ML051180046

License Renewal Application FSER Sections on the Aging Management Reviews for the Reactor Vessels and Their Internal Components

ML051180046
Person / Time
Site:	Browns Ferry
Issue date:	04/19/2005
From:	Matthew Mitchell NRC/NRR/DE/EMCB
To:	Samson Lee NRC/NRR/DRIP/RLEP
References
TAC MC1704, TAC MC1705, TAC MC1706
Download: ML051180046 (27)

Text

April 19, 2005 MEMORANDUM TO: Samson S. Lee, Chief Safety Section License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs FROM:

Matthew A. Mitchell, Chief ( /RA by MAMitchell )

Vessels and Internals Integrity and Welding Section Materials and Chemical Engineering Branch Division of Engineering

SUBJECT:

BROWNS FERRY NUCLEAR LICENSE RENEWAL APPLICATION FINAL SER SECTIONS ON THE AGING MANAGEMENT REVIEWS FOR THE REACTOR VESSELS AND THEIR INTERNAL COMPONENTS (TAC NOS. MC1704, MC1705, AND MC1706)

Tables 3.1.2.1, 3.1.2.2, 3.1.2.3, and 3.1.2.4 of the Browns Ferry Nuclear license renewal application (BFN LRA) includes Tennessee Valley Authority (TVA or the applicant) aging management reviews (AMRs) for the reactor vessels and the reactor vessel internals of the BFN, Units 1, 2, and 3. The staff has completed their review of these AMR items, as well as TVAs responses to requests for additional information (RAIs) that were issued on these AMR items. The staff has determined that TVA has sufficiently addressed the issues (with the exception of the open items) raised by the staff in the RAIs and therefore concludes that the AMR items in Tables 3.1.2.1, 3.1.2.2, 3.1.2.3, and 3.1.2.4 of the BFN LRA are acceptable pending the resolution of the open items. The attachment provides the staffs Final SER (FSER) section on these AMR items.

Docket Nos.: 50-259, 50-260, 50-296

Attachment:

As stated CONTACT:

Ganesh Cheruvenki, DE/EMCB (301) 415-2501

April 19, 2005 MEMORANDUM TO: Samson S. Lee, Chief Safety Section License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs FROM:

Matthew A. Mitchell, Chief ( /RA by MAMitchell )

Vessels and Internals Integrity and Welding Section Materials and Chemical Engineering Branch Division of Engineering

SUBJECT:

BROWNS FERRY NUCLEAR LICENSE RENEWAL APPLICATION FINAL SER SECTIONS ON THE AGING MANAGEMENT REVIEWS FOR THE REACTOR VESSELS AND THEIR INTERNAL COMPONENTS (TAC NOS. MC1704, MC1705, AND MC1706)

Tables 3.1.2.1, 3.1.2.2, 3.1.2.3, and 3.1.2.4 of the Browns Ferry Nuclear license renewal application (BFN LRA) includes Tennessee Valley Authority (TVA or the applicant) aging management reviews (AMRs) for the reactor vessels and the reactor vessel internals of the BFN, Units 1, 2, and 3. The staff has completed their review of these AMR items, as well as TVAs responses to requests for additional information (RAIs) that were issued on these AMR items. The staff has determined that TVA has sufficiently addressed the issues (with the exception of the open items) raised by the staff in the RAIs and therefore concludes that the AMR items in Tables 3.1.2.1, 3.1.2.2, 3.1.2.3, and 3.1.2.4 of the BFN LRA are acceptable pending the resolution of the open items. The attachment provides the staffs Final SER (FSER) section on these AMR items.

Docket Nos.: 50-259, 50-260, 50-296

Attachment:

As stated DISTRIBUTION:

EMCB RF YSanabria, RLEP RidsNrrDe ADAMS ACCESSION No.: ML051180046 DOCUMENT NAME: E:\\Filenet\\ML051180046.wpd INDICATE IN BOX: C=COPY W/O ATTACHMENT/ENCLOSURE, E=COPY W/ATT/ENCL, N=NO COPY OFFICE EMCB:DE EMCB:DE SC:EMCB:DE NAME GCheruvenki BElliot MAMitchell DATE 04/07/05 04/15/05 04/19/05 OFFICIAL RECORD COPY

ATTACHMENT FINAL SAFETY EVALUATION REPORT FOR THE BROWNS FERRY LICENSE RENEWAL APPLICATION EVALUATION OF PLANT-SPECIFIC AGING MANAGEMENT REVIEW (AMR) ITEMS FOR THE REACTOR VESSEL AND REACTOR VESSEL INTERNAL COMPONENTS EVALUATION OF AMR ITEMS DESIGNATED AS BFN IN LRA TABLES 3.1.2.1, 3.1.2.2, 3.1.2.3 AND 3.1.2.4 DOCKET NOS. 50-259, 50-260, 50-296 3.1.3. Aging Management Evaluations in the Generic Aging Lessons Learned (GALL) Report that Are Relied on for License Renewal, For Which GALL Recommends Further Evaluation For component groups evaluated in GALL for which the applicant has claimed consistency with GALL, and for which GALL recommends further evaluation, the staff reviewed the applicants evaluation to determine whether it adequately addressed the issues for which GALL recommended further evaluation.

3.1.3.1 Reactor Vessel, Internals and Reactor Coolant System Summary of Technical Information in the Application.

The applicants specific aging management reviews (AMRs) for components that comprise the Browns Ferry Nuclear (BFN) reactor vessels (RVs), and reactor vessel internals (RVIs) are given in Tables 3.1.2.1 and 3.1.2.2 of the LRA, respectively. These components include the RV attachment welds, other external attachment welds, reactor closure studs and nuts, RV heads, flanges, and shells, RV nozzles and safe ends, RV penetrations, RV internals core shroud and core plate, RV internals core spray lines and spargers, RV internals dry tubes and guide tubes RV internals jet pump assemblies, and bolting of RV vents, drains and the recirculation system.

In Section 3.1.2.1.1 of the LRA, the applicant provided further evaluation of aging management as recommended by the GALL report for the RV, RVIs, and Reactor Coolant System (RCS).

The applicant provided information concerning how it will manage the following aging effects:

Change in material properties and reduction in fracture toughness due to neutron irradiation embrittlement Crack initiation and growth due to stress corrosion, fatigue and cyclic loading Distortion/plastic deformation due to stress relaxation Loss of material due to galvanic, general, crevice, and pitting corrosion Loss of material due to mechanical wear

2 Staff Evaluation In Tables 3.1.2.1 through 3.1.2.4 of the LRA, the applicant provided a summary of AMRs for the RV internals and reactor coolant system and identified which AMRs it considered to be consistent with the GALL report.

In LRA Tables 3.1.2.1 through 3.1.2.4, the staff reviewed additional details of the results of the AMRs for material, environment, aging effect requiring management, and AMP combinations that are not consistent with the GALL Report, or that are not addressed in the GALL Report.

The staff's evaluation is discussed in the following sections.

3.1.3.1.1 Change in material properties and reduction in fracture toughness due to neutron irradiation embrittlement The AMP recommended by the GALL report for managing loss of fracture toughness due to neutron irradiation embrittlement of the RV is XI.M31, RV Surveillance, which complies with the requirements of 10 CFR Part 50, Appendices G and H, and 10 CFR Part 50.61. Loss of fracture toughness due to neutron irradiation embrittlement could occur in the RV. A RV materials surveillance program monitors neutron irradiation embrittlement of the RV. RV surveillance programs may be plant-specific, depending on matters such as the composition of limiting materials, availability of surveillance capsules, and projected fluence levels or may be an integrated surveillance program (ISP) based on the criteria in 10 CFR 50, Appendix H. In accordance with 10 CFR Part 50, Appendix H, an applicant is required to submit its proposed withdrawal schedule for approval prior to implementation.

Section 3.1.2.2.3 of the LRA addressed (1) loss of fracture toughness due to neutron irradiation embrittlement for ferritic materials that have a neutron fluence of greater than 1017 n/cm2 at the end of the license renewal term, and (2) loss of fracture toughness due to irradiation embrittlement of the RV beltline material. Loss of fracture toughness due to neutron irradiation embrittlement for ferritic materials that have a neutron fluence of greater than 1017 n/cm2 at the end of the license renewal term is a TLAA, and the staffs review of the applicants evaluation of this TLAA is documented in Section 4.2. In performing this review, the staff followed the guidance in Section 4.2 of the SRP-LR.

The RV Surveillance Program is mandated by 10 CFR Part 50, Appendix H. The RV Surveillance Program is an integrated surveillance program (ISP) in accordance with 10 CFR Part 50, Appendix H Paragraph III.C that is based on requirements established by the BWR Vessel and Internals Project (BWRVIP). Referencing of BWRVIP activities for license renewal was approved by the NRC in its safety evaluation report (SER) regarding BWRVIP-74 of October 18, 2001. The demonstration of compliance with the required actions of the SE is summarized in Section 3.1.2.2.16 of the LRA. The applicant stated that the AMP B.2.1.28, RV Surveillance Program, as supported by associated TLAA evaluations (Section 4.2 of the LRA),

will manage loss of fracture toughness of RV beltline components due to irradiation embrittlement by addressing the limiting RV beltline shells and welds.

The applicant also stated that BFN RV Surveillance Program is described in UFSAR Section 4.2.6 and is based on BWRVIP-78, BWR Vessel Integrated Surveillance Program Plan and BWRVIP-86, BWR Vessel and Internal Project Updated BWR Integrated Surveillance

3 Program (ISP) Implementation Plan. Use of the BWRVIP-78 and BWRVIP-86 to address a 40 year license period was approved for referencing in the NRCs SER dated February 1, 2000.

Use of the BWRVIP ISP for BFN, Units 2 and 3 was approved by the NRC in its SER dated January 28, 2003. The technical criteria specified in the BWRVIP-78 and BWRVIP-86 were incorporated in the BWRVIP-116, BWR Vessel and Internals Project-Integrated Surveillance Program (ISP)-Implementation for License Renewal. BWRVIP-116 extends the ISP to cover the BWR fleet through an extended period of operation for all units. The applicant committed to implement the requirements of BWRVIP-116, when approved, for all three RVs. Therefore, the applicant did not submit a plant-specific program in its LRA. The details of the staffs review of this aging effect are included in the AMP Section 3.0.3.2 of the staffs SER.

The applicant stated that it will implement the BWRVIP ISP for the period of extended operation, if approved by the NRC for the BWR fleet, as applicable to each RV and will request the approval from the NRC, if necessary, to use the program at applicable RVs for the period of extended operation. This enhancement is scheduled for completion prior to the period of extended operation.

The staff finds that by implementing the ISP program as dictated by the AMP B.2.1.28, the applicant demonstrated that it complies with all the recommendations specified in NUREG-1801, GALL AMP XI.M31 at the BFN units. Therefore, the staff accepts the implementation of AMP B.2.1.28 for managing the aging effect due to loss of fracture toughness due to neutron irradiation embrittlement of the BFN RVs.

Conclusion On the basis of its review, the staff finds that the applicant has appropriately evaluated AMR results involving management of loss of fracture toughness due to neutron irradiation embrittlement as recommended in the GALL report. Since the applicants AMR results are otherwise consistent with the GALL report, the staff finds that the applicant has demonstrated that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21(a)(3).

3.1.3.1.2 Crack Initiation and Growth Due to Thermal and Mechanical Loading or Stress Corrosion Cracking (SCC)

The staffs evaluation of the aging effect due to cyclic loading and fatigue is discussed in XXXX of the SER.

3.1.3 1.2.1 SCC in RV Flange Leak Detection Line and Jet Pump Sensing Line:

SRP-LRA Section 3.1.3.2.4.2 states that the crack initiation and growth due to thermal and mechanical loading or SCC [including intergranular stress corrosion cracking (IGSCC)] could occur in the BWR RV flange detection line and jet pump sensing line. The GALL report recommends a plant-specific AMP be evaluated to mitigate or detect crack initiation and growth due to SCC of the vessel flange detection line and jet pump sensing line.

In Section 3.1.2.2.4 of the applicants LRA, the applicant addressed vessel flange leak detection

4 line which is subjected to SCC. The licensee proposed to use AMP B.2.1.29, One Time Inspection Program, for managing this aging effect.

In RAI-3.1.1-1, the staff requested that the applicant provide information on the plant-specific experience related to cracking due to SCC in vessel flange leak line at the BFN Units, and its method of implementation of the AMP B.2.1.29 at the BFN Units. The staff also requested the applicant provide justification for using one-time inspection in detecting the cracking due to SCC in a timely manner. In its response to RAI-3.1.1-1, by letter dated January 31, 2005, the applicant indicated that in addition to the One-Time Inspection program, AMP B.2.1.4, ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program will be implemented for the BFN RV flange leak detection lines. The applicant stated that it will revise the first paragraph in Section 3.1.2.2.4 of the LRA to include the ISI program as an additional AMP for the RV flange leak detection lines. The applicant stated that the AMR shown in the Table 3.1.2.1 of the LRA will be revised to include the aging effects [i.e., cracking growth from cyclic loading, loss of material due to crevice, pitting and general corrosion, and their associated AMPs (i.e., One-Time Inspection Program and ISI program)] for the carbon steel and low alloy steel RV heads, flanges and shells. The staff finds this response acceptable, because, the proposed AMPs will provide adequate measures in managing the aging effects of the RV flange leak detection lines.

In Section 3.1.2.2.4 of the LRA, the applicant addressed jet pump sensing line which is subjected to SCC. The licensee proposed to use AMPs B.2.1.5, Chemistry Control Program, and B.2.1.29, One Time Inspection Program, for managing this aging effect.

In RAI-3.1.1-2 the staff requested that the applicant provide information on the plant-specific experience related to cracking due to SCC in jet pump sensing line at the BFN units, and its method of implementation of AMP B.2.1.29 at the BFN units. The staff also requested the applicant provide justification for using AMP B.2.1.29 to detect cracking due to SCC in a timely manner.

In its response to RAI-3.1.1-2, by letter dated January 31, 2005, the applicant stated that the jet pump sensing lines have not previously experienced cracking due to SCC, IGSCC or cyclic loading. The jet pump sensing lines inside the RV are not in the scope of license renewal process. According to Section 2.3.12.7 of the BWRVIP-41, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines, inspection of the jet pump sensing lines is continuously occurring during the plant operation. Therefore, if this line fails, plant technical specifications require either a plant shut down or a safety assessment to justify continued operation. Therefore, the failure of the sensing lines inside the RV has no adverse safety consequences and does not need to be included within the scope of license renewal.

However, the applicant agreed to revise the AMR by adding the AMPs (shown below) for the jet pump sensing line penetrations and external lines that are listed in Tables 3.1.2.1 and 3.1.2.4.

The applicant included the AMP B.2.1.11, BWR Reactor Penetration Program, for managing the aging effect related to cracking due to SCC in the jet pump sensing lines penetrations at BFN units. The applicant stated that this AMP is consistent with GALL program XI.M8, BWR Penetrations with no exceptions. AMP B.2.1.11 includes the staffs approved versions of

5 BWRVIP-27, BWR Standby Liquid Control System/Core Plate delta P Inspection and Flaw Evaluation Guidelines, and BWRVIP-49, Instrumentation Penetration Inspection and Flaw Evaluation Guidelines.

The applicant stated that AMP B.2.1.5 will be used at the BFN units. The BFN Chemistry Control Program is based on EPRI Report TR-103515-R2, ( the 2000 Revision of "BWR Water Chemistry Guidelines), which was approved by the staff in February 2000. The staff found the EPRI TR-103515-R2 acceptable because the program is based on updated industry experience and plant-specific and industry-wide operating experience that confirms the effectiveness of the RCS chemistry program. The staff finds that the implementation of AMP B.2.1.5 would be consistent with the GALL AMP XI.M2, and therefore, it is acceptable. In addition, the proposed inspection AMPs would ensure the identification of cracking due to SCC, IGSCC and cyclic loading in a timely manner so that the intended function of the jet pump sensing lines is not sacrificed. Therefore, the staff concludes that by the implementation of the additional AMPs, the aforementioned aging effects of the jet pump sensing lines would be managed effectively during the extended period of operation.

3.1.3 1.2.2 Stainless Steel Reactor Vessel Attachment Welds The AMPs recommended by the GALL report for managing the cracking due to SCC, IGSCC and cyclic loading for the RV attachment welds are XI.M4, BWR Vessel Inner Diameter (ID)

Attachment Welds, and XI.M2, Water Chemistry, which references EPRI Report TR-103515.

In the Table 3.1.2.1 of the LRA, the applicant identified IGSCC as an aging effect for the stainless steel RV attachment welds. The applicant stated that AMP B.2.1.5, Chemistry Control Program, will be used at the BFN units. The BFN Chemistry Control Program is based on EPRI Report TR-103515-R2, (the 2000 Revision of "BWR Water Chemistry Guidelines), which was approved by the staff in February 2000. The staff found EPRI TR-103515-R2 acceptable because the program is based on updated industry experience and plant-specific and industry-wide operating experience that confirms the effectiveness of the RCS chemistry program. The applicant indicated that the vessel attachment welds program is discussed in LRA AMP B.2.1.7, BWR Vessel ID Attachment Welds. AMP B.2.1.7 references AMP B.2.1.4, ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program. The applicants ISI Program is an established aging management program that is based on compliance with the staffs ISI requirements in 10 CFR Part 50.55a. This program has appropriate requirements for inspecting the vessel ID attachment welds. AMP B.2.1.7 also invokes the inspection and evaluation recommendations of BWRVIP-48, Vessel ID Attachment Weld Inspection and Evaluation Guidelines.

In RAI-3.1.2.1-1, the staff requested that the applicant provide the method of implementation of the type and frequency of inspections that are specified in BWRVIP-48, Vessel ID Attachment Welds Inspection and Flaw Evaluation Guidelines. These requirements apply to jet pump raiser brace attachments, core spray piping brackets attachments, steam dryer support and hold down brackets, feedwater spargers, guide rods and surveillance sample holders.

According to Section 2.2.3 of BWRVIP-48, furnace-sensitized stainless steel vessel ID attachment welds are highly susceptible to IGSCC. The staff requested that the applicant

6 identify whether there are any furnace-sensitized stainless steel attachment welds at BFN units, and provide information regarding an augmented inspection program for any existing furnace-sensitized stainless steel attachment welds.

In its response to RAI-3.1.2.1-1, by letter dated January 31, 2005, the applicant stated that all the ID RV attachment welds in BFN units were inspected (type and frequency) in accordance with BWRVIP-48 and ASME Section XI ISI requirements. The applicant indicated that all the ID attachment welds are furnace-sensitized and therefore, an augmented inspection program in accordance with the requirements of BWRVIP-48 will be implemented for all these welds. The staff finds that this type of inspection would ensure that the aforementioned aging effects are properly managed for the extended period of operation. The staff finds that the implementation of AMPs B.2.1.4, B.2.1.5, and B.2.1.7 would be consistent with the GALL AMPs XI.M2 and XI.M4, and therefore, it is acceptable.

3.1.3 1.2.3 Reactor Vessel Nozzles and Safe Ends The AMPs recommended by the GALL report for managing the cracking due to SCC, IGSCC and cyclic loading for the RV nozzlesds are XI.M7, BWR Stress Corrosion Cracking, and XI.M2, Water Chemistry, which references EPRI Report TR-103515.

In Table 3.1.2.1 of the LRA, the applicant indicated that stainless steel materials in the RV nozzle and safe end components, when exposed to treated water environment, experience cracking due to SCC. The applicant invoked AMP B.2.1.10, BWR Stress Corrosion Cracking Program, AMP B.2.1.4, ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program, which is an established aging management program that is based on compliance with the staffs ISI requirements in 10 CFR Part 50.55a. In addition, the applicant indicated that AMP B.2.1.5, Chemistry Control Program, which is based on EPRI Report TR-103515-R2, (the 2000 Revision of "BWR Water Chemistry Guidelines), which was approved by the staff in February 2000.

The staff issued three RAIs related to the nozzle and safe end components. RAI 3.1.2.1-4(A) is addressed in Section 3.1.3.1.5 of this SER, RAI 3.1.2.1-4(B) is addressed in Section 3.1.3 1.2.4, of this SER and RAI 3.1.2.1-4(C) is addressed in this section.

In a RAI-3.1.2.1-4 (C), the staff requested that the applicant identify whether the dissimilar metal welds of reactor vessel nozzles and safe end components have previously experienced cracking due to SCC, IGSCC or cyclic loading, and the extent of cracking. In its response to RAI-3.1.2.1-4(C), by letter dated January 31, 2005, the applicant stated that for the dissimilar metal welds in nozzles and safe end components, and piping, the requirements of ASME Section XI, Subsections IWB, IWC and IWD ISI Program inspections and frequencies in accordance with ASME Section XI, Table IWB-2500-1, examination category B-F would be met.

The applicants BWR IGSCC program inspections and frequencies are in accordance with the normal water chemistry guidelines contained in BWRVIP-75, BWR Vessel and Internals Project (BWRVIP), Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedule.

The applicant implemented alternative examination requirements for IGSCC Category A (as defined in BWRVIP-75) dissimilar metal welds under risk-informed ISI program (previously approved by the staff) for the BFN, Units 2 and 3. The applicant stated that it performed liquid penetrant (PT) and ultrasonic testing (UT) of the dissimilar welds in recirculation inlet and outlet nozzle-to-safe end, core spray nozzle-to-safe end, and pipe-to-safe end, and CRD nozzle-to

7 cap for BFN, Units 2 and 3, and the examination results were acceptable. The applicant stated that for BFN, Unit 1, it performed PT and UT examinations on CRD nozzle-to-cap welds, and the examination results were acceptable. The applicant stated that for BFN, Unit 1, the RCS water chemistry would be improved in accordance with AMP B.2.1.10, and the CRD nozzle-to-safe end welds would be replaced prior to the period of extended operation.

The applicant also claimed that improvements in RCS water chemistry provide mitigative measures to preclude IGSCC in the dissimilar welds in nozzle-to-safe end, pipe-to-safe end and nozzle-to cap components. The staff accepts the proposed program for stainless steel safe ends because it conforms to the recommendations in the BWRVIP-75. However, if the safe ends contain nickel alloy weld metals that are susceptible to SCC, BWRVIP-75 would require more frequent examinations than that specified for BWRVIP-75 Category A welds. In order for the staff to determine whether the applicant has adequately implemented BWRVIP-75, the staff requests that the applicant identify : (1) the weld metal that was used for the butter, nozzle-to-safe end welds, pipe-to-safe end welds and nozzle-to cap welds ; (2) the grade of stainless steel that was used as a safe end; and (3) the examination requirements for butter, nozzle-to-safe end welds, pipe-to-safe end welds and nozzle-to cap welds that are more susceptible to SCC than the BWRVIP-75 Category A welds. This is Open Item 3.1.3.1.2.3-1.

Since the previous examination results for the dissimilar welds in nozzle-to-safe end welds, pipe-to-safe end welds and nozzle-to cap welds were acceptable the staff finds that there is no active degradation thus far, in this system at the BFN, units. The staff found the EPRI TR-103515-R2 acceptable because the program is based on updated industry experience and plant-specific and industry-wide operating experience confirms the effectiveness of the RCS chemistry program. The staff finds that the implementation of AMPs B.2.1.4 and B.2.1.5 would be consistent with the GALL AMPs XI.M2 and XI.M7, and therefore, it is acceptable pending the resolution of the aforementioned open item.

3.1.3 1.2.4 Feedwater Nozzle The AMP recommended by the GALL report for managing cracking due to cyclic loading for the feedwater nozzles is XI.M5, BWR Feedwater Nozzle, which recommends that inspection requirements specified in General Electric (GE) NE-523-A71-0594, Alternate BWR Feedwater Nozzle Inspection Requirements, be implemented for the feedwater nozzles.

The applicant included AMP B.2.1.8 BWR Feedwater Nozzle Program, for managing the aging effect related to cracking due to cyclic loading in the feedwater nozzles at the BFN units.

The applicant stated that the program is consistent with GALL program XI.M5, with no exceptions. The applicant also invoked AMP B.2.1.4 ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program which is an established aging management program that is based on compliance with the staffs ISI requirements in 10 CFR Part 50.55a.

This program has appropriate requirements for inspecting the feedwater nozzle components.

The applicant also stated that the program enhances the ISI specified in ASME Code Section XI with the recommendations of General Electric (GE) NE-523-A71-0594, Alternate BWR Feedwater Nozzle Inspection Requirements. The applicant stated that it implemented the recommendations of NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking to mitigate feedwater nozzle cracking. The applicant also stated that the BFN feedwater nozzles were modified to mitigate cracking by removing the stainless steel cladding and machining the safe end, nozzle bore, and inner bend radius to accept improved

8 double-piston-ring interference-fit spargers with a forged tee design and orificed elbow discharges. The applicant indicated in the LRA that the reactor water cleanup system return lines were routed to both feedwater headers (except BFN, Unit 2 which is only routed to one feedwater header). The applicant stated that changes to plant operating procedures, such as improved feedwater control, to decrease the magnitude and frequency of temperature fluctuations were implemented at BFN, Units 2 and 3. The applicant also indicated that similar improvements will be implemented for the BFN, Unit 1 prior to the period of extended operation.

In RAI 3.1.2.1-4(B), the staff requested that the applicant provide information on the scope and the techniques of the past inspections, the results obtained, applied mitigative methods, repairs, frequency of the inspections and any other relevant information related to the identification of the aging effect in the feedwater nozzles at the BFN units. The staff further requested that the applicant provide information as to how the plant-specific experience related to this aging effect impacts the attributes specified in AMP B.2.1.8.

In its response to RAI 3.1.2.1-4(B), by letter dated January 31, 2005, the applicant stated that it complied with the inspection requirements specified in AMP B.2.1.8 at the BFN units. The applicant stated that it performed UT of the feedwater nozzles at the BFN units the results were acceptable, and no repairs were performed in this system. Therefore, the applicant concluded that the plant-specific experience related to feedwater nozzles has no impact on the attributes specified in AMP B.2.1.8. The staff reviewed the applicants response and finds it acceptable, because the applicant has demonstrated that the actions taken thus far have mitigated cracking in feedwater nozzles.

In RAI B.2.1.8-1, the staff stated that AMP B.2.1.8 references GE report GE-NE-523-A71-0594, which is not the NRC-approved version of the report. The staff requested that the applicant replace references to GE-NE-523-A71-0594 in LRA Appendix A.1.8 and Appendix B.2.1.8 with GE-NE-523-A71-0594-A, Revision 1 which is approved by the staff. In its response to RAI B.2.1.8-1, by letter dated January 31, 2005, the applicant stated that it will revise the LRA to indicate correct GE report.

To RLEP for Resolution: The applicant shall revise A.1.8 of the UFSAR supplement to indicate GE-NE-523-A71-0594-A, Revision 1 which is the applicable document for feedwater nozzle analysis and inspection. This is Confirmatory Item 3.1.3.1.2.4-1.

The staff finds that the implementation of AMPs B.2.1.4 and B.2.1.8, would be consistent with the GALL AMP XI.M5, and therefore, it is acceptable with exception of the aforementioned confirmatory item.

3.1.3 1.2.5 Control Rod Drive (CRD) Return Line Nozzle AMP recommended by the GALL report for managing the cracking due to cyclic loading for the CRD return line nozzle is XI.M6, BWR Control Rod Drive Return Line Nozzle. This AMP recommends that enhanced inspection requirements specified in NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking, should be implemented for the CRD return line nozzles.

In Table 3.1.2.1 of the LRA, the applicant referenced AMP B.2.1.9, BWR Control Rod Drive

9 (CRD) Return Line Nozzle, for managing the aging effect in the CRD return line. The applicant stated that the program is consistent with GALL program XI.M6 with no exceptions. The applicant indicated that inspections that are specified in NUREG-0619, and AMP B.2.1.4, ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program, which is an established aging management program that is based on compliance with the staffs ISI requirements in 10 CFR Part 50.55a. This program has appropriate requirements for inspecting the CRD return line nozzle components.

In RAI B.2.1.9-1, the staff requested that the applicant provide information on the augmented inspection requirements that are specified in the NUREG-0619. The CRD return line nozzle has been capped, and therefore augmented inspection for the nozzle is not required per NUREG-0619. The requirements in NUREG-0619 provide actions to be taken to address cracking in these nozzles. However, the aging effects for the cap and applicable weld are not covered in NUREG-0619. Therefore, the staff requested the applicant address the following issues concerning the cap and weld which provide a pressure boundary function:

RAI B.2.1.9-1(1)- Configuration, location and material of construction of the capped nozzle including the existing base material for the nozzle, piping (if piping remnants exist) and cap material, and any welds.

In its response to RAI B.2.1.9-1(1), by letter dated January 31, 2005, the applicant stated that at BFN, Units 1, 2, and 3 the configuration consists of a stainless steel cap welded to the original carbon steel nozzle with stainless steel weld material. The safe end and corresponding piping were removed from the nozzle.

RAI B.2.1.9-1(2)- Application of the BWRVIP-75 inspection guidelines for this weld and cap.

In its response to RAI B.2.1.9-1(2), by letter dated January 31, 2005, the applicant stated that the requirements of BWRVIP-75 are implemented by AMP B.2.1.10, BWR Stress Corrosion Cracking Program. The CRD return line nozzles welds are currently categorized (BWRVIP-75) as Category D for BFN, Unit 2 and Category C for BFN, Unit 3. The CRD return line nozzles welds are examined by the UT technique at the frequency specified by BWRVIP-75, Table 3-1 for normal water chemistry conditions. The applicant stated that it will implement AMP B.2.1.10 for Unit 1 prior to the period of extended operation.

The staff reviewed the applicants response finds it acceptable provided the applicant includes information in the LRA regarding the category (per BWRVIP-75) of the subject weld in the BFN, Unit 1.

RAI B.2.1.9-1(3)-- The applicability of the event at Pilgrim (leaking weld at capped nozzle, September 30, 2003) to BFN, units. The staff issued Information Notice 2004-08, dated April 22, 2004, which states that the cracking occurred in a 82/182 weld that was previously repaired extensively at the Pilgrim unit. According to NRC IN 2004-08, the Pilgrim CRD return line nozzle is made of SA-508, Class 2 low-alloy steel, while the CRD return line cap is made of Alloy 600. The subject weld is fabricated with Alloy 82/182 material, and the nozzle side of the weld is buttered with Alloy 182 material. In addition, Pilgrim had initial weld deficiencies (lack of fusion) that required weld repair. The staff requested that the applicant provide any plant experience with leakage at the capped nozzle, the past inspection techniques applied, the results obtained, and mitigative strategies imposed. The staff requested that the applicant

10 provide information as to how the plant-specific experience related to this aging effect impacts the attributes specified in AMP B.2.1.9.

In its response to RAI B.2.1.9-1(3), by letter dated January 31, 2005, the applicant stated that the event at Pilgrim was determined to be not applicable to the BFN units. The materials of construction of the nozzle to cap weld at BFN is stainless steel. The BFN welds were completed without recordable indications. Plant experience for BFN, Unit 2 and BFN, Unit 3 indicates that there has been no leakage at the capped CRD return line nozzles. Ultrasonic exams have been performed with no reportable indications. The BFN, Unit 3 capped CRD return line nozzle weld had mechanical stress improvement process (MSIP) performed to mitigate IGSCC, which changed the frequency of inspection. The examination information related to this item is described in RAI B.2.1.9-1(2). The plant-specific experience related to CRD return line nozzle has no impact on the attributes specified in AMP B.2.1.9.

The staff reviewed the applicants response finds it acceptable, in part, because the improved RCS water chemistry and MSIP (for Unit 3) should provide adequate mitigation to preclude IGSCC. However, the staff finds that, unlike weld metal Alloy 182, austenitic stainless steel weld metal (with a minimum delta ferrite) is less susceptible to IGSCC. In addition, low carbon austenitic stainless steel material (L grade) is more resistant to IGSCC than non-L grade austenitic stainless steel.

In order for the staff to determine whether the applicant has adequately implemented BWRVIP-75, for the Category A CRD return line nozzle welds in all BFN units, the staff requests that the applicant identify: (1) the delta ferrite in the weld metal, (2) the grade of stainless steel that was used for the CRD return line cap, (3) the examination requirements for CRD return line welds that meet BWRVIP-75, and (4) plans to implement MSIP in BFN, Units 1 and 2. This is Open Item 3.1.3.1.2.5-1.

The staff finds that the implementation of AMPs B.2.1.9 and AMP B.2.1.4 for the CRD return lines would be consistent with the GALL AMP XI.M6, and therefore, it is acceptable pending the resolution of the aforementioned open item.

3.1.3 1.2.6 Reactor Vessel Penetrations AMPs recommended by the GALL report for managing cracking due to IGSCC for the RV penetrations are XI.M8, BWR Penetration, and XI.M2, Water Chemistry. The GALL AMPs for the RV penetrations include implementation of guidelines specified in BWRVIP-49, Instrumentation Penetration Inspection and Flaw Evaluation Guidelines, and reactor coolant water chemistry in accordance with the guidelines of BWRVIP-29, BWR Water Chemistry Guidelines, (EPRI TR-103515). In addition to these requirements, the GALL program XI.M8, BWR Penetration, recommends that inspection and flaw evaluation guidelines specified in BWRVIP-27, BWR Standby Liquid Control System/Core Plate delta P Inspection and Flaw Evaluation Guidelines, should be implemented for the RV penetrations.

In Table 3.1.2.1, the applicant indicated that nickel alloy and stainless steel materials in the RV penetration components, when exposed to treated water environment experience cracking due to SCC. The applicant included AMP B.2.1.11, BWR Reactor Penetration Program, for managing the aging effect related to cracking due to SCC in the RV penetrations at the BFN

11 units. The applicant stated that this AMP is consistent with GALL program XI.M8 with no exceptions. AMP B.2.1.11 recommends the implementation of the staffs approved versions of BWRVIP-27, BWRVIP-49, and AMP B.2.1.4, ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program. The ISI program is an established aging management program that is based on compliance with the staffs ISI requirements in 10 CFR Part 50.55a.

This program has appropriate requirements for inspecting the RV penetrations (i.e., category B-E for pressure-retaining partial penetration welds; category B-D for full penetration nozzle-to-vessel welds; category B-F for pressure retaining dissimilar metal nozzle-to-safe end welds; and category B-J for similar metal nozzle-to-safe end welds). The extent and schedule of inspection prescribed by the ASME Section XI ISI Program, is designed to maintain structural integrity and ensure that aging effects will be discovered and repaired before the loss of intended function of the component. These inspections can reveal crack initiation and growth and leakage of coolant due to SCC. In addition, the applicant indicated that AMP B.2.1.5 Chemistry Control Program which is based on EPRI Report TR-103515-R2, (the 2000 Revision of "BWR Water Chemistry Guidelines) will be applied at the BFN units. The staff found the EPRI TR-103515-R2 acceptable because the program is based on updated industry experience and plant-specific and industry-wide operating experience confirms the effectiveness of the RCS chemistry program.

In RAI 3.1.2.1-5(B) the staff requested that the applicant provide any previous plant-specific experience regarding the cracking due to SCC, IGSCC in dissimilar metal welds of RV penetrations, and the method and frequency of inspection for managing this aging effect. In its response to RAI 3.1.2.1-5(B), by letter dated January 31, 2005, the applicant stated that the following penetrations are inspected during the ASME Section XI, IWB-2500, Code Category B-P system leakage test during each refuel outage: (1) CRD stub tubes; (2) instrumentation nozzle/nozzle safe ends; (3) standby liquid control nozzles; (4) jet pump instrumentation nozzles; (5) drain line nozzles; and (6) in core monitor housing penetrations.

The applicant also stated that no cracking of the dissimilar metal penetration welds were identified thus far at the BFN units. In addition, the applicant claimed that the improvements in the RCS water chemistry control program would enable the mitigation of IGSCC of the RV penetration welds. The applicant stated that the plant-specific experience related to the RV penetrations has no impact on the attributes of AMP B.2.1.11.

The staff reviewed the response to the RAI 3.1.2.1-5(B) and finds it acceptable, because implementation of the improved water chemistry, and ISI programs as specified in AMP B.2.1.11 would enable the applicant to manage the aging effect due to IGSCC effectively during the extended period of operation, and would be consistent with GALL AMPs XI.M8 and XI.M2.

3.1.3.1.2.7 Reactor Head Closure Studs AMPs recommended by the GALL report for managing the cracking due to SCC for the reactor head closure studs is XI.M3, Reactor Head Closure Studs. This AMP recommends that preventive actions specified in Regulatory Guide 1.65, Materials and Inspections for RV Closure Studs, should be implemented.

In Table 3.1.2.1 of the LRA, the applicant indicates that AMP B.2.1.6 Reactor Head Closure Studs, which is consistent with GALL program XI.M3, will be implemented to monitor the aging effect due to SCC of the reactor head closure studs.

12 The applicant stated that the following requirements will be implemented for the BFN Reactor Head Closure Studs Program.

(a) ISI in conformance with the requirements of the ASME Code Section XI, Subsection IWB, Table IWB 2500-1 (AMP B.2.1.4).

(b) Mitigation of cracking is achieved by complying with the requirements of Regulatory Guide 1.65, Materials and Inspections for RV Closure Studs. The applicant stated that approved lubricants will be used to minimize the potential for cracking of the non-metal-plated reactor head closure studs.

The applicant stated that industry experience indicated that SCC occurred in metal-plated BWR reactor head closure studs. The applicant stated that there are no metal-plated reactor head closure studs in use at the BFN units, and approved lubricants are used to prevent seized studs or nuts. The applicant claimed that with the lack of metal plating and preventive use of approved lubricants, AMP B.2.1.6 has been effective in reducing the probability of SCC of the reactor head closure studs.

The applicant concluded in its LRA that the reactor head closure studs program provides reasonable assurance that aging effect due to cracking in the reactor head closure studs is adequately managed so that their intended functions, consistent with the current licensing basis, are maintained during the period of extended operation.

The staff concludes that the reactor head closure studs at the BFN units are less likely to experience aging effect related to SCC, because these closure studs are not metal plated and approved lubricants are used for their maintenance. The staff finds the implementation of AMP B.2.1.4 is acceptable, because presence of aging effects can be identified by frequent inspections dictated by AMP B.2.1.4. In addition, compliance of RG 1.65 requirements provides adequate assurance in maintaining the integrity of the RV studs at the BFN units. The staff concludes that implementation of the aforementioned requirements provides assurance that the aging effect associated with SCC is adequately managed by the applicant at the BFN units.

3.1.3.1.2.8 Bolting for Reactor Vessel Vents and Drains AMP recommended by the GALL report for managing the aging effects for the bolting in the RV vents and drains is XI.M18, Bolting Integrity. In Table 3.1.2.3 of the LRA, the applicant indicates that AMP B.2.1.16 Bolting Integrity Program, which is consistent with GALL program XI.M18 will be implemented to monitor the aging effects of the bolting in RV vents and drains.

Table 3.1.2.3 of the LRA and AMP B.2.1.16 do not identify SCC as an aging effect for these bolts. Therefore, the staff requested in RAI 3.1.2.3-1(A) that the applicant address the aging effect due to SCC in the bolts of the RV vents and drains.

The applicant stated in its response that SCC can occur in high yield strength (greater than 150 ksi) bolted closures in BWRs when they are exposed to a corrosive environment, typically attributed to leakage of pressure boundary joints or exposure to wetted ambient environments or due to the use of thread lubricant containing MoS2 (molybdenum disulfide). High yield strength, heat-treated alloy steel bolting materials are not specified for flanged connections at BFN. High strength bolting in vendor supplied equipment has not been identified for mechanical components (such as pump casing studs or valve body/bonnet studs) where the

13 material specifications are available. The applicant stated that a review of the BFN operating experience did not identify any instances where mechanical component failure was attributable to SCC of high strength bolting. Therefore, loss of bolting function due to SCC of bolted joints of mechanical equipment is not expected and no aging management is required for the period of extended operation. Since there are no high yield strength bolts in the RV vents and drains at the BFN units, the staff concludes that no AMP is required to monitor the aging effect due to SCC in bolting in reactor vents and drains.

3.1.3.1.2.9 Conclusion On the basis of its review, the staff finds that the applicant has appropriately evaluated AMR results involving management of SCC and IGSCC as recommended in the GALL report (with the exception of the Open Items 3.1.3.1.2.3-1, and 3.1.3.1.2.5-1, and Confirmatory Item 3.1.3.1.2.4-1). Since the applicants AMR results are otherwise consistent with the GALL report, the staff finds that the applicant has demonstrated that the effects of aging will be adequately managed, so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21(a)(3).

AMR Results That are not Addressed in the GALL report 3.1.3 1.3 Distortion/plastic deformation due to stress relaxation and loss of material due to mechanical wear 3.1.3 1.3.1 Reactor head closure studs and nuts; bolting in RV vents, drains and the recirculation system In Table 3.1.2.1 of the LRA, the applicant addressed distortion and plastic deformation due to stress relaxation and loss of material due to mechanical wear as aging effects in reactor head closure studs and nuts. The applicant proposed to use AMP B.2.1.6, Reactor Head Closure Stud Program, which in turn, invokes the requirements of GALL AMP XI.M3, to monitor this aging effect. The applicant reiterates that the aforementioned aging effect is adequately managed by AMP B.2.1.4, ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program. In RAI 3.1.2.1-2, the staff requested that the applicant identify any plant-specific aging effects due to distortion/plastic deformation resulting from stress relaxation and loss of material due to mechanical wear, for the reactor closure studs and nuts at the BFN units.

In its response to RAI 3.1.2.1-2, by letter dated January 31, 2005, the applicant stated that it has not identified any RV closure stud or nut degradation resulting in distortion/plastic deformation due to stress relaxation or loss of material due to mechanical wear. The applicant also stated that no RV closure studs or nuts have been replaced for this reason. Two studs were replaced in BFN, Unit 2 during the BFN, Unit 2 Cycle 4 refueling outage. These were replaced because of physical thread damage. From discussions with plant personnel present during that time period, this damage was the result of impacts during handling and refueling operations and not the result of inservice stress or wear. Based on this, the applicant claimed that there was no impact on the attributes specified in AMP B.2.1.6. The staff concludes that the proposed AMPs B.2.1.4 and B.2.1.6 for the reactor closure studs at the BFN units are consistent with GALL AMP XI.M3, and the subject aging effects are adequately managed by

14 the applicant for the period of extended operation.

In Table 3.1.2.3 of the LRA, the applicant addressed loss of bolting function due to wear as an aging effect in RV vents and drains and the recirculation system. The applicant proposed to use AMP B.2.1.16, Bolting Integrity Program, for monitoring this aging effect, which in turn, invokes the requirements of GALL AMP XI.M18. GALL AMP XI.M18 requires application of ASME Code Section XI Subsection IWB, Table IWB 2500-1 for the bolts that are included in the ASME Section XI program to monitor this aging effect. In addition, the aging effects for the safety-related bolting are mitigated by NUREG-1339, Resolution of Generic Safety Issue 29:

Bolting Degradation Failure in Nuclear Power Plants. For bolts that are not included in the ASME Section XI program, the applicant proposed to use AMP B.2.1.39, System Monitoring Program. The staff concludes that the implementation of AMPs B.2.1.16 and B.2.1.39, and compliance with GALL AMP XI.18 will provide reasonable assurance that loss of bolting function due to wear in RV vents, drains and the recirculation system is adequately managed so that their intended functions, consistent with the current licensing basis, are maintained during the period of extended operation.

In RAI 3.1.2.3-1(B) the staff requested that the applicant provide information on the previous plant-specific experience of loss of bolting function due to wear in the RV vents and drains system at the BFN units. The staff also requested that the applicant provide information on the scope and the techniques of the past inspections, the results obtained, applied mitigative methods, repairs, frequency of the inspections and any other relevant information related to the identification of this aging effect of the bolts in RV vents and drains at the BFN units. In addition, the staff requested that the applicant provide information as to how the plant-specific experience related to this aging effect impacts the attributes specified in AMP B.2.1.16.

In its response to 3.1.2.3-1(B), by letter dated January 31, 2005, the applicant stated that aging effect due to wear was conservatively identified to be an aging effect that requires management for the period of extended operation for pressure boundary bolting in RV vents and drains. The applicant also stated that these bolts are inspected in accordance with AMP B.2.1.4, ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program inspection requirements, and AMP B.2.1.39, System Monitoring Program. The System Monitoring Program performs an entire system inspection once per fuel cycle and includes visual inspections for evidence of material condition and bolting torque relaxation. The System Monitoring Program documents failures in either the maintenance work order or plant corrective action program, as appropriate. The applicant indicated that so far, no instances of RV vents and drains bolting failure due to wear were identified at the BFN units.

In RAI 3.1.2.4-1(A) the staff requested that the applicant provide information on the previous plant-specific experience of loss of bolting function due to stress relaxation in the RV recirculation system at the BFN units. The staff also requested that the applicant provide information on the scope and the techniques of the past inspections, the results obtained, applied mitigative methods, repairs, frequency of the inspections and any other relevant information related to the identification of this aging effect of the RV recirculation bolts at the BFN units. In addition, the staff requested that the applicant provide information as to how the plant-specific experience related to this aging effect impacts the attributes specified in AMP B.2.1.16.

In its response to 3.1.2.4-1(A), by letter dated January 31, 2005, the applicant stated that it

15 inspected the reactor water recirculation pump closure bolting in accordance with the requirements of ASME Section XI, Table IWB-2500-1, Category B-G-1. The inspection methods included visual examination (VT) and ultrasonic testing (UT), and the results were acceptable. Therefore, the applicant did not perform any repair on the reactor recirculation pump closure bolting. The applicant stated that implementation of AMP B.2.1.16, and compliance with the recommendations of NUREG-1339 and EPRI NP-5769 provide adequate assurance that the aging effect due to stress relaxation in the bolting of the RV recirculation system is effectively managed for the extended period of operation.

The staff reviewed the applicants responses to the aforementioned RAIs, and concludes that the implementation of AMPs B.2.1.4, B.2.1.16 and B.2.1.39 is consistent with GALL AMP XI.M18 and the subject aging effects for bolting in RV vents, drains and the recirculation system are adequately managed at the BFN units.

Conclusion On the basis of its review, the staff finds that the applicant has appropriately evaluated AMR results involving management of aging effects (i.e., distortion/plastic deformation due to stress relaxation and loss of material due to mechanical wear) in reactor head closure studs and nuts, and bolting in RV vents, drains and the recirculation system. In addition, the staff finds that the applicant has demonstrated that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21(a)(3).

3.1.3 1.4 Loss of Material due to Galvanic, General, Crevice, and Pitting corrosion In Table 3.1.2.1 of the LRA, the applicant addressed loss of material due to galvanic, general, crevice, and pitting corrosion in (1) reactor head closure studs; (2) RV attachment welds; (3) RV heads, flanges and shells; (4) RV nozzles; (5) RV nozzles and safe ends; (6) RV penetrations; and (7) bolting in RV vents, drains and the recirculation system.

The applicant also identified the implementation of relevant AMPs to manage the aging effects due to galvanic, general, crevice, and pitting corrosion of carbon and low alloy steels, stainless steel and nickel alloy materials when these materials are exposed to the BWR treated water environment. In Table 3.1.2.1 of the LRA, the applicant identifies these aging effects and the relevant AMPs that are associated with each of the aforementioned components. In Table 3.1.2.1 of the LRA, the applicant also included references related to Table IV.A1 of NUREG-1801, Volume 2, for each of the aforementioned components.

NUREG-1801, Volume 2, Table IV.A1, does not identify loss of material due to crevice, general, and pitting corrosion as aging effects in carbon and low alloy steel, stainless steel and nickel alloy materials that are used in the aforementioned RV components when these components are exposed to the BWR treated water environment. General, pitting, and crevice corrosion may occur in stainless steel or nickel alloy components under exposure to aggressive, oxidizing environments. Normally, the presence of elevated dissolved oxygen and/or aggressive ionic impurity concentrations is necessary to create these oxidizing environment in the RCS.

The applicant stated that the AMP B.2.1.5, Chemistry Control Program, will be used at the

16 BFN units. The BFN Chemistry Control Program is based on EPRI Report TR-103515-R2 (the 2000 Revision of "BWR Water Chemistry Guidelines). The staff found EPRI TR-103515-R2 acceptable because the program is based on updated industry experience and plant-specific and industry-wide operating experience confirms the effectiveness of AMP B.2.1.5. In addition, this program provides an acceptable basis for minimizing the dissolved oxygen and ionic impurity concentrations that could otherwise, if left present in high concentrations, lead to an aggressive oxidizing RCS coolant environment which can enhance corrosion of the RV components. Since the applicant has conservatively assumed that loss of material due to general corrosion, pitting corrosion, or crevice corrosion is an applicable aging effect for these RV components, the staff concludes that AMP B.2.1.5 provides a sufficient mitigative strategy for managing this aging effect relative to the recommendations of GALL. The applicant stated that it will invoke AMP B.2.1.4, ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program for the BFN units. The applicants ISI Program is an established aging management program that is based on compliance with the staffs ISI requirements in 10 CFR Part 50.55a. This program has appropriate requirements for inspecting the aforementioned vessel components.

The staff concludes that by implementing AMP B.2.1.4 and AMP B.2.1.5 at the BFN units, the applicant has demonstrated that the effects of aging due to general, pitting and crevice corrosion will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21(a)(3).

Conclusion On the basis of its review, the staff finds that the applicant has appropriately evaluated AMR results involving management of general, crevice and pitting corrosion in reactor head closure studs; RV attachment welds; RV heads, flanges and shells; RV nozzles; RV nozzles and safe ends; RV penetrations; and bolting in RV vents, drains and the recirculation system. The staff finds that the applicant has demonstrated that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21(a)(3).

3.1.3 1.5 Loss of Materials in Low Alloy Steel or Carbon Steel Reactor Vessel Components that are exposed Externally to Inside (Atmospheric) Environments The applicant identifies in Table 3.1.2.1 of the LRA no aging effects, but included references related to Table IV. A1 of NUREG-1801, Volume 2, for carbon and low alloy steel materials of the following RV components exposed to externally to inside (atmospheric) environments.

Other External Attachment Welds to the Reactor Vessel Reactor Vessel Heads, Flanges, and Shell Reactor Vessel Nozzles Reactor Vessel Nozzles and Safe Ends Reactor Vessel Penetrations Reactor Vessel Internals CRD Housing Bolting in Reactor Vessel Vents, Drains and the Recirculation System The staff reviewed the applicants evaluation to determine whether it adequately addressed the

17 issue of uniform corrosion of the carbon and low alloy steel RV components when they are exposed externally to inside (atmospheric) environments. According to the paragraph 3.4.2.2.4 of NUREG-1801, loss of material due to general corrosion can occur on the external surfaces of carbon and low alloy steel RV components exposed to operating temperature less than 2120 F.

Since the operating temperature of the BWR vessel is greater than 2120 F, the loss of material due to general corrosion is not likely to occur in carbon and low alloy steel RV components. In addition, the external surface of the carbon and low alloy steel RV components are exposed to inside (atmospheric) environment which does not contain any aggressive ions resulting in loss of material due to corrosion.

In RAIs 3.1.2-1, 3.1.2.1-4(A) and 3.1.2.1-5(A), the staff requested that the applicant provide an explanation as to why the loss of material due to corrosion is not considered as an aging effect for carbon and low alloy steel vessel attachment welds; vessel heads, flanges, and shells; vessel nozzles and safe ends; vessel penetrations; and bolting in vessel vents, drains for BFN, Unit 1.

In response to RAIs 3.1.2-1, 3.1.2.1-4(A) and 3.1.2.1-5(A), by letter dated January 31, 2005, the applicant indicated that for the BFN, Unit 1,degradation due to corrosion of all the aforementioned RV components would be verified under the BFN, Unit 1 restart program. The applicant has also stated that it will perform further inspection of the subject RV components followed by replacement of the degraded components (if required) that are identified by this inspection. The staff finds that the applicants implementation of inspection and replacement (when necessary) programs provides reasonable assurance that the aging effect due to corrosion of carbon and low alloy steel penetrations for BFN, Unit 1 will be adequately managed so that the intended function(s) will be maintained with the current licensing basis for the extended period of operation.

Therefore, the staff finds that these components do not experience any of the aforementioned aging effects when they are exposed externally to an inside (atmospheric) environment. The staff concludes that the applicants determination of excluding these aging effects in Table 3.1.2.1 of the LRA for the aforementioned RV components is acceptable.

Conclusion On the basis of its review, the staff finds that the applicant has appropriately evaluated AMR results involving management of aging effects due to loss of materials in low alloy and carbon steel external attachment welds to RV, RV heads, flanges and shells, RV nozzles, RV nozzles and safe ends, RV penetrations, RV CRD housings, and bolting in RV vents, drains and the recirculation system that are exposed externally to an inside (atmospheric) environments. The staff finds that the applicant has demonstrated that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21(a)(3).

3.1.3.1.6 Reactor Vessel Internal Components The applicants specific aging management reviews (AMRs) for components that comprise the BFN reactor vessel internals (RVIs) are given in Table 3.1.2.2 of the LRA. These components include the core shroud and core plate, core spray lines and spargers, dry tubes and guide tubes and jet pump assemblies.

18 In Section 3.1.2.1.2 of the LRA, the applicant identified the following aging effects in the RV internal components:

Crack initiation and growth due to stress corrosion, fatigue and cyclic loading Change in material properties and reduction in fracture toughness due to thermal aging and neutron irradiation embrittlement Loss of material due to galvanic, general, crevice, and pitting corrosion 3.1.3.1.6.1 Crack initiation and growth due to stress corrosion cracking, fatigue and cyclic loading The staffs evaluation of the aging effect due to cyclic loading and fatigue is discussed in XXXX of the SER.

AMPs recommended by the GALL report for managing cracking due to IGSCC for the RV internal components are XI.M1, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD, XI.M2, Water Chemistry, and XI.M9, BWR Vessel Internals. AMP XI.M9 includes implementation of guidelines specified in the staff-approved BWRVIP documents for a given component.

In Table 3.1.2.2 of the LRA, the applicant identified SCC as an aging effect in (1) RVIs core shroud and core plate; (2) RVIs core spray and feedwater spargers; (3) RVIs control rod housing and dry tubes and guide tubes; (4) RVIs jet pump assemblies; and (5) RVIs top guide.

In Table 3.1.2.2 of the LRA, the applicant stated that the aging effect due to SCC in the aforementioned components is managed by (1) AMP B.2.1.12, Boiling Water Reactor Vessel Internals Program, (2) AMP B.2.1.4, ASME Code Section XI Subsections IWB, IWC, and IWD ISI Program, and (3) AMP B.2.1.5, Chemistry Control Program. The applicant claimed that continued implementation of these AMPs provides reasonable assurance that the aging effects due to SCC and fatigue and cyclic loading will be managed so that the systems and components within the scope of this program will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

The staff, in RAI 3.1.2.1-6 (A), requested that the applicant provide information on the scope and the techniques of the past inspections, the results obtained, applied mitigative methods, repairs, and frequency of the inspections of the access hole covers (AHCs). In response to RAI 3.1.2.1-6(A), by letter dated January 31, 2005, the applicant stated that the BFN, Unit 1 core shroud AHCs currently have indications of cracking and will be replaced with a bolted design in lieu of a welded design prior to BFN, Unit 1 restart. BFN, Units 2 and 3 AHCs have no reportable indications. In addition, the applicant claimed that the improvements in RCS water chemistry control program, would enable the mitigation of IGSCC of the AHCs. The staff, in RAI B.2.1.12-1(C), requested that the applicant provide information regarding any prior augmented UT for the AHCs as required by Section IV-B1.1.4 of NUREG-1801. The applicant stated that the AHCs are examined in accordance with GE Service Information Letter (SIL) No.

462, Revision 1. The GE SIL allows for inspection of the AHCs either by UT or top-surface visual (VT-1) inspection. The applicant has always used the UT technique as this methodology provides superior flaw detection and allows for a longer reinspection interval. Due to tooling constraints, a top-surface enhanced VT-1 (EVT-1), which is superior to the visual examination guidelines of GE SIL No. 462 was performed for BFN, Unit 3 AHCs. The applicant stated that

19 prior to the period of extended operation, BFN will implement visual inspection of the AHCs and inspection of the AHCs welds by UT unless tooling constraints prohibit performance of a UT. In the event tooling constraints prohibit inspection by UT, then the inspection will be performed by EVT-1. The applicant proposed to inspect the AHCs utilizing the BWR Vessel Internals Program rather than the ASME Section XI ISI Program currently specified in NUREG-1801.

RLEP/EMCB For Resolution: Since Section IV-B1.1.4 of NUREG-1801, requires UT of AHC welds, the staff requests that the applicant revise AMP B.2.1.12 and UFSAR supplement A.1.12 to include UT for the BFN, Units 2 and 3 AHC welds to the maximum extent possible. The staff requests that the applicant identify its previous experience on the extent to which UT was performed on the AHC welds at the BFN units. This is Open Item-3.1.3.1.6.1-1.

The staff, in RAI 3.1.2.1-6 (B), requested that the applicant provide an explanation for excluding the aging effect due to IASCC for the core shroud and core plate. In its response to RAI 3.1.2.1-6 (B), by letter dated January 31, 2005, the applicant stated that it will include aging effect due to IASCC in Table 3.1.2.2 of the LRA, and this aging effect is managed by implementing AMPs B.2.1.12, B.2.1.4, and B.2.1.5. AMP B.2.1.12 in turn invokes inspection requirements specified in BWRVIP-76, Boiling Water Reactor Core Shroud Inspection and Flaw Evaluation Guidelines, which is currently being reviewed by the staff. In AMP B.2.1.12, the applicant stated that it will comply with all the requirements that will be specified in the staffs SER on the BWRVIP-76 report, and will complete all the license renewal action items of this SER when it is issued. The staff reviewed the response and finds it acceptable because the implementation of the inspections as mandated by the ISI program (AMP B.2.1.4) and BWRVIP-76 (pending staffs approval) should identify any cracking due to IASCC in a timely manner so that the intended function of the subject component is maintained during the extended period of operation.

The staff, in RAI 3.1.2.1-6 (C) requested that the applicant provide information regarding the plant-specific experience related to IGSCC cracking of the stainless steel and nickel alloy components in the core shroud and AHCs, and the effective AMP that will be implemented on these systems at the BFN, units. In its response to RAI 3.1.2.1-6 (C), by letter dated January 31, 2005, the applicant stated that indications have been reported in BFN, Unit 1 core shroud welds H-1, H-2, H-3, H-4, and H-5. Core shroud welds H-6 and H-7 have not been examined due to interference from vibration sensing lines. These welds will be UT examined prior to BFN, Unit 1 restart. Indications have been reported in BFN, Unit 2 core shroud welds H-1, H-2, H-3, H-5, H-6, and H-7. Indications have been reported in BFN, Unit 3 core shroud welds H-1, H-2, H-3, H-4, H-5, and H-7. The applicant claimed that the aging effect due to IGSCC is managed by AMPs B.2.1.12, B.2.1.4 and B.2.1.5.

The staff in RAI 3.1.2.1-6 (D), requested that the applicant address the plant-specific experience regarding sudden increases in RCS water conductivity due to a leak in condensate and or reactor water clean up systems, and the impact of these sudden conductivity excursions on the IGSCC of core shroud welds. In its response to RAI3.1.2.1-6 (D), by letter dated January 31, 2005, the applicant stated that there were no increase in conductivity in RCS water due to leaks in condensate and/or reactor water clean up systems in the previous five years.

The staff finds that in the absence of any increase in RCS water conductivity, and with the addition of hydrogen/noble metal to the RCS water, the growth of existing IGSCC in the core shroud welds will be mitigated.

20 The staff, in RAI 3.1.2.1-6 (E), requested that the applicant provide information on verification methods to monitor the effectiveness of the HWC/NMCA program, the methodology of ensuring hydrogen availability in the core shroud region, monitoring of its availability with electro chemical potential (ECP) probes, and the validity of using secondary parameters (i.e., main steam/feedwater oxygen levels) to assess the hydrogen availability at core shroud welds. In its response to RAI3.1.2.1-6 (E), by letter dated January 31, 2005, the applicant stated that a NMCA with a conservative H2/O2 molar ratio is maintained to ensure hydrogen availability in the core shroud region. The applicant stated that it would not utilize ECP probes and, therefore, alternate means are used to monitor ECP. The applicant proposed to use reactor water H2/O2 molar ratio of greater than 4 for power operation. The staff reviewed the response and finds it acceptable because in the absence of ECP measurements, maintaining a H2/O2 molar ratio of greater than 4 would be effective in mitigating IGSCC in core shroud welds at the BFN units.

The staff finds that the implementation of the improved water chemistry, ISI programs in conjunction with the inspection guidelines specified in the BWRVIP-76 report (pending staffs approval) would enable the applicant to manage the aging effect due to IGSCC effectively during the extended period of operation, and would be consistent with AMPs XI.M1, XI.M2 and XI.M9 specified in the GALL report.

The staff, in RAI 3.1.2.2-7 (A), requested that the applicant provide an explanation for excluding the aging effect due to IASCC for the core spray spargers and piping in Table 3.1.2.2 of the LRA. According to Section IV B1.3-a of NUREG-1801, an AMP is required for monitoring IASCC in core spray spargers and piping. In its response to RAI 3.1.2.2-7 (A), by letter dated January 31, 2005, the applicant stated that it will include aging effect due to IASCC in Table 3.1.2.2 of the LRA, and this aging effect is managed by implementing AMPs B.2.1.12, B.2.1.4, and B.2.1.5.

The staff, in RAI 3.1.2.2-7 (B), requested that the applicant provide information on the type and extent of inspections to identify IGSCC and the mitigation techniques for core spray piping and spargers at BFN, Units 2 and 3. In its response to RAI 3.1.2.2-7 (B), by letter dated January 31, 2005, the applicant stated that the inspections (the type and extent) were performed in accordance with the requirements of BWRVIP-18, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines, and AMP B.2.1.4 (ISI program). The applicant stated that thus far no cracking was identified in the core spray system with the following exceptions. The applicant stated that it identified cracking in the elbow-to-shroud pipe and collar-to-shroud welds in downcomer C in BFN, Unit 3 which was subsequently replaced with a bolted piping assembly as a corrective action. The applicant identified cracking in the BFN, Unit 3 core spray sparger adjacent to the T-boxes which was repaired by welded brackets at both T-boxes. The applicant indicated that mitigation of IGSCC in core spray piping and spargers would be achieved by the implementation of HWC/NMCA. The staff, after reviewing BWRVIP-18 concludes that core spray piping and spargers are not adequately protected by the HWC/NMCA. However, implementation of the inspection guidelines as required by BWRVIP-18, and AMP B.2.1.4 (ISI program) does not take credit for HWC/NMCA. Therefore, the staff concludes that the type and extent of inspections mandated by BWRVIP-18 and ISI program should adequately identify cracking (without taking any credit for HWC/NMCA) in core spray piping and spargers in a timely manner so that their intended function is maintained during the period of extended operation. Since the applicant is implementing AMPs B.2.1.4, B.2.1.5 and B.2.1.12 which are consistent with the GALL report AMP XI.M9, the staff finds that the applicant has demonstrated that the effects of aging in core spray piping and spargers will

21 be adequately managed for the period of extended operation.

The staff, in RAI 3.1.2.2-8 (A), requested that the applicant provide an explanation for excluding the aging effect due to IASCC for the CRD housing dry tubes and guide tubes in Table 3.1.2.2 of the LRA. In its response to RAI 3.1.2.2-8 (A), by letter dated January 31, 2005, the applicant stated that it will include aging effect due to IASCC in Table 3.1.2.2 of the LRA, and this aging effect is managed by implementing AMPs B.2.1.12, B.2.1.4, and B.2.1.5. AMP B.2.1.12 in turn invokes inspection requirements specified in BWRVIP-47, Boiling Water Reactor Lower Plenum Inspection and Flaw Evaluation Guidelines. The staff reviewed the response and finds it acceptable because the implementation of the inspections as mandated by the ISI program (AMP B.2.1.4) and BWRVIP-47 should identify any cracking due to IASCC in a timely manner so that the intended function of the subject component is maintained during the extended period of operation.

The staff, in RAI 3.1.2.2-8 (B), requested that the applicant provide information regarding the past plant-specific experience related to IGSCC in the nickel alloy housing guide tubes and dry tubes, and their subsequent replacement with crack-resistant materials at BFN, Units 2 and 3.

The staff also requested that the applicant provide its plan for the replacement of BFN, Unit 1 dry tube and guide tubes. In its response to RAI 3.1.2.2-8 (B), by letter dated January 31, 2005, the applicant stated that all 12 BFN, Unit 2 and BFN, Unit 3 radiation monitor dry tubes were replaced with a crevice free design in the plunger area. Additionally, the material in the plunger area was changed from 304 stainless steel to 304L stainless steel, making the new dry tubes less susceptible to IGSCC.

To RLEP For Resolution: The applicant stated that it will replace all BFN, Unit 1 dry tubes prior to restart. The applicant must commit to replace all BFN, Unit 1 dry tubes prior to restart. This is a commitment which would be either contained in a tracking process for either BFN, Unit 1 restart or license renewal.

Based on this assessment, the applicant finds that the plant-specific experience related to the dry tubes has no impact on the attributes specified in AMP B.2.1.12 and BWRVIP-47. The staff reviewed the applicants response and concludes that the replacement of the dry tubes with more IGSCC resistant material combined with new crevice-free design in all BFN units provides adequate assurance that the aging effect due to IGSCC in these components is adequately managed at the BFN units for the extended period of operation.

The staff, in RAI 3.1.2.2-8 (C), requested that the applicant provide information regarding the plant specific experience related to IGSCC in furnace sensitized stainless steel stub tubes (if any) at the BFN units, and the method and frequency of inspections to identify this aging effect.

In its response to RAI 3.1.2.2-8 (C), by letter dated January 31, 2005, the applicant stated that BFN does not have furnace-sensitized stainless steel stub tubes in BFN, Units 1, 2, and 3, and the stub tubes are manufactured from a nickel alloy. The applicant also stated that there have been no repairs associated with the CRD stub tubes, and improvements in the BWR Chemistry Control Program help mitigate aging and degradation of the lower plenum components. Based on this assessment, the applicant claimed that the plant-specific experience related to the stub tubes has no impact on the attributes specified in AMP-B.2.1.12 and BWRVIP-47 as no degradation has been identified. The staff concurs with the applicants response and finds it acceptable.

22 The staff, in RAI 3.1.2.2-8(D), requested that the applicant provide information regarding the plant-specific experience related to IGSCC cracking in nickel alloy weld metals that were used for the CRD stub tubes, and the method and frequency of inspections to identify this aging effect. In its response to RAI 3.1.2.2-8 (D), by letter dated January 31, 2005, the applicant identified the following locations associated with the lower plenum that have nickel alloy weld metal.

• CRD Housing-to-Stub Tube Weld

• CRD Stub Tube-to-RV Weld

• In-Core Housing-to-RV Lower Head Penetration Weld

• In-Core Guide Tube-to-In-Core Housing Weld The applicant stated that its aging management review does not identify an inspection of the listed welds, and no cracking has been identified at BFN for the listed nickel alloy welds. The applicant also stated that the improvements in the BWR Chemistry Control Program help mitigate aging and degradation of the lower plenum components. Therefore, the applicant claimed that the plant-specific experience related to the lower plenum nickel alloy welds has no impact on the attributes specified in AMP B.2.1.12 and BWRVIP-47 as no degradation has been identified. The staff finds the applicants response acceptable, because the applicant adopted an improved water chemistry program which will help mitigate IGSCC.

The staff concludes that the implementation of the inspection requirements as mandated by the ISI program and the staffs approved BWRVIP-47 report will provide reasonable assurance that IGSCC in the lower plenum welds can be identified in a timely manner, so that the intended function of the subject component is maintained during the extended period of operation.

The staff, in RAI 3.1.2.2-10, requested that the applicant provide an explanation for excluding the aging effect due to IASCC for the top guide. The applicant in its response indicated that it will include IASCC as aging effect for the top guide in Table 3.1.2.2 of the LRA. Section 4.2.8.2 of the staffs SER on TLAA discusses the impact of IASCC and multiple failures of the top guide grid beams at the BFN units.

Conclusion On the basis of its review, the staff finds that the applicant has appropriately evaluated AMR results involving management of cracking due to SCC and IGSCC in the RVIs as recommended in the GALL report (with the exception of the Open Item 3.1.3.1.6.1-1). Since the applicants AMR results are otherwise consistent with the GALL report, the staff finds that the applicant has demonstrated that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21(a)(3).

3.1.3.1.6.2 Change in material properties and reduction in fracture toughness due to thermal aging and neutron irradiation embrittlement AMP recommended by the GALL report for managing the susceptibility of cast austenitic stainless steel (CASS) components to thermal aging embrittlement and neutron irradiation embrittlement is AMP XI.M13.

23 In Table 3.1.2.2 of the LRA, the applicant stated that the aging effects due to change in material properties as a result of thermal and neutron embrittlement of the CASS RVIs jet pump assemblies will be managed by AMP B.2.1.14, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel Program, AMP B.2.1.4, ASME Code Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program, AMP B.2.1.12, Boiling Water Reactor Vessel Internals Program, and the inspection guidelines that are provided in the BWRVIP-41, BWR Jet Pump Assembly Inspection and Flaw evaluation guidelines, which was approved by the staff. The implementation of these programs is consistent with the GALL report AMP XI.M13, and the applicant did not take any exception to the requirements of the GALL. The applicant incorporated a screening criteria which establishes susceptibility of CASS components to thermal aging based on casting method, molybdenum content, and percent ferrite.

The staff, in RAI 3.1.2.2-9, requested that the applicant provide information on the existing (if any) CASS jet pump components, the type of casting, composition of the CASS (i.e.

molybdenum content and delta ferrite values), previous plant-specific experience regarding the cracked components and subsequent inspection of any cracked CASS jet pump components due to neutron and thermal embrittlement, and any technical specification changes related to jet pump components.

In its response to RAI 3.1.2.2-9, by letter dated January 31, 2005, the applicant indicated that the CASS jet pump components were manufactured to ASTM A351, grade CF8. These castings are low molybdenum and the maximum calculated delta ferrite percentage is below 20%. According to Table 2, CASS Thermal Aging Susceptibility Screening Criteria, contained in the May 19, 2000 NRC letter from Christopher Grimes to Douglas J. Walters, materials which have a low molybdenum content and < 20% delta ferrite are not susceptible to thermal aging for statically or centrifugally cast components. The NRC letter from Christopher Grimes to Carl Terry, by letter dated June 5, 2001, states: It is important to note that thermal and/or neutron embrittlement of CASS components becomes a concern only if cracks are present in the components, and that cracking has not been observed in CASS jet pump assembly components. Section 2.4 of the same letter states: Further, the BWRVIP and the NRCs Office of Nuclear Regulatory Research (RES) is engaged in a joint confirmatory research program to determine the effects of high levels of neutron fluence on BWR internals. The applicant has stated in its LRA that for open issues between the BWRVIP and NRC, the applicant will work as part of the BWRVIP to resolve these issues generically. When resolved, the applicant will follow the BWRVIP recommendations resulting from that resolution. The BWR RVIs Program requires inspections of several jet pump assembly welds, which are more susceptible to cracking than the CASS components and will serve as an indication of the potential need for more extensive inspections later in life.

Similar to the CASS jet pump components, the orificed fuel supports (OFS) are also manufactured to ASTM A351, grade CF8. These castings are low molybdenum and the maximum calculated delta ferrite percentage is below 20%. For similar reasons as discussed for the jet pump CASS components, the applicant concluded that no program is needed to manage the effects of thermal/neutron embrittlement of the CASS orificed fuel supports.

The staff agrees with the applicants response related to the implementation of industry recommended monitoring program on the effects of high levels of neutron fluence on the CASS components. The staff concludes that the applicants justification for excluding the CASS jet

24 pumps and OFS components from the AMR for the extended period of operation is acceptable provided that AMPs B.2.1.4 and B.2.1.12 and inspection requirements of BWRVIP-41 are fully implemented for these components. The staff concurs with the applicants claim that continued implementation of these AMPs and the technical guidelines of the BWRVIP-41 report provides reasonable assurance that the aging effects are adequately managed in the RV CASS jet pumps and OFS components.

Conclusion On the basis of its review, the staff finds that the applicant has appropriately evaluated AMR results involving management of aging effects due to change in material properties and reduction in fracture toughness due to thermal aging and neutron embrittlement in the RVIs as recommended in the GALL report. Since the applicants AMR results are otherwise consistent with the GALL report, the staff finds that the applicant has demonstrated that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21(a)(3).

AMR Results That are not Addressed in the GALL Report 3.1.3.1.6.3 Loss of material due to galvanic, general, crevice, and pitting corrosion In Table 3.1.2.2 of the LRA, the applicant addressed loss of material due to galvanic, general, crevice, and pitting corrosion in (1) RVIs core shroud and core plate; (2) RVI core spray piping and spargers; (3) RVIs control rod housing and dry tubes and guide tubes; (4) RVIs jet pump assemblies; and (5) RVIs top guide.

The applicant also identified the implementation of relevant AMPs to manage the aging effects due to galvanic, general, crevice, and pitting corrosion of stainless steel and nickel alloy materials when these materials are exposed to the BWR treated water environment. In Table 3.1.2.2 of the LRA, the applicant included AMP requirements that are specified in NUREG-1801, Vol. 2, Table IV.B1 for each of the aforementioned components. However, NUREG-1801, Vol. 2, Table IV.B1, does not identify loss of material due to crevice, general, and pitting corrosion as aging effects in stainless steel and nickel alloy materials that are used in the aforementioned RV components when these components are exposed to the BWR treated water environment. The staffs evaluation of the AMR related to these aging effects are discussed below:

In Table 3.1.2.2 of the LRA, the applicant stated that the aging effects due to galvanic, general, crevice, and pitting corrosion of stainless steel and nickel alloy materials in the RVIs will be managed by AMP B.2.1.12, Boiling Water Reactor Vessel Internals Program, AMP B. 2.1.5, Chemistry Control Program, and the inspection guidelines that are provided in the following BWRVIP reports for the applicable internal components.

BWRVIP Boiling Water Reactor Core Spray Internal Inspection and Flaw Evaluation Guidelines.

BWRVIP Boiling Water Reactor Core Plate Inspection and Flaw Evaluation Guidelines.

25 BWRVIP Boiling Water Reactor Top Guide Inspection and Flaw Evaluation Guidelines.

BWRVIP Boiling Water Reactor Jet Pump Assembly Inspection and Flaw Evaluation Guidelines.

BWRVIP Boiling Water Reactor Lower Plenum Inspection and Flaw Evaluation Guidelines.

BWRVIP Boiling Water Reactor Core Shroud Inspection and Flaw Evaluation Guidelines.-Staff review is not complete.

The implementation of these additional guidelines and AMPs is consistent with GALL report AMP XI.M9. The applicant claimed that continued implementation of these AMPs provides reasonable assurance that the aforementioned aging effects are adequately managed in the RVIs. The staff concludes that the implementation of AMP B.2.1.5 Chemistry Control Program will provide adequate controls on BWR reactor water chemistry which in turn controls general, pitting and crevice corrosion in RVIs. Furthermore, inspection guidelines that are specified in the aforementioned staff approved (with the exception of BWRVIP-76) BWRVIP reports will provide adequate guidance in performing the necessary inspections so that these aging effects in RVIs are properly identified in a timely manner.

Conclusion On the basis of its review, the staff finds that the applicant has appropriately evaluated AMR results involving management of aging effects due to general crevice and pitting corrosion in RVIs. The staff finds that the applicant has demonstrated that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21(a)(3).