ML17219A186

Safety Evaluation of Relief Request SC-I4R-171 Regarding the Fourth 10-Year Interval of the Inservice Inspection Program

ML17219A186
Person / Time
Site:	Salem
Issue date:	08/17/2017
From:	Richard Ennis Plant Licensing Branch 1
To:	Sena P Public Service Enterprise Group
Ennis R
References
CAC MF9348, CAC MF9349
Download: ML17219A186 (13)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 17, 2017 Mr. Peter P. Sena, Ill President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1AND2- SAFETY EVALUATION OF RELIEF REQUEST SC-14R-171 REGARDING THE FOURTH 10-YEAR INTERVAL OF THE INSERVICE INSPECTION PROGRAM (CAC NOS. MF9348 AND MF9349)

Dear Mr. Sena:

By letter dated March 1, 2017, as supplemented by letter dated June 5, 2017, PSEG Nuclear LLC (PSEG, the licensee) submitted Relief Request SC-14R-171 to the U.S. Nuclear Regulatory Commission (NRC). PSEG proposed an alternative to certain inservice inspection (ISi) requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) for Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2),

PSEG requested to use the alternative inner diameter flaw depth sizing root mean square error criteria in ASME Code Cases N-695-1, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1," and N-696-1, "Qualification Requirements for Mandatory Appendix VIII Piping Examinations Conducted from the Inside Surface,Section XI, Division 1."

The NRC staff has completed its review of the subject relief request as documented in the enclosed safety evaluation. Our safety evaluation concludes that the proposed alternative provides reasonable assurance of the structural integrity and leak-tightness of the subject welds, and that complying with the ASME Code requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.

Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the NRC staff authorizes the use of Relief Request SC-14R-171 for the remainder of the fourth 10-year ISi interval of Salem, Unit No. 1, which started on May 20, 2011, and is scheduled to end on December 31, 2020, and the remainder of the fourth 10-year ISi interval of Salem, Unit 2, which started on November 27, 2013, and is scheduled to end on December 31, 2021.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third party review by the Authorized Nuclear lnservice Inspector

P.Sena If you have any questions concerning this matter, please contact the Salem Project Manager, Mr. Richard Ennis, at (301) 415-1420 or Rick.Ennis@nrc.gov.

Sincerely, James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosure:

Safety Evaluation 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 RELATED TO RELIEF REQUEST SC-14R-171 FOURTH 10-YEAR INTERVAL OF THE INSERVICE INSPECTION PROGRAM PSEG NUCLEAR LLC SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311

1.0 INTRODUCTION

By letter dated March 1, 2017, as supplemented by letter dated June 5, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML17060A477 and ML17156A804, respectively), PSEG Nuclear LLC (PSEG, the licensee) submitted Relief Request SC-14R-171 to the U.S. Nuclear Regulatory Commission (NRC). PSEG proposed an alternative to certain inservice inspection (ISi) requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) for Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2. Specifically, pursuant to Title 10 of the Code of Federal Regulations ( 10 CFR) 50.55a(z)(2), PSEG requested to use the alternative inner diameter (ID) flaw depth sizing root mean square error (RMSE) criteria in ASME Code Cases N-695-1, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1," and N-696-1, "Qualification Requirements for Mandatory Appendix VIII Piping Examinations Conducted from the Inside Surface,Section XI, Division 1."

The subject relief request is for the Salem, Unit No. 1, fourth 10-year ISi interval, which began on May 20, 2011, and is currently scheduled to end on December 31, 2020, and the Salem, Unit No. 2, fourth 10-year ISi interval, which began on November 27, 2013, and is scheduled to end on December 31, 2021.

The licensee's relief request cites requirements in 10 CFR 50.55a(g)(6)(ii)(F) pertaining to Code Case N-770-1. However, in a Federal Register notice dated July 18, 2017 (82 FR 32934), the NRC staff issued a final rule amending its regulations to incorporate by reference recent editions and addenda to the ASME Code and to incorporate by reference a number of code cases. This final rule included a revision to 10 CFR 50.55a(g)(6)(ii)(F) to change the version of ASME Code Case N-770 from N-770-1 to N-770-2. As such, this safety evaluation refers to ASME Code Case N-770-2 in place of ASME Code Case N-770-1.

2.0 REGULATORY EVALUATION

Pursuant to 10 CFR 50.55a(g)(4), "lnservice testing standards requirement for operating plants,

throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, Class 2, and Enclosure

Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions and addenda of the ASME Code that become effective subsequent to editions specified in paragraphs (g)(2) and (3) of 10 CFR 50.55a, and that are incorporated by reference in paragraph (a)(1 )(ii) of 10 CFR 50.55a to the extent practical, within the limitations of design, geometry, and materials of construction of the components.

Pursuant to 10 CFR 50.55a(g)(4)(ii), "Applicable ISi Code: Successive 120-month intervals,"

inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code, incorporated by reference in paragraph (a) of 10 CFR 50.55a, 12 months before the start of the 120-month inspection interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1" (ADAMS Accession No. ML13339A689), when using Section XI, that are incorporated by reference in paragraph (a)(3)(ii) of 10 CFR 50.55a)), subject to the conditions listed in paragraph {b) of 10 CFR 50.55a. However, a licensee whose ISi interval commences during the 12 through 18-month period after July 21, 2011, may delay the update of its Appendix VIII program by up to 18 months after July 21, 2011.

By Federal Register notice 82 FR 32934, dated July 18, 2017 (which became effective on July 18, 2017), the requirements in 10 CFR 50.55a(g)(6)(ii)(F) were revised. Pursuant to 10 CFR 50.55a(g)(6)(ii)(F), "Augmented ISi requirements: Examination requirements for Class 1 piping and nozzle dissimilar-metal butt welds," holders of operating licenses or combined licenses for pressurized-water reactors (PWRs) as of or after August 17, 2017, shall implement the requirements of ASME Code Case N-770-2, "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1," instead of ASME Code Case N-770-1, subject to the conditions specified in paragraphs {g}(6)(ii)(F)(2) through (13) of 10 CFR 50.55a, by the first refueling outage starting after August 17, 2017.

In accordance with 10 CFR 50.55a{z), "Alternatives to codes and standards requirements,"

alternatives to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a, or portions thereof, may be used when authorized by the NRC pursuant to 10 CFR 50.55a(z)(1) or (2). A proposed alternative must be submitted and authorized prior to implementation. In proposing alternatives, the licensee must demonstrate that: ( 1) the proposed alternatives would provide an acceptable level of quality and safety or (2) compliance with the specified requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the NRC to authorize, the alternative requested by the licensee.

3.0 TECHNICAL EVALUATION

3.1 Background By letters dated August 17, 2011 (ADAMS Accession No. ML112140424), and June 23, 2014 (ADAMS Accession No. ML14153A146), the NRC approved implementation of the risk-informed inservice inspection (RI-ISi) program for the Class 1 piping welds (Examination Categories B-F

and B-J) and the Class 2 piping welds (Examination Categories C-F-1 and C-F-2) for the fourth 10-year ISi interval of Salem, Unit Nos. 1 and 2, respectively. The licensee developed the Salem RI-ISi program in accordance with the NRG-approved methodology of Electric Power Research Institute (EPRI) Topical Report (TR)-112657, Revision B-A, "Revised Risk-Informed lnservice Inspection Evaluation Procedure, December 1999 (ADAMS Accession No. ML013470102).

3.2 Components Affected The affected components are ASME Code Class 1 piping welds in the reactor coolant system at Salem, Unit Nos. 1 and 2. These welds consist of the dissimilar metal (OM) butt welds that are fabricated with nickel based alloy materials (e.g., Alloy 82/182) and known to be susceptible to the primary water corrosion cracking (PWSCC) and the similar metal butt welds that are fabricated with austenitic stainless steel materials and known to be not susceptible to PWSCC.

• OM welds - Salem, Unit No. 1 o Four reactor pressure vessel (RPV) outlet (hot leg) nozzle to safe-end OM butt welds have ID of 29 inches and wall thickness of 2.5 inches. The licensee classified three welds as Inspection Item D and one weld as Inspection Item E in accordance with ASME Code Case N-770-2 (Table 1).

29-RC-1110-1 Inspection Item D 29-RC-1120-1 Inspection Item D 29-RC-1130-1 Inspection Item D 29-RC-1140-1 Inspection Item E o Four RPV inlet (cold leg) nozzle to safe-end DM butt welds with ID of 27.5 inches and wall thickness of 2.50 inches. The licensee classified all four welds as Inspection Item Din accordance with ASME Code Case N-770-2 (Table 1).

27 .5-RC-1110-5 Inspection Item D 27.5-RC-1120-5 Inspection Item D 27.5-RC-1130-5 Inspection Item D 27.5-RC-1140-5 Inspection Item D The licensee stated that the materials of construction of the above welds and associated components are low alloy steel nozzles welded to austenitic stainless steel safe-ends by nickel based alloy (e.g., Alloy 82/182) weld metals.

• Similar metal welds - Salem, Unit No. 1 o Four RPV outlet (hot leg) safe-end to pipe butt welds have ID of 29 inches and wall thickness of 2.5 inches. The licensee classified these welds as Examination Category R-A, Item Number R1 .20 (elements not subject to a damage mechanism) in accordance with EPRI TR-112657, Revision B-A (Table 1 of ASME Code Case N-578-1).

29-RC-1110-2 29-RC-1120-2 29-RC-1130-2 29-RC-1140-2

The licensee stated that the materials of construction of the above welds and associated components are austenitic stainless steel safe-ends welded to austenitic stainless steel pipes by austenitic stainless steel weld metals.

• DM welds - Salem, Unit No. 2 o Four RPV outlet (hot leg) nozzle to safe-end DM butt welds have ID of 29 inches and wall thickness of 2.5 inches. The licensee classified all four welds as Inspection Item Din accordance with ASME Code Case N-770-2 (Table 1).

29-RC-1210-1 Inspection Item D 29-RC-1220-1 Inspection Item D 29-RC-1230-1 Inspection Item D 29-RC-1240-1 Inspection Item D o Four RPV inlet (cold leg) nozzle to safe-end DM butt welds have ID of 27.5 inches and wall thickness of 2.50 inches. The licensee classified all four welds as Inspection Item B in accordance with ASME Code Case N-770-2 (Table 1).

27.5-RC-1210-5 Inspection Item B 27.5-RC-1220-5 Inspection Item B 27 .5-RC-1230-5 Inspection Item B 27.5-RC-1240-5 Inspection Item B The licensee stated that the materials of construction of the above welds and associated components are low alloy steel nozzles welded to austenitic stainless steel safe-ends by nickel based alloy (e.g., Alloy 82/182) weld metals.

• Similar metal welds - Salem, Unit No. 2 o Four RPV outlet (hot leg) safe-end to pipe butt welds have ID of 29 inches and wall thickness of 2.5 inches. The licensee classified these welds as Examination Category R-A, Item Number R1 .20 (elements not subject to a damage mechanism) in accordance with EPRI TR-112657, Revision B-A (Table 1 of ASME Code Case N-578-1).

29-RC-1210-2 29-RC-1220-2 29-RC-1230-2 29-RC-1240-2 The licensee stated that the materials of construction of the above welds and associated components are austenitic stainless steel safe-ends welded to austenitic stainless steel pipes by austenitic stainless steel weld metals.

The licensee stated that prior to utilizing the ultrasonic testing (UT) for the examinations of the above welds (i.e., RPV nozzle to safe end welds and safe end to pipe welds at Salem, Unit Nos. 1 and 2), the UT procedures demonstration, equipment, and personnel qualification shall meet applicable Appendix VIII supplements (Section XI).

In its letter dated June 5, 2017, the licensee stated, in part:

The eight cold leg safe end austenitic welds (four on each Salem unit) are not included in the request for relief. The eight cold leg safe end welds contain austenitic stainless steel safe ends which are welded to cast austenitic stainless steel (CASS) elbows .... these eight welds are not currently scheduled for examination during the fourth lnservice Inspection 1O year intervals.

3.3 Applicable Code Edition and Addenda The code of record for the fourth 10-year ISi interval is the 2004 Edition with no addenda of the ASME Code.

3.4 Duration of Relief Request The licensee submitted this relief request for the fourth 10-year ISi interval of Salem, Unit No. 1, which started on May 20, 2011, and is scheduled to end on December 31, 2020, and the fourth 10-year ISi interval of Salem, Unit No. 2, which started on November 27, 2013, and is scheduled to end on December 31, 2021.

3.5 ASME Code Requirements Requirements for the IS/ of OM Welds The ASME Code ISi requirements applicable to the DM welds in this relief request originate in Table IWB-2500-1 (Section XI). However, the regulations under 10 CFR 50.55a{g){6)(ii)(F) mandate augmented inspection in accordance with ASME Code Case N-770-2 with conditions for the DM welds that contain Alloy 82/182. ASME Code Case N-770-2 (Table 1), Inspection Items B, D, and E, require the RPV hot and cold leg nozzle to safe-end DM butt welds to be volumetrically examined by the UT. Footnote number 4 of Case N-770-2 (Table 1) requires that the UT procedures demonstration and personnel qualification meet applicable supplements of Appendix VIII (Section XI).

Requirements for the ISi of Similar Metal Welds The ASME Code ISi requirements applicable to the similar metal welds in this relief request originate in Table IWB-2500-1 (Section XI). The Salem, Unit Nos. 1 and 2, RI-ISi program was developed by the licensee in accordance with the NRG-approved methodology in EPRI TR-112657, Revision B-A, and was authorized by the NRC staff for Salem, Unit Nos. 1 and 2, respectively, in safety evaluations dated August 17, 2011, and June 23, 2014. The RI-ISi program provides an alternative to the ASME Code requirements. In both the ASME Code and the NRC's safety evaluations, it is required that the austenitic welds under this request be volumetrically examined by the UT, and that the UT procedures and personnel be demonstrated and qualified in accordance with applicable supplements of Appendix VIII (Section XI).

Applicable Supplement of ASME Code,Section XI, Appendix VIII For the welds (i.e., DM and austenitic welds) in this relief request, applicable supplements of ASME Code,Section XI, Appendix VIII, are Supplement 10, "Qualification Requirements for Dissimilar Metal Piping Welds"; Supplement 2, "Qualification Requirements for Wrought Austenitic Piping Welds"; and Supplement 14, "Qualification Requirements for Coordinated Implementation of Supplements 10, 2, and 3 for Piping Examinations Performed from the Inside

Surface." In accordance with these supplements, it is required that the UT procedures, equipment, and personnel be qualified for the ID flaw depth sizing, and that the flaw depths estimated by the UT as compared with the true flaw depths do not exceed 0.125 inch RMSE.

An alternative to Supplement 1O is ASME Code Case N-695-1. An alternative to Supplement 14 (i.e., coordinate implementation of Supplements 10, 2, and 3) is ASME Code Case N-696-1. Neither of these code cases have been incorporated by reference into 10 CFR 50.55a by inclusion in NRC Regulatory Guide 1.147, Revision 17.

3.6 Proposed Alternative Performance Demonstration of UT For the UT procedures demonstration, equipment, and personnel qualification, the licensee proposed to use the ID flaw depth sizing RMSE criteria of ASME Code Cases N-695-1 and N-696-1 in lieu of the requirements contained in ASME Code,Section XI, Appendix VIII, Supplements 10 and 14. The proposed alternative ID flaw depth sizing RMSE criteria of ASME Code Cases N-695-1 and N-696-1 are as follows:

a. According to N-695-1, the examination procedures, equipment, and personnel are qualified for ID flaw depth sizing if the flaw depths estimated by the UT as compared with the true flaw depths do not exceed 0.125 inch RMSE for piping less than 2.1 inches in thickness, and 0.250 inch RMSE for piping greater than or equal to 2.1 inches in thickness (i.e., large diameter piping welds).

b. According to N-696-1, the Supplement 2 examination procedures, equipment, and personnel are qualified for ID flaw depth sizing if the flaw depths estimated by the UT, as compared with the true flaw depths, do not exceed, 0.125 inch RMSE for piping less than 2.1 inches in thickness, and 0.250 inch RMSE for piping greater than or equal to 2.1 inches in thickness (i.e., large diameter piping welds) when they are combined with a successful Supplement 10 qualification.

Examination of RPV Nozzle Welds In the event that a flaw is detected in the RPV nozzle to safe end DM welds and/or the safe end to pipe austenitic welds and requires depth sizing, the licensee stated in its proposed alternative that:

• Flaws detected and measured as less than 50 percent through-wall will be sized in accordance the personnel, procedures and equipment qualified to meet the requirements of ASME Code Cases N-695-1 and N-696-1.

• For flaws detected and measured as 50 percent through-wall depth or greater and to remain in service without mitigation or repair, PSEG will submit flaw evaluation(s) for review and approval prior to reactor startup.

The flaw evaluation will include:

1. Information concerning the mechanism that caused the flaw.

2. Information concerning the inside surface roughness and/or profile of the region surrounding the flaw in the examined piping weld.

3. Information concerning areas where UT probe lift-off is observed, if any.

The licensee also stated in its proposed alternative that eddy current testing will be used to determine if flaws are surface connected.

3.7 Basis for Use The licensee stated that although vendors contracted by the licensee have qualified for flaw detection and length sizing for inspections performed from the ID surface of the weld, the UT qualification for ID flaw depth sizing has not yet been successful to meet the ASME Code-required 0.125 inch RMSE. To date, no vendor has been capable of meeting the required 0.125 inch RMSE criterion for flaw depth sizing for inspections performed from the ID surface.

Consequently, a request to use an alternative ID flaw depth sizing RMSE criteria is sought.

The licensee stated that the UT procedures, equipment, and personnel have been qualified in accordance with Appendix VIII for flaw depth sizing from the outer diameter (OD) surface.

Although the OD surface of the welds in this relief request could be accessed (i.e., through the "sandboxes" that covered the reactor vessel inlet and outlet nozzles), there would be a significant radiological dose associated with inspecting the welds from the OD, which would be considered a hardship. The licensee estimated that the radiation dose accrued by personnel performing the volumetric examination of four RPV cold leg nozzle to safe-end welds from the OD at Salem, Unit No. 2, was 3.5 roentgen equivalent man (rem). This estimate does not include any additional dose received by supporting personnel (e.g., maintenance and radiation protection). From the OD inspection of all of the welds in this relief request (24 welds), it is estimated that personnel would unnecessarily receive additional radiation exposure in the order of over 21 rem, which would be a hardship to the licensee.

The licensee stated that performing the examination from the ID reduces the overall exposure to radiation since the ID examination is performed remotely and does not require personnel to access the exterior "sandbox" area of the RPV.

3.8 NRG Staff Evaluation The NRG staff has evaluated this request pursuant to 10 CFR 50.55a(z)(2). The NRG staff focuses on whether (1) compliance with the specified requirements results in a hardship or unusual difficulty, (2) the alternative RMSE is adequate, and (3) the licensee's proposed alternative (i.e., accepting alternative RMSE and obtaining the NRG approval for reactor startup in the event flaws are detected and depth-measured greater than or equal to 50 percent through-wall thickness) provides reasonable assurance of structural integrity and leak-tightness of the welds. The NRG staff finds that if these three criteria are met, then the requirements of 10 CFR 50.55a(z)(2) will also be met.

Hardship In its evaluation, the NRG staff found that requiring the licensee to comply with ASME Code,Section XI, Appendix VIII (Supplements 10 and 14), flaw depth sizing RMSE criteria for performance demonstration of the UT for examinations from the ID surface would result in hardship. The basis for the hardship is as follows:

• For more than a decade, industry has made every effort (repeated attempts and even equipment enhancements) to reduce the ID flaw depth sizing uncertainty, but none have achieved the ASME Code-required RMSE of 0.125 inch.

• The NRG staff confirms that the inherent challenges for reducing the ID flaw depth sizing uncertainty are attributed to geometry and roughness of the welds' ID surface, multiple materials, and microstructural anisotropies. These conditions constitute unusual difficulties.

• Industry has been successful in achieving the ASME Code RMSE of 0.125 inch for the OD flaw depth sizing performance demonstration. The licensee acknowledged that the OD surfaces of the welds in this relief request are accessible for ultrasonic scanning. If the volumetric examinations are performed from OD, personnel who perform a qualified UT from the OD will unnecessarily be exposed to additional radiation dose. Exposure to additional dose constitutes a hardship.

Based on the above, the NRG staff determined that imposing the ASME Code requirements could result in a hardship and unusual difficulty for the facility.

Safety Significance of Alternative RMSE In evaluating the licensee's proposed alternative performance demonstration criteria, the NRG staff assessed whether it appears that the proposed ID flaw depth sizing RMSE of 0.250 inch for large bore piping welds is adequate. The NRG staff found that:

• Since 2002, licensees have submitted relief requests to use alternative RMSE criteria for the UT procedure demonstration and personnel qualifications specifically related to the ID flaw depth sizing in the OM and/or austenitic butt welds in PWR piping. The basis has been that licensees and/or inspection vendors have repeatedly attempted to qualify the UT procedures and personnel for ID flaw depth sizing in accordance with the ASME Code-required RMSE of 0.125 inch, but none have been successful due to materials inhomogeneity/anisotropies and ID surface geometry and conditions.

• In 2011, the NRG and EPRI signed a "Memorandum of Understanding for Nondestructive Examination" to allow review/assessment of Performance Demonstration Initiative (POI) program proprietary data, development of a technical basis to support utilizing alternative RMSE criteria for ID flaw depth sizing, and subsequently, an ASME Code action. In July 2012, the NRG staff, along with EPRI personnel, reviewed the POI program proprietary data used in blind tests. This review was conducted to assess potentials for undersizing the depth of detected flaws, as well as to verify the information and analysis presented by EPRI and industry in a public meeting held between the NRG staff, POI, EPRI, and industry representatives on March 16, 2012 (ADAMS Package Accession No. ML12097A071), and June 19, 2012 (ADAMS Accession Nos. ML12173A517 and ML12173A522). This collective assessment has revealed that the potential exists to undersize and/or oversize depth of flaws detected in the volumetric examinations. Based on this review, the NRG staff determined that until alternative screening criteria have been developed to justify the ID flaw depth sizing uncertainty, alternative RMSE criteria need to be submitted to the NRG for approval through a relief request process prior to implementation at plants.

• In 2013, EPRI published Technical Report No. 3002000612, "Materials Reliability Program: Technical Basis for Change to American Society of Mechanical Engineers (ASME)Section XI Appendix VIII Root-Mean-Square Error Requirement for Qualification of Depth-Sizing for Ultrasonic Testing (UT) Performed from the Inner Diameter (ID) of Large-Diameter Thick-Wall Supplement 2, 10, and 14 Piping Welds (MRP-373),"

recommending the RMSE depth sizing criteria be changed from 0.125 inch (as required by the ASME Code) to 0.250 inch for the large bore PWR piping welds (a nominal wall thickness of at least 2.1 inch) for examinations from ID. This suggests that an inspector qualified with a 0.25 inch RMSE is capable of accurately measuring depth of detected flaws. Subsequently, the industry developed ASME Code Cases N-695-1 and N-696-1 using MRP-373 as a technical basis. On December 31, 2014, ASME approved Code Cases N-695-1 and N-696-1. It should be noted that since these two code cases have not yet been incorporated by reference into 10 CFR 50.55a by inclusion in RG 1.147, Revision 17, the NRC's approval is required prior to their use by licensees.

• In the annual industry/NRC NOE technical information exchange public meetings held on January 8-9, 2014 (meeting summary at ADAMS Accession No. ML14057A752, NRC presentation at ADAMS Accession No. ML14015A036), and January 13-15, 2015 (meeting summary at ADAMS Accession No. ML15026A289, NRC presentation at ADAMS Accession No. ML15013A293, industry presentation at ADAMS Accession No. ML15013A515), the NRC staff acknowledged that a resolution has been reached to change the RMSE ID depth sizing criteria from 0.125 inch to 0.25 inch for PWR piping welds with nominal wall thickness of at least 2.1 inch provided that no flaws greater than or equal to 50 percent through-wall thickness be left in service without NRC approval.

Based on the above collective efforts, the NRC staff determines that the licensee's proposed RMSE is acceptable because it provides reasonable assurance that personnel who perform examinations have been qualified to measure the depth of flaws with a reasonable accuracy.

In addition to the analysis described above, the NRC staff evaluated the safety significance of not allowing detected flaws with depths measured greater than or equal to 50 percent through-wall thickness be left in service without NRC approval. Although the NRC staff finds that the assertions and interpretations documented in MRP-373 (technical basis document for Code Cases N-695-1 and N-696-1) are reasonable for procedures and personnel qualifications, there still exists concern about potentials for flaw undersizing during examinations of the welds due to their ID surface roughness and geometrical surface waviness. Therefore, the NRC staff determines that if a flaw is detected and its depth is measured to be greater than or equal to 50 percent through-wall thickness, and is to remain in service without mitigation or repair, submitting a flaw evaluation, supplemented with the information below, for NRC review and approval prior to reactor startup, provides reasonable assurance of the structural integrity and leak-tightness of the welds in this relief request. The supplemental information required to be submitted with the flaw evaluation is as follows:

• The inner profile of the weld, pipe, and nozzle in the region at and surrounding the flaw,

• An estimate of the percentage of potential surface areas with UT probe lift-off, and

• Information on the mechanism that caused the crack.

Therefore, the NRC staff finds that the licensee's proposed alternative (i.e., use of ASME Code Cases N-695-1 and N-696-1 for performance demonstration and obtaining the NRC approval for restart) provides reasonable assurance of structural integrity and leak-tightness of the welds.

4.0 CONCLUSION

As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of the structural integrity and leak-tightness of the subject welds, and complying with the specified requirement would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the NRC staff authorizes the use of SC-14R-171 for the remainder of the fourth 10-year ISi interval of Salem, Unit No. 1, which started on May 20, 2011, and is scheduled to end on December 31, 2020, and the remainder of the fourth 10-year ISi interval of Salem, Unit No. 2, which started on November 27, 2013, and is scheduled to end on December 31, 2021.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third party review by the Authorized Nuclear lnservice Inspector.

Principal Contributor: A. Rezai Date: August 17, 2017

P. Sena

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 - SAFETY EVALUATION OF RELIEF REQUEST SC-14R-171 REGARDING THE FOURTH 10-YEAR INTERVAL OF THE INSERVICE INSPECTION PROGRAM (CAC NOS. MF9348 AND MF9349) DATED AUGUST 17, 2017 DISTRIBUTION:

PUBLIC JBowen, OEDO RidsNrrDorlLpl1 Resource RidsACRS_MailCTR Resource RidsNrrLALRonewicz Resource RidsNrrDeEpnb Resource RidsNrrPMSalem Resource ARezai, NRR RidsRgn1 MailCenter Resource ADAMS Access1on No.: ML17219A186 *b1y sa f e tty eva uat1on d ate d7/29/2017 OFFICE DORL/LPL 1/PM DORL/LPL 1/LA DE/EPNB/BC* DORL/LPL 1/BC DORL/LPL 1/PM NAME REnnis LRonewicz DAiiey JDanna REnnis DATE 081 /2017 08/08/2017 07/29/2017 08/16/2017 08/17/2017 OFFICIAL RECORD COPY