ML17093A894
| ML17093A894 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/12/2017 |
| From: | Doug Broaddus Special Projects and Process Branch |
| To: | Dent J Entergy Nuclear Operations |
| Lamb J, NRR/DORL/LSPB, 301-415-3100 | |
| References | |
| CAC MF8092, CAC MF8093 | |
| Download: ML17093A894 (11) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. John Dent, Jr.
Site Vice President Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360-5508 May 12, 2017
SUBJECT:
PILGRIM NUCLEAR POWER STATION - RELIEF REQUESTS PNPS-ISl-001 AND PNPS-ISl-002 FOR RELIEF FROM ASME CODE, SECTION XI, VOLUMETRIC EXAMINATION REQUIREMENTS (CAC NOS. MF8092 AND MF8093)
Dear Mr. Dent:
By letter dated June 29, 2016, as supplemented by letter dated January 27, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML16188A269 and ML17037D054, respectively), Entergy Nuclear Operations, Inc. (Entergy or the licensee),
submitted to the U.S. Nuclear Regulatory Commission (NRC) Relief Requests PNPS-ISl-001, PNPS-ISl-002, and PNPS-ISl-003 requesting relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, volumetric examination requirements at the Pilgrim Nuclear Power Station (PNPS).
Specifically, pursuant to Title 1 O of the Code of Federal Regulations (1 O CFR)
Section 50.55a(g)(5)(iii), Entergy requested relief from the "essentially 100 percent" volumetric coverage requirements of ASME Code,Section XI, for the subject welds on the basis that the ASME Code requirement is impractical.
By letter dated January 27, 2017, Entergy withdrew Relief Request PNPS-ISl-003.
The NRC staff concludes, as set forth in the enclosed safety evaluation, that ASME Code examination coverage requirements are impractical for the subject welds listed in Relief Requests PNPS-ISl-001 and PNPS-ISl-002. The NRC staff concludes that the examinations performed, to the extent practical, provide reasonable assurance of structural integrity of the subject components. The NRC staff has further determined that granting Relief Requests PNPS-ISl-001 and PNPS-ISl-002, in accordance with 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, given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Therefore, the NRC staff grants Relief Requests PNPS-ISl-001 and PNPS-ISl-002, pursuant to 10 CFR 50.55a(g)(6)(i),
for the fourth 10-year inservice inspection interval, which started on July 1, 2005, and ended on June 30, 2015, at PNPS.
J. Dent, Jr. All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in the subject request for relief remains applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
Relief Requests PNPS-ISl-004 and PNPS-ISl-005 were completed on April 3, 2017 (ADAMS Accession No. ML17081A563).
If you have any questions, please contact the project manager, John G. Lamb, at (301) 415-3100 or via e-mail at John.Lamb@nrc.gov.
Docket No. 50-293
Enclosure:
Safety Evaluation cc: Listserv Sincerely, IRA by EBrown for/
Douglas A. Broaddus, Chief Special Projects and Process Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUESTS PNPS-ISl-001 AND PNPS-ISl-002 FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL ENTERGY NUCLEAR OPERATIONS, INC.
PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293
1.0 INTRODUCTION
By letter dated June 29, 2016, as supplemented by letter dated January 27, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML16188A269 and ML17037D054, respectively), Entergy Nuclear Operations, Inc. (Entergy or the licensee),
submitted Relief Requests PNPS-ISl-001, PNPS-ISl-002, and PNPS-003 from the "essentially 100 percent" volumetric coverage requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, "Rules tor lnservice Inspection of Nuclear Power Plant Components," tor welds, due to access limitations at the Pilgrim Nuclear Power Station (PNPS). The requests tor relief applies to the fourth 10-year inservice inspection (ISi) interval, which started on July 1, 2005, and ended on June 30, 2015.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 O CFR) 50.55a(g)(5)(iii),
Entergy requested relief from the "essentially 100 percent" volumetric coverage requirements of ASME Code,Section XI, for the subject welds on the basis that the ASME Code requirement is impractical.
By letter dated January 27, 2017, Entergy withdrew Relief Request PNPS-181-003.
By letter dated April 3, 2017 (ADAMS Accession No. ML17081A563}, the U.S. Nuclear Regulatory Commission (NRG) approved Relief Requests PNPS-ISl-004 and PNPS-ISl-005.
2.0 REGULATORY REQUIREMENTS The licensee has requested relief from ASME Code requirements pursuant to 1 O CFR 50.55a(g)(5)(iii). Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall 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 complies with the requirements in the latest edition and addenda of Section XI of the ASME Code, which was incorporated by reference in 1 O CFR 50.55a(b),
Enclosure 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein.
The regulation in 1 O CFR 50.55a(g)(5)(iii) states, in part, that licensees may determine that conformance with certain ASME Code requirements is impractical and that the licensee shall notify the Commission and submit information in support of the determination. Determination 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 ISi 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 120-month inspection interval or subsequent 120-month inspection interval for which relief is sought.
The regulation in 10 CFR 50.55a(g)(6)(i) states that the Commission will evaluate determinations under paragraph (g)(5) of this section that the code requirements are impractical. The Commission may grant such relief and may impose such alternative requirements as it determines are 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 Commission to grant, relief and impose such alternative requirements as it determines are authorized by law, will not endanger life or property or the common defense and security, and are 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.
3.0 TECHNICAL EVALUATION
3.1 Code of Record The PNPS applicable Code of record for these relief requests is the ASME Code,Section XI, 2001 Edition with 2003 Addenda.
The ASME Code,Section XI, 2001 Edition with the 2003 Addenda, Examination Categories B-A and B-D, require 100 percent volumetric examination coverage of the pressure-retaining welds as defined in Table IWB-2500-1 and shown in Figures IWB-2500-2, IWB-2500-3, and IWB-2500-7(a) through (d).
The ASME Code Case N-460, Alternative Examination Coverage for Class 1 and Class 2 Welds,Section XI, Division 1, as approved for use by the NRC in Regulatory Guide 1.147, Revision 17, "lnservice Inspection Code Case Acceptability,Section XI, Division 1" (ADAMS Accession No. ML13339A689), states that a reduction in examination coverage due to part geometry or interference for any ASME Class 1 or 2 weld is acceptable provided that the reduction is less than 10 percent (i.e., greater than 90 percent examination coverage is obtained).
The ASME Code,Section XI, 2001 Edition with the 2003 Addenda, Mandatory Appendix I, Article 1-2110, requires that ultrasonic examination (UT} of reactor vessel nozzle-to-vessel welds greater than 2 inches in thickness shall be qualified by performance demonstration in accordance with Appendix VI 11.
3.2 Relief Request PNPS-ISl-001, Examination Category B-A, Item Numbers 81.12 and 81.21. Pressure-Retaining Welds in Reactor Vessels 3.2.1 Components Covered by Relief Request PNPS-ISl-001 Relief Request PNPS-ISl-001 covers the following:
Table 1: Welds Covered in Relief Request PNPS-ISl-001 Item Weld ID Description Coverage 81.12 RPV-L-1-338A Lower Intermediate Shell 88.40%
Vertical Weld 81.12 RPV-L-1-3388 Lower Intermediate Shell 89.70%
Vertical Weld 81.12 RPV-L-1-338C Lower Intermediate Shell 56.20%
Vertical Weld 81.12 RPV-L-2-338A Lower Shell Vertical Weld 75.80%
81.12 RPV-L-2-3388 Lower Shell Vertical Weld 71.50%
81.12 RPV-L-2-338C Lower Shell Vertical Weld 11.60%
81.12 RPV-L-2-339A Upper Intermediate Shell 18.00%
Vertical Weld 81.12 RPV-L-2-3398 Upper Intermediate Shell 69.00%
Vertical Weld 81.12 RPV-L-2-339C Upper Intermediate Shell 73.20%
Vertical Weld 81.21 RPV-BH-C1 Bottom Head Circumferential 68.20%
Weld 3.2.2 Licensee's Basis for Relief Request PNPS-ISG-001 The licensee determined that it would be impractical to obtain greater than 90 percent coverage of the reactor vessel (RV) shell and head welds listed in Table 1-1 of the licensee's submittal dated June 16, 2016, during UT. These examinations were performed utilizing a Performance Demonstration Initiative Appendix VIII, Supplement 4 and Supplement 6 qualified procedure specific to RV assembly welds.
The geometric configuration of the components limited the accessibility of the volumetric examinations to the outside of the vessel on the RV circumferential head weld and the inside of the vessel on the RV longitudinal shell welds. The 60-degree (
0
) refracted longitudinal (RL) beam angle, which was used in both the axial and circumferential directions of the RV circumferential head weld, was not able to achieve greater than 90 percent of the Code-required volume as required by Code Case N-460. Similarly, the 45° and 70° RL beam angles used in the axial and circumferential directions of the RV longitudinal welds were not able to achieve greater than 90 percent of the Code-required volume. Further information detailing prevention of each component from full examination are listed below and in Attachment 6, Table 1-1 of the licensee's submittal.
Entergy did not propose any alternative examinations, stating that the subject welds were examined to the maximum practical extent. Particularly, the licensee stated that modification and/or replacement of the subject components would be necessary in order to perform significant additional UT examinations.
3.2.3 NRC Staff Evaluation of Relief Request PNPS-181-001, Examination Category B-A, Item Numbers 81.12 and 81.21. Pressure-Retaining Welds in Reactor Vessels As stated by the licensee, the geometric configuration of the Category B-A, Item 81.21 bottom head circumferential weld limits effective volumetric examination to the outside of the vessel.
Utilizing the 60° RL transducer in both the transverse and parallel directions, the licensee was able to obtain 68.2 percent of the total weld coverage and attributed the limitations in coverage to the proximity of the vessel support skirt and thermocouples. The sketch included with the licensee's submittal confirmed the sections of the ASME Code-required examination coverage that were examined and the sections where examination was precluded due to the curvature of the vessel support skirt and the positioning of the thermocouples. No recordable indications were detected during this examination. In a previous examination of this component, three indications were observed but were determined not to be through-wall and were found acceptable to the requirements of Subarticle IWB-3500. These previously recorded indications were observed during the current examination and were still found acceptable to the Subarticle IWB-3500 requirements. Obtaining further examination coverage would require major redesign of the bottom head to move the weld away from the obstructions. The NRC staff reviewed the licensee's submittal and agrees that the examination of weld was conducted to the maximum practical extent.
Due to the configuration of PNPS, volumetric examination of the Category B-A, Item 81.12 RV longitudinal shell welds could only be performed from the inside of the vessel. The examinations were performed along the accessible lengths of the welds utilizing automated full volume 45° T-and P-scans, near surface 70° T-scans clockwise and counter-clockwise, and P-scans up and down. Excluding RPV-L-2-339A and RPV-L-2-3398 upper intermediate shell vertical welds, the full cross-sectional examination area of the remaining welds were obtained; therefore, the impracticality of full examination coverage was based solely on the inaccessible length of the weld. The lower intermediate shell vertical weld examinations were inhibited by jet pump riser braces, guide rod brackets, and specimen holder and support brackets; the lower shell vertical weld examinations were inhibited by the shroud support and associated gusset plates, the shroud tie bar, and the N2K nozzle radius; and the upper intermediate shell vertical welds were inhibited by the feedwater and core spray spargers. Additionally, Entergy performed skewed examinations of the positioning of the RPV-L-2-339A and RPV-L-2-3398 upper intermediate shell vertical welds due to the positioning of surveillance specimen brackets required. No recordable indications were found in any of the examination results. The NRC staff agrees that the examination of welds were conducted to the maximum practical extent and that the examinations performed on the welds would sufficiently detect signs of degradation.
Regarding the RPV-L-2-339A upper intermediate shell vertical weld, Entergy was able to examine 24 percent of the weld length and 71.3 percent of the total cross-sectional area equating to a total composite coverage of 18 percent. The licensee's submittal stated that 81 percent composite coverage of this weld was examined in refueling outage 14 and that the differences in coverage were due to the difference in design of the scanner utilized. The NRC staff requested by RAI that the licensee justify how this drop in coverage provides an equivalent or greater standard of quality and safety. By letter dated January 27, 2017, Entergy attributed the reduction in coverage to the differences in the design of the scanners utilized. The licensee utilized a different vendor from the last inspection and was unaware of the reduction in coverage until the access study was provided to the licensee within weeks of the outage. The scanner utilized in the 2005 examination was nominally 2.0 inches thick compared to the scanner utilized in 2015, which was nominally 3.59 inches thick when horizontally oriented and 3.30 inches thick vertically. The nominal clearance of 3.34 inches under the guide rod is 0.04 inches greater than the nominal thickness of the scanner in the vertical orientation; however, the nominal clearance values do not account for tolerance stack-up or local guide rod deformations which ultimately prevented scanner access under the guide rod.
The only known degradation mechanism for this type of cladded carbon steel weld is neutron irradiation-induced embrittlement; however, this weld is not located in the beltline region.
Therefore, the fracture toughness of the weld and associated base metal are not significantly reduced. The accumulated fluence of this weld was calculated to be 4.28 x 1015 n/cm2 at 54 effective full power years and Pilgrim, which is currently at approximately 31.6 effective full power years, will permanently shut down in June 2019, well before reaching the estimated fluence. Furthermore, cracking in boiling-water reactor vessel welds, including beltline region welds, is uncommon. The other two similar welds in the same shell segment achieved 69 percent and 73.2 percent coverage with no indications. Considering the neutron fluence is well below the 1 O CFR 50 Appendix H threshold value of 1.00 x 1017 n/cm2, service-induced cracking is highly unlikely, and examinations of similar welds in a similar environment produced no recordable indications, the NRC staff finds that there is sufficient evidence that there is no safety issue with the subject weld and it will maintain its integrity for the remaining life of the plant.
Regarding RPV-L-2-338C, automated scanning of the full RPV lower shell vertical weld at 315° was restricted due to the shroud repair lower support hardware clearances. Specifically, the shroud support and associated gusset plates prevent the scanner tool from positioning the transducer to the bottom of the weld, where the RPV vertical weld intersects with the shroud support plate, the shroud tie bar at 315° prevents maneuvering of the scanner tooling to fully access the weld, and N2K nozzle radius prevents even contact of the scanner tooling. The accessible portion of the weld was examined from the top of the weld with the scanner head facing down and covering 15 of the 129.1 weld length inches.
Since the full cross-sectional examination area was obtained, the impracticality of full examination coverage is based solely on the inaccessible length of the weld. Therefore, the licensee was able to examine a total composite coverage of 11.6 percent, which was a drop from the 25 percent coverage obtained in 2005. Lower shell vertical welds RPV-L-2-338A and RPV-L-2-3388 were able to obtain 75.80 percent and 71.50 percent coverage, respectively, during these 2015 examinations and found no recordable indications. With no history of service-induced degradation, the NRC staff finds that there is sufficient evidence that there is no safety issue with this weld and it will maintain its integrity for the remaining life of the plant.
The licensee has shown that achieving essentially 100 percent volumetric examination coverage of the components in this relief request was impractical due to the geometric configuration of the bottom head circumferential weld and the configuration of the plant precluding full examination of the longitudinal shell welds. Based on the volumetric coverages obtained, considering the performance of ultrasonic techniques employed to maximize this coverage, and considering that no relevant indications were found, it is reasonable to conclude that, if significant degradation were occurring, evidence of degradation would be detected by the examinations.
3.3 Relief Request PNPS-ISl-002, Examination Category B-D, Item Number 83.90, Pressure-Retaining Welds in RV Nozzles 3.3.1 Components Covered by Relief Request PNPS-ISl-002 Relief Request PNPS-ISl-002 covers the following:
Table 2: Welds Covered in Relief Request PNPS-ISl-002 Item Weld ID Description Coverage 83.90 RPV-N1A-NV Recirculation Outlet Nozzle-to-Vessel 43.80%
Weld 83.90 RPV-N2A-NV Recirculation Inlet Nozzle-to-Vessel 33.60%
Weld 83.90 RPV-N28-NV Recirculation Inlet Nozzle-to-Vessel 33.60%
Weld 83.90 RPV-N2C-NV Recirculation Inlet Nozzle-to-Vessel 33.60%
Weld 83.90 RPV-N3A-NV Main Steam Nozzle-to-Vessel Weld 37.90%
83.90 RPV-N38-NV Main Steam Nozzle-to-Vessel Weld 37.90%
83.90 RPV-N3C-NV Main Steam Nozzle-to-Vessel Weld 37.90%
83.90 RPV-N3D-NV Main Steam Nozzle-to-Vessel Weld 37.90%
83.90 RPV-N4A-NV Feedwater Nozzle-to-Vessel Weld 49.30%
83.90 RPV-N48-NV Feedwater Nozzle-to-Vessel Weld 49.30%
83.90 RPV-N4C-NV Feedwater Nozzle-to-Vessel Weld 49.30%
83.90 RPV-N4D-NV Feedwater Nozzle-to-Vessel Weld 49.30%
83.90 RPV-N6A-NV Core Spray Nozzle-to-Vessel Weld 42.90%
83.90 RPV-N7A-NV Head Spray Nozzle-to-Vessel Weld 84.40%
83.90 RPV-N9A-NV Jet Pump Instrument Nozzle-to-Vessel 86.30%
Weld 83.90 RPV-N 1 OA-NV Control Rod Return Nozzle-to-Vessel 43.70%
Weld 3.3.2 Licensee's Basis for Relief Request PNPS-ISG-002 While attempting to comply with the ASME Code requirements during the ISi interval for which the request was submitted, the licensee determined that it would be impractical to obtain greater than 90 percent coverage of the RV shell and head welds listed in Attachment 6, Table 2-1 of the submittal during UT examination. The actual examination coverages achieved by the licensee are documented in Attachment 6, Table 2-1 of the licensee's submittal. The licensee stated that these examinations were performed utilizing Entergy-approved procedures specific to ferritic vessels greater than 2 inches in thickness. The licensee stated that no relevant indications were recorded.
Entergy stated in its submittal that the geometric configuration of the Category B-D, Item 83.90 components limits the accessibility of the volumetric examinations to the shell side of the nozzle-to-vessel welds. Sketches included in the licensee's submittal show the geometry of the nozzle-to-vessel weld for each component in the relief request. These sketches also show the insonification angles and ultrasonic beam modes used during the UT examination. In its January 27, 2017, response to the NRC staff's request for additional information (RAI) dated November 22, 2016 (ADAMS Accession No. ML16327A474), the licensee further clarified that the inspection vendor used only 60° degree refracted longitudinal search units for its ASME Code,Section XI, Supplement 4 and 6 examinations, and that various shear wave search units were used for its ASME Code Section XI, Appendix VIII, Supplement 5 examinations. The Entergy submittals indicated that the beam angles used were not able to achieve greater than the ASME Code-required volume.
Entergy did not propose any alternative examinations, stating that the subject welds were examined to the maximum practical extent. Particularly, the licensee stated that modification and/or replacement of the subject components would be necessary in order to perform significant additional UT examinations.
3.3.3 NRC Staff Evaluation for Relief Request PNPS-ISl-002, Examination Category 8-D.
Item Number 83.90, Pressure-Retaining Welds in Reactor Vessel Nozzles As stated by the licensee in its submittal, the geometric configuration of the Category 8-D, Item 83.90 components in this relief request limits the accessibility of the volumetric examinations to the shell side of the nozzle-to-vessel welds. The sketches included with the licensee's submittal show that the geometry of the nozzles will prevent UT of the weld from the nozzle side. In addition, the sketches also show that the weld geometry (blend radius) prevents the transducer from being placed over the weld and limits how close the transducer can be placed adjacent to the weld. These are two factors that limit coverage of the components in this relief request. The NRC staff has reviewed the licensee's submittal, and concludes that the geometric configuration of the components in this relief request limits the accessibility and that the beam angles used were not able to achieve greater than the ASME Code-required volume.
Entergy's submittal stated that examinations were performed using procedures specific to ferritic vessels greater than 2 inches in thickness.Section XI of the ASME Code, Mandatory Appendix I, Article 1-2110, requires that UT of RV nozzle-to-vessel welds greater than 2 inches in thickness shall be qualified by performance demonstration in accordance with Supplements 4, 5, and 6 of Appendix VIII. Supplement 4 contains the qualification requirement for the clad/base metal interface of the reactor vessel. Supplement 5 contains the qualification requirements for nozzle inside radius sections. Supplement 6 contains the qualification requirements for that portion of the RV weld other than clad/base metal interface. As the licensee's submittal did not state that UTs of the components in this relief request were performed in accordance with the ASME Code, the NRC staff requested that the licensee confirm that these examinations were performed in accordance with ASME Code,Section XI, Appendix VIII. In its response dated January 27, 2017, Entergy confirmed that UT of the components was performed in accordance with ASME Code,Section XI, Appendix VIII, with Supplements 4, 5, and 6.
The NRC staff finds that the licensee's response is acceptable since it meets the requirements of ASME Code,Section XI for UT of RV nozzle-to-vessel welds greater than 2 inches in thickness.
Although ASME Code-required volume was not achieved, the volume that was examined had no relevant indications. Entergy performed ultrasonic scans both parallel and transverse to the weld in order to detect both axial and circumferential flaws. Although the geometry of the components limited the coverage of the Supplement 6 scans, the licensee was able to achieve 100 percent coverage of the underclad volume when performing the Supplement 4 scans.
The licensee has shown that it is impractical to meet the ASME Code-required volumetric examination coverage for the components in this relief request due to the geometric configuration of the nozzle-to-vessel welds. Based on the volumetric coverage obtained, considering the performance of ultrasonic techniques employed to maximize this coverage, and considering that no relevant indications were found, it is reasonable to conclude that, if significant degradation were occurring, evidence of degradation would be detected by the examinations.
4.0 CONCLUSION
The NRC staff has reviewed Entergy's submittals and concludes that ASME Code examination coverage requirements are impractical for the subject welds listed in Relief Requests PNPS-ISl-001 and PNPS-ISl-002. The NRC staff concludes that the examinations performed, to the extent practical, provide reasonable assurance of structural integrity of the subject components. The NRC staff has further determined that granting Relief Requests PNPS-ISl-001 and PNPS-ISl-002, in accordance with 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, given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Therefore, the NRC staff grants Relief Requests PNPS-ISl-001 and PNPS-ISl-002, pursuant to 10 CFR 50.55a(g)(6)(i), for the fourth 10-year ISi interval, which started on July 1, 2005, and ended on June 30, 2015, at PNPS.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in the subject request for relief remains applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor: A. Young Date: May 12, 2017
J. Dent, Jr.
SUBJECT:
PILGRIM NUCLEAR POWER STATION - RELIEF REQUESTS PNPS-ISl-001 AND PNPS-ISl-002 FOR RELIEF FROM ASME CODE, SECTION XI, VOLUMETRIC EXAMINATION REQUIREMENTS (CAC NOS. MF8092 AND MF8093) DATED MAY 12, 2017 DISTRIBUTION:
Public LSPB R/F RidsACRS_MailCTR Resource RidsNrrDeEvib Resource RidsNrrDorlLspb Resource RidsNrrLAIBetts Resource RidsNrrLAJBurkhardt Resource RidsNrrPMPilgrim Resource RidsRgn1 MailCenter Resource AYoung, NRR/DE/EVIB JBowen, OEDO ADAMS A ccess1on N ML17093A894
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- 1 1y e-ma1 OFFICE NRR/DORL/LSPB/PM NRR/DORL/LSPB/LAiT NRR/DORL/LSPB/LA NAME Jlamb I Betts JBurkhardt DATE 04/2/17 04/12/17 04/20/17 OFFICE NRR/DE/EVIB/BC*
NRR/DORL/LSPB/BC NAME DRudland EBrown for DBroaddus DATE 03/31/17 05/12/17 OFFICIAL RECORD COPY