L-08-124, Reply to Request for Additional Information for Review of the License Renewal Application and License Renewal Application Amendment No. 5

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Reply to Request for Additional Information for Review of the License Renewal Application and License Renewal Application Amendment No. 5
ML080980212
Person / Time
Site: Beaver Valley
Issue date: 04/02/2008
From: Sena P
FirstEnergy Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-08-124, TAC MD6593, TAC MD6594
Download: ML080980212 (25)
v • d • e


Text

FENOC FirstEnergyNuclear OperatingCompany PeterP. Sena III 724-682-5234 Site Vice President Fax: 724-643-8069 April 2, 2008 L-08-124 10 CFR 54 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit Nos. 1 and 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Reply to Request for Additional Information for Review of the Beaver Valley Power Station, Units 1 and 2. License Renewal Application (TAC Nos. MD6593 and MD6594),

and License Renewal Application Amendment No. 5 Reference 1 provided the FirstEnergy Nuclear Operating Company (FENOC) License Renewal Application for the Beaver Valley Power Station (BVPS). Reference 2 requested additional information regarding Section 4.2 of the BVPS License Renewal Application. This letter provides the FENOC reply to the U.S. Nuclear Regulatory Commission (NRC) request for additional information (RAI). This letter also provides Amendment No. 5 to the BVPS License Renewal Application, including a revised License Renewal future commitment, based on changes resulting from the FENOC reply to the NRC RAI.

Editorial changes were made to RAI questions RAI #2, #3 and #4, based on a RAI clarification discussion with the NRC on March 12, 2008, and are reflected in Attachment 2.

Attachment 1 provides the regulatory commitment list. Attachment 2 provides the FENOC reply to the NRC RAI. The Enclosure provides Amendment No. 5 to the BVPS License Renewal Application.

If there are any questions or if additional information is required, please contact Mr. Clifford I. Custer, Fleet License Renewal Project Manager, at 724-682-7139.

Xiop0

Beaver Valley Power Station, Unit Nos. 1 and 2 L-08-124 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on April X) , 2008.

S in c erely , / 0* "7 Peter P. Sena III

References:

1. FENOC Letter L-07-113, "License Renewal Application," August 27, 2007.
2. NRC Letter, "Request for Additional Information for the Review of the Beaver Valley Power Station, Units 1 and 2, License Renewal Application (TAC Nos.

MD6593 and MD6594)," March 5, 2008.

Attachments:

1. Regulatory Commitment List
2. Reply to Request for Additional Information Regarding Beaver Valley Power Station, Units 1 and 2, License Renewal Application, Section 4.2.2

Enclosure:

Amendment No. 5 to the BVPS License Renewal Application cc: Mr. K. L. Howard, NRC Project Manager Mr. S. J. Collins, NRC Region I Administrator cc: w/o Attachments or Enclosure Dr. P. T. Kuo, NRC Director, Division of License Renewal Mr. D. L. Werkheiser, NRC Senior Resident Inspector Ms. N. S. Morgan, NRR Project Manager Mr. D. J. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)

ATTACHMENT I L-08-124 Regulatory Commitment List Page 1 of 1 The following list identifies those actions committed to by FirstEnergy Nuclear Operating Company (FENOC) for Beaver Valley Power Station (BVPS) Unit Nos. I and 2 in this document. Any other actions discussed in the submittal represent intended or planned actions by FENOC. They are described only as information and are not Regulatory Commitments. Please notify Mr. Clifford I. Custer, Project Manager, Fleet License Renewal, at (724) 682-7139, of any questions regarding this document or associated Regulatory Commitments.

Re-ulatory Commitment Due Date

1. WCAP-15571, Supplement 1, September 30, 2008 Section 6.2, will be corrected to show that the data for the surveillance program weld material are deemed "not credible", and that the data were used with a qa margin of 28 OF. The corrected WCAP will be submitted to the NRC.
2. The Unit 2 Pressure-Temperature Limits September 30, 2008 Report (PTLR), currently Revision 2, will be updated to include the results of the Capsule X analysis documented in WCAP-16527-NP and WCAP-16527-NP, Supplement 1. The updated PTLR will be submitted to-the NRC.

ATTACHMENT 2 L-08-124 Reply to Request for Additional Information Regarding Beaver Valley Power Station, Units 1 and 2, License Renewal Application, Section 4.2.2 Page 1 of 17 Question RAI #1 Tables 4.2-5 and 4.2-6 of the iBeaver Valley License Renewal Application (LRA) lists Chemistry Factor (CF) and pressurized thermal shock (PTS) reference temperature (RTPTS) values for several reactor vessel (RV) beltline materials that were calculated in accordance with both Regulatory Positions (RPs) 1.1 and 2.1 of Regulatory Guide (RG) 1.99, Rev. 2. Only one of these two RPs may be used for determining the actual CF, adjusted reference temperature (ART), and RTPTS value for each material. Therefore, please indicate which RP (RP 1.1 or RP 2.1) was used in determining the actual CF, ART, and RTPTS values for the RV beltline materials. Please provide justification for the selection of RP 1.1 or 2.1 for each material, based on factors such as surveillance data credibility or non-credibility, conservatism of RP 2.1 data, or other factors (i.e., NRC recommendation that non-credible surveillance data be used for calculating the CF for limiting plate B6903-1 with full a, margin of 17 °F).

RESPONSE RAI #1 WCAP-1 5571, Supplement 1, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program" (Reference 1),

documents the BVPS Unit 1 end-of-license-extended (EOLE) analysis for PTS.

WCAP-1 6527-NP, Supplement 1, "Analysis of Capsule X from First Energy Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program" (Reference 2),

documents the BVPS Unit 2 EOLE analysis for PTS. In these analyses, the chemistry factors (CFs) were determined using Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Revision 2 (Reference 3), Positions 1.1 and 2.1.

Position 1.1 uses the Tables from the Regulatory Guide along with the best estimate copper and nickel weight percents, which are presented in LRA Table 4.2-5 (Reference 4) and LRA Table 4.2-6 (Reference 5). Position 2.1 uses surveillance capsule data.

In accordance with Regulatory Position 1.1, the values of 0A are 28 °F for welds and 17 OF for base metal, except that aA need not exceed 0.50 times the mean value of ARTNDT. In accordance with Regulatory Position 2.1, when there are credible surveillance data, cA is the lesser of /2 ARTNDT, or 14 OF for welds, or 8.5 °F for base metal. In using Regulatory Position 2.1, when there are non-credible surveillance data, FENOC uses 28 OF for welds and 17 °F for base metal. In a February 12, 1998, NRC-Industry Meeting, the NRC recommended that non-credible surveillance data be used for calculating the CF for limiting plate B6903-1 with full 0 A margin of 17 OF. This

L-08-124 Page 2 of 17 methodology was approved by the NRC in a Safety Evaluation dated February 20, 2002 (Reference 6).

In accordance with Regulatory Guide 1.99, Revision 2, if Position 2.1 results in a higher value of ART (or RTpTs) than Position 1.1, the ART (or RTpTs) calculated per Position 2.1 is used. However, if Position 2.1 results in a lower value of ART (or RTPTS) than Position 1.1, either value of ART (or RTPTS) may be used. For conservatism, if Position 2.1 results in a lower value of ART (or RTPTS) than Position 1.1, FENOC uses the ART (or RTPTS) calculated per Position 1.1 (References 1, 2, 7 and 8).

Surveillance capsule data from BVPS Unit 1 and Unit 2, St. Lucie Unit 1, and Fort Calhoun Unit 1 were used to determine the Regulatory Position 2.1 CF (References 1, 2, 7 and 8). Surveillance material data credibility or non-credibility are as follows (References 9 and 10):

Unit 1

  • Lower Shell Plate B6903-1: non-credible BVPS Unit 1surveillance data
  • Intermediate to Lower Shell Girth Weld 11-714 (Heat 90136): credible St. Lucie Unit 1 surveillance data
  • Intermediate Shell Longitudinal Weld 19-714 A&B (Heat 305424): non-credible BVPS Unit 1 surveillance data
  • Lower Shell Longitudinal Weld 20-714 A&B (Heat 305414): non-credible Fort Calhoun Unit 1 surveillance data Unit 2
  • Intermediate Shell Plate B9004-2: credible BVPS Unit 2 surveillance data

" Lower Shell Longitudinal Weld 101-142 A&B (Heat 83642): credible BVPS Unit 2 surveillance data

  • Intermediate Shell Longitudinal Weld 101-124 A&B (Heat 83642): credible BVPS Unit 2 surveillance data
  • Intermediate to Lower Shell Girth Weld 101-171 (Heat 83642): credible BVPS Unit 2 surveillance data

References:

1. WCAP-1 5571, Supplement 1, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," July 2007.

L-08-124 Page 3 of 17

2. WCAP-1 6527-NP, Supplement 1, "Analysis of Capsule X from First Energy Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program,"

July 2007.

3. Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"

Rev. 2.

4. LRA Table 4.2-5, "Unit 1 Beltline Region PTS Data for EOLE."
5. LRA Table 4.2-6, "Unit 2 Beltline Region PTS Data for EOLE."
6. Safety Evaluation By The Office of Nuclear Reactor Regulation Related to Amendment No. 249 to Facility Operating License No. DPR-66, Pennsylvania Power Company, Ohio Edison Company, FirstEnergy Nuclear Operating Company, Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, February 20, 2002 (ML020510501).
7. BVPS Unit 1, "Pressure and Temperature Limits Report," Rev. 4, September 18, 2007 (ML072670026).
8. BVPS Unit 2, Pressure and Temperature Limits Report, Rev. 2 (ML071790328).
9. WCAP-1 5571, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," Rev. 0.
10. WCAP-1 6527-NP, "Analysis of Capsule X from First Energy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program,"

Rev. 0.

L-08-124 Page 4 of 17 Question RAI #2 For Beaver Valley, Unit 1 (Beaver Valley 1), the staff noted several discrepancies among LRA Section 4.2.2, LRA Table 4.2-5, Appendix D of WCAP-1 5571, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit I Reactor Vessel Radiation Surveillance Program, Rev. 0" (Surveillance Data Credibility Analysis), and WCAP-15571, Supplement 1, regarding the application of surveillance data for determining the RTpTs value for intermediate Shell Longitudinal Weld 19-714 (Heat 305424).

First, Page D-5 of Appendix D of WCAP-15571 states: "The surveillance weld

[Heat 305424] has two out of four data points outside the 28 OF scatter band.

Hence, the surveillance data is not credible." Please reconcile this statement with the statement in the second paragraph of Page 4.2-6 of the LRA, which indicates that the "data for the Unit 1 surveillance program weld material is deemed credible" and the similar statement in Section 6.1 of Supplement I to WCAP-15571 indicating that "the data for the surveillance program weld material is deemed credible." It should be noted in reconciling these statements that Intermediate Shell Longitudinal Weld 19-714 (Heat 305424) is represented in the Beaver Valley I surveillance program. However, a review of Appendix D of WCAP-15571 shows that two of the four surveillance data points (Capsules "V" and "Y") fall outside of the 28 OF ARTNDT scatter band. Therefore, the surveillance weld data is not credible, and the CF for weld 19-714 must be determined using RP 1.1.

Second, the second paragraph of Page 4.2-6 of the LRA states that the data for the Beaver Valley I surveillance program weld material was "used with a a,&

margin of 14 OF." Likewise, Section 6.1 of Supplement I to WCAP-15571 indicates that "The [surveillance program weld] data was used with a aA margin of 14 OF."

Please reconcile these statements with the 28 OF value for UA presented in Table 4.2-5 for Intermediate Shell Longitudinal Weld 19-714, based on RP 2.1. It should be noted that, per RG 1.99, Rev. 2, the correct value for a,&is 28 OF, based on RP 1.1, and the use of RP 2.1 is not permitted for this weld because the surveillance data is not credible. Furthermore, the use of a a& value of 14 OF is not permitted because the surveillance data is not credible.

RESPONSE RAI #2 WCAP-15571, Supplement 1 (Reference 1), documents the BVPS Unit 1 EOLE analysis for PTS. In this analysis, the chemistry factors (CFs) were determined using Regulatory Guide 1.99 Revision 2 (Reference 2), Positions 1.1 and 2.1. Position 1.1 uses the Tables from the Regulatory Guide along with the best estimate copper and nickel weight percents, which are presented in LRA Table 4.2-5 (Reference 3).

Position 2.1 uses surveillance capsule data.

L-08-124 Page 5 of 17 As documented in Appendix D (Page D-5) of WCAP-15571 (Reference 4), the surveillance weld 19-714 [Heat 305424] has two out of four data points outside the 28 OF scatter band and therefore, the surveillance data are deemed non-credible. In accordance with Regulatory Position 1.1, the values of 0c are 28 OF for welds and 17 OF for base metal, except that uA need not exceed 0.50 times the mean value of ARTNDT.

In accordance with Regulatory Position 2.1, when there are credible surveillance data, 0

A is the lesser of 1/2 ARTNDT or 14 OF for welds, or 8.5 °F for base metal. In using Regulatory Position 2.1, when there are non-credible surveillance data, FENOC uses 28 OF for welds and 17 OF for base metal. In the February 12, 1998 NRC-Industry Meeting, the NRC recommended that non-credible surveillance data be used for calculating the CF for limiting plate B6903-1 with full cu margin of 17 OF. This methodology was approved by the NRC in a Safety Evaluation dated February 20, 2002 (Reference 5). Therefore, LRA Table 4.2-5 correctly shows the qa value for surveillance weld 19-714 [Heat 305424] as 28 °F for Regulatory Positions 1.1 and 2.1. In addition, Table 7 of WCAP-15571 Supplement 1 correctly shows the ciA value for surveillance weld 19-714 [Heat 305424] as 28 OF for Regulatory Positions 1.1 and 2.1.

The statements in the second paragraph of Page 4.2-6 of the LRA (Reference 6) and similar statements in Section 6.1 of WCAP-1 5571, Supplement 1, indicate that the data for the Unit 1 surveillance program weld material are deemed credible, and that the data were used with a uA value of 14 OF. These statements are incorrect, and corrections will be made as follows:

1. WCAP-1 5571, Supplement 1, Section 6.2, will be corrected to show that the data for the surveillance program weld material are deemed not credible, and that the data were used with a ca margin of 28 OF. The corrected WCAP will be submitted to the NRC. See Attachment 1 to this letter, "Regulatory Commitment List,"

Commitment No. 1.

2. LRA Section 4.2.2, "Pressurized Thermal Shock," 2 nd paragraph of Page 4.2-6, is revised to indicate that the data for the surveillance program weld material are deemed not credible, and that the data were used with a an,& margin of 28 OF. See the revised LRA text in the Enclosure to this letter.

References:

1. WCAP-1 5571, Supplement 1, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," July 2007.
2. Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"

Rev. 2.

3. LRA Table 4.2-5, "Unit 1 Beltline Region PTS Data for EOLE."

L-08-124 Page 6 of 17

4. WCAP-1 5571, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," Rev. 0.
5. Safety Evaluation By The Office of Nuclear Reactor Regulation Related to Amendment No. 249 to Facility Operating License No. DPR-66, Pennsylvania Power Company, Ohio Edison Company, FirstEnergy Nuclear Operating Company, Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, February 20, 2002 (ML020510501).
6. LRA Section 4.2.2, Page 4.2-6.

L-08-124 Page 7 of 17 Question RAI #3 Intermediate-to-Lower Shell Circumferential Weld 11-714 (Heat 90136) and Lower Shell Longitudinal Weld 20-714 (Heat 305414) are not represented in the Beaver Valley I surveillance program. However, CF and RTpTs values based on RP 2.1 were reported in Table 4.2-5 for these welds. Please confirm whether the heats for these welds are represented in the surveillance programs for St. Lucie (Heat 90136) and Fort Calhoun (Heat 305414).

The staff reviewed Revision 4 of the Beaver Valley 1 Pressure-Temperature Limits Report (PTLR). Table 5.2-4a of the Beaver Valley 1 PTLR provides a note indicating that the use of the St. Lucie and Fort Calhoun surveillance capsule data was approved by the NRC by letter dated February 20, 2002. Per your response to RAI Question 1, please indicate whether this surveillance data was used (per RP 2.1) in the determination of the actual CF, RTPTS, and ART values for these welds. Please verify whether these surveillance data sets were deemed credible in accordance with RG 1.99, Rev. 2, and provide references for the documents where the analyses for determining credibility (or non-credibility) may be found.

Please indicate whether the CFs for these welds based on RP 2.1 (84.8 for Intermediate-to-Lower Shell Circumferential Weld 11-714 and 223.9 for Lower Shell Longitudinal Weld 20-714) were calculated by adjusting the measured ARTNDT values by the ratio of the CF for the vessel weld to the CF for the surveillance weld, as prescribed in RG 1.99, Rev. 2. If the ARTNDT values were properly adjusted for determining these CF values, please provide the CF ratio adjustment factors for these welds or a reference for the document where these adjustment factors may be obtained. If the ARTNDT values were not adjusted for determining these CF values, please modify the Beaver Valley 1 PTLR and LRA Table 4.2-5 to include CF calculations based on RP 2.1 for these welds that account for this adjustment.

RESPONSE RAI #3 As documented in Attachment D of WCAP-1 5571 (Reference 1), weld 11-714 (Heat 90136) is represented in the surveillance program for St. Lucie Unit 1, and the surveillance data are deemed credible in accordance with Regulatory Guide 1.99 Revision 2 (Reference 2). The credibility analysis of the St. Lucie Unit 1 surveillance data is found in WCAP-1 5446 (Reference 3). As documented in WCAP-1 5571, Supplement 1 (Reference 4), the St. Lucie Unit 1 surveillance data for weld 11-714 (Heat 90136) were used to determine the Regulatory Position 2.1 chemistry factor (CF) that is reported in LRA Table 4.2-5 (Reference 5).

L-08-124 Page 8 of 17 As documented in Attachment D of WCAP-1 5571 (Reference 1), the Lower Shell Longitudinal Weld 20-714 A&B (Heat 305414) is represented in the surveillance program for Fort Calhoun Unit 1, and the surveillance data are deemed non-credible in accordance with Regulatory Guide 1.99, Revision 2 (Reference 2). The credibility analysis was performed in Attachment D of WCAP-1 5571, and the data are deemed non-credible. As documented in WCAP-1 5571, Supplement 1 (Reference 4), the Fort Calhoun Unit 1 surveillance data for weld 20-714 A&B (Heat 305414) were used to determine the Regulatory Position 2.1 chemistry factor (CF) that is reported in LRA Table 4.2-5 (Reference 5).

The latest BVPS Unit 1 PTLR is Revision 4 (Reference 6). This revision changed the applicability of the heatup and cooldown requirement to 30 effective full power years (EFPY). WCAP-16799-NP (Reference 7) generated the P-T limit curves for 30 EFPY, and the results were reported in the BVPS Unit 1 PTLR, Revision 4. Section 2 of WCAP-1 6799-NP notes that, in the calculations of chemistry factors, the ratio was applied to account for chemistry differences between the vessel weld material and the surveillance weld material, and the temperature differences between BVPS Unit 1 and the St. Lucie and Fort Calhoun plants. In addition, Section 2 of WCAP-16799-NP notes that the details of the ratio procedures have been previously documented in WCAP-15570 (Reference 8), and remained unchanged for this analysis. Table 4-12 of WCAP-15570 provides the CF calculation for BVPS Unit 1 weld 11-714 (Heat 90136) using St.

Lucie Unit 1 surveillance data, and BVPS Unit 1 weld 20-714 A&B (Heat 305414) using Fort Calhoun Unit 1 surveillance data. Table 4-12 notes that the CF ratio adjustment factors for St. Lucie and Fort Calhoun data are 1.17 and 0.993, respectively.

References:

1. WCAP-1 5571, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," Rev. 0.
2. Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"

Rev. 2.

3. WCAP-1 5446, "Analysis of Capsule 2840 from the Florida Power & Light Company St. Lucie Unit 1 Reactor Vessel Radiation Surveillance Program," September 2000 (ML003756770).
4. WCAP-1 5571, Supplement 1, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," July 2007.
5. LRA Table 4.2-5, "Unit 1 Beltline Region PTS Data for EOLE."
6. BVPS, Unit 1, "Pressure and Temperature Limits Report," Rev. 4, September 18, 2007 (ML072670026).

L-08-124 Page 9 of 17

7. WCAP-1 6799-NP, "Beaver Valley Power Station Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," Rev. 1.
8. WCAP-1 5570, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," Rev. 2 (ML011870482).

L-08-124 Page 10 of 17 Question RAI #4 Table 5.2-5 of the Beaver Valley 1 PTLR (Rev. 4) states that the CF for Lower Shell Plate B6903-1 is 147.2 (based on RP 1.1). This is the incorrect CF for this plate, per the February 12,1998 NRC-Industry meeting, where the NRC recommended that the non-credible surveillance data for this specific plate be used along with a full a,&of 17 OF for RTPTs and ART calculations. LRA Section 4.2.2 accurately reflects that the non-credible surveillance data and full a& of 17 OF were used to arrive at a 54 effective full power year (EFPY) RTPTS value of 275.7 OF, based on a RP 2.1 CF value of 149.2. Furthermore, Table 5.2-7 of the PTLR provides ART calculations for this limiting plate that are based on the correct CF value of 149.2 and states that these calculations are based on the non-credible plate surveillance data and full a,&of 17 OF. Please modify Table 5.2-5 of the Beaver Valley I PTLR to reflect the correct CF (149.2) for Lower Shell Plate B6903-1. The application of surveillance data and selection of CFs for calculation of RTPTs and ART values in the Beaver Valley I PTLR should be consistent with the LRA.

RESPONSE RAI #4 Table 5.2-5 of the BVPS Unit 1 PTLR (Reference 1) does not require modification to show a chemistry factor (CF) of 149.2 for lower shell plate B6903-1. Table 5.2-5 applies to CFs that were taken from Table 1 (welds) and Table 2 (base metal) of Regulatory Guide 1.99, Revision 2 (Reference 3). Therefore, Table 5.2-5 correctly shows a CF of 147.2 for lower shell plate B6903-1.

Table 5.2-4 of the BVPS Unit 1 PTLR (Reference 1) applies to CFs that were calculated using surveillance data. Therefore, Table 5.2-4 correctly shows a CF of 149.2 for lower shell plate B6903-1.

As documented in WCAP-1 5571, Supplement 1 (Reference 3) and the BVPS Unit 1 PTLR (Reference 1), the CFs were determined using Regulatory Guide 1.99, Revision 2 (Reference 2), Positions 1.1 and 2.1. Position 1.1 chemistry factors are taken from Table 1 (welds) and Table 2 (base metal) of the Regulatory Guide. Position 2.1 chemistry factors are calculated using surveillance data.

References:

1. BVPS Unit 1, "Pressure and Temperature Limits Report," Rev. 4, September 18, 2007 (ML072670026; also see ML071790328, Enclosure 2).
2. Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"

Rev. 2.

L-08-124 Page 11 of 17

3. WCAP-1 5571, Supplement 1, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," July 2007.

L-08-124 Page 12 of 17 Question RAI #5 LRA Section 4.2.2 states that a neutron flux management program is in place at Beaver Valley I for ensuring that the limiting material would meet the PTS screening requirements of 10 CFR 50.61 at the end of the current 40-year license term. Please verify whether the limiting material is projected to exceed the 270 OF screening limit of 10 CFR 50.61 in the year 2033 (43.87 EFPY) under this same flux management program. If the current flux management program will not maintain the limiting material below the PTS screening limit until 2033 (43.87 EFPY), please discuss any additional measures that are required to ensure that the limiting material does not exceed the PTS screening limit until 2033 (43.87 EFPY).

LRA Section 4.2.2 states that documentation of a flux reduction program for Beaver Valley I will be submitted in accordance with the requirements of 10 CFR 50.61. The staff requests that you provide a formal commitment to submit the appropriate documentation of your program for maintaining the limiting RV beltline plate (Plate B6903-1) below the 10 CFR 50.61 PTS screening criterion through the end of the period of extended operation (54 EFPY). This commitment must include a schedule for submitting this documentation relative to the projected date (year 2033, 43.87 EFPY) when the limiting material will exceed 270 °F PTS screening limit.

Note: For the Palisades LRA, which documented a similar situation regarding PTS, the licensee provided Long Term Commitment No. 3, which reads as follows:

At the appropriatetime, priorto exceeding the PTS screening criteria,Palisadeswill select the optimum alternative to manage PTS in accordance with the NRC regulationsand make relevant submittals to obtain NRC review and approval.

RESPONSE RAI #5 As documented in WCAP-1 5571, Supplement 1(Reference 1), the limiting plate material in the BVPS Unit I Reactor Vessel beltline is the lower shell plate B6903-1, with a projected EOLE (60 years of operation) RTPTS value of 275.71F using the BVPS 2Unit 1 surveillance capsule data for 54 EFPY (equivalent to a fluence of 6.09E19 n/cm (E > 1.0 MeV)). If mitigating actions are not implemented, the 10 CFR 50.61 (Reference 2) screening criteria of 270 OF for lower shell plate B6903-1 will be reached at a fluence level of 4.961 E19 n/cm 2 (E > 1.0 MeV), which is equivalent to 43.87 EFPY.

At the end of Cycle 18 (9/23/2007), BVPS Unit 1 had accrued 21.00 EFPY. Assuming a 90% capacity factor from the start of Cycle 19 (10/14/2007), the Unit 1 Reactor Vessel

L-08-124 Page 13 of 17 is projected to reach the PTS screening criteria of 270 'F on the limiting plate (B6903-1) in the year 2033.

FENOC provides a license renewal future commitment to submit the appropriate documentation of the BVPS Unit 1 Flux Reduction Plan for maintaining the limiting Reactor Vessel beltline plate (Plate B6903-1) below the 10 CFR 50.61 PTS screening criteria through the end of the period of extended operation (54 EFPY). The license renewal future commitment in LRA, Appendix A, Table A.4-1, Item Number 24, is revised to read:

"Priorto exceeding the PTS screening criteriafor BVPS Unit 1, FENOC will select a flux reduction measure to manage PTS in accordance with the requirements of 10 CFR 50.61. A flux reduction plan will be submitted for NRC review and approval at least I yearpriorto implementation of the flux reduction measure."

See the revised license renewal future commitment text in the Enclosure to this letter.

LRA Section 4.2.2, "Pressurized Thermal Shock," (Unit 1) page 4.2-6, paragraph 5, and Section A.2.2.2, "Pressurized Thermal Shock," (Unit 1) page A.2-4, 4 th paragraph, are revised to read:

"Therefore, a sensitivity assessment of available flux reduction measures was completed. The sensitivity assessment included several fuel management scenarios (such as low leakage core design, low power peripheralfuel assemblies, reinsertion of hafnium rods, and the use of part length shielded assemblies)and several assumed capacity factors up to 98 percent. Several flux reduction options are available which would maintain the limiting plate below the PTS screening criterion to the EOLE.

The flux reduction program will be managed under the Reactor Vessel Integrity Program(Section B.2.35). Priorto exceeding the PTS screening criteriafor BVPS Unit 1, FENOC will select a flux reduction measure to manage PTS in accordance with the requirements of 10 CFR 50.61. A flux reduction plan will be submitted for NRC review and approvalat least I year priorto implementation of the flux reduction measure."

See the revised LRA text in the Enclosure to this letter.

References:

1. WCAP-1 5571, Supplement 1, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," July 2007.
2. 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events."

L-08-124 Page 14 of 17 Question RAI #6 The staff noted that the 54 EFPY Upper Shelf Energy (USE) value for the limiting plate at Beaver Valley I (Plate B6903-1) may be slightly less than 50 ft-lbs (lowest allowable USE value at EOL per 10 CFR Part 50, Appendix G), when calculated using Figure 2 from RG 1.99, Rev. 2, without applying the surveillance data.

Please discuss why the surveillance data were deemed credible for determining the USE value for this plate, based on the five credibility criteria specified in Section B of RG 1.99, Rev. 2.

RESPONSE RAI #6 Section B (Criterion 3) of Regulatory Guide 1.99, Revision 2 (Reference 1), reads as follows:

"When there are two or more sets of surveillance data from one reactor, the scatterof ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28 *Ffor welds and 17 'F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scattershould not exceed twice those values.

Even if the data fail this criterion for use in shift calculations,they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82 (Ref. 1);"

As documented in Appendix D of WCAP-15571 (Reference 2), the scatter for the surveillance data of plate B6903-1 is small enough to permit the determination of the upper shelf energy (USE), unambiguously. As documented in WCAP-1 5571, Supplement 1 (Reference 3), surveillance data were used to determine the projected EOLE (60-years of operation) USE value for plate B6903-1. As discussed in Section 6.2 and shown on Figure 1 of WCAP-15571, Supplement 1, measured drops in USE determined from surveillance data (Capsules V, U, W and Y) for surveillance plate B6903-1 is plotted on Figure 2 of Regulatory Guide 1.99, Revision 2, with a horizontal line drawn parallel to the existing lines as the upper bound of all data. From Figure 1 of WCAP-1 5571, Supplement 1, the decrease in USE (%) for surveillance plate B6903-1 is interpreted as 38 %. Therefore, the projected EOLE (60-years of operation) USE for plate B6903-1 is 51.5 ft-lb, and maintains USE throughout the life of the vessel of no less than 50 ft-lb in accordance with 10 CFR 50, Appendix G.

References:

1. Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"

Rev. 2.

L-08-124 Page 15 of 17

2. WCAP-1 5571, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," Rev. 0.
3. WCAP-1 5571, Supplement 1, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," July 2007.

L-08-124 Page 16 of 17 Question RAI #7 A review of Appendix D of WCAP-1 6527-NP indicates that the data for the surveillance weld at Beaver Valley 2 (Heat No. 83642) were not adjusted by the ratio of the CF for the vessel weld to the CF for the surveillance weld, as prescribed in RG 1.99, Rev. 2. Please verify that the copper and nickel content of the surveillance weld is identical to that for all vessel welds at Beaver Valley 2. If there is a difference between the chemistry of the surveillance weld and that of the vessel welds, please modify LRA Section 4.2.2, LRA Table 4.2-6, and the Beaver Valley 2 PTLR to account for the CF ratio adjustment.

RESPONSE RAI #7 Appendix D of WCAP-1 6527-NP (Reference 1) addresses only the credibility evaluation of the surveillance materials. Hence, this credibility evaluation for the surveillance weld material is independent of the CF for the vessel weld material, and, therefore, the ratio factor is not applied to the surveillance weld ARTNDT values.

WCAP-1 5677-NP (Reference 2) is the source document for development of the latest BVPS Unit 2 PTLR, Rev. 2 (Reference 3). Per Tables 4-5 and 4-6 of WCAP-15677-NP, the best estimate copper and nickel values for the surveillance and vessel weld materials differ, resulting in different CF values of 38.0°F for the surveillance weld and 34.4°F for the vessel welds. The resulting ratio is 0.905. In WCAP-15677-NP, this ratio was set to 1.0 for conservatism, as it results in a greater ARTNDT shift.

References:

1. WCAP-1 6527-NP, "Analysis of Capsule X from First Energy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program,"

Rev. 0.

2. WCAP-1 5677-NP, "Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," August 2001.
3. Beaver Valley Power Station, Unit 2, "Pressure and Temperature Limits Report,"

Rev. 2 (ML071790328).

L-08-124 Page 17 of 17 Question RAI #8 LRA Section 4.2.2, LRA Table 4 2-6, and WCAP-16527-NP, Supplement I all incorporate data from the evaluation of surveillance capsules "U", "V", "W" and "X" for Beaver Valley 2. However the Beaver Valley 2 PTLR (Rev. 2) only incorporates data from the evaluation surveillance capsules "U", "V" and "W".

As the Beaver Valley 2 PTLR forms part of the basis for the Beaver Valley LRA, the staff requests that you update the Beaver Valley 2 PTLR to incorporate the results from the evaluation of surveillance capsule "X". The application of surveillance data and selection of CFs for calculation of RTPTs and ART values in the Beaver Valley 2 PTLR should be consistent with the LRA.

RESPONSE RAI #8 The latest BVPS, Unit 2, PTLR is Revision 2 (Reference 1), and was associated with the implementation of the BVPS Improved Standard Technical Specification Conversion License Amendment for Unit 2. This PTLR only incorporated data from the evaluation of surveillance capsules U, V andýW.

FENOC commits to update the Unit 2 PTLR, currently Revision 2, with the results of Capsule X analysis documented in WCAP-16527-NP (Reference 2) and WCAP-16527-NP, Supplement 1 (Reference 3). The revised PTLR will be submitted to the NRC no later than September, 30, 2008. See Attachment 1 to this letter, "Regulatory Commitment List," Commitment #2.

References:

1. Beaver Valley Power Station, Unit 2, "Pressure and Temperature Limits Report,"

Rev. 2 (ML071790328).

2. WCAP-1 6527-NP, "Analysis,of Capsule X from First Energy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program,"

Rev. 0.

3. WCAP-1 6527-NP, Supplement 1, "Analysis of Capsule X from First Energy Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program,"

July 2007.

IENCLOSURE Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2 Letter L-08-124 Amendment No. 5 to the BVPS License Renewal Application Page 1 of 5 Sections Affected 4.2.2 A.2.2.2 Table A.4-1 The Enclosure identifies the correction by Affected Section, License Renewal Application (LRA) Page No., and Affected Paragraph and Sentence. The count for the affected paragraph, sentence, bullet, etc. starts at the top of the affected page. Below each section the reason for the change is identified, and the sentence affected is printed in italics with deleted text lined-out and added text underlined.

Enclosure L-08-124 Page 2 of 5 License Renewal Affected Application Paragraph Affected Section Page No. and Sentence Section 4.2.2 Page 4.2-6 2 nd Paragraph, 4 th and 5 th Sentences The LRA Unit 1 pressurized thermal shock (PTS) text needs to be revised to indicate that the data for the Unit 1 surveillance program weld material (heat 305424) are deemed "not credible", and the data were used with a cA margin of 28 OF instead of 14 OF. The 2 nd paragraph on LRA page 4.2-6 is revised to read, "Usingthe prescribedPTS Rule (10 CFR 50.61) methodology, RTpTs values were generated for beltline and extended beitline region materialsof the BVPS Unit I Reactor Vessel for fluence values at EOLE (54 EFPY). The data for the surveillance program plate materialare deemed not credible. Therefore, the data were used with a UA (standarddeviation forARTNDT) margin of 17°F. The data for the Unit 1 surveillanceprogram weld material(heat 305424) are deemed not credible. Therefore, the data were used with a UA margin of 4-4-288°F. The surveillance capsule materials are representative of the actual vessel plates and intermediate shell longitudinal weld. Chemistry factor values for the B VPS Unit 1 beltline region materialswere based on Position 1.1 and 2.1 of Regulatory Guide 1.99 [Reference 4.2-61. Additionally, chemistry factor values for the BVPS Unit 1 extended beltline materialswere based on Position 1.1 of Regulatory Guide 1.99."

Enclosure L-08-124 Page 3 of 5 License Renewal Affected Application Paragraph Affected Section Page No. and Sentence Section 4.2.2 Page 4.2-6 5 th Paragraph, Last Sentence The LRA Unit 1 PTS text needs to be revised to change the license renewal future commitment regarding submittal to the NRC of the documentation of the Unit 1 Flux Reduction Program to be more specific, as requested in the NRC request for additional information, RAI #5, of this letter. The 5 th paragraph on LRA page 4.2-6 is revised to read, "Therefore, a sensitivity assessmentof available flux reduction measures was completed. The sensitivity assessment included several fuel management scenarios (such as low leakage core design, low power peripheralfuel assemblies, reinsertionof hafnium rods, and the use of part length shielded assemblies) and several assumed capacity factors up to 98 percent. Several flux reduction options are available which would maintain the limiting plate below the PTS screening criterion to the EOLE. The flux reduction program will be managed under the Reactor Vessel Integrity Program (Section B.2.35). Priorto exceeding the PTS screening criteria for BVPS Unit 1, FENOC will select a flux reduction measure to manage PTS in accordance with the reguirements of 10 CFR 50.61. A flux reduction plan will be submitted for NRC review and approvalat least 1 year priorto implementation of the flux reduction measure. Documentation of a flux redu.tion program for-Unit 1 will be submittd in accordance with th ...... ts of 10 CFR 50.61."

Enclosure L-08-124 Page 4 of 5 License Renewal Affected Application Paragraph Affected Section Page No. and Sentence Section A.2.2.2 Page A.2-4 4 th Paragraph, Last Sentence The LRA Unit 1 PTS text in Appendix A, "Updated Final Safety Analysis Report Supplement," needs to be revised to change the license renewal future commitment regarding submittal to the NRC of the documentation of the Unit 1 Flux Reduction Program to be more specific, as requested in the NRC request for additional information, RAI #5, of this letter. The 4 th paragraph on LRA, Appendix A, page A.2-4, is revised to read, "Therefore, a sensitivity assessment of available flux reduction measures was completed. The sensitivity assessment included several fuel management scenarios (such as low leakage core design, low power peripheralfuel assemblies, reinsertion of hafnium rods, and the use of part length shielded assemblies)and several assumed capacity factors up to 98 percent. Several flux reduction options are available which would maintain the limiting plate below the PTS screening criterion to the EOLE. The flux reduction program will be managed underthe Reactor Vessel Integrity Program. Priorto exceedinq the PTS screening criteria for BVPS Unit 1, FENOC will select a flux reduction measure to manage PTS in accordance with the reguirements of 10 CFR 50.61. A flux reduction plan will be submitted for NRC review and approval at least 1 year priorto implementation of the flux reduction measure.

Dou,,*ntation of a flux rcduction program for Ui*t wil be submitted in accordanee with tho rqIrements of 10 CFR 50.61."

Enclosure L-08-124 Page 5 of 5 License Renewal Affected Application Paragraph Affected Section Page No. and Sentence Table A.4-1 Page A.4-9 Item Number 24 License renewal future commitment in LRA Appendix A.4, "Unit 1 License Renewal Commitments," Table A.4-1, "Unit 1 License Renewal Commitments,"

Item Number 24, needs to be revised to be more specific regarding the Unit 1 Flux Reduction Program approach and requirements for submittal to the NRC, as requested in the NRC request for additional information, RAI #5, of this letter.

License renewal future commitment, Item Number 24 of Table A.4-1, is revised to read:

Item ImplementationRead Related LRA A Item Number Commitment Impeme Schedule Source CmetNo./

Section Comments 24 Prior to exceeding the PTS A flux reduction LRA A.2.2.2 screening criteria for BVPS plan will be 4.2.2 Unit 1, FENOC will select a flux submitted at least reduction measure to manage 1 year prior to PTS in accordance with the implementation of requirements of 10 CFR 50.61. the flux reduction A flux reduction plan will be measure.

submitted for NRC review and approval. in accordance whththe DoUm.ntation of a flu ...... ements- E reductionR pro~gram for Unit 1 OG-5O6 be submitted inAccordance Oillw with the requirements -of 10 CER 50671.

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