/T limit curves for Columbia, it has been determined that the proposed P/T limit curves satisfy the requirements of: (a) 10 CFR 50.60(a), Acceptance Criteria for Fracture Prevention Measures for Light Water Nuclear Power Reactors for Normal Operation; (b) Appendix G to 10 CFR Part 50, Fracture Toughness Requirements; and (c) Appendix G to Section XI of the ASME Code for 33.1 EFPY.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Washington State official was notified of the proposed 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 facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may 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 no public comment on such finding (69 FR 53102; August 31, 2004). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

Based on the NRC staffs review and evaluation of the licensees proposed P/T limit curves for Columbia, it has been determined that the proposed P/T limit curves satisfy the requirements of: (a) 10 CFR 50.60(a), Acceptance Criteria for Fracture Prevention Measures for Light Water Nuclear Power Reactors for Normal Operation; (b) Appendix G to 10 CFR Part 50, Fracture Toughness Requirements; and (c) Appendix G to Section XI of the ASME Code for 33.1 EFPY.

On the basis of the above regulatory and technical evaluations of the licensees justifications for Technical Specification changes, the NRC staff has concluded that the licensees proposed Technical Specification changes are acceptable.

Principal Contributors: J. Terrell L. Lois Date: May 12, 2005