ML13130A144: Difference between revisions
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
||
| Line 18: | Line 18: | ||
=Text= | =Text= | ||
{{#Wiki_filter:UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 May 24,2013 Mr. James E. Lynch Site Vice President Northern States Power Company -Minnesota Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch, MN 55089-9642 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND REQUEST FOR ADDITIONAL INFORMATION RELATED TO REACTOR VESSEL INTERNALS PROGRAM SUBMITTAL FOR FULFILLMENT OF LICENSE RENEWAL COMMITMENT 25 (TAC NOS. MF0052 AND MF0053) | {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 24,2013 Mr. James E. Lynch Site Vice President Northern States Power Company - Minnesota Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch, MN 55089-9642 | ||
==SUBJECT:== | |||
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION RELATED TO REACTOR VESSEL INTERNALS PROGRAM SUBMITTAL FOR FULFILLMENT OF LICENSE RENEWAL COMMITMENT 25 (TAC NOS. MF0052 AND MF0053) | |||
==Dear Mr. Lynch:== | ==Dear Mr. Lynch:== | ||
By letter dated October 1, 2012 (Agencywide Documents Access and Management System (ADAMS)Accession No. ML12276A041), as supplemented by letters dated March 7, 2013, and March 22, 2013 (ADAMS Accession Nos. ML13084A378 and ML13067A284, respectively), Northern States Power Company, a Minnesota corporation (NSPM, the licensee), doing business as Xcel Energy, submitted an aging management program (AMP) for the reactor vessel internals at Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2. Materials Reliability Program-227-A report, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," and its supporting reports were used as technical bases for developing the PINGP AMP. The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with Mr. Dale Vincent of your staff on May 7, 2013, it was agreed that you would provide a response to this request with 30 days of the date of this letter. The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. | |||
J. E. Lynch -If circumstances result in the need to revise the requested response date, please contact me at (301) 415-4037. | By letter dated October 1, 2012 (Agencywide Documents Access and Management System (ADAMS)Accession No. ML12276A041), as supplemented by letters dated March 7, 2013, and March 22, 2013 (ADAMS Accession Nos. ML13084A378 and ML13067A284, respectively), | ||
Sincerely, Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306 Request for Additional cc w/encl: Distribution via REQUEST FOR ADDITIONAL INFORMATION (RAI) PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. SO-282 AND SO-306 By letter dated October 1, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12276A041), as supplemented by letters dated March 7, 2013 (ADAMS Accession No. ML13067A284) and March 21, 2013 (ADAMS Accession No. ML13084A378), Northern States Power Company, a Minnesota corporation (the licensee), doing business as Xcel Energy, submitted an aging management program (AMP) for the reactor vessel internals (RVI) at Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2. Materials Reliability Program (MRP)-227-A report, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," and its supporting reports were used as technical bases for developing the PINGP AMP. The Nuclear Regulatory Commission (NRC) staff reviewed this report and issued a final safety evaluation (SE) on December 16, 2011 (ADAMS Accession No. ML11308A770). | Northern States Power Company, a Minnesota corporation (NSPM, the licensee), doing business as Xcel Energy, submitted an aging management program (AMP) for the reactor vessel internals at Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2. Materials Reliability Program-227-A report, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," and its supporting reports were used as technical bases for developing the PINGP AMP. | ||
Based on the review of PINGP's AMP conducted thus far, the staff has developed this first request for additional information (RAI). The staff may issue additional RAls based on the resolution of Action Items 1 and 2 addressed in the staffs SE for the MRP-227-A report. RAI-1: Historically, the following materials used in the PWR RVI components were known to be susceptible to some of the aging degradation mechanisms that are identified in the MRP-227-A report. In this context, the NRC staff requests that the licensee confirm that these materials are not currently used in the RVI components at PINGP, Units 1 and 2. (1) Nickel base alloys -lnconeI600; Weld Metals -Alloy 82 and 182 Alloy X-7S0 (excluding control rod guide tube split (2) Alloy A-286 ASTM A 4S3 Grade 660, Condition A or B (3) Stainless steel type 347 material (excluding baffle-former bolts) (4) Precipitation hardened (PH) stainless steel materials | The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with Mr. Dale Vincent of your staff on May 7, 2013, it was agreed that you would provide a response to this request with 30 days of the date of this letter. | ||
The NRC staff requests that the licensee provide information regarding the extent of aging degradation (if any) that has occurred thus far in all of the RVI components. | The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. | ||
J. E. Lynch - 2 If circumstances result in the need to revise the requested response date, please contact me at (301) 415-4037. | |||
Sincerely, | |||
~=te~~ | |||
Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306 | |||
==Enclosure:== | |||
Request for Additional Information cc w/encl: Distribution via ListServ | |||
REQUEST FOR ADDITIONAL INFORMATION (RAI) | |||
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. SO-282 AND SO-306 By letter dated October 1, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12276A041), as supplemented by letters dated March 7, 2013 (ADAMS Accession No. ML13067A284) and March 21, 2013 (ADAMS Accession No. | |||
ML13084A378), Northern States Power Company, a Minnesota corporation (the licensee), doing business as Xcel Energy, submitted an aging management program (AMP) for the reactor vessel internals (RVI) at Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2. | |||
Materials Reliability Program (MRP)-227-A report, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," and its supporting reports were used as technical bases for developing the PINGP AMP. The Nuclear Regulatory Commission (NRC) staff reviewed this report and issued a final safety evaluation (SE) on December 16, 2011 (ADAMS Accession No. ML11308A770). Based on the review of PINGP's AMP conducted thus far, the staff has developed this first request for additional information (RAI). The staff may issue additional RAls based on the resolution of Action Items 1 and 2 addressed in the staffs SE for the MRP-227-A report. | |||
RAI-1: Historically, the following materials used in the PWR RVI components were known to be susceptible to some of the aging degradation mechanisms that are identified in the MRP-227-A report. In this context, the NRC staff requests that the licensee confirm that these materials are not currently used in the RVI components at PINGP, Units 1 and 2. | |||
(1) Nickel base alloys -lnconeI600; Weld Metals - Alloy 82 and 182 and Alloy X-7S0 (excluding control rod guide tube split pins) | |||
(2) Alloy A-286 ASTM A 4S3 Grade 660, Condition A or B (3) Stainless steel type 347 material (excluding baffle-former bolts) | |||
(4) Precipitation hardened (PH) stainless steel materials 4 and 1S-S (S) Type 431 stainless steel material RAI-2: Condition 7 of Revision 1 of the NRC staffs December 16, 2011, SE, stipulates that the licensee shall include a summary of the operating experience related to the aging degradation in the RVI components. The NRC staff requests that the licensee provide information regarding the extent of aging degradation (if any) that has occurred thus far in all of the RVI components. | |||
Specifically, include the operating history of the following components at PINGP, Units 1 and 2: | Specifically, include the operating history of the following components at PINGP, Units 1 and 2: | ||
* baffle-former bolts | * baffle-former bolts | ||
| Line 33: | Line 54: | ||
* core barrel bolting, and | * core barrel bolting, and | ||
* thermal shields. | * thermal shields. | ||
Provide a summary that includes a list of RVI components that have been inspected thus far under the American Society of Mechanical Engineers (ASME) Code, Section XI, inservice inspection program, and the inspection results. This list shall include any RVI component categorized under the "Existing" inspection category in the MRP-227-A report. RAI-3: According to Section A.1.4 in MRP-175, "Materials Reliability Program: PWR Internal Aging Degradation Mechanism Screening Threshold Values," the susceptibility to stress corrosion cracking (SCC) in nickel-based Alloy X-750 PWR RVI components depends on the type of heat treatment that is performed on the alloy. High temperature heat treatment (HTH) processes that are used on Alloy X-750 components offer better resistance to SCC than the other age hardened heat treatment processes. | |||
Additionally, Appendix A of the MRP-227-A report identified, as a part of the industry's operational experience, that the clevis insert assembly in Alloy X-750 bolting in one operating unit failed due to primary water stress corrosion cracking (PWSCC). Therefore, the staff requests that the licensee provide information related to the type of heat treatment process that was used for the Alloy X-750 materials used in the RVI components at PINGP. If Alloy X-750 material is used for clevis insert bolting or for any other RVI components at PINGP, Units 1 and 2, confirm that HTH treatment was performed on this material. | -2 Provide a summary that includes a list of RVI components that have been inspected thus far under the American Society of Mechanical Engineers (ASME) Code, Section XI, inservice inspection program, and the inspection results. This list shall include any RVI component categorized under the "Existing" inspection category in the MRP-227-A report. | ||
If the clevis insert bolting did not undergo HTH treatment, discuss your plans to inspect these bolts (in addition to the inspections to monitor aging due to wear) for identifying PWSCC. RA14: In Enclosure 1, Page 2, of the licensee's October 1, 2012, submittal, the licensee indicated that the control rod guide tube (CRGT) cards will be inspected no later than two refueling outages from the beginning of the license renewal period (Le., the period of extended operation). | RAI-3: According to Section A.1.4 in MRP-175, "Materials Reliability Program: PWR Internal Aging Degradation Mechanism Screening Threshold Values," the susceptibility to stress corrosion cracking (SCC) in nickel-based Alloy X-750 PWR RVI components depends on the type of heat treatment that is performed on the alloy. High temperature heat treatment (HTH) processes that are used on Alloy X-750 components offer better resistance to SCC than the other age hardened heat treatment processes. Additionally, Appendix A of the MRP-227-A report identified, as a part of the industry's operational experience, that the clevis insert assembly in Alloy X-750 bolting in one operating unit failed due to primary water stress corrosion cracking (PWSCC). Therefore, the staff requests that the licensee provide information related to the type of heat treatment process that was used for the Alloy X-750 materials used in the RVI components at PINGP. If Alloy X-750 material is used for clevis insert bolting or for any other RVI components at PINGP, Units 1 and 2, confirm that HTH treatment was performed on this material. If the clevis insert bolting did not undergo HTH treatment, discuss your plans to inspect these bolts (in addition to the inspections to monitor aging due to wear) for identifying PWSCC. | ||
The licensee also stated that it would perform inspections on the CRGT cards to assess the wear of these cards. The staff requests that the licensee provide the following information: | RA14: In Enclosure 1, Page 2, of the licensee's October 1, 2012, submittal, the licensee indicated that the control rod guide tube (CRGT) cards will be inspected no later than two refueling outages from the beginning of the license renewal period (Le., the period of extended operation). The licensee also stated that it would perform inspections on the CRGT cards to assess the wear of these cards. The staff requests that the licensee provide the following information: | ||
(1) The number of cards that are planned to be inspected (2) The inspection results (3) How the criteria for maximum allowed wear was established (4) The licensee's corrective actions, if any, and (5) The licensee's plan for subsequent inspections of this component during the period of extended operation. | (1) The number of cards that are planned to be inspected (2) The inspection results (3) How the criteria for maximum allowed wear was established (4) The licensee's corrective actions, if any, and (5) The licensee's plan for subsequent inspections of this component during the period of extended operation. | ||
J. E. Lynch If circumstances result in the need to revise the requested response date, please contact me at (301) 415-4037. | |||
J. E. Lynch -2 If circumstances result in the need to revise the requested response date, please contact me at (301) 415-4037. | |||
* via memo OFFICE NRR/LPL3-1/PM NRR/LPL3-21LA NRRlEVIB/BC NRRlLPL3-1/BC NRRlLPL3*1/PM NAME TWengert SRohrer SRosenberg* | Sincerely, Ira! | ||
RCarlson TWengert DATE 05/17/13 05/13/13 04/15/13 05/24/13 05/24/13 OFFICIAL RECORD COpy}} | Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306 | ||
==Enclosure:== | |||
Request for Additional Information cc w/encl: Distribution via ListServ DISTRIBUTION: | |||
PUBLIC LPL3-1 R/F RidsNrrDorlLpl3-1 Resource RidsNrrDorlDpr Resource RidsNrrPMPrairielsland Resource RidsNrrLABTully Resource RidsAcrsAcnw_MailCTR Resource GCheruvenki, NRR RidsRgn3MailCenter Resource RidsNrrDeEvib Resource ADAMS Accession Number: ML13130A144 | |||
* via memo OFFICE NRR/LPL3-1/PM NRR/LPL3-21LA NRRlEVIB/BC NRRlLPL3-1/BC NRRlLPL3*1/PM NAME TWengert SRohrer SRosenberg* RCarlson TWengert DATE 05/17/13 05/13/13 04/15/13 05/24/13 05/24/13 OFFICIAL RECORD COpy}} | |||
Latest revision as of 18:24, 4 November 2019
| ML13130A144 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/24/2013 |
| From: | Thomas Wengert Plant Licensing Branch III |
| To: | Jeffery Lynch Northern States Power Co |
| Wengert T | |
| References | |
| TAC MF0052, TAC MF0053 | |
| Download: ML13130A144 (5) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 24,2013 Mr. James E. Lynch Site Vice President Northern States Power Company - Minnesota Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch, MN 55089-9642
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION RELATED TO REACTOR VESSEL INTERNALS PROGRAM SUBMITTAL FOR FULFILLMENT OF LICENSE RENEWAL COMMITMENT 25 (TAC NOS. MF0052 AND MF0053)
Dear Mr. Lynch:
By letter dated October 1, 2012 (Agencywide Documents Access and Management System (ADAMS)Accession No. ML12276A041), as supplemented by letters dated March 7, 2013, and March 22, 2013 (ADAMS Accession Nos. ML13084A378 and ML13067A284, respectively),
Northern States Power Company, a Minnesota corporation (NSPM, the licensee), doing business as Xcel Energy, submitted an aging management program (AMP) for the reactor vessel internals at Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2. Materials Reliability Program-227-A report, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," and its supporting reports were used as technical bases for developing the PINGP AMP.
The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with Mr. Dale Vincent of your staff on May 7, 2013, it was agreed that you would provide a response to this request with 30 days of the date of this letter.
The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources.
J. E. Lynch - 2 If circumstances result in the need to revise the requested response date, please contact me at (301) 415-4037.
Sincerely,
~=te~~
Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosure:
Request for Additional Information cc w/encl: Distribution via ListServ
REQUEST FOR ADDITIONAL INFORMATION (RAI)
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. SO-282 AND SO-306 By letter dated October 1, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12276A041), as supplemented by letters dated March 7, 2013 (ADAMS Accession No. ML13067A284) and March 21, 2013 (ADAMS Accession No.
ML13084A378), Northern States Power Company, a Minnesota corporation (the licensee), doing business as Xcel Energy, submitted an aging management program (AMP) for the reactor vessel internals (RVI) at Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2.
Materials Reliability Program (MRP)-227-A report, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," and its supporting reports were used as technical bases for developing the PINGP AMP. The Nuclear Regulatory Commission (NRC) staff reviewed this report and issued a final safety evaluation (SE) on December 16, 2011 (ADAMS Accession No. ML11308A770). Based on the review of PINGP's AMP conducted thus far, the staff has developed this first request for additional information (RAI). The staff may issue additional RAls based on the resolution of Action Items 1 and 2 addressed in the staffs SE for the MRP-227-A report.
RAI-1: Historically, the following materials used in the PWR RVI components were known to be susceptible to some of the aging degradation mechanisms that are identified in the MRP-227-A report. In this context, the NRC staff requests that the licensee confirm that these materials are not currently used in the RVI components at PINGP, Units 1 and 2.
(1) Nickel base alloys -lnconeI600; Weld Metals - Alloy 82 and 182 and Alloy X-7S0 (excluding control rod guide tube split pins)
(2) Alloy A-286 ASTM A 4S3 Grade 660, Condition A or B (3) Stainless steel type 347 material (excluding baffle-former bolts)
(4) Precipitation hardened (PH) stainless steel materials 4 and 1S-S (S) Type 431 stainless steel material RAI-2: Condition 7 of Revision 1 of the NRC staffs December 16, 2011, SE, stipulates that the licensee shall include a summary of the operating experience related to the aging degradation in the RVI components. The NRC staff requests that the licensee provide information regarding the extent of aging degradation (if any) that has occurred thus far in all of the RVI components.
Specifically, include the operating history of the following components at PINGP, Units 1 and 2:
- baffle-former bolts
- baffle-edge bolts
- baffle-former assembly
- clevis insert bolts
- core barrel bolting, and
- thermal shields.
-2 Provide a summary that includes a list of RVI components that have been inspected thus far under the American Society of Mechanical Engineers (ASME) Code,Section XI, inservice inspection program, and the inspection results. This list shall include any RVI component categorized under the "Existing" inspection category in the MRP-227-A report.
RAI-3: According to Section A.1.4 in MRP-175, "Materials Reliability Program: PWR Internal Aging Degradation Mechanism Screening Threshold Values," the susceptibility to stress corrosion cracking (SCC) in nickel-based Alloy X-750 PWR RVI components depends on the type of heat treatment that is performed on the alloy. High temperature heat treatment (HTH) processes that are used on Alloy X-750 components offer better resistance to SCC than the other age hardened heat treatment processes. Additionally, Appendix A of the MRP-227-A report identified, as a part of the industry's operational experience, that the clevis insert assembly in Alloy X-750 bolting in one operating unit failed due to primary water stress corrosion cracking (PWSCC). Therefore, the staff requests that the licensee provide information related to the type of heat treatment process that was used for the Alloy X-750 materials used in the RVI components at PINGP. If Alloy X-750 material is used for clevis insert bolting or for any other RVI components at PINGP, Units 1 and 2, confirm that HTH treatment was performed on this material. If the clevis insert bolting did not undergo HTH treatment, discuss your plans to inspect these bolts (in addition to the inspections to monitor aging due to wear) for identifying PWSCC.
RA14: In Enclosure 1, Page 2, of the licensee's October 1, 2012, submittal, the licensee indicated that the control rod guide tube (CRGT) cards will be inspected no later than two refueling outages from the beginning of the license renewal period (Le., the period of extended operation). The licensee also stated that it would perform inspections on the CRGT cards to assess the wear of these cards. The staff requests that the licensee provide the following information:
(1) The number of cards that are planned to be inspected (2) The inspection results (3) How the criteria for maximum allowed wear was established (4) The licensee's corrective actions, if any, and (5) The licensee's plan for subsequent inspections of this component during the period of extended operation.
J. E. Lynch -2 If circumstances result in the need to revise the requested response date, please contact me at (301) 415-4037.
Sincerely, Ira!
Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosure:
Request for Additional Information cc w/encl: Distribution via ListServ DISTRIBUTION:
PUBLIC LPL3-1 R/F RidsNrrDorlLpl3-1 Resource RidsNrrDorlDpr Resource RidsNrrPMPrairielsland Resource RidsNrrLABTully Resource RidsAcrsAcnw_MailCTR Resource GCheruvenki, NRR RidsRgn3MailCenter Resource RidsNrrDeEvib Resource ADAMS Accession Number: ML13130A144
- via memo OFFICE NRR/LPL3-1/PM NRR/LPL3-21LA NRRlEVIB/BC NRRlLPL3-1/BC NRRlLPL3*1/PM NAME TWengert SRohrer SRosenberg* RCarlson TWengert DATE 05/17/13 05/13/13 04/15/13 05/24/13 05/24/13 OFFICIAL RECORD COpy