LR-N17-0079, Supplemental Information for Response to Request for Additional Information, Aging Management Program Plan for Reactor Vessel Internals

Supplemental Information for Response to Request for Additional Information, Aging Management Program Plan for Reactor Vessel Internals

ML17123A070
Person / Time
Site:	Salem
Issue date:	05/03/2017
From:	Davison P Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF5149, CAC MF5150, LR-N17-0079
Download: ML17123A070 (7)

Text

PSEG Nuclear LLC P. 0. Box 236, Hancocks Bridge, New Jersey 08038**0236 MAY 0 3 2017 PSEG Nuclear LLC 10 CFR 54 LR-N17-0079 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311

Subject:

Supplemental Information for Response to Request for Additional Information, Re: Aging Management Program Plan for Reactor Vessel Internals (CAC Nos. MF5149 and MF5150)

References:

1. NRC letter to PSEG, "Salem Nuclear Generating Station, Unit Nos. 1 and 2 -

Request for Additional Information Re: Aging Management Program Plan for Reactor Vessel Internals (TAC Nos. MF5149 and MF5150)," dated March 31, 2015 (ADAMS Accession No. ML15069A181)

2. PSEG letter to NRC, "Response to Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Request for Additional Information Re: Aging Management Program Plan for Reactor Vessel Internals (TAC Nos. MF5149 and MF5150),"

dated May 28, 2015 (ADAMS Accession No. ML15148A426)

3. PSEG letter to NRC, "Supplemental Information for Response to Request for Additional Information, RAI RAI-11, Re: Aging Management Program Plan for Reactor Vessel Internals (CAC Nos. MF5149 and MF5150),"dated January 13, 2017 (ADAMS Accession No. ML17013A251)

4. Pressurized Water Reactor Owners Group letter to NRC. "PWR Owners Group Submittal of PWROG-14048-P, Revision 1, "Functionality Analysis: Lower Support Columns," dated March 1, 2017 (ADAMS Accession No. ML17066A266)

In response to the Reference 1 letter, PSEG Nuclear LLC (PSEG) provided additional information in Reference 2 concerning the reactor vessel internals aging management program for Salem Nuclear Generating Station (Salem). Attachment 1 of this letter provides an updated response for request for additional information RAI-7.

In the Reference 3 letter, PSEG committed to submit a copy of the technical report containing the lower support column flaw tolerance analysis applicable to the Salem Units 1 and 2. The technical report associated with tasks related to PWROG-14048-P has been revised and submitted to the NRC as detailed in Reference 4, and it contains the flaw tolerance analysis

MAY 0 3 2017 Page 2 LR-N17-0079 applicable to the Salem Units 1 and 2. Based on the results of the flaw tolerance analysis, the Salem Reactor Vessel Internals Lower Support Columns will remain functional through their respective period of extended operation. This completes the commitment contained in Reference 3.

There are no regulatory commitments contained in this letter.

Should you have any questions regarding this submittal, please contact Ms. Tanya Timberman at 856-339-1426.

Sincerely,

-u Paul J. D ison Vice President Nuclear Engineering : Response to Request for Additional Information cc: Mr. D. Dorman, Administrator, Region I, NRC Ms. C. Parker, Project Manager, NRC NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Chief, NJBNE Salem Commitment Tracking Coordinator Corporate Commitment Tracking Coordinator

LR-N17-0079 Attachment 1 Response to Request for Additional Information

LR-N17-0079 Attachment 1 Response to Request for Additional Information Regarding Aging Management Program Plan for Reactor Vessel Internals Salem Generating Station, Units 1 and 2 Docket Nos. 50-272 and 50-311 Regarding AI 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."1 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 toTE 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 staffs 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.

PSEG Response to RAI-7a PSEG originally submitted its aging management plan (AMP) for Salem Units 1 and 2 by letter dated August 11, 2014 (Reference 1). In this submission, assessment for the potential of thermal embrittlement (TE) and irradiation embrittlement (IE) of Cast Austenitic Stainless Steel (CASS) reactor vessel internals components was based upon the guidelines developed by the industry in MRP-191 (Reference 2). The relevant thresholds were: greater than 20% ferrite content, calculated from chemical compositions using Hull's factors (Reference 3), forTE, and neutron fluences greater than 1 dpa (6. 7x1020 n/cm2 E>1MeV) for IE. PSEG applied these screening criteria to the CF-8 grade materials that were used in the construction of the Salem Units 1 and 2 reactor vessel internals.

As a result of finding that all ninety-six (96) Lower Core Support Caps (LCSC) for Salem Unit 2 had chemical compositions which yielded ferrite contents of less than 20%, these components were screened out forTE. For Salem Unit 1, however, Certified Material Test Records (CMTRs) could only be located for twenty-three (23) of the 96 LCSCs. On this basis, a conservative assumption was made that the components, for which CMTRs could not be located, could possibly have ferrite contents greater than 20% and these components were 1 ADAMS Accession Number ML14163A112 1

LR-N17-0079 Attachment 1 screened in forTE. For both Salem Units, all LCSCs were screened in for IE for those regions in which the neutron fluence was above 1 dpa.

The screening guidelines contained in "NRC Position on Aging Management of CASS Reactor Vessel Internal Components" (References 4 and 5) are more restrictive than those recommended by MRP-191. In 2016, the NRC Staff issued further revised screening guidelines for loss of toughness in CASS in the safety evaluation (SE) of BWRVIP-234 (Reference 6). In these guidelines, the NRC Staff recognized that low molybdenum grades of CASS had previously been unnecessarily penalized for loss of toughness when overall guidelines had been issued covering high molybdenum grades (CF-8M and CF-3M) as well as low molybdenum grades. For low molybdenum grades CF-3 and CF-8, the NRC Staff issued revised screening guidelines for loss of toughness in Table A1 of Reference 6 which is reproduced below as Table 1.

Table 1: Table A1 and Notes of NRC SE of BWRVIP-234 Screening of CF-3 and CF-8 RVI Components with Neutron Exposure between 0.00015 and 1 dpa0

+

Casting Method Further Evaluation Delta Ferrite %

Yes > 20%

static No :520%

Yes >25%

centrifugal ++

No :525%

0 All CASS materials need further evaluation above 1 dpa neutron fluence.

+Estimate delta ferrite content from chemical composition with Hull's equivalent factors. If chemical composition is unknown, further evaluation is required.

++ Upper limit for validity of ferrite screening of CASS from N UREG/CR-4513, Rev. 1.

The SE stated that the NRC Staff recognized that "the criteria proposed by industry (20 percent ferrite and 1 dpa) have been shown to project significant margin on toughness and, therefore, safety when compared to the measured embrittlement for the CF-3 and CF-8 materials." These updated guidelines for CF-3 and CF-8 are identical to those used in the original screening for the LCSCs in Salem Unit 1 and 2.

While the original assessment of the LCSCs assumed conservative chemical compositions for the LCSC with missing CMTRs in Salem Unit No. 1, a statistical approach developed by the Pressurized Water Reactor Owners Group (PWROG) (Reference 7), and subsequently approved by the NRC (Reference 8), would now identify these components as most probably having ferrite contents below the 20% ferrite content screening threshold.

Therefore, these components would now not be considered to be potentially susceptible to TE.

Again, these components would not require further evaluation under the guidelines of Table 1 since the original assessments were conservative with respect to the expectations forTE.

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LR-N17-0079 Attachment 1 PSEG Response to RAI-7b The lower support column bodies are defined consistently in MRP-227-A (Reference 9) and Table C-2 of the Salem Units 1 and 2 AMP submittal (Reference 1 ) as an expansion component linked to the control rod guide tube (CRGT) lower flange weld. According to Table 4-3 of (Reference 9) and Table C-1 of (Reference 1 ), the CRGT lower flange weld is managed for age related degradation including IE and TE. PSEG is maintaining the aging management plan outlined in Table C-2 of (Reference 1 ) for the lower support columns for Salem Units 1 and 2.

This plan requires inspection of the lower support column bodies, including the caps, by enhanced visual (EVT-1) as an expansion component from the CRGT lower flange weld.

The basis for this aging management approach is as follows: Section 3.3.7 of the safety evaluation (SE) of MRP-227-A identifies the concerns related to the effects of TE and IE on the lower support column bodies and states that applicants/licensees shall perform plant-specific analysis or evaluation to demonstrate that the MRP-227-A recommended inspections will ensure functionality of the lower support columns until the next scheduled inspections. The SE goes on to provide three possible approaches that would be accepted. One of which is a functionality analysis. The functionality analysis applicable to Salem Units 1 and 2 is documented in PWROG-14048-P, Revision 1 (Reference 10), submitted to the NRC for information only under OG-1 7-62 (Reference 1 1 ). Evaluations conducted in PWROG-1 4048-P, Revision 1 , using conditions consistent with or bounding of Salem Units 1 and 2 design basis conditions, show that there is low likelihood of failure even when considering potential effects of IE and TE. Furthermore, this report also shows significant redundancy exists in this structure such that functionality can be maintained should unexpected failures of a small number of columns randomly occur. Therefore, the MRP-227-A recommended inspections will ensure functionality of the lower support columns during the period of extended operation.

References:

1. PSEG Nuclear, LLC Letter, LR-N14-0183, "Submittal of PWR Vessel Internals Inspection Plans for Aging Management of Reactor Internals at Salem Generating Station, Units 1 and 2," August 1 1, 201 4 (ADAMS Accession No. ML14224A667).

2. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-1 91), TR-1 0 1 3234, EPRI, Palo Alto, CA: 2006.

3. F. C. Hull "Delta Ferrite and Martensite Formation in Stainless Steels" Welding Journal May 1 973, pp1 93-203.

4. Nuclear Regulatory Commission Document, "NRC Position on Aging Management of CASS Reactor Vessel Internal Components," June 1 1 , 2014 (ADAMS Accession Number ML14163A112).

5. Nuclear Regulatory Commission Document, "NRC Position on Aging Management of CASS Reactor Vessel Internal Components," June 23, 201 4. (ADAMS Accession Number ML14174A719).

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LR-N17-0079

6. Nuclear Regulatory Commission Letter, "Final Safety Evaluation of the BWRVIP-234:

Thermal Aging and Neutron Embrittlement Evaluation of Cast Austenitic Stainless Steel for BWR Internals (TAC No. ME5060)" June 22, 2016 (ADAMS Accession No. ML16096A002).

7. PWR Owner's Group Report PWROG-15032-NP Revision 0, "PA-MSC-1288 Statistical Assessment of the PWR RV Internals CASS Materials," November 11, 2015.

8. Nuclear Regulatory Commission Document, "Office of Nuclear Reactor Regulations Staff Assessment of the Pressurized Water Reactor Owner's Group Report PWROG-15032-NP, Revision 0, "PA-MSC-1288 Statistical Assessment of PWR RV Internals CASS Materials," September 6, 2016 (ADAMS Accession No. ML16250A001).

9. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), EPRI, Palo Alto, CA: 2011. 1022863.

10. PWR Owner's Group Report, PWROG-14048-P, Revision 1, " Functionality Analysis:

Lower Support Columns", February 2017.

11. PWR Owner's Group Letter, OG-17-62, "Submittal of PWROG-14048-P, Revision 1,

" Functionality Analysis: Lower Support Columns" to the NRC for Information Only (PA MSC-1103)," March 1, 2017.

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