ML23139A147

Relief Request Associated with Fourth Interval In-service Inspection Limited Examinations of Weld Coverage

ML23139A147
Person / Time
Site:	Salem
Issue date:	06/05/2023
From:	Hipolito Gonzalez Plant Licensing Branch 1
To:	Carr E Public Service Enterprise Group
Kim J
References
EPID L-2022-LLR-0066
Download: ML23139A147 (1)

Text

June 5, 2023 Mr. Eric Carr President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT 2 - RELIEF REQUEST ASSOCIATED WITH FOURTH INTERVAL INSERVICE INSPECTION LIMITED EXAMINATIONS OF WELD COVERAGE (EPID L-2022-LLR-0066)

Dear Mr. Carr:

By letter dated September 27, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22270A326), as supplemented by letter dated February 1, 2023 (ML23032A426), PSEG Nuclear LLC (PSEG, the licensee) requested relief from the examination coverage requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, at Salem Generating Station, Unit 2 (Salem, Unit 2), for the fourth 10-year inservice inspection (ISI) interval.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii),

the licensee requested relief from the required examination coverage and to use alternative requirements (if necessary) for ISI of the vessel and piping welds on the basis that the ASME Code requirements are impractical. The U.S. Nuclear Regulatory Commission (NRC or the Commission) staff has determined that it is impractical to satisfy the ASME Code-required greater than 90 percent examination coverage for the subject welds due to material and component configuration. The NRC staff finds that the proposed examination coverage provides reasonable assurance of structural integrity or leak tightness of the subject component. The NRC staff determines that granting the relief request pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Therefore, the NRC staff grants the use of this relief request (i.e., S2-I4R-211) for the duration of the fourth 10-year ISI interval at Salem, Unit 2, which began on November 27, 2013, and ended on December 31, 2021.

All other requirements of the ASME Code,Section XI, for which an alternative was not specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

E. Carr If you have any questions, please contact the Salem Project Manager, James Kim, at 301-415-4125 or by email to James.Kim@nrc.gov.

Sincerely, Digitally signed by Hipolito Hipolito J. J. Gonzalez Date: 2023.06.05 Gonzalez 12:12:33 -04'00' Hipólito J. González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-311

Enclosure:

Safety Evaluation cc: Listserv

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST S2-I4R-211 REGARDING LIMITED EXAMINATIONS OF WELD COVERAGE PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC SALEM NUCLEAR GENERATING STATION, UNIT 2 DOCKET NO. 50-311

1.0 INTRODUCTION

By letter dated September 27, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22270A326), as supplemented by letter dated February 1, 2023 (ML23032A426), PSEG Nuclear LLC (the licensee) requested relief from the examination coverage requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, at Salem Generating Station, Unit 2 (Salem, Unit 2), for the fourth 10-year inservice inspection (ISI) interval.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii),

the licensee requested relief from the required examination coverage and to use alternative requirements (if necessary) for ISI of the vessel and piping welds on the basis that the ASME Code requirements are impractical.

2.0 REGULATORY EVALUATION

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for In-service Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest Edition and Addenda of Section XI of the ASME Code, incorporated by reference in 10 CFR 50.55a(b), 18 months prior to the start of the 120-month interval, subject to the conditions listed in paragraph (b) of Section 50.55a.

The regulation in 10 CFR 50.55a(b)(2)(xv)(A) requires that, when applying Supplement 2, Qualification Requirements for Wrought Austenitic Piping Welds, to ASME Code,Section XI, Enclosure

Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems, the following examination coverage criteria be met:

1. Piping must be examined in two axial directions, and when examination in the circumferential direction is required, the circumferential examination must be performed in two directions, provided access is available. Dissimilar metal welds must be examined axially and circumferentially.

2. Where examination from both sides is not possible, full coverage credit may be claimed from a single side for ferritic welds. Where examination from both sides is not possible on austenitic welds or dissimilar metal welds, full coverage credit from a single side may be claimed only after completing a successful single-sided Appendix VIII demonstration using flaws on the opposite side of the weld. Dissimilar metal weld qualifications must be demonstrated from the austenitic side of the weld, and the qualification may be expanded for austenitic welds with no austenitic sides using a separate add-on performance demonstration. Dissimilar metal welds may be examined from either side of the weld.

The regulation in 10 CFR 50.55a(b)(2)(xvi)(B) requires in part that, examinations performed from one side of a ferritic or stainless steel pipe weld must be conducted with equipment, procedures, and personnel that have demonstrated proficiency with single-sided examinations.

To demonstrate equivalency to two-sided examinations, the demonstration must be performed to the requirements of Appendix VIII, as conditioned by paragraph 10 CFR 50.55a(b)(2)(xvi)(B) and paragraph 10 CFR 50.55a(b)(2)(xv)(A).

The regulation in 10 CFR 50.55a(g)(5)(iii) states:

If the licensee has determined that conformance with a Code requirement is impractical for its facility the licensee must notify the NRC [Nuclear Regulatory Commission or Commission] and submit, as specified in § 50.4, information to support the determinations. Determinations of impracticality in accordance with this section [50.55a] must be based on the demonstrated limitations experienced when attempting to comply with the code requirements during the inservice inspection interval for which the request is being submitted. Requests for relief made in accordance with this section [50.55a] must be submitted to the NRC no later than 12 months after the expiration of the initial or subsequent 120-month inspection interval for which relief is sought.

The regulation at 10 CFR 50.55a(g)(6)(i) states:

The Commission will evaluate determinations under paragraph (g)(5) of this section [50.55a] that code requirements are impractical. The Commission may grant such relief and may impose such alternative requirements as it determines is authorized by law, and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

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 staff to grant the relief requested by the licensee.

3.0 TECHNICAL EVALUATION

3.1 Applicable Code Edition and Addenda The licensee stated that the fourth 10-year ISI interval at Salem, Unit 2, was based on the ASME Code,Section XI, 2004 Edition, as modified by 10 CFR 50.55a. The licensee further stated that the requirements of the ASME Code,Section XI, Appendix VIII, and use of the Performance Demonstration Initiative (PDI) requirements at Salem, Unit 2, were in accordance with the ASME Code,Section XI, 2001 Edition, for the limited examinations contained in the relief request as conditioned by 10 CFR 50.55a(b)(2)(xv) and 10 CFR 50.55a(b)(2)(xxiv).

3.2 Duration of Proposed Alternative This request for relief is for Salem, Unit 2, fourth 10-year ISI interval, which began on November 27, 2013, and ended on December 31, 2021.

3.3 Reactor Vessel Shell Welds 3.3.1 ASME Code Component(s) Affected The affected components are Class 1 reactor vessel shell welds having Examination Categories B-A and B-B with associated Item numbers of B1.11, B1.12, B1.21, B1.22, and B2.12 per Table IWB-2500-1 of the ASME Code,Section XI, 2004 Edition as shown in the table 1 below.

Table 1 of the relief request provides detailed information regarding the subject welds. The licensee examined the subject welds in inspection period 3 of the fourth ISI interval during refueling outage S2R25 in 2021.

Table 1. Examination Categories B-A and B-B Welds Examination Item No. Class 1 Weld Examination Coverage Requirements Category B-A B1.11 To include essentially 100% of the Reactor Vessel Circumferential Shell Welds B-A B1.12 To include essentially 100% of the Reactor Vessel Longitudinal Shell Welds B-A B1.21 To include essentially 100% of the Reactor Vessel Circumferential Head Welds B-A B1.22 To include essentially 100% of the Reactor Vessel Meridional Head Welds B-B B2.12 To include essentially 100% of the Pressurizer Shell-to-Head Longitudinal Welds 3.3.2 Applicable Code Requirements The licensee stated that ASME Code Case N-460, Alternative Examination Coverage for Class 1 and Class 2 WeldsSection XI, Division 1, as approved in Regulatory Guide 1.147, Revision 20, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, specifies that essentially 100 percent examination coverage equates to more than 90 percent of the examination volume or required surface area of each weld where the reduction in coverage

is due to interference by another component or part geometry. The licensee invoked Code Case N-460 for the required coverage associated with the welds in the relief request.

3.3.3 Impracticality and Burden of Compliance The licensee stated that details of examination restrictions and reductions in required examination coverage are provided in Attachment 1 to the relief request. The licensee further stated that when examined, the subject welds did not receive the required code volume or surface area coverage due to their component design configurations or interference. These conditions resulted in scanning or surface area access limitations that prohibited obtaining essentially 100 percent examination coverage of the required examination volumes or surface areas, but when this situation occurred 100 percent coverage of the accessible volumes or surface areas of each weld was obtained.

The licensee stated that to comply with the code required examination volumes or surface areas, the welds and their associated components would have to be physically modified and/or disassembled beyond their current design. According to the licensee, to replace these items with items of alternate configurations or materials to enhance examination coverage would require unique redesign and fabrication. Because these items are in the Class 1 boundaries and for the Class 1 items that form a part of the reactor coolant pressure boundary, their redesign and fabrication would be an extensive effort based on the limitations that exist. The licensee stated that it could not identify the improvements to the UT examination techniques that could achieve additional coverage. The licensee stated that radiography is impractical due to the amount of work being performed in the areas on a 24-hour basis when the welds are available for examination and other burden.

3.3.4 Proposed Alternative In lieu of satisfying essentially 100 percent examination coverage per the requirements of the ASME Code,Section XI, the licensee proposed the percentage of examination coverage achieved in the recent examination as shown in the relief request and is summarized in table 2 below. As part of the proposed alternative, the licensee stated that it will continue to perform periodic system pressure tests and associated VT-2 visual examinations in accordance with requirements of Examination Category B-P in Table IWB-2500-1 of the ASME Code,Section XI, for Class 1 pressure retaining welds and items each refueling outage. The licensee further stated that it will conduct required UT examinations to the maximum extent possible as required by the ASME Code,Section XI.

Table 2. Examination Coverage Achieved for Examination Category B-A and B-B Welds Component Weld Description Coverage Identification (ID) Obtained (Percent) 2-RPV-10442 (W13) Lower Shell to Lower Head Circumferential Weld 75.8 2-RPV-1442B (W02) Upper Shell @ 180°, Longitudinal Weld 72.2 2-RPV-3442A (W11) Lower Shell @ 60° Longitudinal Weld 83.6 2-RPV-3442B (W09) Lower Shell @ 180° Longitudinal Weld 83.6 2-RPV-3442C (W10) Lower Shell @ 300° Longitudinal weld 83.6 2-RPV-3443 (W14) Lower Head Disc to Peel Segments 27.9 Circumferential Weld

Component Weld Description Coverage Identification (ID) Obtained (Percent) 2-RPV-1443C (W31) Meridional Weld @ 30° Lower Head Weld 80.7 2-RPV-1443D (W32) Meridional Weld @ 90° Lower Head Weld 87.9 2-RPV-1443E (W33) Meridional Weld @ 150° Lower Head Weld 77.2 2-RPV-1443F (W34) Meridional Weld @ 210° Lower Head Weld 87.9 2-RPV-1443A (W35) Meridional Weld @ 270° Lower Head Weld 85.2 2-RPV-1443B (W36) Meridional Weld @ 330° Lower Head Weld 80.8 2-PZR-LONG D Pressurizer Longitudinal Weld Shell D 75 3.3.5 Basis for Use The licensee explained that it has performed the weld examinations to the maximum extent possible for each of the welds identified with limitations in Attachment 1 to the relief request.

The licensee stated that the Class 1 Examination Category B-A, reactor vessel welds, Examination Category B-B, and pressurizer longitudinal weld shell D within the scope of the relief request are all located inside the containment. The licensee explained that even though their examination did not meet the essentially 100 percent code required volume coverage requirement, there is instrumentation in place to assure that early detection of any reactor coolant system (RCS) pressure boundary leakage is identified. The licensee contended that this is accomplished by the leakage detection instrumentation inside the containment where the RCS leakage detection instrumentation is required to be operable. The instrumentation consists of monitoring of containment floor drain sump level to determine flow rate, containment cooler condensate flow rate increases, and airborne gaseous radioactivity increases. These instruments are used to quantify any unidentified leakage from the RCS and to meet Section 3.4.7.2 of the plant technical specifications surveillance requirements that have a limiting condition for operation (LCO) as shown below:

a. No pressure boundary leakage,

b. 1 gallons per minute (GPM) unidentified leakage,

c. 150 gallons per day primary-to-secondary leakage through any one steam generator,

d. 10 GPM identified leakage from the RCS 3.4 Piping Welds 3.4.1 Background In safety evaluation dated June 23, 2014 (ML14153A146), the NRC authorized the licensee to implement a risk informed (RI)-ISI program as an alternative to the ASME Code,Section XI, ISI requirements for Class 1 (i.e., Examination Category B-F and B-J) and Class 2 (i.e.,

Examination Category C-F-1 and C-F-2) piping welds in the fourth 10-year ISI interval of Salem, Unit 2. The licensee developed Salems RI-ISI program in accordance with the NRC-approved methodology of Electric Power Research Institute (EPRI) Topical Report (TR)-112657, Revision B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure, (ML013470102). In the first and second periods of the fourth 10-year ISI interval, Salems RI-ISI program had been implemented in accordance with NRC safety evaluation dated June 23, 2014. In the third period of the fourth 10-year ISI interval, the licensee has chosen to implement a RI-ISI program in accordance with ASME Code Case N-716-1, Alternative Classification and

Examination Requirements,Section XI. ASME Code Case N-716-1, has been incorporated by reference into 10 CFR 50.55a via inclusion in RG 1.147, Revision 20.

3.4.2 Component Affected ASME Code Class 1 piping welds governed by Salems RI-ISI program and shown in Table 1 of to S2-I4R-211, Revision 0 are affected. The licensee categorized each weld in accordance with Salems RI-ISI program (i.e., EPRI TR-112657, Revision B-A in the first and second periods of the fourth 10-year ISI interval, and ASME Code Case N-716-1 in the third period of the fourth 10-year ISI interval) as follows:

Weld ID 10-SJ-1211-16 in the safety injection (SI) system piping was categorized as Examination Category R-A, Item Numbers R1.11 (i.e., subject to thermal fatigue) and R1.16 (i.e., subject to intergranular or transgranular stress corrosion cracking (IGSCC or TGSCC)).

Weld ID 3-CV-1241-13, 3-CV-1241-14, 3-CV-1231-16, and 2-CV-1275-44 in the chemical and volume control system (CVCS) piping were categorized as Examination Category R-A, Item No. R1.11 (i.e., subject to thermal fatigue).

In Table 1 of Attachment 1 to S2-I4R-211, Revision 0, the licensee also provided for each weld, a description and its associated components, examination limitations, materials of construction, exposed temperature and pressure, and percent examination coverage obtained.

3.4.3 ASME Code Requirement The ASME Code requirements applicable to Class 1 piping welds originate in ASME Code,Section XI, Table IWB-2500-1. Alternative to the Code requirements is Salems RI-ISI program (i.e., NRC Safety Evaluation dated June 23, 2014, and ASME Code Case N-716-1). In both the ASME Code requirements and the Salems RI-ISI program, the piping welds in S2-I4R-211, Revision 0 are required to be volumetrically examined during each 10-year ISI interval, and the extent of examination is as follows:

Note (3) of Table 1 in ASME Code Case N-716-1 states, in part, that the extent of examination is essentially 100 percent (i.e., greater than 90 percent) coverage of the required examination volume or area when the entire examination volume or area cannot be examined due to interference by another component or part geometry.

ASME Code Case N-460 states, in part, that a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided that the reduction in coverage for that weld is less than 10 percent (i.e., acceptable coverage is essentially 100 percent or greater than 90 percent) when the entire examination volume or area cannot be examined due to interference by another component or part geometry.

3.4.4 Impracticality of Compliance The licensee stated that it was not possible to obtain greater than 90 percent of the ASME Code required examination volume of the subject pipe welds due to the geometric configuration and material type of the components which limited scan paths of the ultrasonic probes. The ultrasonic testing (UT) was a single-sided examination for all pipe welds. In Table 1 and Figures 1.14, 1.15 ,1.16, 1.17, and 1.18 of Attachment 1 to S2-I4R-211, Revision 0, the licensee

described and illustrated the limitations that prevented ultrasonic scanning of the pipe-to-valve, valve-to-elbow, elbow-to-branch connection, and pipe-to-branch connection welds.

The licensee stated that the burden caused by compliance includes major modification of plant components which include redesign and replacement of the welds and associated components.

3.4.5 Basis for Relief The licensee stated that it scanned each piping weld in the axial and circumferential directions using the applicable ultrasonic probe angles (i.e., refracted angles of 45, 60, and/or 70 degrees) and the ultrasonic wave modes (i.e., refracted shear and longitudinal waves) as described and shown in Figures 1.14, 1.15, 1.16, 1.17, and 1.18 of Attachment 1 to S2-I4R-211, Revision 0.

The UT was performed to the maximum extent possible utilizing personnel qualified and procedures demonstrated in accordance with ASME Code,Section XI, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems, with conditions in 10 CFR 50.55a. No unacceptable indications were identified in the volume examined.

The licensee stated that during the UT of these welds, although it extended the beam path into the volume on the opposite side of the weld centerline (far-side) to examine to the extent practical the other side of the weld, no credit was claimed for this percent coverage obtained on the far side and was not included in the percent coverage reported in Table 1 of Attachment 1 to S2-I4R-211, Revision 0. No unacceptable indications were identified in the far-side volume examined.

The licensee provided fatigue cumulative usage factor (CUF) at the 2-inch socket weld (Weld ID 2-CV-1275-44) location, between the 2-inch check valve 2CV76 and 4-inch pressurizer spray line, based on the 5 cycles of the controlling design transient, Inadvertent Auxiliary Spray, and the actual annual end of 2021 cycle count for inadvertent auxiliary spray transient which is only one. To further demonstrate that the actual CUF is much less than the design value, the licensee performed screening for limiting locations at the pressurizer spray line and determined that the pressurizer spray nozzle had a higher CUF than the 2-inch weld location. The licensee included the pressurizer spray nozzle location in its online fatigue monitoring program. As of December 31, 2021, the actual CUF at the pressurizer spray nozzle location is 0.087.

The licensee stated that the piping welds had been subjected to the ASME Code,Section XI, system leakage test and associated visual examination, VT-2, in the fourth 10-year ISI interval.

No through-wall leak was identified in any of the piping welds examined.

3.4.6 Proposed Alternative In Table 1 of Attachment 1 to S2-I4R-211, Revision 0, the licensee reported the percent coverage achieved for each weld examined. Table 3 below shows the percent coverage obtained.

Table 3. ASME Code Coverage for Examination Category R-A Welds System Component ID Coverage Obtained (Percent)

SI 10-SJ-1211-16 50 CVCS 3-CV-1241-13 45.5 CVCS 3-CV-1241-14 45.5 CVCS 3-CV-1231-16 48.8 CVCS 2-CV-1275-44 35 The licensee proposed the above alternative coverage in lieu of the required essentially 100 percent coverage.

3.5 NRC Staff Evaluation 3.5.1 Vessel Shell Welds The NRC staff reviewed the examination locations, coverage maps and calculations, and examination discussions in Attachment 1 to the relief request that are related to subject Examination Categories B-A and B-B Welds. The NRC staff evaluated the weld examination in terms of nondestructive examination (NDE) methods, examination coverage, examination results, defense-in-depth measures, and impracticality and burden of compliance.

NDE Methods The NRC staff noted that article I-2110, Reactor Vessels, of the ASME Code,Section XI, Appendix I, requires that ultrasonic examination procedures, equipment and personnel used to detect and size flaws for various RPV welds greater than 2 inches in thickness shall be qualified by performance demonstration in accordance with the ASME Code,Section XI, Appendix VIII.

The NRC staff determined that the licensee followed the ASME Code,Section XI, Appendix VIII, requirements. The licensee used the PDI requirements in accordance with the 2001 Edition of the ASME Code,Section XI, for the weld examinations as conditioned by 10 CFR 50.55a(b)(2)(xv) and 10 CFR 50.55a(b)(2)(xxiv). The NRC staff noted that the PDI program is an industry program for the inspections of nuclear plant welds and is maintained by EPRI. The NRC staff has approved the PDI program to satisfy the provisions of the ASME Code,Section XI, Appendix VIII as conditioned in 10 CFR 50.55a(b)(2)(xv) and 10 CFR 50.55a(b)(2)(xxiv). The NRC staff determined that the volumetric examination of the subject reactor vessel welds under Examination Category B-A has satisfied the ASME Code,Section XI, Appendix VIII, and therefore is acceptable.

With regard to the examination of the pressurizer weld, the NRC staff noted that article I-2120, Other Vessels, of the ASME Code,Section XI, Appendix I, requires that for vessels other than reactor greater than 2-inch thickness, which would include the pressurizer, ultrasonic examination shall be conducted in accordance with the ASME Code,Section V, Nondestructive Examination, Article 4, Ultrasonic Examination Methods for Welds, as supplemented by Table I-2000-1 of the ASME Code,Section XI. The NRC staff noted that the volumetric examination of pressurizer weld 2-PZR-LONG D was performed in accordance with Article 4 of the ASME Code,Section V, as supplemented by Table I-2000-1 as shown in Attachment 1 to the relief request. The NRC staff verified that the licensee followed the ASME Code,Section V,

Article 4, to examine the subject pressurizer weld; therefore, the examination of the subject pressurizer weld is acceptable.

Examination Coverage The examination coverage of the subject vessel welds ranges from a minimum of 27.9 percent to a maximum of 87.9 percent of the required weld volume as shown in Table 1 of Attachment 1 to the relief request. The minimum 27.9 percent examination coverage occurred on the lower head disc to the peel segments circumferential weld 2-RPV-3443. The NRC staff understands that this low examination coverage was caused by the inference of the incore instrumentation nozzles at the bottom of the reactor vessel. However, the NRC staff questioned whether as part of the ultrasonic examination of weld 2-RPV-3443, the licensee visually inspected the surface of the weld that was not ultrasonically examined to determine any surface degradation. By letter dated February 1, 2023, the licensee stated that it performed VT-3 visual examination on the inside surface of the entire lower reactor vessel head and incore instrument nozzle welds including weld 2-RPV-3443 as part of the ASME Code,Section XI, Category B-N examinations during refueling outage S2R25. The licensee also performed bare metal visual examination on the outside of the reactor pressure vessel lower head instrument nozzles and bottom head for evidence of leakage during refueling outage S2R25. The licensee stated that it performs this examination every other outage in accordance with 10 CFR 50.55a and ASME Code Case N-722-1, Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 MaterialsSection XI, Division 1. The licensee reported that no degraded conditions were identified during these examinations. The NRC staff finds that even though the examination coverage for weld 2-RPV-3443 is only 27.9 percent of required volume, the licensee has performed the required visual examination with respect to Examination Category B-N of reactor vessel internal components and ASME Code Case N-722-1 of the exterior of the bottom head of the reactor vessel. As such, the licensee has performed additional visual examinations to monitor the structural integrity of weld 2-RPV-3443. Therefore, the NRC staff finds that even though the examination coverage was only 27.9 percent, the licensee has performed additional visual examination to monitor its structural integrity.

As shown in Attachment 1 to the relief request, the NRC staff determined that for each of the vessel welds, the licensee performed examination following the required weld figures in the ASME Code,Section XI, IWB-2500. The NRC staff further determined that the licensee followed the corresponding Code Required Volume as shown on various figures in IWB-2500. Therefore, the NRC staff determined that the licensee has satisfied the provisions of the ultrasonic examination in terms of examining required weld volumes. Considering the cited limitations, the NRC staff determined that the coverage obtained for each of the vessel welds in Attachment 1 to the subject relief request is acceptable, considering various obstructions and limitations.

Examination Results to the relief request discusses the indications that were detected in vessel welds 2-RPV-10442, 2-RPV-1442B, 2-RPV-3442B, and 2-RPV-3442C. The licensee stated that these indications originated during the manufacturing process. The NRC staff questioned why these indications were not detected and reported in previous examinations. By letter dated February 1, 2023, the licensee explained that discrepancies in flaw detection would most likely be attributed to advancements in examination equipment and procedures during the last 20 years. The licensee stated that it used different technology, recording methodology, and procedure requirements in the last reactor vessel weld examination performed in 2002 than currently employed. According to the licensee, the primary contributor to the difference between

the number of indications recorded in the previous examination and the current examination is a difference in recording thresholds and use of phased array UT techniques. The licensee stated that these four vessel welds are scheduled for subsequent UT examination during the current 5th ISI interval in refueling outage S2R31. The NRC staff notes that the indications detected in the subject reactor vessel welds are within the acceptance standards of the ASME Code,Section XI. The NRC staff finds that the licensee has used advanced UT technology in the recent inspection such that small indications were detected. In addition, the licensee will perform subsequent UT examination of these four vessel welds in the 5th ISI. Therefore, the NRC staff finds that the licensee has appropriately monitored the subject vessel welds to ensure their structural integrity.

Defense-in-Depth Measures The Salem, Unit 2, has RCS leakage detection systems that can be considered as a defense-in-depth measure to monitor potential leakage from subject welds. The RCS leakage detection instrumentation inside the containment is regulated by the units technical specifications. The NRC staff determined that the instrumentation consists of monitoring of containment floor drain sump level to determine flow rate, containment cooler condensate flow rate increases, and airborne gaseous radioactivity increases. These instruments are used to quantify any unidentified leakage from the RCS and to meet the plant Technical Specifications surveillance requirements that have a limiting condition for operation (LCO) in Section 3.4.7.2.

The NRC staff noted that the limiting leakage rates for the LCO of Section 3.4.7.2 as discussed above are part of licensing basis and, therefore, are acceptable.

The NRC staff finds that the licensee will perform a system leakage test and associated VT-2 visual examination during plant startup in accordance with the ASME Code,Section XI, IWA-5000. The system leakage test is an acceptable defense-in-depth measure to detect any potential leakage from the subject vessel welds thereby monitors their structural integrity.

Impracticality and Burden of Compliance The NRC staff determined that essentially 100 percent examination coverage could not be achieved for the subject vessel welds because of various obstructions and weld locations. For example, the low examination coverage of the lower head disc to peel segments circumferential weld 2-RPV-3443 is caused by the obstruction of the incore instrumentation nozzles at the bottom head of the reactor vessel. The examination of several longitudinal welds of the RPV shell was hampered by the obstruction from the core support lugs or the nozzle boss. The examination coverage for the pressurizer weld 2-PZR-LONG D was limited because of the obstruction from the permanent support ring. The NRC staff recognizes that to achieve essentially 100 percent examination coverage, the licensee would have to modify the subject vessel welds or remove the obstructions which would be a burden to the licensee.

The NRC staff determined that based on the examination coverage obtained for the subject vessel welds, if significant service-induced degradation were occurring, there is reasonable assurance that evidence of degradation would be detected by the examination coverages achieved.

The NRC staff determined that the volumetric examinations for the subject vessel welds were performed to the maximum extent practical. The NRC staff further determined that obtaining the ASME Code-required examination volume for the subject vessel welds is impractical because of

the coverage limitations and that the modifications necessary to obtain the required coverage would impose a burden upon the licensee.

The NRC staff finds that there is reasonable assurance that the structural integrity of the subject vessel welds will be maintained because (1) evidence of safety significant service-induced degradation in subject vessel welds, if it were to occur, would be detected in the examination coverages achieved, (2) the licensee performed the required pressure testing, which included visual examination for evidence of leakage, in accordance with the ASME Code,Section XI, IWA-5000, and (3) Salem, Unit 2, has RCS leakage detection systems to monitor the subject vessel welds for leakage.

3.5.2 Piping Welds The NRC staff has evaluated the piping welds in S2-I4R-211, Revision 0 pursuant to 10 CFR 50.55a(g)(6)(i). The NRC staffs evaluation focused on: (1) whether a technical justification exists to support the determination that the ASME Code requirement is impractical, (2) that imposition of the Code-required inspections would result in a burden to the licensee without a commensurate increase in safety, and (3) that the licensees proposed alternative (accepting the reduced inspection coverage in this case) provides reasonable assurance of structural integrity and leak tightness of the subject welds. The NRC staff finds that if these three criteria are met that the requirements of 10 CFR 50.55a(g)(6)(i), (i.e., granting the requested relief will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility) will also be met.

As described in the submittal and shown in Table 1 and Figures 1.14, 1.15, 1.16, 1.17, and 1.18 of Attachment 1 to S2-I4R-211, Revision 0, the predominant limitation that prevented the licensees UT to achieve essentially 100 percent coverage of the ASME Code-required volume were design and configuration of the piping welds and associated components. As an example, pipe-to-valve, valve-to-elbow, elbow-to-branch, or pipe-to-branch connection configurations restricted the UT to single-sided scanning only. Radiographic testing (RT) was not practical due to the piping weld configurations and increased exposure of personnel to radiation when the water must be drained before the RT is performed. Therefore, the NRC staff finds that a technical justification exists to support the determination that achieving essentially 100 percent coverage is impractical.

The licensee proposed that making these piping welds accessible for inspection from both sides would require replacement or significant modification of the weld and associated components.

The NRC staff finds that replacing or reconfiguring the components is the only reasonable means to achieve the required coverage of these piping welds, and that replacement or reconfiguration of the components constitutes a burden on the licensee without a commensurate increase in safety.

The NRC staff considered whether the licensees proposed alternative provided reasonable assurance of structural integrity and leak tightness of the piping welds in table 3 based on:

(1) the examination coverage achieved, and (2) safety significance of unexamined areas and unachievable coverage (e.g., the presence or absence of known active degradation mechanisms and essentially 100 percent coverage achieved for similar welds in similar environments subject to similar degradation mechanisms).

In evaluating the licensees proposed alternative coverage, the NRC staff assessed whether the licensee obtained as much coverage as reasonably possible, and the methods used for achieving the reported coverage. From review of submittal, the NRC staff verified that:

The licensee examined the subject piping welds using the appropriate equipment, ultrasonic modes of propagation, probe angles, frequencies, and scanning directions to obtain maximum coverage.

The licensee calculated the coverage in a reasonable manner.

The personnel and UT procedures utilized for the examination were qualified as required by the ASME Code,Section XI, and 10 CFR 50.55a.

The coverage was limited by geometric configuration of the components that limited access to the required examination volume.

The licensee did not identify any unacceptable indications in the volume scanned by UT.

Therefore, the NRC staff found that the licensee made every effort to obtain as much coverage as reasonably possible with the ASME Code-required UT.

In addition to the coverage analysis described above, the NRC staff evaluated the safety significance of the unexamined areas of welds and unachievable coverage. From review of submittal and Figures 1.14, 1.15, 1.16, 1.17, and 1.18 of Attachment 1 to S2-I4R-211, Revision 0, the NRC staff verified that:

The licensees UT has covered, to the extent possible, the regions (i.e., the weld root and the heat affected zone (HAZ) of the base material near the inner diameter surface of the joint) that are typically susceptible to higher stresses and, therefore, potential degradation.

For the stainless steel welds, the coverage obtained was limited to the volume up to the weld centerline (near-side). The NRC staff notes that claiming coverage for the volume on the opposite side of the weld centerline (far-side) requires meeting the 10 CFR 50.55a(b)(2)(xv)(A)(2) far-side UT qualifications, which has not been demonstrated in any qualification attempts to date. During scanning, the licensees UT inspected the far-side volume by the Best Effort examination and did not identify any unacceptable indications in the volume scanned. The licensee did not take credit for the coverage achieved from the Best Effort examination.

For the socket weld, the licensees calculated fatigue CUF based on cycle count for inadvertent auxiliary spray transient remains below the ASME Code,Section III limit.

Therefore, the NRC staff determined that based on the coverage achieved by the qualified UT to the extent possible, the supplemental Best Effort examinations, and the examination of the weld root and its HAZ to the extent possible, it is reasonable to conclude that if significant service induced degradation had occurred, evidence of it would have been detected by the examinations that the licensee performed.

In this analysis, the NRC staff also found that, in addition to the required volumetric examinations, the subject piping welds have received the ASME Code,Section XI, Table IWB-2500-1, Examination Category B-P required system leakage test and associated visual examination, VT-2, every refueling outage. Despite reduced coverage of the required examination volume, the NRC staff finds that the VT-2 inspection will provide additional assurance that any pattern of degradation, if it were to occur, would be detected and the licensee will take appropriate correction actions.

4.0 CONCLUSION

The NRC staff has determined that it is impractical to satisfy the ASME Code-required greater than 90 percent examination coverage for the subject welds due to material and component configuration. The NRC staff finds that the proposed examination coverage provides reasonable assurance of structural integrity or leak tightness of the subject component. The NRC staff determines that granting the relief request pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Therefore, the NRC staff grants the use of this relief request (i.e., S2-I4R-211, Revision 0) for the duration of the fourth 10-year ISI interval at Salem, Unit 2, which began on November 27, 2013, and ended on December 31, 2021.

All other ASME Code,Section XI requirements for which relief has not been specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

Principal Contributors: Eric Palmer, NRR Ali Rezai, NRR John Tsao, NRR Date: June 5, 2023

ML23139A147 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DNRL/NVIB/BC NAME JKim KEntz ABuford DATE 5/23/2023 5/22/2023 5/4/2023 OFFICE NRR/DNRL/NPHP/BC NRR/DORL/LPL1/BC NAME MMitchell HGonzález DATE 5/4/2023 6/5/23