View source for LR-N21-0066, Submittal of Relief Request Associated with the Fourth Inservice Inspection (ISI) Interval Limited Examinations

{{Adams
| number = ML21314A579
| issue date = 11/10/2021
| title = Submittal of Relief Request Associated with the Fourth Inservice Inspection (ISI) Interval Limited Examinations
| author name = Richard M
| author affiliation = PSEG Nuclear, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000272
| license number = DPR-070
| contact person = 
| case reference number = LR-N21-0066
| document type = Letter type:LR, Request for Code Relief or Alternative
| page count = 115
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:PSEG Nuc lear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG NuclearLLC 10 CFR 50.55a(g)(5)(iii)
LR-N21-0066 November 10, 2021 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station Unit 1 Renewed Facility Operating License Nos. DPR-70 NRC Docket No. 50-272

==Subject:==
Submittal of Relief Request Associated with the Fourth Inservice Inspection (ISI)
Interval Limited Examinations In accordance with 10 CFR 50.55a, Codes and standards, paragraph (g)(5)(iii), PSEG Nuclear LLC (PSEG) hereby requests NRC approval of the enclosed request for the fourth 10-year ISI interval for the Salem Generating Station Unit 1 which ended on December 31, 2020. The relief request addresses limitations for examinations performed in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components, for Class 1 and 2 components.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this matter, please contact Brian Thomas at 856-339-2022.
Sincerely, Digitally signed by Montgome Montgomery, Richard ry, Richard Date: 2021.11.10 16:42:22 -05'00' Richard Montgomery Manager - Licensing PSEG Nuclear LLC Enclosure 1: Salem Generating Station Unit 1 American Society of Mechanical Engineers (ASME) Section XI 10 CFR 50.55a(g)(5)(iii) Request for Relief Number S1-I4R-210

LR-N21-0066                                              10 CFR 50.55a(g)(5)(iii)
Page 2 November 10, 2021 cc:  Mr. D. Lew, Administrator, Region I, NRC Mr. J. Kim, Project Manager, NRC NRC Senior Resident Inspector, Salem Ms. A Pfaff, Manager NJBNE PSEG Corporate Commitment Tracking Coordinator Site Commitment Tracking Coordinator

LR-N21-0066                                                    10 CFR 50.55a(g)(5)(iii)
Enclosure 1 Salem Generating Station Unit 1 American Society of Mechanical Engineers (ASME) Section XI 10 CFR 50.55a(g)(5)(iii) Request for Relief Number S1-I4R-210

Salem Generating Station Unit 1 American Society of Mechanical Engineers (ASME) Section XI 10 CFR 50.55a Request for Relief Number S1-I4R-210, Revision 0 In Accordance with 10 CFR 50.55a(g)(5)(iii)
                               --Inservice Inspection Impracticality--
: 1.       ASME Code Component(s) Affected The Salem Generating Station (SGS) Unit 1 Class 1 and 2 welds with limited examinations that are included in this request for relief are for the Fourth Ten-Year Inservice Inspection Interval.
The content of this request includes the insights gained from guidance provided in Reference 1 and the following Code Classes, Examination Categories, and Item Numbers apply.
Code Classes:                   1 and 2 Examination Categories:         B-A, B-B, C-B, and R-A Item Numbers:                   B1.11, B1.12, B1.21, B1.22, B2.11, B2.40, C2.21, R1.11, R1.16, R1.20
: 2.       Applicable Code Edition and Addenda The Fourth 10-Year Inservice Inspection (ISI) Interval at Salem Unit 1 was based on the ASME Boiler and Pressure Vessel Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2004 Edition, (Reference 3) as modified by 10 CFR 50.55a. The Appendix VIII requirements and use of the Performance Demonstration Initiative (PDI) requirements at Salem Unit 1 were in accordance with the 2001 Edition of Section XI, (Reference 4) for the limited examinations contained in this request as conditioned by 10 CFR 50.55a(b)(2)(xv) and 10 CFR 50.55a(b)(2)(xxiv).
The Salem Unit 1 Fourth 10-Year ISI Interval ended on December 31, 2020.
Page 2 of 113
: 3. Applicable Code Requirements Exam         Item No.         Class 1 Weld Examination Coverage Requirements Cat.
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 Exam         Item No.         Class 2 Weld Examination Coverage Requirements Cat.
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).
Exam         Item No. Class 1 and Class 2 Piping Welds / Risk-Informed Inservice Cat.                             Inspection Program Coverage Requirements R-A           R1.11    To include essentially 100% of the examination location potentially subject to thermal fatigue.
R-A         R1.11/16   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 previously defined in 10 CFR 50.55a(g)(6)(ii)(A)(2), now removed, and as stated in ASME Code Case N-460 (Reference 5) as approved in Regulatory Guide 1.147, Revision 19 (Reference 6), essentially 100% equates to more than 90% 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. Code Case N-460 was invoked for the required coverage associated with the welds in this request.
Limited Class 1 and Class 2 Welds/Risked-Informed Inservice Inspection Programs Class 1 and Class 2 piping welds selected for examination during the fourth interval under the Risk-Informed Inservice Inspection (RI-ISI) Programs used at Salem Unit 1 were examined during the first and second period in accordance with EPRI Topical Report TR-112657, Rev. B-A methodology (Reference 7) which was supplemented by ASME Code Case N-578-1 (Reference 8) and included the piping weld (elements) selected for examination under Category R-A. The use of these documents was based on a request for alternative S1-I4R-105 Page 3 of 113

(Reference 9). During the fourth interval, third period the implementation of the RI-ISI Program was based on Code Case N-716-1(Reference 10) as approved in Regulatory Guide 1.147.
RI-ISI Program Limited Examination Evaluations When limited piping weld examinations are identified under EPRI TR-112657 Rev. B-A supplemented by Code Case N-578-1 RI-ISI Programs, an evaluation is performed per Note 3 of Table 1 of Code Case N-578-1. Since Salem Unit 1 implemented Code Case N-716-1 in the third period, an evaluation required by Note 3 of Table 1 of Code Case N-716-1 for all limited examinations in the RI-ISI Program for the entire fourth interval is provided below. Table 1 Note 3 of Code Case N-578-1 and Table 1 Note 3 of Code Case N-716-1 are similar and states in part .When the required examination volume or area cannot be examined due to interference by another component or part geometry, limited examination shall be evaluated for acceptability.
Acceptance of limited examinations or volumes shall not invalidate the results of the change-in-risk evaluation. Salem Unit 1 has evaluated all of the limited examinations for the fourth interval.
Evaluation of RI-ISI Limited Examinations N-716-1, Table 1 Note 3 requires when the required examination volume or area cannot be examined due to interference by another component or part geometry, limited examinations shall be evaluated for acceptability. Acceptance of limited examinations or volumes shall not invalidate the results of the change-in-risk evaluation. Areas with acceptable limited examination and their bases, shall be documented.
PSEG has reviewed each instance of limited coverage and took the appropriate steps (e.g.,
relief requests) consistent with its impact on the basis of the N-716 application. This is done using the delta risk values. N-716-1 change in risk acceptance criteria for each system is 1E-7 Core Damage Frequency (CDF) and 1E-8 Large Early Release Frequency (LERF) and the total increase should be less than 1E-6 (CDF) and 1E-7 (LERF). The With POD Credit column is the official result that must meet these criteria (Note: POD - probability of detection).
Below is a summary of the cumulative effect of the limited examinations.
This Request identifies six Unit 1 piping welds with limited examination coverage (90% or less). Welds 2-CV-1175-36, 14-PS-1131-2, 4-PS-1131-29, 10-SJ-1111-17, 6-SJ-1131-17, and 6-SJ-1131-20 are addressed below, by system.
o   Chemical and Volume Control (CVC) (Weld 2-CV-1175-36)
To conservatively quantify the effect on delta risk for weld 2-CV-1175-36, the delta risk is adjusted by not crediting the CVC weld with limited examination coverage. The below table provides the delta risk with weld 2-CV-1175-36 credited and with weld 2-CV-1175-36 not credited. The columns under heading Examined are the delta risk with the weld included, the columns under heading Not Examined is the adjusted delta risk after removing the weld. As can be seen, the change is negligible and well below the acceptable limits of 1E-07 for CDF and 1E-08 for LERF. Therefore, a limited examination of this CVC weld is acceptable. Note that weld 2-CV-1175-36 had previously been accepted with limited examination coverage of 50% under the previous RI-ISI Program based on Code Case N-578 and Request for Alternative S1-I4R-105 (Reference 13) as approved by the NRC in the 3rd 10-year ISI Interval (Reference 11).
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CDF Impact        I       LERF Impact w/ POD         I          w/ POD                   Source of Delta Risk Examined Latest interval periodic update CVC               -7.44E-09                 -7.44E-10 I                            required by N-716-1, Section 7 Not Examined Adjusted delta risk using the
                  -6.00E-09                 -6.00E-10 I                            periodic update delta risk o  Reactor Coolant System (RC) (Welds 14-PS-1131-2 and 4-PS-1131-29) 4-PS-1131-29 Note that this weld had previously been accepted with limited examination coverage of 50% under the deterministic Section XI Program (Reference 11).
Because this weld was examined during the 2nd ISI interval with limited coverage of 50%, as part of the traditional ISI program, then again in the 4rd interval with 50% limited coverage as part of the risk-informed program, there is no change in delta risk between the traditional ISI program and the risk-based ISI program caused by limited coverage.
14-PS-1131-2 To conservatively quantify the effect on delta risk for this PS weld, the delta risk is adjusted by not crediting this PS weld with limited examination coverage. The below table provides the delta risk with this PS weld credited and with this PS weld not credited. The columns under heading Examined are the delta risk with the weld included, the columns under heading Not Examined is the adjusted delta risk after removing the weld. As can be seen, the change is negligible and well below the acceptable limits of 1E-07 for CDF and 1E-08 for LERF.
Therefore, a limited examination of this PS weld is acceptable. Note that weld 14-PS-1131-2 had previously been accepted with limited examination coverage of 83.3% under the previous RI-ISI Program based on Code Case N-578 and Request for Alternative S1-I4R-105 (Reference 13) as approved by the NRC in the 3rd 10-year ISI Interval (Reference 11).
CDF Impact        I       LERF Impact w/ POD         I          w/ POD                   Source of Delta Risk Examined Latest interval periodic update RC                -6.27E-09                 -6.27E-10 I                            required by N-716-1, Section 7 Not Examined Adjusted delta risk using the
                  -4.83E-09                 -4.83E-10 I                            periodic update delta risk o  Safety Injection (SJ) (Welds 10-SJ-1111-17, 6-SJ-1131-17, and 6-SJ-1131-20)
To conservatively quantify the effect on delta risk for the three SJ welds, the delta risk is adjusted by not crediting the three SJ welds with limited examination coverage. The below table provides the delta risk with the three SJ welds credited and with the three SJ welds not credited. The columns under heading Examined are the delta risk with the three SJ welds included, the columns under heading Not Examined is the adjusted delta risk after removing the three Page 5 of 113

SJ welds. As can be seen, the change is negligible and well below the acceptable limits of 1E-07 for CDF and 1E-08 for LERF. Therefore, a limited examination of the three SJ welds is acceptable.
CDF Impact        I       LERF Impact w/ POD         I          w/ POD                  Source of Delta Risk Examined Latest interval periodic update SJ                 2.24E-09                  2.24E-10 I                           required by N-716, Section 7 Not Examined Adjusted delta risk using the 3.72E-09                  3.73E-10 I                           periodic update delta risk
* Cumulative Effect to Delta Risk for the One CVC Weld, Two RC Welds and the Three SJ Welds The impact of the limited examination on the delta risk for the individual systems are addressed above. The cumulative effect on the total delta risk is provided below.
The columns under heading Examined are the delta risk with the six welds included, the columns under heading Not Examined is the adjusted delta risk after removing the six welds. As can be seen, the change is negligible and well below the acceptable limits of 1E-07 for CDF and 1E-08 for LERF. Therefore, a limited examination of the one CVC Weld, two RC Welds and three SJ Welds is acceptable.
CDF Impact        I       LERF Impact w/ POD         I          w/ POD                  Source of Delta Risk Examined CVC Latest interval periodic update RC                 -1.45E-08                 -1.45E-09 SJ                                I                           required by N-716, Section 7 Not Examined
                            -1.01E-08                 -1.01E-09          Adjusted delta risk using the I                           periodic update delta risk
: 4.      Impracticality of Compliance The Construction permit for Salem Unit 1 was issued on September 25, 1968 and falls under the provisions of 10 CFR 50.55a(g)(1), which were applied to components (including supports) that must meet the requirements of paragraphs (g)(4) and (g)(5) to the extent practical.
Components that are part of the reactor coolant pressure boundary and their supports must meet the requirements applicable to components that are classified as ASME Code Class 1.
Other safety-related pressure vessels, piping, pumps and valves, and their supports must meet the requirements applicable to components that are classified as ASME Code Class 2 or Class
: 3. Therefore, 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 fitting 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. A typical example of such a Page 6 of 113

condition is a valve-to-pipe weld where essentially 100% of the code required volume cannot be examined from the valve side of the weld and where a plant modification would be needed to provide this coverage. Details of examination restrictions and reductions in required examination coverage are provided in Attachment 1.
When examined, the welds listed in Attachment 1 of this request 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% examination coverage of the required examination volumes or surface areas, but when this situation occurred 100% coverage of the accessible volumes or surface areas of each weld was obtained.
: 5.       Burden Caused by Compliance Burden Caused by Compliance To comply with the code required examination volumes or surface areas for 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 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.
For the Class 1, Examination Category B-A, Reactor Pressure Vessel Shell Welds, Item No.
B1.11 (1-PRV-10042) limitations were caused by the Core Guide Lugs.
For the Class 1, Examination Category B-A, Reactor Pressure Vessel Shell Weld, Item No.
B1.12 (1-RPV-1042B) limitations were caused by the outlet nozzle boss at 7&deg;.
For the Class 1, Examination Category B-A, Reactor Vessel Head Circumferential Weld, Item No. B1.21 (1-RPV-4043) limitations were caused by the Incore Nozzles.
For the Class 1, Examination Category B-A, Reactor Vessel Head Meridional Welds, Item No.
B1.22 (1-RPV-4043A, 1-RPV-4043B, 1-RPV-4043C, 1-RPV-4043D, 1-RPV-4043E, and 1-RPV-4043F) limitations were caused by Incore Nozzles.
For the Class 1, Examination Category B-B, Pressurizer Shell-to-Head Circumferential Weld, Item No. B2.11 (1-PZR-21) limitations were caused by permanent vessel support ring, weld pads, and insulation support straps in the weld location. These straps were loosened to make some areas of the weld accessible.
For Class 1, Examination Category B-B, Steam Generator Tubesheet-to-Head Weld, Item No.
B2.40 (13-STG-11) limitations were caused by four Steam Generator Vessel Supports and a data plate.
For Class 2 Examination Category C-B, Nozzle-to-Shell Welds, Item No. C2.21 (16-BFN-2111-
: 1) limitations were caused by the Steam Generator insulation support ring.
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For the Class 1 and Class 2 piping welds examined per the RI-ISI Programs the limitations listed in this request are typically limited by their design configurations or materials. The configurations of these welds or their materials only allow UT examination coverage from one side of the weld or limited coverage from a specific area or areas of one side of the weld and thus they would also require a design modification or replacement to obtain the required examination coverage.
Overall, it is not possible to obtain UT interrogation of greater than 90% of the required code examination volume or surface areas for the welds in this request without extensive weld or component design modifications. Examinations have been performed to the maximum extent possible and 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.
The examination techniques used for welds in this request for relief were reviewed to determine if additional coverage could be achieved by improving those techniques. None could be identified and the examinations have been performed to the maximum extent possible.
Therefore, SGS Unit 1 has determined that obtaining essentially 100% coverage is not feasible and is impractical without adding additional burden consisting of significant redesign work, increased radiation exposure, and/or potential damage to the plant or the component itself.
: 6.       Proposed Alternative and Basis for Use Proposed Alternative
: 1) Periodic system pressure tests and VT-2 visual examinations will continue to be performed in accordance with ASME 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.
Basis for Use 10 CFR 50.55a(g)(4) recognizes 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.
When a component is found to have conditions, which limit the required examination volume or surface area, SGS 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. SGS Unit 1 has performed the weld Page 8 of 113

examinations listed in this request to the maximum extent possible for each of the welds identified with limitations in Attachment 1.
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 SGS 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 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 Based upon the extent of the required surface examination area or UT examination volume achieved for each of the welds within this request for relief, and coupled with applicable leakage monitoring, and required system pressure tests with VT-2 visual examinations, no further action can be taken by SGS Unit 1 at this time to improve these examinations without applying impractical options. Therefore, the proposed alternative in this request for relief will provide assurance of an acceptable level of quality and safety by providing reasonable assurance of structural integrity.
: 7. Duration of Proposed Alternative This request for relief is for Salem Unit 1, Fourth 10-Year ISI Interval, which began on May 20, 2011 and ended on December 31, 2020.
: 8.       Precedents Note: Industry requests for relief due to impracticality associated with limited weld examinations are common. Some of the more recent NRC approvals of requested relief that are aligned with Reference 1 are:
: 1) Millstone Power Station, Unit No. 2 - Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval (EPID L-2020-LLR-0024, EPID L-2020-LLR-0025, and EPID L-2020-LLR-0026) dated December 10, 2020 (Accession No. ML20312A001).
: 2) Millstone Power Station, Unit No. 3 - Relief Request for Limited Coverage Examinations Performed in the Third 10-Year Inservice Inspection Interval (EPID L-2020-LLR-0027 through EPID L-2020-LLR-0032) dated December 10, 2020 (Accession No. ML20312A002).
Page 9 of 113
: 3) Perry Nuclear Power Plant, Unit No. 1 - Relief Request IR-062 Regarding Weld Examination Coverage for the Third Inservice Inspection Interval (EPID L-2020-LLR-0067) dated September 18, 2020 (Accession No. ML20252A026).
: 4) Nine Mile Point Nuclear Station, Unit 2 - Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (EPID-L-2019-LLR-0099) dated June 2, 2020 (Accession No. ML20141L053).
: 5) Beaver Valley Power Station, Unit 2 - Relief Request 2-TYP-3-B3.110-1, 2-TYP     C2.21-1, 2-TYP-3-C1.30-1, and 2-TYP-3-RA-1 Regarding Weld Examination Coverage for the Third Inservice Inspection Interval (EPID L-2019-LLR-0082) dated April 28, 2020 (Accession No. ML20080J789).
: 6) Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Relief Request RE:
Limited Examination Coverage During Fourth 10-Year Inservice Inspection Interval (EPID L-2019-LLR-0103) dated April 14, 2020 (Accession No. ML20097D644).
: 7) Beaver Valley Power Station, Unit No. 1 - Relief Requests 1-TYP-4-C2.21-1 and 1-TYP-4-RA-1 Regarding Weld Examination Coverage for the Fourth Inservice Inspection Interval (EPID L-2019-LLR-0083) dated March 26, 2020 (Accession No. ML20079F816).
: 8) Grand Gulf Nuclear Station, Unit 1 - Relief Request GG-ISI-023, Examination Coverage of Class 1 Piping and Vessel Welds (EPID L-2018-LLR-0197) dated October 11, 2019 (Accession No. ML19266A586).
: 9) Joseph M. Farley Nuclear Plant, Units 1 and 2 - RE: Relief from Impractical American Society of Mechanical Engineers Code Requirements (FNP-ISI-RR-02) (EPID L-2018-LLR-0195) dated August 8, 2019 (Accession No. ML19213A194).
: 10) Hope Creek Nuclear Generating Station - Issuance of Relief Request No. HC-I3R-08, Revision 0, RE: Relief from the Requirements of the ASME Code (EPID L-2018-LLR-0124) dated July 11, 2019 (Accession No. ML19136A026).
: 11) James A FitzPatrick Nuclear Power Plant - Issuance of Relief Request I4R-22 RE:
Relief from the Requirements of the ASME Code (EPID L-2018-LLR-0103) dated June 21, 2019 (Accession No. ML19135A444).
: 12) Monticello Nuclear Generating Plant - Request for Relief for Coverage of Nozzle-to-Vessel Weld Examinations (EPID L-2018-LLR-0072) dated June 12, 2019 (Accession No. ML19148A694).
: 13) LaSalle County Station, Units 1 and 2 - Relief Request I3R-15, Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME Code), Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components associated with the Third 10-Year Inservice Inspection Interval (EPID-L-2018-LLR-0123) dated May 20, 2019 (Accession No. ML19121A317)
: 14) Watts Bar Nuclear Plant, Unit 1 - Relief Request 1-ISI-21 from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (EPID L-2018-LLR-0090) dated April 12, 2019 (Accession No. ML19071A009).
: 15) Surry Power Station Unit No. 2 - Relief Requests Regarding Examination Coverage for Pressurizer Nozzle Inner Radius Sections and Certain Stainless Steel Piping Welds (S2-I5-LMT-C01 and S2-I5-LMT-P01) (EPID No. L-2018-LLR-0091) dated March 15, 2019 (Accession No. ML19043A824).
: 16) Surry Power Station Unit No. 1 - Relief Requests Regarding Examination Coverage for Pressurizer Nozzle Inner Radius Section and Certain Stainless Steel Piping Welds (S1-I5-LMT-C01 and S1-I5-LMT-P01) (EPID L-2018-LLR-0041) dated February 28, 2019 (Accession No. ML18331A060).
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: 9. References
: 1. NRC presentation Coverage Relief Requests, Industry/NRC NDE Technical Information Exchange Public Meeting January 13-15, 2015, (ADAMS Accession No. ML15013A266).
: 2. NRC presentation Coverage Relief Request Update, 2020 Industry/NRC NDE Technical Information Exchange Public Meeting. (ADAMS Accession No. ML20009E155)
: 3. ASME Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2004 Edition.
: 4. ASME Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems, 2001 Edition.
: 5. ASME Code Case N-460, Alternative Examination Coverage for Class 1 and Class 2 Welds Section XI, Division 1.
: 6. NRC Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 19, Dated October 2019. (ADAMS Accession No.:
ML19128A244).
: 7. EPRI Topical Report TR-112657 Rev. B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure, Dated: December 1999, (ADAMS Accession No.:
ML20205N012).
: 8. ASME Code Case N-578-1, Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B Section XI, Division 1.
: 9. NRC SER Salem Generating Station - Safety Evaluation of Relief Request to continue using a Risk-Informed Inservice Inspection Program for Salem Nuclear Generating Station, Unit No. 1 (TAC NO. ME4918) (ADAMS Accession No. ML112140424).
: 10. ASME Code Case N-716-1, Alternative Classification and Examination Requirements, Section XI, Division 1.
: 11. NRC SER, Salem Nuclear Generating Station, Unit No. 1 - Safety Evaluation of Relief Request No. S1-I3R-114 for Third 10-Year Interval Inservice Inspection (TAC No. ME8565), dated April 22, 2013 (ADAMS Accession No. ML13071A215).
: 12. PSEG Letter to NRC, Inservice Inspection (ISI) Program Plan, Fourth Ten-Year Interval, dated August 11, 2011 (ADAMS Accession No. ML112232141)
: 13. PSEG Letter to NRC, Request for Authorization to Continue using a Risk-Informed Inservice Inspection Alternative to the ASME Boiler and Pressure Vessel Code Section XI requirements for Class 1 and 2 Piping, dated October 21, 2010 (ADAMS Accession No. ML103060462)
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Attachment 1 Salem Generating Station Unit 1 American Society of Mechanical Engineers (ASME) Section XI 10 CFR 50.55a Request for Relief Number S1-I4R-210, Revision 0 In Accordance with 10 CFR 50.55a(g)(5)(iii)
                               --Inservice Inspection Impracticality--
Introduction This attachment contains figures and tables as applicable that are used to depict the limitations and calculations used for obtained coverage, materials and product forms, with ultrasonic examination angles and wave modes used, any limited surface examinations and the examination results for the welds associated with this request for relief, including any applicable previous examination history used. The following Table 1 for SGS Unit 1 identifies the welds within the scope of this request and summarizes the extent of examination coverage achieved for each weld.
Many of the welds listed were previously examined with different approved procedures and techniques during the span of the Third 10-Year ISI Interval and therefore not all the coverage calculations used are identical, but they are based on the actual NDE data reports that were provided for the examinations completed.
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TABLE 1 - Unit 1 SGS WELDS WITH LIMITED EXAMINATIONS Seq. Number /   Class,       Weld      Material   Material  Pipe Size   Examination  Examination Limitations         Normal Operating    Applicable Tables Weld       Category  Description    1 and        2        and/or       Code               and Results               Conditions           and Figures Identification and Item                Product    Product Thickness      Coverage                                 (Pressure/Temperature)
Number        No.                    Form       Form                  Obtained2 1.1          1        Reactor     SA-302     SA-302 8.3 - 5.38      64.4%       Exam was limited due to           2235 psig         Tables 1.1-1 and 1-RPV-10042      B-A    Vessel Lower    Gr. B      Gr. B                               the close proximity of the         575&deg;F                  1.1-2 B1.11   Shell to Lower                                                     Core Guide Lugs                               Figures 1.1-1, 1.1-2, Head, Circ                                                                                                           and 1.1-3 weld 1.2          1        Reactor     SA-302     SA-302      10.75       70.9%       Exam was limited due to           2235 psig       Tables 1.2-1, 1.2-2, 1-RPV-1042B      B-A    Vessel Upper    Gr. B      Gr. B                                the configuration of the          575&deg;F            1.2-3, and 1.2-4 B1.12     Shell at 7&deg;,                                                    Outlet Nozzle Boss                             Figures 1.2-1, 1.2-2 Long. Weld                                                                                                            and 1.2-3 1.3          1         Reactor    SA-302     SA-302       5.38       27.9%       Exam was limited due to           2235 psig         Tables 1.3-1 and 1-RPV-4043      B-A    Vessel Lower    Gr. B      Gr. B                              the proximity of the Incore         575&deg;F                  1.3-2 B1.21    Head Disc to                                                           Nozzles                                  Figures 1.3-1, 1.3-2 Peel                                                                                                             and 1.3-3 Segments, Circ Weld 1.4          1         Reactor    SA-302     SA-302       5.38       85.2%       Exam was limited due to           2235 psig       Tables 1.4-1, 1.4-2, 1-RPV-1043A      B-A        Vessel      Gr. B      Gr. B                              the proximity of the Incore         575&deg;F                and 1.4-3 B1.22     Meridional                                                            Nozzles                                  Figures 1.4-1, 1.4-2, Weld at 270&deg;,                                                                                                       1.4-3 and 1.4-4 Lower Head 1.5          1         Reactor    SA-302     SA-302       5.38       80.8%       Exam was limited due to           2235 psig         Tables 1.5-1 and 1-RPV-1043B      B-A        Vessel      Gr. B      Gr. B                              the proximity of the Incore         575&deg;F                  1.5-2 B1.22     Meridional                                                            Nozzles                                  Figures 1.5-1, 1.5-2, Weld at 330&deg;,                                                                                                       1.5-3 and 1.5-4 Lower Head 1.6          1         Reactor    SA-302     SA-302       5.38       80.7%       Exam was limited due to           2235 psig         Tables 1.6-1 and 1-RPV-1043C      B-A        Vessel      Gr. B      Gr. B                              the proximity of the Incore         575&deg;F                  1.6-2 B1.22     Meridional                                                            Nozzles                                  Figures 1.6-1, 1.6-2, Weld at 30&deg;,                                                                                                       1.6-3 and 1.6-4 Lower Head 1.7          1         Reactor    SA-302     SA-302       5.38       87.9%       Exam was limited due to           2235 psig         Tables 1.7-1 and 1-RPV-1043D      B-A        Vessel      Gr. B      Gr. B                              the proximity of the Incore         575&deg;F                  1.7-2 B1.22     Meridional                                                            Nozzles                                  Figures 1.7-1, 1.7-2, Weld at 90&deg;,                                                                                                       1.7-3 and 1.7-4 Lower Head Page 13 of 113

TABLE 1 - Unit 1 SGS WELDS WITH LIMITED EXAMINATIONS Seq. Number /    Class,       Weld      Material     Material   Pipe Size  Examination    Examination Limitations        Normal Operating    Applicable Tables Weld      Category Description      1 and           2      and/or       Code                and Results               Conditions           and Figures Identification and Item                Product      Product Thickness      Coverage                                  (Pressure/Temperature)
Number       No.                      Form         Form                Obtained2 1.8         1        Reactor      SA-302       SA-302      5.38       77.2%        Exam was limited due to           2235 psig        Tables 1.8-1, 1.8-2 1-RPV-1043E      B-A         Vessel       Gr. B        Gr. B                             the proximity of the Incore          575&deg;F             1.8-3, 1.8-4, B1.22     Meridional                                                                Nozzles                                       and 1.8-5 Weld at 150&deg;,                                                                                                       Figures 1.8-1, 1.8-2, Lower Head                                                                                                           1.8-3 and 1.8-4 1.9         1        Reactor      SA-302       SA-302      5.38       87.9%        Exam was limited due to           2235 psig            Table 1.9-1 1-RPV-1043F      B-A         Vessel       Gr. B        Gr. B                             the proximity of the Incore          575&deg;F         Figures 1.9-1, 1.9-2, B1.22     Meridional                                                                Nozzles                                    1.9-3 and 1.9-4 Weld at 210&deg;,
Lower Head 1.10         1      Pressurizer    SA-302       SA-216       4.6      42.15%        Exam was limited due to           2235 psig           Table 1.10-1 1-PZR-21      B-B       Shell J to     Gr. B         Gr.                              the proximity of insulation          655&deg;F            Figures 1.10-1, B2.11   Upper Head,                    WCC                               support straps, permanent                          1.10-2, 1.10-3, and Circ Weld                                                        vessel support ring and                                 1.10-4 welded pads.
1.11         1          Steam      SA-533       SA-508       5.5       68.5%        Exam was limited due to           2235 psig           Table 1.11-1 13-STG-11      B-B      Generator        Gr A       Cl. 2a                            the proximity of four steam           607&deg;F            Figures 1.11-1, B2.40    Lower Head        Cl 2                                           generator supports and a                             1.11-2, 1.11-3, to Tube                                                               data plate.                                  1.11-4, 1.11-5, Sheet, 13SG                                                                                                         1-11.6, 1.11-7, and 1.11-8 1.12         2          Steam      SA-508       SA-533        4        80.2%        Exam was limited due to            860psig            Table 1.12-1 16-BFN-2111-1     C-B      Generator      Cl. 2a        Gr. A                                  the proximity of the            425&deg;F            Figures 1.12-1, C2.21     Feedwater                     Cl. 2                           insulation ring at 180&deg; from                           1.12-2, 1.12-3, Nozzle to                                                         the nozzle boss. This                              1.12-4, 1.12-5, Shell,                                                    limited all scans angles and                            1.12-6, 1.12-7, 16-BF-2111                                                               directions.                                1.12-8 and 1.12-9 1.13         1       Chemical       A-376        A-403        2         50%      This was a one-sided exam            2520 psig          Figures 1.13-1, 2-CV-1175-36     R-A     Volume and      TP316       WP316     Schedule                     due to the pipe to tee            494&deg;F            1.13-2, 1.13-3, R1.11   Control - Pipe                            160                            configuration.                                  and 1.13-4 to Tee1 1.14         1        Reactor      SA-182       SA-216       14        71.7%       The circumferential scans          2235 psig          Figures 1.14-1, 14-PS-1131-2     R-A       Coolant       TP316          Gr. Schedule                  were limited to 90% of the           655&deg;F            1.14-2, 1.14-3, R1.11      System -      (Weld         WCC        160                      required volume due to                              1.14-4, 1.14-5, Nozzle to    filler 309     (Weld                            permanent welded lugs, the                              1.14-6, 1.14-7, Safe-end1        SS)      filler 309                           axial scans were limited to                         1.14-8 and 1.14-9 SS)                               53.4% due to the welded brackets and the nozzle taper.
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TABLE 1 - Unit 1 SGS WELDS WITH LIMITED EXAMINATIONS Seq. Number /         Class,          Weld       Material     Material   Pipe Size  Examination     Examination Limitations         Normal Operating     Applicable Tables Weld           Category      Description      1 and           2      and/or       Code                and Results                 Conditions           and Figures Identification     and Item                     Product      Product Thickness      Coverage                                    (Pressure/Temperature)
Number              No.                         Form         Form                Obtained2 1.15                1          Reactor       SA-182       SA-216        4        85.9%        The circumferential scans           2520 psig       Tables 1.15-1, 1.15-2, 4-PS-1131-29           R-A         Coolant        TP316          Gr. Schedule                    were limited to 71.8% of              494&deg;F             and 1.15-3 R1.11        System -        (Weld         WCC        160                    the required volume due to                              Figures 1.15-1, Safe-end to    filler 309     (Weld                                    the nozzle radius                                 1.15-2, 1.15-3, Nozzle1         SS)      filler 309                           configuration (CS & CCW).                             1.15-4, and 1.15-5 SS)                               The axial scans were not limited, 100% of the code required volume was achieved upstream and downstream.
1.16                1           Safety        A-376        A-351       10         50%        This was a one-sided exam              650 psi          Figures 1.16-1, 10-SJ-1111-17           R-A         Injection      TP316      Gr. CF8    Schedule                     due to the pipe to valve             120&deg;F           1.16-2, 1.16-3, R1.11/16    System - Pipe                              160                            configuration.                                1.16-4, and 1.16-5 to Valve (11SJ56) 1.17                1           Safety        A-376        A-403        6        80.8%       Exam was limited due to a             600 psig          Figures 1.17-1, 6-SJ-1131-17           R-A         Injection      TP316       WP316     Schedule                   permanent pipe restraint.              400&deg;F         1.17-2, 1.17-3, and R1.16      System - Pipe                              160                                                                                1.17-4 to Elbow1 1.18                1           Safety        A-376        A-403        6         80%          Exam was limited 8.5 of           2235 psig          Figures 1.18-1, 6-SJ-1131-20           R-A         Injection      TP316       WP316     Schedule                      upstream side of weld               607&deg;F         1.18-2, 1.18-3, and R1.20      System - Pipe                              160                       obstructed by restraint                                  1.18-4 to Elbow1                                                                  bracket.
NOTES: 1. Containment RCS Leakage Detection Applies
: 2. Ultrasonic (UT) Examination, Phased Array UT Examination (PAUT) and Surface Examination by Liquid Penetrant (PT) or Magnetic Particle (MT).
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1.1     Weld 1-RPV-10042 Lower Shell to Lower Head, Circ. Weld Figure 1.1-1 Weld 1-RPV-10042 (Extracted from Reference 12 DWG A-1)
This weld was UT examined in Inspection Period 3, during the RFO27 refueling outage in 2020.
The NDE data came from UT Report No.: VEN-20-021. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-1. The corresponding CRV as shown on that Figure is E-F-G-H. This volume was examined with 45&deg;S and 45&deg;L wave transducers.
The UT examination was limited by the proximity of the Core Guide Lugs resulting in total UT coverage of 64.4% as described in Tables 1.1-1 and 1.1-2 combined with Figures 1.1-2 and 1.1-
: 3. No recordable indications were detected during this scan.
Section XI Appendices and Supplements used for this UT examination were Appendix VIII, Supplements 4 and 6.
Note: No laminations exist on the RPV shell that could interfere with the angle beam examinations performed on this weld.
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Table 1.1-1 Weld 1-RPV-10042 Scan Coverage and Scan Summary Table 1.1-2 Weld 1-RPV-10042 Scan Coverage and Scan Summary Page 17 of 113

Figure 1.1-2 Weld 1-RPV-10042 (Examination Location and Coverage Map)
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Figure 1.1-3 Weld 1-RPV-10042 (Examination Location and Coverage Map)
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1.2     Weld 1-RPV-1042B Upper Shell @7&deg;, Longitudinal Weld Seam Figure 1.2-1 Weld 1-RPV-1042B (Extracted from Reference 12 DWG A-1)
This weld was UT examined in Inspection Period 3, during the RFO27 refueling outage in 2020.
The NDE data came from UT Report No.: VEN-20-010. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-2. The corresponding CRV as shown on that Figure is A-B-C-D. This volume was examined using 45&deg;S and 45&deg;L wave transducers.
The UT examination was limited by the proximity of the Outlet Nozzle Boss resulting in total UT coverage of 70.9% as described in Tables 1.2-1 and 1.2-2 combined with Figures 1.2-2 and 1.2-
: 3. One recordable flaw was detected during this examination. The flaw was determined to be subsurface and acceptable as shown in Tables 1.2-3 and 1.2-4.
Section XI Appendices and Supplements used for this UT examination were Appendix VIII, Supplement 4 and 6.
Note: No laminations exist on the RPV shell that could interfere with the angle beam examinations performed on this weld.
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Table 1.2-1 Weld 1-RPV-1042B Scan Coverage and Scan Summary Table 1.2-2 Weld 1-RPV-1042B Scan Coverage and Scan Summary Page 21 of 113

Table 1.2-3 Weld 1-RPV-1042B Flaw Evaluation Summary Sheet Table 1.2-4 Weld 1-RPV-1042B Flaw Evaluation Summary Sheet Page 22 of 113

Figure 1.2-2 Weld 1-RPV-1042B (Examination Location and Coverage Map)
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Figure 1.2-3 Weld 1-RPV-1042B (Examination Location and Coverage Map)
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1.3     Weld 1-RPV-4043 - Lower Head Disc to Peel Segments, Circ. Weld Figure 1.3-1 Weld 1-RPV-4043 (Extracted from Reference 12 DWG A-1)
This weld was UT examined in Inspection Period 3, during the RFO27 refueling outage in 2020.
The NDE data came from UT Report No.: VEN-20-022. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-3. The corresponding CRV as shown on that Figure is A-B-C-D. This volume was examined using 45&deg;S and 45&deg;L wave transducers.
The UT examination was limited by the proximity of the Incore Nozzles resulting in total UT coverage of 27.9% as described in Tables 1.3-1 and 1.3-2 combined with Figures 1.3-2 and 1.3-
: 3. No recordable indications were detected during this examination.
Section XI Appendices and Supplements used for this UT examination were Appendix VIII, Supplements 4 and 6.
Note: No laminations exist on the RPV lower head that could interfere with the angle beam examinations performed on this weld.
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Table 1.3-1 Weld 1-RPV-4043 Scan Coverage and Scan Summary Table 1.3-2 Weld 1-RPV-4043 Scan Coverage and Scan Summary Page 26 of 113

Figure 1.3-2 Weld 1-RPV-4043 (Examination Location and Coverage Map)
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Figure 1.3-3 Weld 1-RPV-4043 (Examination Location and Coverage Map)
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1.4     Weld 1-RPV-1043A - Meridional Weld at 270&deg;, Lower Head Figure 1.4-1 Weld 1-RPV-1043A (Extracted from Reference 12 DWG A-1)
This weld was UT examined in Inspection Period 3, during the RFO27 refueling outage in 2020.
The NDE data came from UT Report No.: VEN-20-035. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-3. The corresponding CRV as shown on that Figure is E-F-G-H. This volume was examined using 45&deg;S and 45&deg;L wave transducers.
The UT examination was limited due to the proximity of the Incore Nozzles resulting in total UT coverage of 85.2% as described in Tables 1.4-1 and 1.4-2 combined with Figures 1.4-2, 1.4-3 and 1.4-4. There was one recordable indication classified as subsurface flaw. This flaw is characteristic of slag inclusion from the welding process during fabrication. The flaw was evaluated as acceptable as shown in Table 1.4-3.
Section XI Appendices and Supplements used for this UT examination were Appendix VIII, Supplements 4 and 6.
Note: No laminations exist on the RPV lower head that could interfere with the angle beam examinations performed on this weld.
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Table 1.4-1 Weld 1-RPV-1043A Scan Coverage and Scan Summary Table 1.4-2 Weld 1-RPV-1043A Scan Coverage and Scan Summary Page 30 of 113

Table 1.4-3 Weld 1-RPV-1043A Flaw Evaluation Summary Sheet Page 31 of 113

Figure 1.4-2 Weld 1-RPV-1043A (Examination Location & Coverage Map)
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Figure 1.4-3 Weld 1-RPV-1043A (Examination Location & Coverage Map)
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Figure 1.4-4 Weld 1-RPV-1043A (Examination Location & Coverage Map)
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1.5     Weld 1-RPV-1043B - Meridional Weld at 330&deg;, Lower Head Figure 1.5-1 Weld 1-RPV-1043B (Extracted from Reference 12 DWG A-1)
This weld was UT examined in Inspection Period 3, during the RFO27 refueling outage in 2020.
The NDE data came from UT Report No.: VEN-20-036. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-3. The corresponding CRV as shown on that Figure is E-F-G-H. This volume was examined using 45&deg;S and 45&deg;L wave transducers.
The UT examination was limited due to the proximity of the Incore Nozzles resulting in total UT coverage of 80.8% as described in Tables 1.5-1 and 1.5-2 combined with Figures 1.5-2, 1.5-3 and 1.5-4. No recordable indications were detected during this examination.
Section XI Appendices and Supplements used for this UT examination were Appendix VIII, Supplements 4 and 6.
Note: No laminations exist on the RPV lower head that could interfere with the angle beam examinations performed on this weld.
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Table 1.5-1 Weld 1-RPV-1043B Scan Coverage and Scan Summary Table 1.5-2 Weld 1-RPV-1043B Scan Coverage and Scan Summary Page 36 of 113

Figure 1.5-2 Weld 1-RPV-1043B (Examination Location & Coverage Map)
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Figure 1.5-3 Weld 1-RPV-1043B (Examination Location & Coverage Map)
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Figure 1.5-4 Weld 1-RPV-1043B (Examination Location & Coverage Map)
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1.6     Weld 1-RPV-1043C - Meridional Weld at 30&deg;, Lower Head Figure 1.6-1 Weld 1-RPV-1043C (Extracted from Reference 12 DWG A-1)
This weld was UT examined in Inspection Period 3, during the RFO27 refueling outage in 2020.
The NDE data came from UT Report No.: VEN-20-031. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-3. The corresponding CRV as shown on that Figure is E-F-G-H. This volume was examined using 45&deg;S and 45&deg;L wave transducers.
The UT examination was limited due to the proximity of the Incore Nozzles resulting in total UT coverage of 80.7% as described in Table 1.6-1 combined with Figures 1.6-2, 1.6-3 and 1.6-4.
No recordable indications were detected during this examination.
Section XI Appendices and Supplements used for this UT examination were Appendix VIII, Supplements 4 and 6.
Note: No laminations exist on the RPV lower head that could interfere with the angle beam examinations performed on this weld.
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Table 1.6-1 Weld 1-RPV-1043C Scan Coverage and Scan Summary Table 1.6-2 Weld 1-RPV-1043C Scan Coverage and Scan Summary Page 41 of 113

Figure 1.6-2 Weld 1-RPV-1043C (Examination Location & Coverage Map)
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Figure 1.6-3 Weld 1-RPV-1043C (Examination Location & Coverage Map)
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Figure 1.6-4 Weld 1-RPV-1043C (Examination Location & Coverage Map)
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1.7     Weld 1-RPV-1043D - Meridional Weld at 90&deg;, Lower Head Figure 1.7-1 Weld 1-RPV-1043D (Extracted from Reference 12 DWG A-1)
This weld was UT examined in Inspection Period 3, during the RFO27 refueling outage in 2020.
The NDE data came from UT Report No.: VEN-20-032. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-3. The corresponding CRV as shown on that Figure is E-F-G-H. This volume was examined using 45&deg;S and 45&deg;L wave transducers.
The UT examination was limited due to the proximity of the Incore Nozzles resulting in total UT coverage of 87.9% as described in Table 1.7-1 combined with Figures 1.7-2, 1.7-3 and 1.7-4.
No recordable indications were detected during this examination.
Section XI Appendices and Supplements used for this UT examination were Appendix VIII, Supplements 4 and 6.
Note: No laminations exist on the RPV lower head that could interfere with the angle beam examinations performed on this weld.
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Table 1.7-1 Weld 1-RPV-1043D Scan Coverage and Scan Summary Table 1.7-2 Weld 1-RPV-1043D Scan Coverage and Scan Summary Page 46 of 113

Figure 1.7-2 Weld 1-RPV-1043D (Examination Location & Coverage Map)
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Figure 1.7-3 Weld 1-RPV-1043D (Examination Location & Coverage Map)
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Figure 1.7-4 Weld 1-RPV-1043D (Examination Location & Coverage Map)
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1.8     Weld 1-RPV-1043E - Meridional Weld at 150&deg;, Lower Head Figure 1.8-1 Weld 1-RPV-1043E (Extracted from Reference 12 DWG A-1)
This weld was UT examined in Inspection Period 3, during the RFO27 refueling outage in 2020.
The NDE data came from UT Report No.: VEN-20-033. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-3. The corresponding CRV as shown on that Figure is E-F-G-H. This volume was examined using 45&deg;S and 45&deg;L wave transducers.
The UT examination was limited due to the proximity of the Incore Nozzles resulting in total UT coverage of 77.2% as described in Tables 1.8-1 and 1.8-2 combined with Figures 1.8-2, 1.8-3 and 1.8-4. There were two recordable indications detected during this examination. The indications are classified as subsurface welding process indications. Each recordable flaw was evaluated for acceptance see Tables 1.8-3, 1.8-4, and 1.8-5.
Section XI Appendices and Supplements used for this UT examination were Appendix VIII, Supplements 4 and 6.
Note: No laminations exist on the RPV lower head that could interfere with the angle beam examinations performed on this weld.
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Table 1.8-1 Weld 1-RPV-1043E Scan Coverage and Scan Summary Table 1.8-2 Weld 1-RPV-1043E Scan Coverage and Scan Summary Page 51 of 113

Table 1.8-3 Weld 1-RPV-1043E Flaw Evaluation Summary Sheet Table 1.8-4 Weld 1-RPV-1043E Flaw Evaluation Summary Sheet Page 52 of 113

Table 1.8-5 Weld 1-RPV-1043E Flaw Evaluation Summary Sheet Page 53 of 113

Figure 1.8-2 Weld 1-RPV-1043E (Examination Location & Coverage Map)
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Figure 1.8-3 Weld 1-RPV-1043E (Examination Location & Coverage Map)
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Figure 1.8-4 Weld 1-RPV-1043E (Examination Location & Coverage Map)
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1.9     Weld 1-RPV-1043F - Meridional Weld at 210&deg;, Lower Head Figure 1.9-1 Weld 1-RPV-1043F (Extracted from Reference 12 DWG A-1)
This weld was UT examined in Inspection Period 3, during the RFO27 refueling outage in 2020.
The NDE data came from UT Report No.: VEN-20-034. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-3. The corresponding CRV as shown on that Figure is E-F-G-H. This volume was examined using 45&deg;S and 45&deg;L wave transducers.
The UT examination was limited due to the proximity of the Incore Nozzles resulting in total UT coverage of 87.9% as described in Table 1.9-1 combined with Figures 1.9-2, 1.9-3 and 1.9-4.
No recordable indications were detected during this examination.
Section XI Appendices and Supplements used for this UT examination were Appendix VIII, Supplements 4 and 6.
Note: No laminations exist on the RPV lower head that could interfere with the angle beam examinations performed on this weld.
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Table 1.9-1 Weld 1-RPV-1043F Scan Coverage and Scan Summary Page 58 of 113

Figure 1.9-2 Weld 1-RPV-1043F (Examination Location & Coverage Map)
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Figure 1.9-3 Weld 1-RPV-1043F (Examination Location & Coverage Map)
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Figure 1.9-4 Weld 1-RPV-1043F (Examination Location & Coverage Map)
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1.10    Weld 1-PZR Shell J to Upper Head, Circ. Weld Figure 1.10-1 Weld 1-PZR-21 (Extracted from Reference 12 DWG A-3)
This weld was UT examined in Inspection Period 2, during the RFO24 refueling outage in 2016.
The NDE data came from UT Report No.: UT-16-040. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-1. The corresponding CRV as shown on that Figure is E-F-G-H. This volume was examined using 60&deg;L and 60&deg;RL wave transducers.
The UT examination was limited due to the proximity of the support rings and 1 set of weld pads Page 62 of 113

resulting in total UT coverage of 42.15% as described in Table 1.10-1 combined with Figures 1.10-2, 1.10-3 and 1.10-4. No recordable indications were detected during this examination.
This examination was performed using Section XI Appendix VIII, Supplements 4 and 6 per Appendix I, I-2600 which states For components to which Appendix VIII is not applicable, examinations procedures, personnel, and equipment qualified in accordance with Appendix VIII may be applied provided such components, materials, sizes, and shapes are within the scope of the qualified examination procedure. The procedure utilized was 54-ISI-805-008 which was qualified through Appendix VIII, Supplements 4 and 6. Demonstration of the procedure included the detection of cracks at various orientations for both single and dual sided access. Personnel used for this examination were qualified through Performance Demonstration to perform examination of this procedure for Supplements 4 and 6 for single and dual sided access using qualified equipment specified within Procedure 54-ISI-805/PDI-UT-6. Demonstration was performed on a thickness of 6.88. Although pressurizer components are not applicable to Appendix VIII, the component, materials, sizes and shapes were within the scope of the qualified examination procedure. The thickness of 1-PZR-21 weld is 4.6. Prior to the examination, requirements of IWA-2240 were met to the satisfaction of the Authorized Nuclear Inservice Inspector (ANII).
Note: No laminations exist on the Pressurizer that could interfere with the angle beam examinations performed on this weld.
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Multiple scan limitations due to insulation support straps in the weld location. These straps were loosened to make some areas of the weld accessible. There are also many areas around the circumference of the weld that block access of the weld. Most of these areas were encountered from 175 to 290.
Table 1.10-1 Weld 1-PZR-21 Scan Coverage and Scan Summary Page 64 of 113

Figure 1.10-2 Weld 1-PZR-21 Scan Coverage and Scan Summary Figure 1.10-3 Weld 1-PZR-21 Scan Coverage and Scan Summary Page 65 of 113

Figure 1.10-4 Weld 1-PZR-21 (Weld Thickness and Contour)
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1.11    Weld 13-STG Lower Head to Tube Sheet, 13SG Figure 1.11-1 Weld 13-STG 11 (Extracted from Reference 12 DWG A-6)
This weld was UT examined in Inspection Period 3, during the RFO27 refueling outage in 2020.
The NDE data came from UT Report No.: UT-20-054. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-6. The corresponding CRV as shown on that Figure is E-F-G-H. This volume was examined using 60&deg;L and 60&deg;RL wave transducers. The UT examination was limited due to the proximity of the supports and a name Page 67 of 113

plate resulting in total UT coverage of 68.5% as described in Table 1.11-1 combined with Figures 1.11-2, 1.11-3, 1.11-4, 1.11-5, 1.11.6, 1.11-7 and 1.11-8. No recordable indications were detected during this examination. This weld was previously approved for limited examination for the 3rd Interval with a UT coverage of approximately 67% (Reference 11).
Section XI, Subsection IWA General Requirements, Article IWA-2000 Examination and Inspection, IWA-2200 Alternative Examinations states Alternative examination methods, a combination of methods, or newly developed techniques may be substituted for the methods specified in this Division, provided the Inspector is satisfied that the results are demonstrated to be equivalent or superior to those of the specified method.
Section XI, Appendix I, Article I-2000 Examination Requirements, I2100 Vessels Greater than 2 in. states in I-2120 Other Vessels: All other vessels (such as Steam Generators) greater than 2 in. in thickness shall be conducted in accordance with Article 4 of Section V.
Procedure 54-ISI-805-008 has been qualified through Appendix VIII, Supplement 4 and
: 6. Demonstration of this procedure included the detection of cracks at various orientations for both single and dual sided access. Personnel used for this examination have been qualified through Performance Demonstration (PDI) to perform examination of this procedure for Supplement 4 and 6 for single and dual sided access using the qualified equipment specified within Procedure 54-ISI-805/PDI-UT-6. Demonstration was performed to a thickness of 11.06 inches.
Although steam generator components are not applicable to Appendix VIII, the component materials, sizes, and shapes are within the scope of the qualified examination procedure. Also the code specified method for examination has not changed and will not be changed using procedure 54-ISI-805-008 for this examination.
Prior to the examination, requirements of IWA-2240 were met to the satisfaction of the Inspector.
Note: No laminations exist on the Steam Generator that could interfere with the angle beam examinations performed on this weld.
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Table 1.11-1 Weld 13-STG-11 Scan Coverage and Scan Summary Page 69 of 113

Figure 1.11-2 Weld 13-STG-11 Scan Coverage and Scan Summary Figure 1.11-3 Weld 13-STG-11 Scan Coverage and Scan Summary Page 70 of 113

Figure 1.11-4 Weld 13-STG-11 Scan Coverage and Scan Summary Figure 1.11-5 Weld 13-STG-11 Scan Coverage and Scan Summary Page 71 of 113

Figure 1.11-6 Weld 13-STG-11 Scan Coverage and Scan Summary Figure 1.11-7 Weld 13-STG-11 Scan Coverage and Scan Summary Page 72 of 113

Figure 1.11-8 Weld 13-STG-11 (Weld Thickness and Contour)
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1.12   Weld 16-BFN-2111 Nozzle-to-Shell, 16-BF-2111 Figure 1.12-1 Weld 16-BFN-2111-1 (Extracted from Reference 12 DWG A-4)
This weld was UT and MT examined in Inspection Period 3, during the RFO27 refueling outage in 2020. The NDE data came from UT Report No.: VEN-20-003. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWC-2500-4(a). The corresponding CRV as shown on that Figure is C-D-E-F. This volume was examined using 0&deg;L, 35&deg;S, 45&deg;S, and 60&deg;S wave transducers. The UT examination was limited due to the proximity of the Steam Page 74 of 113

Generator Insulation Support Ring at 180&deg; from the nozzle boss resulting in total UT coverage of 80.2% as described in Table 1.12-1 combined with Figures 1.12-2, 1.12-3, 1.12-4, 1.12-5, 1.12.6, 1.12-7. 1.12-8, and 1.12-9. No recordable indications were detected during this examination. This weld was previously approved for limited examination for the 3rd Interval with a UT coverage of approximately 70.6% (Reference 11).
The surface examination was also limited due to the Steam Generator Insulation Support Ring resulting in a total surface area of 85.2%. This weld was previously approved for limited examination for the 3rd Interval with an MT coverage of approximately 85.3% (Reference 11).
Section XI Appendix I, I-2400 was used for this UT examination. This required Article 4 of Section V, as supplemented by Table I-2000-1 to be used.
Note: No laminations exist on the Steam Generator that could interfere with the angle beam examinations performed on this weld.
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Table 1.12-1 Weld 16-BFN-2111-1 Scan Coverage and Scan Summary Page 76 of 113

Figure 1.12-2 Weld 16-BFN-2111-1 Scan Coverage and Scan Summary Page 77 of 113

Figure 1.12-3 Weld 16-BFN-2111-1 Scan Coverage and Scan Summary Figure 1.12-4 Weld 16-BFN-2111-1 Scan Coverage and Scan Summary Page 78 of 113

Figure 1.12-5 Weld 16-BFN-2111-1 Scan Coverage and Scan Summary Figure 1.12-6 Weld 16-BFN-2111-1 Scan Coverage and Scan Summary Page 79 of 113

Figure 1.12-7 Weld 16-BFN-2111-1 Scan Coverage and Scan Summary Page 80 of 113

Figure 1.12-8 Weld 16-BFN-2111-1 (Photo)
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Figure 1.12-9 Weld 16-BFN-2111-1 (Thickness and Contour)
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1.13   Weld 2-CV-1175 Pipe to Tee Figure 1.13-1 Weld 2-CV-1175-36 (Extracted from Reference 12 DWG A-12)
This weld was UT examined in Inspection Period 2, during the RFO24 refueling outage in 2016.
The NDE data came from UT Report No.: UT-16-022. The UT Code Required Volume (CRV) was determined based on EPRI TR-112657 Rev. B-A, Figure 4-1. The corresponding CRV as shown on that Figure is A-B-C-D. This volume was examined using 45&deg;S, 60&deg;S and 70&deg;S wave transducers. The UT examination was limited to a one-sided exam due to the configuration of the pipe to tee resulting in total UT coverage of 50% as described in Figures 1.13-2, 1.13-3 and 1.13-4. No recordable indications were detected during this examination. This weld was previously approved for limited examination for the 3rd Interval with a UT coverage of approximately 50% (Reference 11).
This examination was performed in accordance with Request for Alternative S1-I4R-105 as approved by the Nuclear Regulatory Commission (NRC) on August 17, 2001 (Reference 9).
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The examination satisfied 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).
Figure 1.13-2 Weld 2-CV-1175-36 Scan Coverage and Scan Summary Page 84 of 113

Figure 1.13-3 Weld 2-CV-1175-36 (Photo)
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Figure 1.13-4 Weld 2-CV-1175-36 (Thickness and Contour)
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1.14    Weld 14-PS-1131 Nozzle-to-Safe-End
                                                                                         ~,J pS C
Figure 1.14-1 Weld 14-PS-1131-2 (Extracted from Reference 12 DWG A-18)
This weld was UT examined in Inspection Period 2, during the RFO25 refueling outage in 2017.
The NDE data came from UT Report No.: UT-17-021. The UT Code Required Volume (CRV) was determined based on EPRI TR-112657 Rev. B-A, Figure 4-2. The corresponding CRV as shown on that Figure is A-B-C-D. This volume was examined using 45&deg;S, 35&deg;L, 45&deg;L, and 60&deg;L wave transducers. The UT examination was limited due to the configuration of the nozzle taper and permanent welded lugs resulting in total UT coverage of 71.7% as described in Figures 1.14-2, through 1.14-9. No recordable indications were detected during this examination. This weld was previously approved for limited examination for the 3rd Interval with a UT coverage of approximately 83.3% (Reference 11).
This examination was performed in accordance with Request for Alternative S1-I4R-105 as approved by the Nuclear Regulatory Commission (NRC) on August 17, 2001 (Reference 9).
The examination satisfied 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).
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Figure 1.14-2 Weld 14-PS-1131-2 Scan Coverage and Scan Summary Page 88 of 113

Figure 1.14-3 Weld 14-PS-1131-2 Scan Coverage and Scan Summary Page 89 of 113

Figure 1.14-4 Weld 14-PS-1131-2 Scan Coverage and Scan Summary Page 90 of 113

Figure 1.14-5 Weld 14-PS-1131-2 Scan Coverage and Scan Summary Page 91 of 113

Figure 1.14-6 Weld 14-PS-1131-2 Scan Coverage and Scan Summary Page 92 of 113

Figure 1.14-7 Weld 14-PS-1131-2 Scan Coverage and Scan Summary Page 93 of 113

Figure 1.14-8 Weld 14-PS-1131-2 (Photo)
Figure 1.14-9 Weld 14-PS-1131-2 (Thickness and Contour)
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1.15   Weld 4-PS-1131 Safe-End-to-Nozzle Figure 1.15-1 Weld 4-PS-1131-29 (Extracted from Reference 12 DWG A-21)
This weld was UT examined partially in Period 2 during the RFO25 refueling outage in 2018 and completed in RFO26 refueling outage in 2019. The NDE data came from UT Report No.: VEN-19-005. This is a dissimilar metal weld. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-8(c). The corresponding CRV as shown on that Figure is C-D-E-F. This volume was examined using phased array transducers. The UT examination was limited due to the configuration of the nozzle radius configuration resulting in total UT coverage of 85.9% as described in Figures 1.15-3, through 1.15-5. One recordable indication was detected during this examination. This indication plotted to the component ID and into the Page 95 of 113

cladding under the weld volume. This weld does not contain susceptible material to PWSCC.
True flaw type indications such as cracking provide the following characteristics:
* Indications provide substantial and unique echo-dynamic travel (walk) through the time base
* Several areas of unique amplitude peaks are observed throughout the indication length
* Indications show evidence of flaw tip signals.
For the indication recorded, none of the characteristics listed above were observed during this examination. In addition, past radiographs were reviewed and identified no areas containing fabrication flaws. As identified in Section XI, IWB-3514.1(d)(1); Surface flaws that do not penetrate through the nominal clad thickness into the base metal need not be compared with standards of IWB-3514.1(a). Conservatively, the final NDE evaluation was performed as if the indication was located in the pressure retaining base material and the results were found acceptable for continued service per Section XI acceptance criteria identified in IWB-3500, Table IWB-3514-2. Therefore, based on the information furnished above, the indication was determined to be acceptable (See Tables 1.15-1, 1.15-2, 1.15-3 and Figure 1.15-2).
This examination was performed in accordance with Request for Alternative S1-I4R-105 as approved by the Nuclear Regulatory Commission (NRC) on August 17, 2001 (Reference 9).
The examination satisfied 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).
Table 1.15-1 Weld 4-PS-1131-29 UT Indication Data Sheet Page 96 of 113

Table 1.15-2 Weld 4-PS-1131-29 UT Indication Data Sheet Figure 1.15-2 Weld 4-PS-1131-29 UT Indication Plot Sheet Page 97 of 113

Table 1.15-3 Weld 4-PS-1131-29 ASME XI (2004) Table IWB-3514-2 Page 98 of 113

Figure 1.15-3 Weld 4-PS-1131-29 Scan Coverage and Scan Summary Page 99 of 113

Figure 1.15-4 Weld 4-PS-1131-29 (Thickness and Contour)
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Figure 1.15-5 Weld 4-PS-1131-29 (Photo)
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1.16    Weld 10-SJ-1111 Pipe to Valve (11SJ56)
Figure 1.16-1 Weld 10-SJ-1111-17 (Extracted from Reference 12 DWG A-57)
This weld was UT examined in Inspection Period 2, during the RFO25 refueling outage in 2017.
The NDE data came from UT Report No.: UT-17-029. The UT Code Required Volume (CRV) was determined based on EPRI TR-112657 Rev. B-A, Figure 4-2. The corresponding CRV as shown on that Figure is C-D-E-F. This volume was examined using 45&deg;S, 60&deg;S and 60&deg;L wave transducers. The UT examination was limited to a one-sided examination due to the configuration of the pipe to valve configuration resulting in total UT coverage of 50% as described in Figures 1.16-2, through 1.16-5. No recordable indications were detected during this examination.
Page 102 of 113

This examination was performed in accordance with Request for Alternative S1-I4R-105 as approved by the Nuclear Regulatory Commission (NRC) on August 17, 2001 (Reference 9).
The examination satisfied 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).
Figure 1.16-2 Weld 10-SJ-1111-17 Scan Coverage and Scan Summary Figure 1.16-3 Weld 10-SJ-1111-17 Scan Coverage and Scan Summary Page 103 of 113

Figure 1.16-4 Weld 10-SJ-1111-17 (Photo)
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Figure 1.16-5 Weld 10-SJ-1111-17 (Thickness and Contour)
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1.17    Weld 6-SJ-1131 Pipe to Elbow Figure 1.17-1 Weld 6-SJ-1131-17 (Extracted from Reference 12 DWG A-67)
This weld was UT examined in Inspection Period 2, during the RFO25 refueling outage in 2017.
The NDE data came from UT Report No.: UT-17-026. The UT Code Required Volume (CRV) was determined based on EPRI TR-112657 Rev. B-A, Figure 4-11. The corresponding CRV as shown on that Figure is C-D-E-F. This volume was examined using 45&deg;S, and 60&deg;S wave transducers. The UT examination was limited due to a permanent pipe restraint pad resulting in total UT coverage of 80.8% as described in Figures 1.17-2, through 1.17-4. No recordable indications were detected during this examination.
This examination was performed in accordance with Request for Alternative S1-I4R-105 as approved by the Nuclear Regulatory Commission (NRC) on August 17, 2001 (Reference 9).
The examination satisfied 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).
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Figure 1.17-2 Weld 6-SJ-1131-17 Scan Coverage and Scan Summary Page 107 of 113

Figure 1.17-3 Weld 6-SJ-1131-17 (Photo)
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Figure 1.17-4 Weld 6-SJ-1131-17 (Thickness and Contour)
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1.18    Weld 6-SJ-1131 Pipe to Elbow Figure 1.18-1 Weld 6-SJ-1131-20 (Extracted from Reference 12 DWG A-67)
This weld was UT examined in Inspection Period 3, during the RFO26 refueling outage in 2019.
The NDE data came from UT Report No.: VEN-19-077. The UT Code Required Volume (CRV) was determined based on Section XI, Figure IWB-2500-8(c). The corresponding CRV as shown on that Figure is C-D-E-F. This volume was examined using 60&deg;S and 60&deg;L wave transducers.
The UT examination was limited 8.5 of the upstream side of the weld due to a restraint bracket resulting in total UT coverage of 80% as described in Figures 1.15-2, through 1.15-4. No recordable indication was detected during this examination.
This examination was performed in accordance with Code Case N-716-1 as approved by the Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.147 Rev. 19. The examination satisfied 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).
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Figure 1.18-2 Weld 6-SJ-1131-20 Scan Coverage and Scan Summary Page 111 of 113

Figure 1.18-3 Weld 6-SJ-1131-20 (Thickness and Contour)
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Figure 1.18-4 Weld 6-SJ-1131-20 (Photo)
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