ML22130A791

Issuance of Relief Request No. S1-I4R-210 Fourth Inservice Inspection Interval Limited Examinations

ML22130A791
Person / Time
Site:	Salem
Issue date:	05/24/2022
From:	James Danna Plant Licensing Branch 1
To:	Carr E Public Service Enterprise Group
Kim J
References
EPID L-2021-LLR-0085
Download: ML22130A791 (14)

Text

May 24, 2022 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 NO. 1 - ISSUANCE OF RELIEF REQUEST NO. S1-I4R-210 RE: FOURTH INSERVICE INSPECTION INTERVAL LIMITED EXAMINATIONS (EPID L-2021-LLR-0085)

Dear Mr. Carr:

By letter dated November 10, 2021 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML21314A579), as supplemented by letter dated March 21, 2022 (ML22080A177), 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 Nuclear Generating Station Unit No. 1 (Salem Unit 1).

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.55a(g)(5)(iii), the licensee requested relief on the basis that the required examination coverage is impractical due to physical obstructions and limitations imposed by design, geometry, and materials of construction of the subject components.

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject request and concluded that it is impractical for the licensee to comply with the requirements of the ASME Code,Section XI in the examination of subject welds. The NRC staff finds that requiring the licensee to perform a design modification to obtain ASME Code-required coverage would result in a significant burden. The NRC staff further determines that the licensee-obtained examination coverage in combination with required system leakage tests and leakage detection capability of the reactor coolant system leakage detection systems will provide reasonable assurance of the structural integrity and leak tightness of the subject welds. The NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Accordingly, the NRC staff determines that granting relief 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.

Therefore, the NRC grants the use of Relief Request S1-I4R-210 at Salem Unit 1 for the fourth inservice inspection interval, which ended on December 31, 2020.

All other ASME Code,Section XI, requirements for which an alternative was not specifically requested and authorized 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 James James G. G. Danna Date: 2022.05.24 Danna 14:55:43 -04'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-272

Enclosure:

Safety Evaluation cc: Listserv

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO. S1-I4R-210 REGARDING ALTERNATIVE EXAMINATION FOR WELDS PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 DOCKET NO. 50-272

1.0 INTRODUCTION

By letter dated November 10, 2021 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML21314A579), as supplemented by letter dated March 21, 2022 (ML22080A177), 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, Rules for Inservice Inspection of Nuclear Power Plant Components, at Salem Nuclear Generating Station, Unit No. 1 (Salem Unit 1).

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.55a(g)(5)(iii), ISI [Inservice Inspection] program update: Notification of impractical ISI Code requirements, the licensee requested relief on the basis that the required examination coverage is impractical due to physical obstructions and limitations imposed by design, geometry, and materials of construction of the subject components.

2.0 REGULATORY EVALUATION

Pursuant to 10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants, 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, 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), Use and conditions on the use of standards, 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein.

Enclosure

The regulation in 10 CFR 50.55a(b)(2)(xv)(A), Specimen set and qualification: First provision, states that, when applying Supplement 2 (Qualification Requirements for Wrought Austenitic Piping Welds) to ASME Code,Section XI, 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), Ferritic and stainless steel piping examinations:

Second provision, states 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 this paragraph and paragraph [10 CFR 50.55a(b)(2)(xv)(A)].

The regulation in 10 CFR 50.55a(g)(5)(iii) states that, If the licensee has determined that conformance with a code requirement is impractical for its facility, the licensee shall notify the NRC [U.S. Nuclear Regulatory Commission] and submit, as specified in § 50.4, information to support the determinations. Determinations of impracticality in accordance with this section 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 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), Impractical ISI requirements: Granting of relief, states that, The Commission will evaluate determinations under paragraph (g)(5) of

[10 CFR 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 ASME Code Component(s) Affected The affected components are welds in ASME Code Classes 1 and 2 under the examination categories of B-A, B-B, C-B, and R-A. The item numbers are B1.11, B1.12, B1.21, B1.22, B2.11, B2.40, C2.21, R1.11, R1.16, and R1.20. The ASME Code Class 1 and 2 welds with limited examinations for relief are for the fourth 10-year ISI interval. The individual welds are identified in Table 1 of Attachment 1 to the relief request.

3.2 Applicable Code Edition and Addenda The fourth 10-year ISI interval at Salem Unit 1 was based on the ASME Code,Section XI, 2004 Edition, as modified by 10 CFR 50.55a. The licensee used the requirements of the ASME Code,Section XI, Appendix VIII and the Performance Demonstration Initiative (PDI) program in accordance with the 2001 Edition of the ASME Code,Section XI, for the limited examinations as conditioned by 10 CFR 50.55a(b)(2)(xv),Section XI condition: Appendix VIII specimen set and qualification requirements, and 10 CFR 50.55a(b)(2)(xxiv),Section XI condition: Incorporation of the performance demonstration initiative and addition of ultrasonic examination criteria.

3.3 Applicable Code Requirements Examination Item No Weld Examination Coverage Requirements Category B-A B1.11 To include essentially 100% examination of the reactor vessel circumferential shell welds.

B-A B1.12 To include essentially 100% examination of the reactor vessel longitudinal shell welds.

B-A B1.21 To include essentially 100% of the accessible length of all the reactor vessel circumferential head welds.

B-A B1.22 To include essentially 100% of the accessible length of all the reactor vessel meridional head welds.

B-B B2.11 To include essentially 100% of the pressurizer shell-to-head circumferential welds.

B-B B2.40 To include essentially 100% of the steam generator tubesheet-to-head welds.

C-B C2.21 To include the examination volume of the Pressure Vessel Nozzle Inside Radius Section as depicted in the applicable figure shown in Figures IWC-2500-4(a), (b), or (d).

R-A R1.11 To include essentially 100% of the examination location potentially subject to thermal fatigue.

Examination Item No Weld Examination Coverage Requirements Category R-A R1.11/1.16 To include essentially 100% of the examination location potentially subject to thermal fatigue and intergranular stress corrosion cracking.

R-A R1.16 To include essentially 100% of the examination locations potentially subject to intergranular stress corrosion cracking.

R-A R1.20 To include essentially 100% of the examination location with no degradation mechanism.

As stated in ASME Code Case N-460, Alternative Examination Coverage for Class 1 and Class 2 WeldsSection XI, Division 1, as approved in Regulatory Guide (RG) 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI Division 1, Revision 19, dated October 2019 (ML19138A244), essentially 100 percent 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 used Code Case N-460 for the required coverage associated with the welds in this relief request.

The licensee examined the welds covered in the risk-informed inservice inspection (RI-ISI) programs in accordance with Electric Power Research Institute Topical Report (TR)-112657, Revision Final Report, Revised Risk-Informed Inservice Inspection Evaluation Procedure, April 1999 (ML20205N012), which was supplemented by ASME Code Case N-578-1, Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B,Section XI, Division 1. The licensees examination included the piping weld (elements) selected for examination under Category R-A. The licensee stated that the use of these documents was based on an NRC-approved Relief Request S1-I4R-105 at Salem Unit 1 as shown in the NRC safety evaluation dated August 17, 2011 (ML112140424). During the fourth interval, third period, the implementation of the RI-ISI program was based on ASME Code Case N-716-1 Alternative Classification and Examination Requirements,Section XI, Division 1, as approved in RG 1.147.

3.4 Impracticality and Burden of Compliance Table 1 of Attachment 1 of the relief request submittal dated November 10, 2021, presents the percentage of coverage of the required code volume or surface area. The licensee stated, in part, in its relief request submittal, that, although the design of the plants has provided access for examinations to the extent practical, component design configurations resulting in examination limitations such as those from support interference, geometric configurations of welds and materials such as fittings or valve bodies made of cast stainless steel may not allow the full required examination volume or surface area coverage with the latest techniques available.

The licensee also stated that a valve-to-pipe weld cannot be examined from the valve side of the weld and where a plant modification would be needed to provide this coverage.

The licensee also stated, in part, in its relief request submittal that it has examined the subject welds to the maximum extent possible and, that, obtaining essentially 100% coverage for the welds listed in this request for relief, the welds and their associated components would have to be physically

modified and/or disassembled beyond their current design. Overall components and fittings associated with the [subject] welds listed in this request are constructed of standard design items and materials meeting typical national standards that specify required configurations and dimensions. 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 and 2 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 further stated, in part, in its relief request submittal 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. Using radiography would result in numerous work-related stoppages and increased exposure due to the shutdown and startup of other work in the areas. The water may need to be drained from systems or components where radiography is performed, which increases the radiation dose rates over a much broader area than the weld being examined. There is significant impracticality associated with the performance of weld or area modifications or the use of radiography in order to increase the examination coverage.

3.5 Proposed Alternative In lieu of achieving essentially 100 percent examination coverage, the licensee proposed the examination coverage achieved in the field for the subject welds as shown in Table 1 of to the relief request.

1) Perform periodic system pressure tests and VT-2 visual examinations will continue to be performed in accordance with ASME Code Section Xl, Examination Category B-P, for Class 1 pressure retaining welds and items each refueling outage and Examination Category C-H for Class 2 pressure retaining welds and items each inspection period of Table IWB-2500-1 and Table IWC-2500-1, respectively.

2) Conduct required examinations to the maximum extent possible as required by ASME Section XI or the RI-ISI Programs.

3.6 Basis for Use The licensee stated in its relief request submittal that the regulation in 10 CFR 50.55a(g)(4) requires that throughout the service life of a nuclear power facility, components which are classified as ASME Code Class 1, Class 2 and Class 3 must meet the requirements set forth in the ASME Code to the extent practical within the limitations of design, geometry and materials of construction of the welds and items described in Attachment 1 to the relief request. When a component is found to have conditions, which limit the required examination volume or surface area, Salem Unit 1 is required to submit this information to the enforcement and regulatory authorities having jurisdiction at the plant site. This request for relief has been written to address areas where these types of conditions exist and where the required amount of coverage was reduced below the minimum acceptable. Salem Unit 1 has performed

the weld examinations listed in this request to the maximum extent possible for each of the welds identified with limitations in Attachment 1 to the relief request.

The licensee further stated in its relief request submittal that, The Class 1 Examination Category B-A, Head-to-Flange Weld, the Class 1 Examination Category B-B, Pressurizer Shell J to Upper Head, Pressurizer Longitudinal Weld Shell D, and Steam Generator Lower Head-to-Tubesheet Weld, the Class 1 Risk-Informed Piping Welds, and the Class 2 Examination Category C-B, Steam Generator Feedwater Nozzle-to-Vessel Weld within the scope of this request are all located inside the containment. Even though their examination did not meet the essentially 100% 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. 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 particulate and gaseous radioactivity increases. These instruments are used to quantify any unidentified leakage from the RCS and to meet the [Salem]

Unit 1 Technical Specifications (TS) Surveillance Requirements that have a Limiting Condition for Operation (LCO) in TS 3.4.6.2 stating that RCS Operational Leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM [gallons per minute] UNIDENTIFIED LEAKAGE,

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

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System.

3.7 Duration of Proposed Alternative This relief request is for Salem Unit 1, fourth 10-year ISI interval, which began on May 20, 2011, and ended on December 31, 2020.

3.8 NRC Staff Evaluation The NRC staff reviewed the examination locations, coverage maps and calculations, and examination discussions in Attachment 1 of the relief request. The NRC staff evaluated the weld examination in terms of nondestructive examination (NDE) methods, examination coverage, RI-ISI welds, examination results, defense-in-depth measures, impracticality and burden of compliance.

NDE Methods to the relief request discusses the NDE methods used in the examination of the subject welds. For the welds in the reactor vessel, the NRC staff verified that the licensee used the ASME Code,Section XI, Appendix VIII, Supplements 4 and 6, in accordance with the 2001 Edition of the ASME Code,Section XI, as conditioned by 10 CFR 50.55a(b)(2)(xv) and 10 CFR 50.55a(b)(2)(xxiv). For welds in the pressurizer and steam generator, the licensee used the PDI program to qualify personnel in addition to follow the requirements of Appendix VIII,

Supplements 4 and 6. The licensee indicated that although pressurizer and steam generator components are not applicable to Appendix VIII, the component, materials, sizes and shapes examined were within the scope of the qualified examination procedure. In addition, the licensee used the ASME Code,Section V, Nondestructive Examination, Article 4, Ultrasonic Examination Methods for Welds, to examine steam generator nozzle-to-shell weld, 16-BFN-2111-1.

For piping welds, the licensee performed the examination in accordance with its request for Alternative S1-I4R-105 as approved by the NRC on August 17, 2011. The licensee examined the pipe welds based on the requirements of Appendix VIII with qualified personnel, procedures and equipment to the 2001 Edition of Appendix VIII as conditioned by 10 CFR 50.55a(b)(2)(xv).

The NRC staff finds that the examination of subject welds has satisfied (1) the requirements of Appendix VIII with qualified personnel, procedures and equipment to the 2001 Edition of Appendix VIII as conditioned by 10 CFR 50.55a(b)(2)(xv), (2) the PDI program, and (3) the ASME Code,Section V.

Examination Coverage As shown in Attachment 1 to the relief request, the NRC staff determined that for each of the vessel and piping welds, the licensee followed the appropriate weld figures in the corresponding ASME Code,Section XI, Figures IWB-2500-x and IWC-2500-x to examine the subject welds.

The NRC staff further determined that the licensee examined the corresponding Code-required volume and Code-required surface area as shown in diagrams in various Figures IWB-2500-X and IWC-2500-X except that the examination of the subject welds could not satisfy the essentially 100 percent coverage. The NRC staff focused on the following two welds that achieved significant low percentage of examination coverage.

The NRC staff noted that Section 1.3 of Attachment 1 states that the examination coverage achieved for the circumferential weld of the reactor vessel lower head disc to peel segments, 1-RPV-4043, was 27.9 percent because the examination was limited due to the proximity of the reactor vessel incore nozzles. The NRC staff questioned whether the licensee could have achieved a higher percentage of coverage. By letter dated March 21, 2022, the licensee explained that all examination scan areas were modeled in 3-D space to achieve maximum coverage with two different ultrasonic testing (UT) head configurations. The licensee considered robotic manipulator limitations and physical obstructions during the scan plan generation process. The licensee further explained that the coverage map in Figure 1.3-3 of Attachment 1 is a 2-D plan view, which does not show all incore penetrations or the ability of the robotic manipulator to reach these areas successfully without damage to the robotic manipulator or plant equipment. During outage preparations, the scan limitations are determined by using a 3-D model to run the planned scans in a simulation mode. This allows the scan areas which are obtainable to be determined and takes into account the location of all incore nozzles, the transducer UT head configuration, and any manipulator limitations which may exist. The licensee stated that to increase the coverage, it did take credit for single side coverage when applicable and it performed the examination to the maximum extent possible. The NRC staff determined that based on the map of incore penetrations in Figure 1.3-3 of Attachment 1 and the licensees explanation, the examination coverage is significantly restricted. If the licensee were to achieve essentially 100 percent coverage, the licensee would need to remove the incore penetrations which is impractical. The NRC staff finds that the existing 27.9 percent coverage is acceptable based on the impracticality determination per 10 CFR 50.55a(g)(5)(iii).

The NRC staff noted that Section 1.10 of Attachment 1 states that the examination coverage achieved for circumferential weld 1-PZR-21 of pressurizer shell J to upper head is 42.15 percent because the weld examination was limited due to the proximity of insulation support straps, permanent vessel support ring and welded pads. By letter dated March 21, 2022, the licensee stated that no additional coverage can be credited or achieved beyond the 42.15 percent examination coverage. However, the licensee explained that as a comparison, it also examined the only other similar weld, 1-PZR-1, which is the lower head to shell A circumferential weld during the fourth ISI 10-year interval. For weld 1-PZR-1, the licensee achieved 96.8 percent examination coverage and did not detect any recordable indications. The licensee stated that the removal of the insulation support straps to achieve higher examination coverage for weld 1-PZR-21 would require cutting and welding in elevated dose area with limited access to perform work. This would be a personnel safety hazard. The licensee further stated that the removal of the permanent vessel support ring would require cutting and welding to the pressurizer resulting in a considerable hardship and possibly inducing additional stresses to the vessel. The NRC staff determined that based on the licensees description, it is impractical to remove the insulation support straps and vessel support ring at the pressurizer to achieve higher examination coverage. Therefore, the existing 42.15 percent examination coverage is acceptable based on the impracticality determination per 10 CFR 50.55a(g)(5)(iii).

The NRC staff determined that the coverage calculation for each weld in Attachment 1 to the subject relief request acceptable based on the impracticality determination per 10 CFR 50.55a(g)(5)(iii).

RI-ISI Welds For the piping welds covered under the RI-ISI program such as those under Examination Category R-A and Item Nos. R1.11, R1.60, and R1.20, the licensee performed a probabilistic risk assessment to determine the changes in core damage frequency (CDF) and large early release frequency (LERF) because of limited coverage. The NRC staff determined that to quantify the effect on delta risk for the RI-ISI welds, the delta risk was adjusted by not crediting the RI-ISI weld with limited examination coverage. The licensee calculated the delta risk considering the case where the weld is credited (included in the analysis) and not credited (removed from the analysis). The NRC staff determined that for all the RI-ISI welds covered in the relief request, the changes in CDF and LERF values are negligible and are below the acceptable limits of 1E-07 for CDF and 1E-08 for LERF. Therefore, the NRC staff determined that a limited examination coverage for the RI-ISI welds is acceptable in terms of risk to the plant safety.

Examination Results to the relief request shows that most of welds covered under the relief request have no recordable indications except the following welds.

The NRC staff noted that Section 1.2 of Attachment 1 discusses a recordable subsurface indication in the upper shell at 7 degrees (°), longitudinal seam weld 1-RPV-1042B. The licensee evaluated this indication as shown in Tables 1.2-3 and 1.2-4 of Attachment 1. By letter dated March 21, 2022, the licensee stated that the indication in W02 (1-RPV-1042B) runs parallel to the axis of the weld. The length of the indication detected in W02 (1-RPV-1042B) is 0.85 inches with a depth (flaw height) of 0.13 inches. The distance between the indication and the nearest outside unclad surface is 0.35 inches. The licensee explained that the flaw is characteristic of a slag inclusion from the welding process during fabrication, and the indication

was found for the first time during this examination. The licensee stated that the examination performed in 2001 (the third 10-year ISI interval) was different in technology, recording methodology, and procedure requirements. The main 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 ultrasonic testing techniques. The licensee stated that weld 1-RPV-1042B is currently scheduled to be examined during the fifth ISI 10-year interval refueling outage S1R33 (fall 2029).

The NRC staff noted that Section 1.4 of Attachment 1 discusses a subsurface indication in lower head meridional weld 1-RPV-1043A at 270 degrees. The licensee stated that this flaw is characteristic of slag inclusion from the welding process during fabrication. The licensee evaluated the indication as shown in in Table 1.4-3 of Attachment 1. By letter dated March 21, 2022, the licensee stated that the indication is parallel to the weld 1-RPV-1043A axis. The licensee further stated that it detected the indication for the first time during the examination and, therefore, there was no comparison to previous results. The licensee indicated that meridional weld 1-RPV-1043A is currently scheduled to be examined during the fifth ISI 10-year interval refueling outage S1R33 (fall 2029).

The NRC staff noted that Section 1.8 of Attachment 1 discusses two subsurface indications in lower head meridional weld 1-RPV-1043E, at 150° during this examination. The licensee evaluated each recordable indication as shown in Tables 1.8-3, 1.8-4, and 1.8-5 of . By letter dated March 21, 2022, the licensee stated that These are two separate indications. Both are located circumferentially at approximately the same degree location; however, they are separated in the vertical dimension by approximately 16 degrees. The licensee noted that since the head is spherical, the vertical and horizontal dimensions are both specified in degrees, Alpha and Theta, respectively. The licensee further stated that as a result, these two indications are separated by approximately 16° (approximately 25 ]inches]) in the vertical (Alpha) dimension. The licensee explained that the flaws are located within the weld and are indicative of fabrication flaws typical of small slag inclusions. The licensee stated that these indications were detected for the first time during this examination, and there was no comparison to previous results. The licensee further stated that meridional weld 1-RPV-1043E is currently scheduled to be examined during the fifth ISI 10-year interval refueling outage S1R33 (fall 2029).

The NRC staff noted that Section 1.15 of Attachment 1 discusses a recordable indication in Weld 4-PS-1131-29 Safe-End-to-Nozzle. The licensee evaluated this indication in Tables 1.15-1, 1.15-2, 1.15-3 and Figure 1.15-2 of Attachment 1. The licensee explained that this indication is not a true flaw based on the following:

1. This weld does not contain susceptible material to PWSCC.

2. The indication does not contain any following typical cracking characteristics. Specifically, typical flaws will provide substantial and unique echo-dynamic travel through the time base. Flaw areas of unique amplitude peaks will be observed throughout the indication length. Flaw indications will show evidence of flaw tip signals.

3. Past radiographs were reviewed and identified no areas containing fabrication flaws.

4. Final NDE evaluation determined that the indication was located in the base material and the results were found acceptable for continued service per ASME Code Section XI acceptance criteria identified in IWB-3500, Table IWB-3514-2.

The NRC staff finds that (1) the indications in welds 1-RPV-1043A, 1-RPV-1042B and 1-RPV-1043E are caused by the welding process rather than by service conditions and they will be reexamined in the fifth ISI interval in fall 2029 to determine the condition of the indications, (2) the licensee has appropriately dispositioned the indication in weld 4-PS-1131-29, (3) the licensee has appropriately evaluated the indications to show the acceptance of these indications for continued operation, and (4) the licensee used appropriate phased array ultrasonic testing technique in its examinations. Therefore, the NRC staff finds that the licensee has appropriately dispositioned these indications.

Defense-in-Depth Measures The NRC staff noted that Salem Unit 1 has RCS leakage detection systems to monitor potential leakage from subject welds. The RCS leakage detection instrumentation inside the containment is controlled by the units Technical Specifications. These leakage detection instruments are used to quantify any unidentified leakage from the RCS and to meet the Salem Unit 1 TSs surveillance requirements that have a LCO in TS 3.4.6.2.

The NRC staff further noted that the ASME Code,Section XI, IWA-5000 requires licensees to perform a system leakage test during plant startup. The NRC staff determined that the licensee did not ask relief from the IWA-5000 requirement in the subject relief request. As such, the licensee will perform the system leakage test in accordance with the ASME Code,Section XI, IWA-5000 to detect any potential leakage from the subject welds and is, therefore, acceptable.

Impracticality and Burden of Compliance As shown in the weld configurations and UT beam coverage maps in Attachment 1 to the relief request, the NRC staff determined that essentially 100 percent examination coverage could not be achieved for the subject welds because of various obstructions and weld locations. For example, the permanent insulation support ring blocked the accessibility for the examination of the pressurizer weld. The incore penetrations at the bottom of the reactor vessel limited examination coverage of the meridian welds in the lower reactor vessel head. For many piping welds, the coverage was limited because the UT examination could only be performed from one side of the weld instead of both sides (i.e., a single side examination). Therefore, the NRC staff recognizes that to achieve essentially 100 percent examination coverage, the licensee would have to modify the welds or remove the obstructions which would be a burden to the licensee.

Summary The NRC staff concluded that based on the examination coverage obtained for the subject 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 welds were performed to the maximum extent practical. The NRC staff further determined that obtaining the ASME Code-required examination volume for the subject 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 concluded that there is reasonable assurance that the structural integrity of the subject welds will be maintained because (1) evidence of service-induced degradation in the subject welds, if it were to occur, would be detected in the examination coverages achieved,

(2) the licensee will continue to perform the required pressure testing, which includes visual examination for evidence of leakage, in accordance with the ASME Code,Section XI, IWA-5000, (3) the licensees probabilistic risk assessment has shown that the risk of not achieving essentially 100 percent examination coverage for the RI-ISI welds is insignificant to the plant safety, and (4) the licensee has RCS leakage detection systems in the containment and the plant technical specifications prescribe acceptance criteria for leak rates to monitor the subject welds for leakage.

4.0 CONCLUSION

As set forth above, the NRC staff determines that it is impractical for the licensee to comply with the requirements of the ASME Code,Section XI in the examination of subject welds. The NRC staff finds that requiring the licensee to perform a design modification to obtain ASME Code-required coverage would result in a significant burden. The NRC staff further determines that the licensee obtained examination coverage in combination of required system leakage tests and leakage detection capability of the RCS leakage detection systems will provide reasonable assurance of the structural integrity and leak tightness of the subject welds.

The NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Accordingly, the NRC staff determines that granting relief 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. Therefore, the NRC grants the use of Relief Request S1-I4R-210 at Salem Unit 1 for the fourth ISI interval, which ended on December 31, 2020.

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

Principal Contributors: Karen Sida, NRR Bart Fu, NRR John Tsao, NRR Date: May 24, 2022

ML22130A791 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DNRL/NVIB/BC NAME JKim KEntz (PBlechman for) ABuford (JTsao for)

DATE 5/10/2022 5/11/2022 5/02/2022 OFFICE NRR/DNRL/NPHP/BC NRR/DORL/LPL1/BC NAME MMitchell JDanna DATE 5/02/2022 5/24/2022