ML15069A181

Request for Additional Information Aging Management Program Plan for Reactor Vessel Internals (TAC Nos. MF5149 and MF5150)

ML15069A181
Person / Time
Site:	Salem
Issue date:	03/31/2015
From:	Carleen Parker Plant Licensing Branch 1
To:	Joyce T Public Service Enterprise Group
Parker C, NRR/DORL/LPL1-2, 415-1603
References
TAC MF5150, TAC MF5749
Download: ML15069A181 (7)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 31, 2015 Mr. Thomas Joyce President and Chief Nuclear Officer PSEG Nuclear P.O. Box 236, N09 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1AND2- REQUEST FOR ADDITIONAL INFORMATION RE: AGING MANAGEMENT PROGRAM PLAN FOR REACTOR VESSEL INTERNALS (TAC NOS. MF5149 AND MF5150)

Dear Mr. Joyce:

By letter dated August 11, 2014, 1 PSEG Nuclear LLC (PSEG) submitted two reports for its License Renewal Reactor Vessel Internals Aging Management Program at Salem Nuclear Generating Station (Salem), Units 1 and 2, for U.S. Nuclear Regulatory Commission (NRC) staff review. Topical report MRP-227-A, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," 2 and its supporting reports were used as the technical bases for developing the Salem, Units 1 and 2, aging management programs. The NRC staff issued a final safety evaluation of topical report MRP-227-A on December 16, 2011. 3 The NRC staff has determined that additional information is needed to complete its review of the submittal. The specific questions are found in the enclosed request for additional information.

The draft questions were sent via electronic transmission on March 6, 2015, to Mr. Paul Duke of your staff. The draft questions were sent to ensure the questions were understandable, the regulatory basis was clear, and to determine if the information was previously docketed. On March 20, 2015, Mr. Brian Thomas of your staff indicated that the licensee will submit a response within 60 days of this letter.

1 Agencywide Documents Access and Management System (ADAMS) Accession Number ML14224A667.

2 ADAMS Accession No. ML12017A194.

3 ADAMS Accession No. ML11308A770.

T. Joyce If you have any questions, please contact me at 301-415-1603 or via e-mail at Carleen. SandersParker@nrc.gov.

Sincerely, Carleen J. Parker, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION OFFICE OF NUCLEAR REACTOR REGULATION AGING MANAGEMENT PROGRAM PLAN FOR REACTOR VESSEL INTERNALS SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2 PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NOS. 50-272 AND 50-311 By letter dated August 11, 2014, 1 PSEG Nuclear LLC (PSEG) submitted two reports for its License Renewal Reactor Vessel Internals Aging Management Program at Salem Nuclear Generating Station (Salem), Units 1 and 2, for U.S. Nuclear Regulatory Commission (NRC) staff review. Topical report MRP-227-A, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," 2 and its supporting reports were used as the technical bases for developing the Salem, Units 1 and 2, aging management programs (AMPs). The NRC staff issued a final safety evaluation of topical report MRP-227-A on December 16, 2011. 3 To complete its review, the NRC staff requests a response to the questions below.

The action items (Als) in some of the requests for additional information (RAls) refer to the licensee's action items (LAls) in Section 4 of the NRC's final safety evaluation of MRP-227-A.

RAl-1 Historically, the following materials used in the pressurized-water reactor (PWR) reactor vessel internal (RVI) components were known to be susceptible to some of the aging degradation mechanisms that are identified in the MRP-227-A topical report. In this context, the NRC staff requests that the licensee provide a list of any additional RVI components (not listed in MRP-227-A and MRP-191, Revision 0, "Material Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design"4 ), which are manufactured from the following materials. If any of these materials are identified as an additional RVI component at Salem, Unit 1 or 2, please provide information on the type of aging effect that was detected, and the type of AMP implemented on these components.

a) Nickel base alloys-lnconel 600; Weld Metals-Alloy 82 and 182 and Alloy X-750 1

Agencywide Documents Access and Management System (ADAMS) Accession Number ML14224A667.

2 ADAMS Accession No. ML12017A194.

3 ADAMS Accession No. ML11308A770.

4 ADAMS Accession No. ML091910130.

Enclosure

b) Stainless steel type 347 material (excluding baffle-former bolts) c) Precipitation hardened (PH) stainless steel materials-17-4 and 15-5 d) Type 431 stainless steel material e) Alloy A-286, ASTM A 453 Grade 660, Condition A or B RAl-2 The NRC staff requests the licensee to provide a list of RVI components at Salem, Units 1 and 2, that have been inspected under the American Society of Mechanical Engineers Code,Section XI In-service Inspection (ISi) program, thus far, and the inspection results. Please include any RVI component categorized under the "Existing" inspection category in the MRP-227-A topical report.

RAl-3 According to Section A.1.4 in MRP-175, "Materials Reliability Program: PWR Internal Aging Degradation Mechanism Screening Threshold Values,"5 susceptibility to stress-corrosion cracking (SCC) in nickel-based Alloy X-750 PWR RVI components depends on the type of heat treatment that is performed on the alloy. High temperature heat treatment (HTH) processes that are used on Alloy X-750 components offer better resistance to sec than the other age hardened heat treatment processes. Licensee determination of the heat treatment applied to its Alloy X-750 PWR RVI components would appear to be a critical parameter in ensuring the licensee's AMP will adequately manage the potential effects of aging. Additionally, Appendix A of topical report MRP-227-A, which summarizes the operational experience due to age-related degradation mechanisms, addressed that Alloy X-750 used for the clevis insert bolt assembly in one unit failed due to primary water stress-corrosion cracking (PWSCC).

Therefore, the NRC staff requests that the licensee provide information related to the type of heat treatment process that was used for the Alloy X-750 clevis insert bolting at Salem, Units 1 and 2. If the existing clevis insert bolts at Salem, Units 1 and 2, did not undergo an HTH process, please indicate if these bolts have been inspected, or if there are plans to inspect these bolts (in addition to the inspections for monitoring aging due to wear) for degradation due to PWSCC.

RAl-4 In MRP-2013-025, "MRP-227-A Applicability Template Guideline,"6 report MRP has identified two generic questions that all Combustion Engineering, Inc. and Westinghouse design plants referencing topical report MRP-227-A must address to close Al 1 related to plant-specific applicability of the topical report. If the answer to either or both questions is yes, then further 5

ADAMS Accession No. ML061880278.

6 ADAMS Accession No. ML13322A454.

evaluation will be necessary to demonstrate the applicability of MRP-227-A to Salem, Units 1 and 2. The NRC staff therefore requests the following information:

1. Do the Salem, Units 1 and 2, RVI components have non-weld or bolting austenitic stainless steel components with 20 percent cold work or greater, and if so, do the affected components have operating stresses greater than 30 ksi? In particular, the plant-specific information on the extent of cold work on its RVI components. The licensee can apply "Option 1" or "Option 2," as addressed in Appendix A of the MRP-2013-025 report. If "Option 2" is applicable to Salem, Units 1 and 2, the licensee should list plant-specific RVI components that have been exposed to cold work equal to or greater than 20 percent. Plant-specific information related to this issue as addressed in "Option 2" in Appendix A, should be provided.

2. Have Salem, Units 1 and 2, ever utilized atypical design or fuel management that could make the assumptions of MRP-227-A regarding core loading/core design non-representative for that plant, including power changes/uprates? The following guidelines provided by MRP should be followed. The licensee is requested to use the MRP document dated October 14, 2013, MRP-2013-025, and it can apply "Option 1" or "Option 2," as addressed in Appendix B of the MRP-2013-025 report.

Option 1:

Salem, Units 1 and 2, comply with the MRP-227-A assumptions regarding core loading/core design. Neutron fluence and heat generation rates are concluded to be Option A or Option B.

Option A: acceptable based on the following assessment to the limiting MRP guidance threshold values.

Option B: unacceptable based on an assessment to the limiting MRP guidance threshold values.

If Option A as addressed under "Option 1" is applicable, the following plant-specific values should be submitted: (a) active fuel to fuel alignment plate distance; (b) average core power density; and (c) heat generation figure of merit.

If Option B under "Option 1" is applicable to Salem, Units 1 and 2, the licensee should justify the usage of its fuel management program.

Option 2:

Salem, Units 1 and 2, do not comply with the MRP-227-A assumptions regarding core loading/core design. The licensee should provide a technical justification for the application of MRP-227-A criterion to Salem, Units 1 and 2.

RAl-5 In response to Al 3, the licensee indicated that the control rod guide tube (CRGT) support pins that were fabricated from Alloy X-750 material were replaced by strain-hardened 316 stainless steel (SS) material. The licensee also indicated that Salem, Units 1 and 2, are consistent with the requirements in Table 4-9 of topical report MRP-227-A. Table 4-9 of topical report MRP-227-A, however, does not include CRGT support pins. As indicated in Section 3.2.5.3 of the NRC staff's safety evaluation of topical report MRP-227-A, licensees shall evaluate the adequacy of their existing programs to manage aging degradation during the period of extended operation (PEO) for both Alloy X-750 and type 316 SS support pins.

Please provide the plant specific evaluation of the existing program under which the replacement CRGT support pins are currently inspected. In addition, please provide an evaluation of the adequacy of the existing inspection program to ensure that the aging degradation is adequately managed during the PEO for the 316 SS CRGT support pins.

RAl-6 In response to Al 5, the licensee referenced a Westinghouse letter with proposed acceptance criteria for 304 SS hold down springs. Please provide an explanation of the methodology for developing the acceptance criteria for the measurement of loss of compressibility of the 304 SS hold down springs.

RAl-7 Regarding Al 7, there is new NRC staff guidance on the threshold limits for thermal embrittlement (TE) and irradiation embrittlement (IE) of Cast Austenitic Stainless Steel (CASS). The bases for the NRC staff's new consensus on the threshold limits are described in "NRC Position on Aging Management of CASS Reactor Vessel Internal Components."7 a) Please address any difference between the new guidance and the evaluation performed for Salem, Units 1 and 2. In particular, please address, the new screening guidelines of CASS materials for loss of fracture toughness of highly irradiated components (i.e.,

components susceptible to IE), in addition to TE. If any changes to the evaluation are necessary, please submit the re-evaluation, if not, please explain why not. This evaluation could affect some of the CASS components that are listed as susceptible to TE in Tables 6-2 of Attachments 1 and 2 of the submittal for Salem, Units 1 and 2, respectively, especially the lower support column caps as described in the next paragraph.

b) The NRC staff's initial review indicated that, in addition to TE, the lower support column caps (which is one of the pieces that comprises the lower support column body as explained in Section 6.2. 7 in Attachments 1 and 2 of the August 11, 2014, letter) are susceptible to IE. Please provide an explanation of how aging degradation due to TE and IE of the lower support column bodies is being managed and will be managed during the PEO.

7 ADAMS Accession NumberML14163A112.

ML15069A181 OFFICE LPL 1-2/PM LPL 1-2/LA DE/EVIB/BC DLR/RARB/BC LPL 1-2/BC LPL 1-2/PM NAME CParker ABaxter SRosenberg DMorey DBroaddus CParker DATE 03/23/2015 03/18/2015 03/27/2015 03/25/2015 03/31/2015 03/31/2015