GO2-04-164, License Amendment Request to Revise the Reactor Pressure Vessel Material Surveillance Program

License Amendment Request to Revise the Reactor Pressure Vessel Material Surveillance Program

ML042880448
Person / Time
Site:	Columbia
Issue date:	09/23/2004
From:	Atkinson D Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GO2-04-164
Download: ML042880448 (24)

Text

ENERGY NJ NORTHWEST People *Vision

• Solutions P.O. Box 968

• Richland, WA

• 99352-0968 September 23, 2004 G02-04-164 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397; LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM

References:

(1) Letter G02-04-032 dated March 5, 2004, from DK Atkinson (Energy Northwest) to U.S. Nuclear Regulatory Commission, "Schedule for Requesting Revision of Technical Specification P/T Curves and Adoption of the BWRVIP Integrated Surveillance Program" (2) Letter dated February 1, 2002, from Mr. William Bateman (NRC) to Mr.

Carl Terry (BWRVIP Chairman), "Safety Evaluation Regarding EPRI Proprietary Reports "BWR Vessels and Internals Project, BWR Integrated Surveillance Program Plan (BWRVIP-78)" and "BWRVIP-86: BWR Vessels and Internals Project, BWR Integrated Surveillance Program Implementation Plan" (3) NRC Regulatory Issue Summary 2002-05, "NRC Approval of Boiling Water Reactor Pressure Vessel Integrity Surveillance Program," dated April 8, 2002.

(4) Letter G02-04-107 dated June 9, 2004, from DK Atkinson (Energy Northwest) to NRC, "License Amendment Request to Revise Technical Specification 3.4.11, Reactor Coolant System (RCS)

Pressure/Temperature (P/T) Limits"

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Energy Northwest hereby requests an amendment to the Columbia Generating Station (Columbia) Operating License NPF-21. The proposed amendment would revise Columbia's licensing basis by replacing the current plant-specific Reactor Pressure Vessel (RPV) material surveillance program with the Boiling Water Reactor Vessels and Internals Project (BWRVIP) Integrated Surveillance Program (ISP). Specifically, the proposed amendment would revise Columbia's Final Safety Analysis Report (FSAR) to include participation in the ISP as described in the program document BWRVIP-86-A, "BWR Vessel and Internals Project Updated BWR Integrated Surveillance Program (ISP) Implementation Plan," dated October 2002.

fV

i LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM Page 2 of 3 A condition for participating inthe ISP program set forth in the NRC Safety Evaluation approving the BWRVIP ISP (Reference 2) is that each licensee is to provide information, with their license amendment request, regarding the specific neutron fluence methodology to be implemented for their facility. The information submitted must be sufficient to determine that the methodology is NRC approved and that neutron fluence methodology compatibility is addressed as it applies to the comparison of neutron fluences calculated for the RPV versus the neutron fluences calculated for surveillance capsules in the ISP, which are designated to represent the RPV. A license amendment request (Reference 4) was previously submitted to the NRC for review and approval that will implement a neutron fluence methodology that satisfies this condition. Compatibility of the fluence methodology being implemented by Energy Northwest is addressed in paragraph 4.2 of Attachment 1 to this letter.

The NRC has issued a Safety Evaluation (SE) (Reference 2) approving the BWRVIP ISP as an acceptable alternative to all existing BWR plant-specific RPV surveillance programs for the purpose of maintaining compliance with 10 CFR 50 Appendix H, "Reactor Vessel Material Surveillance Program Requirements," through the end of current facility 40-year operating licenses. On April 8, 2002 the NRC issued Regulatory Issue Summary (RIS) 2002-05, UNRC Approval of Boiling Water Reactor Pressure Vessel Integrated Surveillance Program," (Reference 3). RIS 2005-05 states that licensees who elect to participate in the ISP shall submit a license amendment request to incorporate this program into their licensing basis. This license amendment request is submitted in accordance with the guidance contained in References 2 and 3.

Similar requests have been approved for the AmerGen Energy Company's Clinton Power Station, Unit 1, by NRC letter dated August 12, 2003 (TAC No. MB6998), and for Exelon Generation Company's Quad Cities Nuclear Power Station, Units 1 and 2, by NRC letter dated August 28, 2003 (TAC Nos. MB7008 and MB7009).

Energy Northwest has evaluated the proposed amendment pursuant to the criteria of 10 CFR 50.92(c) and has determined the proposed amendment warrants a no significant hazards consideration.

Energy Northwest requests approval of the proposed amendment by October 2005.

Energy Northwest also requests a 60-day implementation period upon approval of this request. There are no new commitments associated with this submittal.

LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM Page 3 of 3 If you have any questions or require additional information regarding this matter, please contact Mr. DW Coleman, Regulatory Programs Manager, at (509) 377-4342.

Respectfully, DK Atkinson Vice President, Technical Services Mail Drop PE08 Attachments:

1. Evaluation of the Proposed Amendment

2. Marked-up Affected Pages from the Columbia Generating Station Final Safety Analysis Report

3. List of Regulatory Commitments cc: BS Mallet - NRC - RIV RN Sherman - BPA/1399 WA Macon - NRC - NRR TC Poindexter - Winston & Strawn NRC Sr. Resident Inspector - 988C JO Luce - EFSEC RR Cowley - WDOH

STATE OF WASHINGTON)

Subject:

Request for Amendment, Final Safety Analysis Report Methodology Change COUNTY OF BENTON )

1,DK Atkinson, being duly sworn, subscribe to and say that I am the Vice President, Technical Services, for ENERGY NORTHWEST, the applicant herein; that I have the full authority to execute this oath; that I have reviewed the foregoing; and that to the best of my knowledge, information, and belief that the statements made in it are true.

DAT E Z3, 2004 DK Atkinson Vice President, Technical Services On this date personally appeared before me DK Atkinson, to me known to be the individual who executed the foregoing instrument, and acknowledged that he signed the same as his free act and deed for the uses and purposes herein mentioned.

GIVEN under my hand and seal this 2_Z day of St9kanz 2004 zzm d.. i'a Notary ublic in and for the STATE OF WASHINGTON Residing at 664&' id v . .

My Commission expires __z____

LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM Attachment I Page 1 of 7 EVALUATION OF THE PROPOSED AMENDMENT

1.0 DESCRIPTION

This submittal is a request to amend Operating License NPF-21 for Columbia Generating Station (Columbia). The proposed amendment would revise Columbia's licensing basis by replacing the current plant-specific Reactor Pressure Vessel (RPV) material surveillance program with the Boiling Water Reactor Vessels and Internals (BWRVIP) Integrated Surveillance Program (ISP). Specifically, the proposed amendment would revise Columbia's Final Safety Analysis Report (FSAR) to include participation in the ISP as described in the program document BWRVIP-86-A, "BWR Vessel and Internals Project Updated BWR Integrated Surveillance Program (ISP)

Implementation Plan," dated October 2002.

A condition for participating in the ISP program set forth in the NRC Safety Evaluation approving the BWRVIP ISP (Reference 2) isthat each licensee is to provide information, with their license amendment request regarding the specific neutron fluence methodology to be implemented for their facility. The information submitted must be sufficient to determine that the methodology is NRC approved and that neutron fluence methodology compatibility is addressed as it applies to the comparison of neutron fluences calculated for the RPV versus the neutron fluences calculated for surveillance capsules in the ISP, which are designated to represent the RPV. A license amendment request (Reference 3) was previously submitted to the NRC for review and approval that will implement a neutron fluence methodology consistent with Regulatory Guide (RG) 1.190 (Reference 5). Use of this methodology satisfies this condition. Compatibility of the fluence methodology being implemented by Energy Northwest is addressed in paragraph 4.2 of Attachment 1 to this letter.

2.0 PROPOSED CHANGE

The Columbia Generating Station FSAR, Section 5.3.1.6, "Material Surveillance,"

describes the current plant-specific RPV material surveillance program. A proposed revision to this section is provided in Attachment 2 to document Columbia's adoption of the BWRVIP ISP and to incorporate BWRVIP-86-A. Excerpts from other sections of Columbia's FSAR related to the methodology for determination of RPV neutron fluence that are being revised are also provided in Attachment 2. Following NRC approval of this license amendment request, the FSAR will be updated to incorporate the changes associated with implementation of the ISP program identified in Attachment 2 in accordance with 10 CFR 50.71(e).

LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM Attachment I Page 2 of 7

3.0 BACKGROUND

10 CFR 50, Appendix H requires monitoring the beltline regions of RPVs with a material surveillance program that complies with the American Society for Testing and Materials (ASTM) E 185, except as modified by Appendix H. The ASTM protocol provides guidelines for designing a minimum surveillance program, selecting materials, and evaluating test results for light-water cooled nuclear power reactors. It also provides recommendations for the minimum number of surveillance capsules and their withdrawal schedule. 10 CFR 50, Appendix H requires that the proposed surveillance capsule withdrawal schedule be submitted to and approved by the NRC prior to implementation.

Energy Northwest's RPV material surveillance program for Columbia was developed in accordance with 10 CFR 50, Appendix H. The program is described in Columbia's FSAR Section 5.3.1.6, "Material Surveillance." The current Columbia RPV surveillance capsule withdrawal schedule is contained in FSAR Table 5.3-8.

The BWRVIP ISP was developed in response to an issue regarding the potential lack of adequate unirradiated baseline Charpy V-notch (CVN) data for one or more materials in plant-specific RPV surveillance programs at several Boiling Water Reactors (BWRs).

This lack of baseline properties would inhibit licensees ability to effectively monitor changes in fracture toughness properties of RPV materials in accordance with 10 CFR 50, Appendix H. The BWRVIP ISP, as approved by the NRC in Reference 1, resolves this issue.

Implementation of the BWRVIP ISP will provide additional benefits. When the original surveillance materials were selected for plant-specific surveillance programs, the state of knowledge concerning RPV material response to irradiation, post-irradiation and subsequent effects on fracture toughness was different than it is today. As a result, many utilities did not include what would be identified today as the plant's limiting RPV materials in their surveillance programs. The BWRVIP effort to identify and evaluate materials from other BWRs, which may better represent a facility's limiting materials, should improve the overall evaluation of BWR RPV embrittlement. Also, the inclusion of testing of Boiling Water Reactor Owners Group (BWROG) Supplemental Surveillance Program (SSP) capsules will improve the overall quality of the data being used to evaluate BWR RPV embrittlement.

The benefits of implementing the ISP also include the following:

• Costs, occupational radiation exposure and outage times of the BWR fleet will be reduced due to elimination of the need for some units (including Columbia) to remove surveillance material specimens, and;

• Implementation of the ISP is expected to reduce the cost of surveillance testing and analysis because materials that are of little or no value will no longer be tested.

LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM Attachment I Page 3 of 7

4.0 TECHNICAL ANALYSIS

BWRVIP-86-A (Reference 2) describes the implementation plan for an ISP to support operation of all domestic BWR RPVs through the completion of each facility's current 40-year operating license that complies with 10 CFR 50, Appendix H. In a Safety Evaluation (SE) dated February 1, 2002 (Reference 1), the NRC concluded that the ISP proposed by the BWRVIP, if implemented in accordance with specific conditions, is an acceptable alternative to existing BWR plant-specific RPV surveillance programs. The NRC SE requires that each licensee electing to participate in the ISP provide the following:

1. Information regarding what specific neutron fluence methodology will be implemented as part of participation in the ISP, and;

2. Address neutron fluence methodology compatibility as it applies to the comparison of neutron fluences calculated for its RPV versus the neutron fluences calculated for surveillance capsules in the ISP which are designated to represent its RPV.

4.1 Fluence Methodology Energy Northwest has used the NRC-approved General Electric NEDC 32983P-A, "General Electric Methodology for Reactor Pressure Fast Neutron Flux Evaluations," to calculate the most recent fluence values for Columbia. The methodology is in accordance with the recommendations of Regulatory Guide 1.190 (Reference 5) and was approved by the NRC in a letter dated October 27, 2003 (Reference 4). This methodology was utilized to support a proposed revision to Columbia's RPV pressure-temperature limit curves that was submitted to the NRC for review and approval by letter dated June 9, 2004 (Reference 3). Columbia's FSAR is being revised, as shown in Attachment 2, to reflect that an NRC approved fluence methodology will be used which conforms with RG 1.190. Use of an NRC approved fluence methodology satisfies the first condition contained within the NRC SE (Reference 1).

4.2 Fluence Methodology Compatibility At an August 29, 2002 workshop regarding the establishment and implementation of the BWRVIP RPV Integrated Surveillance Program, the NRC staff stated that neutron fluence methodology compatibility is satisfied if the surveillance capsules and the RPVs are evaluated with an NRC approved methodology that complies with RG 1.190.

BWRVIP-86-A requires the evaluation of ISP capsule fluences to be performed using a methodology that is consistent with the guidance of RG 1.190. Columbia's FSAR is being revised, as shown in Attachment 2, to include the requirement to use an NRC-

LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM Page 4 of 7 approved fluence methodology that is consistent with RG 1.190. This satisfies the second condition contained within the NRC SE (Reference 1).

In accordance with the current plant-specific RPV material surveillance program, Columbia's first surveillance capsule has been withdrawn and tested. Under the BWRVIP ISP, Columbia is not identified as a host plant. The representative materials for Columbia's limiting RPV plate and weld materials, and their associated withdrawal schedules, are identified in Reference 2. Thus, in accordance with the ISP, future withdrawal and testing of Columbia's surveillance capsules will be permanently deferred.

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Determination Energy Northwest is proposing to revise the licensing basis for Columbia Generating Station by replacing the current plant-specific RPV material surveillance program with the BWRVIP ISP. This change is acceptable because implementation of the proposed ISP at Columbia meets the criteria specified in 10 CFR 50, Appendix H, Paragraph IIl.C, "Requirements for an Integrated Surveillance Program."

In accordance with 10 CFR 50.92(c), a proposed change to the operating license involves a no significant hazards consideration if operation of the facility in accordance with the proposed change would not: 1) involve a significant increase in the probability or consequences of any accident previously evaluated; 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety. Energy Northwest has evaluated the proposed changes to the Columbia Generating Station FSAR using the three criteria set forth in 10 CFR 50.92(c) and has determined that they warrant a no significant hazards consideration as described below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change implements an ISP program that meets the requirements of 10 CFR 50, Appendix H, Paragraph III.C, "Requirements for an Integrated Surveillance Program." The proposed ISP program ensures the same level of RPV integrity as Columbia's current material surveillance program.

Implementation of the proposed ISP is not a precursor or initiator of any previously evaluated accident. No physical changes to Columbia Generating Station are involved with the proposed change. The proposed change will not cause the RPV or interfacing systems to be operated outside of any design limit

LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM Attachment I Page 5 of 7 or testing limit, and will not alter any assumptions or initial conditions previously used in evaluating the radiological consequences of an accident.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibilityof a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change revises the licensing basis for Columbia Generating Station to reflect participation in the BWRVIP ISP. The NRC has approved the ISP as an acceptable material surveillance program pursuant to 10 CFR 50, Appendix H, paragraph III.C. No physical changes to the plant are associated with the proposed change. No changes in design or operation of any system, structure, or component will be made as a result of the proposed change. The ISP is an alternative monitoring program and cannot create a new failure mode or a new or different kind of accident from any previously evaluated.

Therefore, the proposed change does create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in the margin of safety?

Response: No Compliance with RPV material surveillance program requirements specified in 10 CFR 50, Appendix H and the fracture toughness requirements contained in 10 CFR 50, Appendix G ensures an adequate margin of safety exists in the fracture toughness of RPV beltline ferritic materials during any condition of normal operation, anticipated operational occurrence, and system hydrostatic tests.

Implementation of the proposed ISP has been evaluated to meet the requirements of 10 CFR 50, Appendix H and this margin of safety is not impacted. Compliance with the requirements of 10 CFR 50, Appendix G will not be affected by this proposed change.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

In conclusion, based on the considerations discussed above; (1) there is a reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations, and; (3)the issuance of the

LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM Page 6 of 7 amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.2 Applicable Regulatory Requirements 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for light water nuclear power reactors for normal operation,' requires that all light water power reactors, with certain exceptions, must meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in 10 CFR 50 Appendices G and H. 10 CFR 50 Appendix G, "Fracture Toughness Requirements," specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary, including RPVs. 10 CFR 50 Appendix H, "Reactor Vessel Material Surveillance Program Requirements," requires licensees to implement an RPV material surveillance program in order to monitor changes in the fracture toughness properties in the reactor beltline region which result from exposure of these materials to neutron irradiation and the thermal environment.

10 CFR 50 Appendix H, Paragraph III.C, "Requirements for an Integrated Surveillance Program," provides specific criteria upon which approval of an ISP shall be based. An ISP is an alternative method to a plant specific material surveillance program. Appendix H Paragraph III.C states that in an ISP, "the representative materials chosen for surveillance for a reactor are irradiated in one or more other reactors that have similar design and operating features." In Reference 1 the NRC documented that the BWRVIP ISP met the criteria specified in Appendix H, Paragraph III.C provided that all licensees use one or more compatible neutron fluence methodologies acceptable to the NRC staff to determine capsule and RPV neutron fluences. Inaddition, the NRC required a plant-specific license amendment to be submitted by each licensee wishing to adopt the ISP confirming their incorporation of the ISP into their licensing basis.

Conformance with the NRC General Design Criteria (GDC) for Nuclear Power Plants, Appendix A, to 10 CFR 50, is described in Section 3.1 of Columbia's FSAR. In particular, GDC 31, "Fracture Prevention of Reactor Coolant Pressure Boundary,"

requires, in part, that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions, the boundary will behave in a non-brittle manner and the probability of a rapidly propagating fracture is minimized. Adoption of the ISP as described herein does not conflict with Columbia's FSAR statement of conformance with GDC 31.

Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and, (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM Page 7 of 7

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve; (i) a significant hazards consideration; (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite; or, (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. Letter from Mr. William Bateman, (NRC), to Mr. Carl Terry (BWRVIP Chairman),

"Safety Evaluation Regarding EPRI Proprietary Reports "BWR Vessels and Internals Project, BWR Integrated Surveillance Program Plan (BWRVIP-78)" and "BWRVIP-86: BWR Vessels and Internals Project, BWR Integrated Surveillance Program Implementation Plan," dated February 1, 2002.

2. BWRVIP-86-A, "BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan," Final Report, October 2002.

3. Letter (GO2-04-107) from DK Atkinson (Energy Northwest) to NRC, "License Amendment Request to Revise Technical Specification 3.4.11 Reactor Coolant System (RCS) Pressure/Temperature (P/T) Limits," dated June 9, 2004.

4. Letter from SA Richards (USNRC) to JF Klapproth, "Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891)," MFN 01-050, dated September 14, 2001.

5. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

6. NRC Regulatory Issue Summary 2002-005, "NRC Approval of Boiling Water Reactor Pressure Vessel Integrated Surveillance Program," dated April 8, 2002.

LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM Page 1 of 12 Marked-up Affected Pages from the Columbia Final Safety Analysis Report 1.) LDCN-FSAR-04-033; Marked-up FSAR pages FSAR Page FSAR SectionlTable Title 5.3-6 5.3.1.6 Materials Surveillance 5.3-7 5.3.1.6 Materials Surveillance (con't) 5.3.1.6.1 Positioning of Surveillance Capsules 5.3-17 5.3.3.7 Inservice Surveillance 5.3.4 References 5.3-21 Table 5.3-1 10 CFR 50 Appendix G Matrix (con't) 5.3-31 Table 5.3-8 Capsule Withdrawal Schedule 5.3-32 Table 5.3-8 10 CFR 50 Appendix H Matrix 5.3-33 Table 5.3-8 10 CFR 50 Appendix H Matrix (con't) 2.) LDCN-FSAR-04-005; Excerpts of FSAR Changes FSAR Section/Table Title 4.3.2.8 Vessel Irradiations 4.3.4 References 1.6 Material Incorporated by Reference Table 1.6-1 Topical Reports 1.8 Conformance to NRC Regulatory Guides

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 curve A shown in Figure 5.3-1. The predicted shift in the RTNDT temperature was determined using the methodology outlined in Regulatory Guide 1.99, Revision 2.

Technical Specification 3.10.1 allows inservice leak and hydrostatic testing to be performed in Mode 4 when the metallurgical characteristics of the reactor pressure vessel require testing at temperatures greater than 200'F, given specified Mode 3 Limiting Conditions for Operations are met. This exemption is only applicable provided reactor coolant temperature does not exceed 2750 F.

5.3.1.5.2.5 Operating Limits During Heatup, Cooldown, and Core Operation. The fracture toughness analysis was done for the normal heatup or cooldown rate of 1000 F/hr. The temperature gradients and thermal stress effects corresponding to this rate were included. The results of the analysis are operating limits defined by Figure 5.3-1. Curves A, B, and C give the limits for hydrotest, nonnuclear heating, and nuclear heating. The minimum boltup temperature of 80'F is based on an RTNDT at 20'F for a shell plate which connects to the lower flange (Heat and Slab No. C-1307-2); above 80'F the core beltline plate (Heat and Slab No.

C-1272-1), which has an initial RTNDTof 28 0 F, is most limiting for inservice h ic or leak pressure tests (curve A). The feedwater nozzles, which have an RTNDT Fl re more restrictive than the core beltline at lower pressures during nonnuclear and nPmg Ap:A YA.EWD IW (curves B and C). 0F c.- ArA-4--coS 4 4I,oa-c+-- ol 5.3.1.5.2.6 Reactor Vessel Annealing. Inplace annealing of the reactor vessel to counteract radiation embrittlement is unnecessary because beitline material adjusted reference temperature of the NDT is well within the 10 CFR 50 Appendix G 200'F screening limit.

5.3.1.6 Material Surveillance The materials surveillance program monitors changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from exposure to neutron f513-01 i-nA;intin~n andr thprrnml -ntirnnrent_ _

.C r& L -r--

r§8 IC- M~-EN ALS 5u 'dEIILA-t.c-Materials for It IsenI: reactor beltline region. The specimens ed from a plate actually used in the beltline region and a weld typical of those in region and thus represent base metal, weld metal, and the 0 transition zone between base metal and weld. The plate and wel e eated in a manner which simulates the actual heat treatment perf on the core re ates of the C mleteddvessel. - r-os9 (Traw e S 51 +4--IpnDvia~ A Djf

!(UTDIL AuL OuPpcW.i1m iSN .tAT101o L _ _

FS -4E MATMALs 5uP-VEU.A sCZ PP 4itA-i 5.3-6

A Insert for LDCN-FSAR-04-033 Mark-ups (for paragraph 5.3.1.6):

5.3-01 Insert:

The CGS plant-specific RPV materials surveillance program is replaced by the NRC approved BWR Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP), as described in BWRVIP-86-A (Reference 5.3.4-2). The NRC approved the ISP for the industry in their safety evaluation dated February 1, 2002 (Reference 5.3.4-3). The ISP meets the requirements of 10 CFR 50, Appendix H.

The current surveillance capsule withdrawal schedule for the representative materials for the CGS vessel is based on the latest approved version of BWRVIP-86-A (Reference 5.3.4-2). No capsules from the CGS vessel are included in the ISP. The withdrawal of capsules for the CGS plant-specific surveillance program is permanently deferred by participation in the ISP. Capsules from other plants will be removed and tested in accordance with the ISP implementation plan. The results from these tests will provide the necessary data to monitor embrittlement for the CGS vessel.

.- Ip COLUMBIA GENERATING STATION .

FINAL SAFET ANALYSIS REPORT

)

addressed the requirements of Appendix H to 10 CFR 50 and s ports te current application t

of Regulatory Guide 1.99. -oz 5.3.1.6.1 Positio *n CCf ueandMethodIoAt ent F rA

/ -r- 5par-t _ .- LA-#-&CE 4 Surveilnc seien capsules are located at three azimuths at a common elevation in the core be/tline region. The sealed capsules are not attached to the vessel but are in welded capsule holders. The capsule holders are mechanically restrained by capsule holder brackets as shown I in Figure 5.3-4. The capsule holder brackets allow the capsule holder to be removed at any desired time in the life of the plant for specimen testing. A positive spring-loaded locking device is provided to retain the capsules in position throughout any anticipated event during the lifetime of the vessel.

The capsule holder brackets are designed, fabricated, and analyzed to the requirements of the ASME B&PV Code Section III. The surveillance brackets are welded to the clad material which surfaces the pressure vessel walls. As attached, the brackets do not have to comply with specifications of the ASME Code.

LDCN-99-086 5.3-7

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 5.3.3.7 Inservice Surveillance Inservice inspection of the reactor pressure vessel is in accordance with the requirements as discussed in Section 5.2.4. The vessel was examined once prior to startup to satisfy the preoperational requirements of IS-232 or the ASME Code,Section XI. Subsequent inservice inspection will be scheduled and performed in accorda nce wi ents of53- 06 10 CFR 50.55a subparagraph (g). {5 eruo 5 3. 1., Fop, Desc*4P-notJ oF=

5ea THce lM~f-llktS SUKVEILL"C PF-4P4M.

The materials surveillance prog monitors changes in the fracture toughness properties of ferritic materials in the reactor ye 1 beltline region resulting from their exposure to neutron irradiation and thermal environment.. _ = e n th recto veselandperodicllyforImpct estng perting procedures -will be modi fiein accordance with test results to ensure adequate brittle fracture control.

Material surveillance programs and inservice inspection programs are in accordance with applicable ASME Code requirements and provide assurance that brittle fracture control and pressure vessel integrity will be maintained throughout the service lifetime of the reactor pressure vessel.

?,EFEREwl CG7F5o SM=i7 -? 5--

5.3-17

Insert for LDCN-FSAR-04-033 Mark-ups:

5.3-07 Insert (new paragraph):

5.

3.4 REFERENCES

5.3.4-1 WPPSS-ENT-089, "WNP-2 RPV Surveillance Program", Current Revision.

5.3.4-2 BWRVIP-86-A, "BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan," Final Report, October 2002.

5.3.4-3 Letter from U.S. NRC to C. Terry (BWRVIP), "Safety Evaluation Regarding EPRI Proprietary Reports 'BWR Vessel and Internals Project, BWR Integrated Surveillance Program Plan (BWRVIP-78)' and

'BWRVIP-86: BWR Vessel and Internals Project, BWR Integrated Surveillance Program Implementation Plan,"' dated February 1, 2002.

Table 5.3-1 10 CFR 50 Appendix G Matrix (Continued)

~'eze~,~ L~w-*,4-o4 -cO*.

Comply 7 z 4 I Appendix G Yes/No

• 1 43 Paragraph Topic or N/A Alt e Actions or Comments

/ 2 . 913 lV.B (continued) Beltline plates were teste wienf hujnal CVNs at + 100 F only. The minimum values are for eafrC127(1 15% Cu; 34, 26, 30, 31, 34, 30 ft-lb; 10 and 40 % she at Heat C1273- 1 (0.14% Cu; n0 33, 33, 30, 30, 34, 35 ft-Rb; shear at + 10F). Beltline welds were tested with CVNs at IOF or -20 y. Lowest5 ues are found for Heat 04P046/Lot D217A27A (0. Cu;: , 39, 40 ft-lb; ;D, 20 and 30% shear at -20°F). Heat C3 t(

LA 35, 39, 40 ft-lb; 60% shear at +10°F) and He D211A27A z (0.09% Cu; 29, 30, 31, 36, 38 ft-lb; 30 and 40% shear at -20°F). VW Because of the preceding relatively low test temperatures and Cu c,

contents, it is anticipated that end-of-life upper shelf CVN values would be in excess of 50 ft-lb. z~

IV.C Requirements for Annealing when RTm > 200 N/A 53 O 0

V.A Requirements for Material Surveillance Program V.B Conditions for Continued Operation V.C Alternative if V.B Cannot be Satisfied

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 6.13- 10

/ /I l

/ Ta5.- /

//Ca ~hrawae/ule I

/ Ve~~svel Azimut./ Withd~rawpfime' I Capsule Locati/ Lead Far IC 3 to symmetry 1capsul are expected 8 have th lead 1 1)b/ 0V=I X t,4 2

y fCh nges this schedujeo withd~ga of surveilla epeimens car e made withog prior approval 10CR5 ppdxH erqie etion .3, ad Gee Lett 9- I and t aF ty E tiation ReportR for License brindmn10,dae J e 15, 1992. accorda e with the SE once this approva as been obtaine is table ust be upd d. This rticular sched was approved by NRC to Energ orthwest letter dat luly 21,93////

b psule two ated at 1200 zuth had fallen of nor to the May ,1989, in-ye v

xaminatio . The 120° sule was last seen i osition during th ASME Sectio I Preserv' Exam'inatio PSI). This capsule s subsequently rstalled dunn e 1991 outa Since this psule missed some ea in life exposur the 1200 caps was placedi standby mo rather than being pull at 24 EFPY.

' Capsule located at 3000 azirnut was withdrawn f testing in Ap of 1996. The speci were tested and reco tituted, then the psule was rein lied in May o 1997.

This, psule is now consider to be in standby LDCN-99-CS6.99-000 5.3-3 1

Table(3) 10 CFR 50 Appendix H Matrix Comply Appendix H Ycs/No Paragraph Topic or N/A Alternativc Actions or Comments I Introduction II.A Fluence 10'7 n/cm t II.B Standards Requirements (ASTM) for Surveillance LA 'g,3-12 ,

Cj~

wL II.C.1 Surveillance Specimen Shall be Taken for Locations Alongside the Fracture Test Specimens (Section M.B of Appendix G)

II.C.2 Locations of Surveillance Capsules l{ )

H.C.3.a Withdrawal Schedule of Capsules, RTNDT < 100F lI.C.3.b Withdrawal Schedule of Capsules, RTD= <2001F N/A II.C.3.c Withdrawal Schedule of Capsules, RTmar >200°F N/A ° B B B El '.

I .,

6.13-11 Table8 10 CFR 50 Appendix H Matrix (Continued)

Appendix H Comply Appendix H YeslNo Paragraph .

Topic or N/A Altei native Actions or Comments III.A Fracture Toughness Testing Requirements of Yes Requirements for postirr Specimens 0n Method of Determining Adjusted Reference Yes _

Temperature for Base Metal, HAZ, and Weld Is- ;P.

Metal 08 92 t Reporting Requirements of Test Results Yes Reportin reuiements o

Requirement for Dosimetry Measurement Yes -Mgmtry reiremen uz

I 0

Reporting Requirements of Pressure/Temperature Yes ussion o pres Linits reouirements is found in z CZ >

B 3z 00 0 a "o W 2.) LDCN-FSAR-04-005; Excerpts of Changes Excerpts of changes pertaining to RPV fluence calculation methodology. Note:

These changes are on hold pending approval of a proposed revision to Columbia's RPV pressure-temperature limit curves that was submitted to the NRC by letter dated June 9, 2004 (Reference 3 on Attachment 1).

Changes to Section 4.3.2.8 Vessel Irradiations Replace Section:

(First paragraph of replaced section provided below)

The reactor pressure vessel (RPV) irradiation calculation provides a best-estimate prediction of the fluence rather than a conservative prediction as was the case with earlier methods. The methodology for neutron flux calculation conforms to Licensing Topical Report (LTR) NEDC-32983P-A (Reference 4.3-13). In general, the methodology described in the LTR adheres to the guidance in Regulatory Guide 1.190 for neutron flux evaluation and was approved by the U.S. NRC in the Safety Evaluation Report (SER) for referencing in licensing actions.

Changes to 4.

3.4 REFERENCES

Add:

4.3-13 GE Nuclear Energy, "Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations", NEDC-32983P-A, December 2001.

Changes to Section 1.6 Material Incorporated by Reference Table 1.6-1 Topical Reports Add:

NEDC-32983P-A General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations (December 2001)

Changes to Section 1.8 CONFORMANCE TO NRC REGULATORY GUIDES 1.8.2 NUCLEAR STEAM SUPPLY SYSTEM SCOPE OF SUPPLY EVALUATION Add:

Regulatory Guide 1.190, Revision 0. March 2001 Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence Regulatorv Guide Intent:

This regulatory guide has been developed to provide state-of-the-art calculations and measurement procedures that are acceptable to the NRC staff for determining pressure vessel fluence.

Application Assessment:

Assessed capability in design.

Compliance or Alternative Approach Statement:

The methodology for neutron flux calculation for the CGS reactor vessel conforms to Licensing Topical Report (LTR) NEDC-32983P-A. In general, the methodology described in the LTR adheres to the guidance in Regulatory Guide 1.190 for neutron flux evaluation and was approved by the U.S. NRC in the Safety Evaluation Report (SER) for referencing in licensing actions.

General Compliance or Alternate Assessment:

Reference compliance assessment for Regulatory Guide 1.99.

Specific Evaluation

Reference:

See 4.3.2.8.

Similar Application

Reference:

Similar application is used for Browns Ferry Nuclear Plant, Units 2 and 3, reactor vessels.

LICENSE AMENDMENT REQUEST TO REVISE THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM Page 1 of 1 List of Regulatory Commitments Regulatory Commitment Due Date None NIA