ML13017A467

Approval of Request for Change to the Reactor Vessel Surveillance Capsule Removal Schedule

ML13017A467
Person / Time
Site:	Fort Calhoun
Issue date:	01/31/2013
From:	Lynnea Wilkins Plant Licensing Branch IV
To:	Cortopassi L Omaha Public Power District
Wilkins L
References
TAC ME8219
Download: ML13017A467 (12)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 31, 2013 Mr. Louis P. Cortopassi Site Vice President and Chief Nuclear Officer Omaha Public Power District Fort Calhoun Station 9610 Power Lane, Mail Stop FC-2-4 Omaha, NE 68008

SUBJECT:

FORT CALHOUN STATION, UNIT 1 - REQUEST FOR APPROVAL OF PROPOSED CHANGES TO THE REACTOR PRESSURE VESSEL SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES (TAC NO. ME8219)

Dear Mr. Cortopassi:

By letter dated February 6,2012, as supplemented by letter dated May 30,2012, Omaha Public Power District, the licensee for Fort Calhoun Station, Unit 1 (FCS), submitted a request for revision to the reactor vessel (RV) material surveillance capsule withdrawal schedule for FCS.

The proposed changes were submitted pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix H, "Reactor Vessel Material surveillance Program Requirements,"

Section III.B.3, which requires that: (1) withdrawal schedules be submitted, as specified in 10 CFR 50.4, and (2) the proposed schedule must be approved by the U.S. Nuclear Regulatory Commission (NRC) prior to implementation.

The NRC staff has reviewed the proposed withdrawal schedule for FCS, and determined that the changes to the schedule will continue to meet the RV surveillance capsule withdrawal schedule criteria in American Society for Testing and Materials (ASTM) E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," and is in compliance with 10 CFR Part 50, Appendix H. The NRC staff, therefore, concludes that the RV withdrawal schedule, as proposed in the licensee's letter dated May 30, 2012, is acceptable for implementation.

L. Cortopassi - 2 The NRC staff's safety evaluation is enclosed. If you have any questions, please contact me at 301-415-1377 or via e-mail at Iynnea.wilkins@nrc.gov.

Sincerely,

~a prOje~t E. Wilkins, Plant Licensing Branch IV Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosure:

As stated cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REACTOR VESSEL SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION. UNIT 1 DOCKET NO. 50-285

1.0 INTRODUCTION

By letter dated February 6, 2012 (Reference 1), as supplemented by letter dated May 30, 2012 (Reference 2), Omaha Public Power District, the licensee for Fort Calhoun Station, Unit 1 (FCS), submitted a request for revision to the reactor vessel (RV) material surveillance capsule withdrawal schedule for FCS. The proposed changes were submitted pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements,"Section III.B.3, which requires that: (1) withdrawal schedules be submitted, as specified in 10 CFR 50.4, and (2) the proposed schedule must be approved by the U.S. Nuclear Regulatory Commission (NRC) prior to implementation.

The licensee's submittals indicate the proposed withdrawal schedule reflects the planned coordinated U.S. pressurized-water reactor (PWR) reactor vessel surveillance program, the technical basis for which is included in the Electric Power Research Institute (EPRI), "Materials Reliability Program: Coordinated PWR Reactor Vessel Surveillance Program (CRVSP)

Guidelines (MRP-326)," 1022871, Final Report, December 2011 (Reference 3).

2.0 REGULATORY EVALUATION

The regulations in 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation," require light-water reactors (LWRs) to meet the RV materials surveillance program requirements set forth in Appendix H to 10 CFR Part 50.

Appendix H to 10 CFR Part 50 provides the NRC staff's criteria for the design and implementation of RV material surveillance programs for operating LWRs. The rule, in part, requires RV surveillance program designs and withdrawal schedules to meet the requirements of the edition of American Society for Testing and Materials (ASTM) Standard Practice E185, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactors," that is current on the issue date of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) to which the RV was purchased, although later editions of ASTM E 185 may be used inclusive of the 1982 Edition of ASTM E 185 Enclosure

-2 (ASTM E185-82). The rule also requires proposed RV surveillance programs to be submitted to the NRC and approved prior to implementation. The applicable criteria in ASTM E185-82 are discussed in Section 3.1 of this safety evaluation (SE). The FCS RV Surveillance Program was designed to ASTM E185-66, per Section 4.5.3 of the FCS Updated Safety Analysis Report (USAR), Revision 16 (Reference 4).

In Section 3.1.2.3.1, "Reactor Vessel Integrity Program," of NUREG-1782, "Safety Evaluation Report Related to the License Renewal of Fort Calhoun Station, Unit 1," October 2003 (Reference 5), the NRC staff approved the FCS Reactor Vessel Integrity Program, which controls the RV surveillance capsule removal schedule, and found the program was consistent, with enhancements, with Section XI.M31 , "Reactor Vessel Surveillance," of NUREG-1801, Revision 0, "Generic Aging Lessons Learned (GALL) Report," Volume 2, July 2001 (GALL Report) (Reference 6). The actual current removal schedule was approved by NRC letter to the licensee dated May 2,2002 (Reference 7). For plants applying for license renewal, the Section XI.M31 of the GALL Report provides guidance for RV surveillance programs for the period of extended operation. Criteria 5 and 6 of Section XI.M31 of the GALL Report relate to surveillance capsule withdrawal schedules.

Criterion 5 of Section XI.M31 of the GALL Report recommends that if an applicant has a surveillance program that consists of capsules with a projected fluence of less than the 60-year fluence at the end of 40 years, at least one capsule is to remain in the reactor vessel and is tested during the period of extended operation. The applicant may either delay withdrawal of the last capsule or withdraw a standby capsule during the period of extended operation to monitor the effects of long-term exposure to neutron irradiation.

Criterion 6 of Section XI.M31 of the GALL Report recommends, in part, that if an applicant has a surveillance program that consists of capsules with a projected fluence exceeding the 60-year fluence at the end of 40 years, the applicant withdraws one capsule at an outage in which the capsule receives a neutron fluence equivalent to the 60-year fluence and tests the capsule in accordance with the requirements of ASTM E185. Additional recommendations apply under Criterion 6 if capsules are either left in the RV after 40 years or for monitoring RV exposure conditions if all capsules have been removed.

The FCS Reactor Vessel Integrity Program meets Criterion 5 of Section XI.M31 of the GALL Report since no capsules received the 60-year equivalent fluence prior to the expiration of the original license. FCS USAR Section 15.2.19 summarizes the Reactor Vessel Integrity Program as it relates to managing the effects of aging and states, in part, that The program includes revising the FCS surveillance capsule removal schedule in order to optimize the program through the end of the period of extended operation.

By letter dated June 7,2001, via Amendment No. 199, the NRC staff approved an integrated surveillance program for FCS (Reference 8). Under the integrated surveillance program, surveillance data from a Japanese PWR, Mihama, Unit 1, is used to determine the pressurized thermal shock reference temperature (RT PTS) value for weld wire heat combination 12008/27204. Based on surveillance data for this heat combination from Mihama, Unit 1, the NRC staff concluded that weld wire heat combination 12008/27204 was no longer the limiting

-3 material for FCS. Weld heat combination 12008/13253, which is included in one FCS surveillance capsule, became the limiting material with regard to pressurized thermal shock (PTS) for FCS. The integrated surveillance program also uses data from the Diablo Canyon Power Plant, Unit 1, and Palisades Plant RV surveillance programs for tandem weld wire heat 27204/27204. Also, in Reference 8, the NRC staff concluded the FCS RV would remain below the PTS screening criteria through the end of the original license period. In NUREG-1782 (Reference 5), the NRC staff concluded that the FCS RV would remain below the PTS screening criteria through the end of the period of extended operation, and also concluded that the upper shelf energy (USE) was acceptably projected through the end of the period of extended operation. In NUREG-1782, the NRC staff compared the FCS USE projections using Regulatory Position 1.2 of Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988 (Reference 9), with surveillance data for the FCS beltline weld material heats from Donald C. Cook Nuclear Plant, Unit 1, Diablo Canyon Power Plant, Unit 1, Salem Nuclear Generating Station, Unit 2, and Mihama, Unit 1. However, FCS does not credit a formal integrated surveillance program for the USE evaluation.

3.0 TECHNICAL EVALUATION

3.1 Evaluation Criteria of ASTM Standard Practice E185-82 Table 1 of ASTM E185-82 requires that either a minimum of three, four, or five surveillance capsules be removed from the RVs, based on the limiting amount of the shift in the reference transition of the nil-ductility temperature (limiting LlRTNDT) that is projected to occur at the clad to-base metal interface (RV inner wall) location of the RV at the end-of-licensed plant life (EOl).

ASTM E185-82 establishes the following criteria for determining the minimum number of capsules that are to be removed in accordance with a withdrawal schedule and the number of capsules that are to be tested:

• For plants with projected RT NDT shifts (i.e., LlRT NDT) less than 100 degrees Fahrenheit (OF) (56 degrees Celsius (OC)), three capsules are required to be removed from the RV and the first two capsules are required to be tested (for dosimetry, tensile-ductility, Charpy-V impact toughness, and alloying chemistry).

• For plants with projected LlRT NDT between 100 of (56°C) and 200 of (111°C),

four surveillance capsules are to be removed from the RV and the first three capsules are required to be tested.

• For plants with projected LlRT NDT above 200 of (111°C), five surveillance capsules are required to be removed from the RV and the first four capsules are required to be tested.

• Standard Practice ASTM E185-82 permits the last scheduled surveillance capsules in three-, four-, or five-capsule withdrawal schedules to be removed without the implementation of testing. However, licensees who opt to pull their final required capsules without the implementation of testing are required by the Standard Practice to hold the capsules in storage.

4 Table 1 of ASTM Standard Practice E185-82 also provides specific criteria for removal of surveillance capsules. The removal times are based on criteria that the surveillance capsules be removed after a certain amount of power operation has elapsed or at various times when the RV shell is projected to achieve certain levels of neutron fluence. The intent of the Standard Practice is to achieve a set of testing data over a range of neutron fluences for the RV that bounds the current life of the plant. Of key importance are the removal criteria for the final capsule required for capsule withdrawal. For the final capsule that is required for removal, ASTM E 185-82 requires that the capsule be removed at a time when the neutron fluence projected for the capsule is between the limiting fluence value projected for the RV at the EOl and two times that value.

Per "Evaluation of Reactor Vessel Surveillance Data Pertinent to the Fort Calhoun Reactor Vessel 8eltline Materials, CEN-636, Revision 02, Final Report," dated July 19, 2000, included as Attachment D to OPPD's letter to NRC dated August 3, 2000 (Reference 10), the maximum shift in the RTPTS value, which is identical to the maximum shift of the RT NDT at the RV inner wall, is 267 of at the end of the period of extended operation for weld heat number 27204. This shift value was essentially confirmed by the staff via its independent calculation of the RTPTS values for FCS at the end of the period of extended operation in Reference 5, which were only 1 of higher than the licensee's values for the weld materials. Since the maximum ARTNDT for the FCS RV is greater than 200 of, the ASTM E 185-82 criteria would require five surveillance capsules to be withdrawn, and four of the five to be tested.

3.2 Changes Proposed to the Withdrawal Schedule for FCS In the enclosure to its letter dated February 6, 2012, the licensee provided Tables 1 and 2, showing the current withdrawal schedule (Table 1) and the proposed withdrawal schedule for the remaining capsules only (Table 2). In its letter dated May 30, 2012, in response to an NRC staff request for additional information (RAI) dated May 3, 2012 (Reference 11), the licensee provided a revision of Table 2. Each table provides the capsule number, capsule location in terms of azimuthal location in degrees, capsule lead factor (ratio of peak RV inside surface flux to capsule flux), capsule removal EFPY (actual for those already removed, estimated for those planned to be removed), and neutron fluence (actual or estimated). Table 1 below combines the information from Table 1 of Reference 1 and Table 2 of Reference 2 and provides the capsule number, azimuthal location, lead factor, and the current and proposed surveillance capsule withdrawal schedules in terms of calendar year, EFPY and neutron fluence.

The licensee also provided in Reference 1 the projected RV peak fluence at 60 calendar years (48 EFPY) of 3.5x10 19 n/cm 2 (Energy (E) >1 mega electron volt (MeV>> and the projected RV peak fluence at 80 calendar years (67 EFPY) of 4.72x1019 n/cm 2 (E>1 MeV).

3.3 NRC Staff Evaluation In the initial proposed withdrawal schedule in Reference 1, one capsule (W-275S) was to be withdrawn with a neutron fluence significantly less (approximately 86 percent) than the projected EOl fluence of the peak RV inner wall location, and thus did not meet the ASME E185-82 recommendation that the fourth capsule in a five capsule schedule be withdrawn at 15 EFPY, or at the time when the accumulated neutron fluence of the capsule corresponds to the approximate EOl fluence at the RV inner wall location, whichever comes first. Therefore, by

-5 letter dated May 3, 2012, in RAI #1, the NRC staff requested that the licensee provide a technical justification for withdrawing Capsule W-275S at a fluence less than the approximate RV inner wall EOl fluence. In its response dated May 30,2012, the licensee provided a revised capsule withdrawal schedule. Table 1 shows the current NRC-approved withdrawal schedule and the revised proposed schedule from the licensee's letter dated May 30, 2012, and also shows the data for the three previously withdrawn capsules, which was not included in the submittal.

In its response to RAI #1, the licensee also stated that the purpose of [supplemental capsule W-275S] is to obtain limiting weld surveillance results that are FCS-specific, rather than the representative results provided by the original FCS capsules (per Reference 3, Capsule W-275S contains material of actual limiting weld wire heat, tandem heat 12008/132523, while the original capsules contain no weld wire heats matching those in the FCS RV). The licensee also stated in the RAI response that Capsule W-275S is not intended to represent the fourth or "fifth capsule of [an ASTM E185-82-compliant schedule] but is included in the surveillance program as an integral part of the FCS Integrated Surveillance Program 1. The licensee also noted that the proposed schedule represents an enhancement of the current schedule in that a high-fluence capsule will be withdrawn sooner (more than 10 years before the end of the extended license), thus providing meaningful surveillance data for managing loss of fracture toughness due to neutron embrittlement of the RC beltline materials.

Table 1 - FCS Current and Proposed Withdrawal Schedule Current Schedule Proposed Schedule Fluence Fluence 2 2 (n/cm ) (n/cm )

With (E >1.0 MeV) With (E >1.0 MeV)

Location Lead Removed drawal (Actuall Removal drawal (Actual!

Capsule (degrees) Factor EOC EFPY Estimated) Year EFPY Estimated)

W-225 225 1.53 3 (1977) 2.5 5.1x1018 3 (1977) ~ 5.1x10" ,

W-265 265 1.07 7(1983) 5.9 9.0x1018 7(1983) . 9.0x1018 W-275 275 1.05 14(1993) 13.6 1.38x10 19 14 (1993) 13.6 1.38x1019 W-45 45 1.51 Standby 2023 37.8 4.41x10'9 W-85 85 1.17 Standby W-95 95 1.17 Planned 48 2033 3.39x10'9 W-225S 225 1.12 Standby W-265S 265 0.97 Standby W-275S 275 - Planned 33.6 1.719x10'9 2027 41.7 2.42x10'9 The NRC staff reviewed the proposed withdrawal schedule for FCS against the criteria of ASTM E185-82. Capsule W-45 is being changed from a standby capsule to be withdrawn at a neutron 1 The NRC staff notes that Reference 8, which approved the integrated surveillance program, mainly addressed surveillance data from other plants for tandem weld wire heats 12008/27204 and 27204/27204, and does not address any data from other plant's surveillance programs for tandem heat 12008/13523

-6 fluence approximately equivalent to the RV inner wall fluence at 80 calendar years (67 EFPY).

Capsule W-275S will be withdrawn at a fluence of 2.42 x 1019 n/cm 2 (E > 1 MeV) rather than the previously planned fluence of 1.719x10 19 n/cm 2 (E > 1 MeV). The resulting schedule is somewhat unusual because the capsule receiving the highest fluence (capsule W-45) will be withdrawn before the last capsule due to the higher lead factor of capsule W-45. However, Capsule W-45 would fulfill the ASTM E185-82 recommendation that the last capsule of a 5 capsule schedule be withdrawn at a fluence not less than once or greater than twice the peak EOl vessel fluence. Current EOl for FCS is 60 calendar years (48 EFPY), at which the peak RV inner wall fluence is estimated to be 3.5x1019 n/cm 2 (E > 1 MeV). The proposed withdrawal fluence of Capsule W-45 of 4.41 x1 019 n/cm 2 (E > 1 MeV) is 1.26 times the EOl (60-year) peak RV inner wall fluence, thus meeting the ASTM E185-82 recommendation.

The previous proposed schedule provided no lead factor for Capsule W-275S (implying capsule fluence is identical to the peak RV inner wall fluence). In addition, the NRC staff estimated a lower neutron fluence at the proposed withdrawal EFPY in the original submittal. Therefore, by letter dated May 3, 2012, in RAI #2, the staff requested the licensee to provide the details of the fluence projection for capsule W-275S. By letter dated May 30, 2012, in response to RAI #2, the licensee clarified that no lead factor was reported in the original submittal because this capsule was installed at the end of Cycle 14 in 1993, rather than in 1973 as in the case for the other capsules, so there is no integrated lead factor for Capsule W-275S for the entire life of the plant. However, the licensee indicated the lead factor for Capsule W-275S is 1.52 since its installation. The licensee indicated that Capsule W-275S has a nominal fast neutron flux of

2. 74x10 10 n/cm 2 (E > 1 MeV) resulting in a prOjected fluence at 47.2 EFPY of 2.905x1 019 n/cm 2 (E> 1 MeV). However, in the revised withdrawal schedule provided in the RAI #1 response, the licensee changed the proposed withdrawal date of Capsule W-275S to 41.7 EFPY corresponding to a fluence of 2.42x1 0 19 n/cm 2 (E > 1 MeV). In the revised schedule, a different capsule (W-95) will have a withdrawal fluence equivalent to the end-of-life RV inner wall fluence.

The NRC staff concludes that the proposed revised schedule will meet the intent of the recommendations of ASTM E 185-82, with respect to the withdrawal schedule of the capsules.

In particular, although chronologically in terms of EFPY Capsule W-95 will be the fifth and last capsule to be withdrawn, this capsule will be withdrawn at a neutron fluence approximately equal to the RV inner wall fluence (3.5x10 19 n/cm 2 , E> 1 MeV) at the end of life (60 calendar years). This essentially meets the recommendation of E185-82 for the fourth capsule in a five capsule schedule, to be withdrawn at 15 EFPY or at the time when the accumulated neutron fluence of the capsules corresponds to the approximate EOl fluence at the RV inner wall location, whichever comes first. Since it is no longer possible for a capsule to be withdrawn at 15 EFPY, it is acceptable that this capsule be withdrawn before exceeding the maximum fluence recommended by ASTM E185-82 for withdrawal of the fourth capsule. Although Capsule W-45 is actually the fourth capsule chronologically in terms of EFPY to be withdrawn, the proposed withdrawal neutron fluence is not less than once nor greater than twice the peak EOl RV inner wall fluence, thus meeting the recommendation of ASTM E 185-82 for the fifth capsule in a five-capsule schedule. The staff also notes that the FCS USAR indicates that the RV surveillance program was designed to ASTM E185-66, which is less prescriptive with respect to the capsule withdrawal timing. ASTM E185-66 requires that surveillance specimens be withdrawn at three different times and that one of the data points obtained "shall correspond to the neutron exposure of the component near the end of its design life." Therefore, the fact that the proposed withdrawal schedule deviates somewhat from the recommendations of ASTM

-7 E185-82 is acceptable, because FCS surveillance program continues to be consistent with the version of the standard current at the time the RV was designed.

Since three capsules will be withdrawn and tested during the period of extended operation, the proposed withdrawal schedule meets the recommendation of Criterion 5 of NUREG-1801, Revision 0, which recommends that if an applicant has a surveillance program that consists of capsules with a projected fluence of less than the 60-year fluence at the end of 40 years, at least one capsule is to remain in the reactor vessel and is tested during the period of extended operation. Therefore, the proposed withdrawal schedule is consistent with the Reactor Vessel Integrity Program approved by the NRC staff in Reference 4, and described in FCS USAR Section 15.2.19.

The NRC staff also notes that capsule W-95 and Capsule W-45 do not contain the limiting weld material for FCS, but can provide valuable higher fluence data for the industry coordinated surveillance program documented in MRP-326 (Reference. 2).

As previously noted, FCS relies on an integrated surveillance program to monitor embrittlement of tandem weld heat 12008/27204, which is the second most limiting material in the FCS RV with regard to PTS. For the limiting material with respect to PTS in the FCS RV, tandem weld wire heat 12008/13253, the licensee has not implemented an integrated surveillance program.

Supplemental capsule W-275S at FCS contains tandem weld heat 12008/13253 and will provide some data on this heat, and could potentially become part of an integrated surveillance program. The capsules designated as "standby" (W-85, W-225S, and W-265S) do not contain the limiting weld materials and have relatively low lead factors, so would not provide either valuable plant-specific embrittlement data or accelerated high-fluence data for the industry database. Based on the above, the selection of the capsules to be withdrawn in the proposed schedule for FCS is appropriate.

The NRC staff concludes that the proposed withdrawal schedule is acceptable because it is meets the requirements of ATSM E 185-66, the version of that standard that was current at the issue date of the ASME Code to which the FCS RV was purchased, thus meets the requirements of 10 CFR 50, Appendix H. In addition, the proposed schedule meets the intent of the recommendations of ASTM E185-82. Finally, the proposed withdrawal dates for both the fourth and fifth capsules will provide data at fluences greater than or equal to 3.0x1 0 19 n/cm 2 that supports the overall industry goal, with which the NRC staff agrees, of acquiring more high-fluence data.

4.0 CONCLUSION

The NRC staff has reviewed the licensee's proposed surveillance capsule withdrawal schedule for FCS provided in Table 2 of Reference 2, and provided in Table 1 of this SE, and has determined that the proposed schedule is in compliance with 10 CFR Part 50, Appendix H. The NRC staff, therefore. concludes that the RV surveillance capsule withdrawal schedule, as proposed in the Table 2 of Reference 2, and reproduced in Table 1 of this SE, is acceptable for implementation. This SE does not affect the withdrawal schedule for any surveillance capsules in the FCS Integrated Surveillance Program installed in plants other than FCS.

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5.0 REFERENCES

1. Herman, J. B., Omaha Public Power District, letter to U.S. Nuclear Regulatory Commission, "Request for Change in Reactor Vessel Surveillance Capsule Removal Schedule," dated February 6,2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML120410132).

2. Goodell, J. R, Omaha Public Power District, letter to U.S. Nuclear Regulatory Commission, "OPPD Response to NRC RAI Regarding Request for Change in Reactor Vessel Surveillance Capsule Removal Schedule," dated May 30, 2012 (ADAMS Accession No. ML12152A192).

3. Electric Power Research Institute, "Materials Reliability Program: Coordinated PWR Reactor Vessel Surveillance Program (CRVSP) Guidelines {MRP-326)," 1022871, Final Report, December 2011 (ADAMS Accession Nos. ML12040A314 and ML12040A315).

4. Updated Safety Analysis Report for Fort Calhoun Station, Revision 16 (not publicly available).

5. U.S. Nuclear Regulatory Commission, NUREG-1782, "Safety Evaluation Report Related to the License Renewal of Fort Calhoun Station, Unit 1," October 2003 (ADAMS Accession No. ML033020438).

6. U.S. Nuclear Regulatory Commission, NUREG-1801, Revision 0, "Generic Aging Lessons Learned (GALL) Report," Volume 2, July 2001 (GALL Report) (ADAMS Accession No. ML012060545).

7. Dembek, S., U.S. Nuclear Regulatory Commission, letter to R T. Ridenoure, Omaha Public Power District, "Fort Calhoun Station, Unit No.1 - Reactor Vessel Surveillance Capsule Removal Schedule Change {TAC No. MB3422)," dated May 2,2002 (ADAMS Accession No. ML021070609)

8. Wang, A. B., U.S. Nuclear Regulatory Commission, letter to S. K. Gambhir, Omaha Public Power District, "Fort Calhoun Station, Unit No. 1 - Issuance of Amendment Deletion of Section 3.D, "License Term" {TAC No. MA9690)," dated June 7,2001 (ADAMS Accession No. ML011580518).

9. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988 (ADAMS Accession No. ML003740284).

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10. Omaha Public Power District, "Evaluation of Reactor Vessel Surveillance Data Pertinent to the Fort Calhoun Reactor Vessel Beltline Materials, CEN-636, Revision 02, Final Report," dated July 19, 2000, Attachment D to letter to U.S. Nuclear Regulatory Commission from W. G. Gates, OPPD, "Application for Amendment of Operating License (OPPD Letter No. LlC-00-0064)," dated August 3, 2000 (ADAMS Accession No. ML003738600).

11. Wilkins, L. E., U.S. Nuclear Regulatory Commission, letter to David J. Bannister, Omaha Public Power District, "Fort Calhoun Station, Unit No. 1 - Request for Additional Information Regarding Proposed Change in Reactor Vessel Surveillance Capsule Removal Schedule (TAC No. ME8219)," dated May 3,2012 (ADAMS Accession No. ML121080125).

Principal Contributor: Jeffrey C. Poehler Date: January 31, 2013

L. Cortopassi - 2 The NRC staff's safety evaluation is enclosed. If you have any questions, please contact me at 301-415-1377 or via e-mail at Iynnea.wilkins@nrc.gov.

Sincerely, IRA!

Lynnea E. Wilkins, Project Manager Plant licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosure:

As stated cc w/encl: Distribution via Listserv DISTRIBUTION:

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