ML091380115: Difference between revisions
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Inspections for stress corrosion cracking will be performed by | Inspections for stress corrosion cracking will be performed by | ||
visual examination with a magnified resolution as described in 10 CFR 50.55a(b)(2)(xxi)(A) or with ultrasonic methods. | visual examination with a magnified resolution as described in 10 CFR 50.55a(b)(2)(xxi)(A) or with ultrasonic methods. | ||
[Revised in letter dated 2/6/2009 in response to RAI B2.1.22-1 Follow Up question] | [Revised in letter dated 2/6/2009 in response to RAI B2.1.22-1 Follow Up question] | ||
U1 - 8/9/2013 U2 - 10/29/2014 19 The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program will be enhanced as follows: Program implementing procedures will be revised to ensure the components and structures subject to inspection are clearly identified. Program inspection procedures will be enhanced to include the parameters corrosion and wear where U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 6 | U1 - 8/9/2013 U2 - 10/29/2014 19 The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program will be enhanced as follows: Program implementing procedures will be revised to ensure the components and structures subject to inspection are clearly identified. Program inspection procedures will be enhanced to include the parameters corrosion and wear where U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 6 | ||
Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S omitted.20A Metal-Enclosed Bus Program will be implemented. Program features will be as described in LRA Section B2.1.26. | Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S omitted.20A Metal-Enclosed Bus Program will be implemented. Program features will be as described in LRA Section B2.1.26. | ||
U1 - 8/9/2013 U2 - 10/29/2014 21 Number Not Used | U1 - 8/9/2013 U2 - 10/29/2014 21 Number Not Used | ||
[Revised in letter dated 3/27/2009] | [Revised in letter dated 3/27/2009] | ||
22The Nickel-Alloy Penetration Nozzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors Program will be enhanced as follows: The program will require that any deviations from implementing the appropriate required inspection methods of the NRC First Revised Order EA-03-009, "Issue of Order Establishing Interim Inspection | 22The Nickel-Alloy Penetration Nozzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors Program will be enhanced as follows: The program will require that any deviations from implementing the appropriate required inspection methods of the NRC First Revised Order EA-03-009, "Issue of Order Establishing Interim Inspection | ||
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will be entered into the Corrective Action Program and evaluated for impact on structural integrity and identification of | will be entered into the Corrective Action Program and evaluated for impact on structural integrity and identification of | ||
additional actions that may be warranted. | additional actions that may be warranted. | ||
[Added in letter dated 4/6/09 in response to Follow Up RAI B2.1.38]U1 - 8/9/2013 U2 - 10/29/2014 42 During the two consecutive refueling outages following refueling cavity leak repairs in each Unit (scheduled for refueling outages | [Added in letter dated 4/6/09 in response to Follow Up RAI B2.1.38]U1 - 8/9/2013 U2 - 10/29/2014 42 During the two consecutive refueling outages following refueling cavity leak repairs in each Unit (scheduled for refueling outages | ||
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areas where reactor cavity leakage had been observed | areas where reactor cavity leakage had been observed | ||
previously to confirm that leakage has been resolved. The inspection results will be documented. If refueling cavity leakage is again identified, the issue will be entered into the Corrective Action Program and evaluated for identification of additional actions to mitigate leakage and monitor the condition of the containment vessel and internal structures. | previously to confirm that leakage has been resolved. The inspection results will be documented. If refueling cavity leakage is again identified, the issue will be entered into the Corrective Action Program and evaluated for identification of additional actions to mitigate leakage and monitor the condition of the containment vessel and internal structures. | ||
[Added in letter dated 4/6/09 in response to Follow Up RAI B2.1.38]U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 16 Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S 43Preventive maintenance requirements will be implemented to require periodic replacement of rubber flexible hoses in the | [Added in letter dated 4/6/09 in response to Follow Up RAI B2.1.38]U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 16 Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S 43Preventive maintenance requirements will be implemented to require periodic replacement of rubber flexible hoses in the | ||
Diesel Generators and Support System that are exposed to fuel oil or lubricating oil internal environments. | Diesel Generators and Support System that are exposed to fuel oil or lubricating oil internal environments. | ||
[Added in letter dated 4/6/09 in response to RAI 3.3.2-8-1] | [Added in letter dated 4/6/09 in response to RAI 3.3.2-8-1] | ||
U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 17 1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121 April 6, 2009 L-PI-09-047 10 CFR 54 U S Nuclear Regulatory Commission | U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 17 1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121 April 6, 2009 L-PI-09-047 10 CFR 54 U S Nuclear Regulatory Commission | ||
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Administrator, Region III, USNRC License Renewal Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC | Administrator, Region III, USNRC License Renewal Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC | ||
Prairie Island Indian Communi ty ATTN: Phil Mahowald Minnesota Department of Commerce NSPM Response to Follow Up RAI B2.1.38 1Follow Up RAI B2.1.38 In letter L-PI-08-098, dated December 5, 2008, "Responses to NRC Requests for Additional Information Dated November 5, 2008 Regarding Application for Renewed Operating Licenses," the applic ant submitted responses to th e staff's RAIs. In addition, the applicant provided information during a publ ic meeting on March 2, 2009. The staff has reviewed the information provided, but needs additional information to complete its review. Please provide the following informa tion to supplement the initial response: | Prairie Island Indian Communi ty ATTN: Phil Mahowald Minnesota Department of Commerce NSPM Response to Follow Up RAI B2.1.38 1Follow Up RAI B2.1.38 In letter L-PI-08-098, dated December 5, 2008, "Responses to NRC Requests for Additional Information Dated November 5, 2008 Regarding Application for Renewed Operating Licenses," the applic ant submitted responses to th e staff's RAIs. In addition, the applicant provided information during a publ ic meeting on March 2, 2009. The staff has reviewed the information provided, but needs additional information to complete its review. Please provide the following informa tion to supplement the initial response: | ||
: a. Due to the leakage path of borated water along the bottom of the steel containment vessel, the vessel/concrete interface may remain wetted after refueling outages. During the public meet ing, the applicant discussed its plan to remove concrete from sump C, below the reactor vessel, to physically assess the condition of the steel vessel, concrete , and rebar. Please provide more detailed information about this plan including acti ons and schedule. Discuss what other actions will be taken, prior to the period of extended operation, to adequately assess the condition of the steel vessel. | : a. Due to the leakage path of borated water along the bottom of the steel containment vessel, the vessel/concrete interface may remain wetted after refueling outages. During the public meet ing, the applicant discussed its plan to remove concrete from sump C, below the reactor vessel, to physically assess the condition of the steel vessel, concrete , and rebar. Please provide more detailed information about this plan including acti ons and schedule. Discuss what other actions will be taken, prior to the period of extended operation, to adequately assess the condition of the steel vessel. | ||
: b. In the absence of a commitment to fix the leakage prior to the period of extended operation, please explain how the IWE progr am, or a plant-spec ific program, will address the leakage and will ensure that t hese aging effects will be effectively managed, especially in the inaccessible r egions, during the period of extended operation. | : b. In the absence of a commitment to fix the leakage prior to the period of extended operation, please explain how the IWE progr am, or a plant-spec ific program, will address the leakage and will ensure that t hese aging effects will be effectively managed, especially in the inaccessible r egions, during the period of extended operation. | ||
: c. Provide the chemical properties of the "white deposits" found on the concrete surfaces and address the possibilit y of calcium hydroxide Ca(OH) 2 leaching from concrete. | : c. Provide the chemical properties of the "white deposits" found on the concrete surfaces and address the possibilit y of calcium hydroxide Ca(OH) 2 leaching from concrete. | ||
NSPM Response to Follow Up RAI B2.1.38 Response to Part a | NSPM Response to Follow Up RAI B2.1.38 Response to Part a | ||
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of components serviced by closed-cycle cooling water when the systems or components are opened during scheduled maintenance or surveillance ac tivities. Inspections are performed to identify the pres ence of aging effects and to confirm the effectiveness of the chemistry controls. Visual inspection of component internals will be used to detect loss of | of components serviced by closed-cycle cooling water when the systems or components are opened during scheduled maintenance or surveillance ac tivities. Inspections are performed to identify the pres ence of aging effects and to confirm the effectiveness of the chemistry controls. Visual inspection of component internals will be used to detect loss of | ||
material and heat transfer degr adation. Enhanced visual or volumetric examination tec hniques will be used to detect cracking. | material and heat transfer degr adation. Enhanced visual or volumetric examination tec hniques will be used to detect cracking. | ||
[Revised in letter dated 1/20/2009 in response to RAI 3.3.2 01] U1 - 8/9/2013 U2 - 10/29/2014 B2.1.9 7 The Compressed Air Monitoring Program will be enhanced as follows: | [Revised in letter dated 1/20/2009 in response to RAI 3.3.2 01] U1 - 8/9/2013 U2 - 10/29/2014 B2.1.9 7 The Compressed Air Monitoring Program will be enhanced as follows: | ||
Station and Instrument Ai r System air quality will be monitored and maintained in accordance with | Station and Instrument Ai r System air quality will be monitored and maintained in accordance with | ||
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U1 - 8/9/2013 U2 - 10/29/2014 B2.1.20 17 An Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program will be implemented. Program features will be as described in LRA Section B2.1.21 U1 - 8/9/2013 U2 - 10/29/2014 B2.1.21 18 An Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program will be implemented. Program features will be as described in LRA section B2.1.22. Inspections for stress corrosion cracking will be performed by visual examination with a magni fied resolution as described in 10 CFR 50.55a(b)(2)(xxi)(A) or with ultrasonic methods. [Revised in letter dated 2/6/2009 in response to RAI B2.1.22-1 Follow Up question] | U1 - 8/9/2013 U2 - 10/29/2014 B2.1.20 17 An Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program will be implemented. Program features will be as described in LRA Section B2.1.21 U1 - 8/9/2013 U2 - 10/29/2014 B2.1.21 18 An Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program will be implemented. Program features will be as described in LRA section B2.1.22. Inspections for stress corrosion cracking will be performed by visual examination with a magni fied resolution as described in 10 CFR 50.55a(b)(2)(xxi)(A) or with ultrasonic methods. [Revised in letter dated 2/6/2009 in response to RAI B2.1.22-1 Follow Up question] | ||
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.22 19 The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program will be enhanced as follows: Program implementing procedures will be revised to ensure the components and structures subject to inspection are clearly identified. Program inspection procedures will be enhanced to include the parameters corrosion and wear where U1 - 8/9/2013 U2 - 10/29/2014 B2.1.23 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 7 Commitment Number Commitment Implementation Schedule Related LRA Section Number omitted. 20 A Metal-Enclosed Bus Program will be implemented. Program features will be as described in LRA Section B2.1.26. | U1 - 8/9/2013 U2 - 10/29/2014 B2.1.22 19 The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program will be enhanced as follows: Program implementing procedures will be revised to ensure the components and structures subject to inspection are clearly identified. Program inspection procedures will be enhanced to include the parameters corrosion and wear where U1 - 8/9/2013 U2 - 10/29/2014 B2.1.23 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 7 Commitment Number Commitment Implementation Schedule Related LRA Section Number omitted. 20 A Metal-Enclosed Bus Program will be implemented. Program features will be as described in LRA Section B2.1.26. | ||
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.26 21 Number Not Used | U1 - 8/9/2013 U2 - 10/29/2014 B2.1.26 21 Number Not Used | ||
[Revised in letter dated 3/27/2009] | [Revised in letter dated 3/27/2009] | ||
22 The Nickel-Alloy Penetration No zzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors Program will be enhanced as follows: The program will require that any deviations from implementing the appropria te required inspection methods of the NRC First Revised Order EA-03-009, "Issue of Order Establishing Interim Inspection | 22 The Nickel-Alloy Penetration No zzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors Program will be enhanced as follows: The program will require that any deviations from implementing the appropria te required inspection methods of the NRC First Revised Order EA-03-009, "Issue of Order Establishing Interim Inspection | ||
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report the revised CUFs and CUFs adjusted for environmental effects at these locations as an amendment to the PINGP LRA. Conforming changes to LR A Section 4.3.3, "PINGP EAF Results," will also be included in that amendment to reflect analysis results and remove references to stress-based | report the revised CUFs and CUFs adjusted for environmental effects at these locations as an amendment to the PINGP LRA. Conforming changes to LR A Section 4.3.3, "PINGP EAF Results," will also be included in that amendment to reflect analysis results and remove references to stress-based | ||
fatigue monitoring. | fatigue monitoring. | ||
[Added in letter dated 1/9/2009 in response to RAI 4.3.1.1-1] April 30, 2009 4.3.3 37 NSPM will revise procedures for excavation and trenching controls and archaeological, cult ural and historic resource protection to identify sensitiv e areas and provide guidance for ground-disturbing activities. The procedures will be revised to | [Added in letter dated 1/9/2009 in response to RAI 4.3.1.1-1] April 30, 2009 4.3.3 37 NSPM will revise procedures for excavation and trenching controls and archaeological, cult ural and historic resource protection to identify sensitiv e areas and provide guidance for ground-disturbing activities. The procedures will be revised to | ||
include drawings and illustrations to assist users in identifying 8/9/2013 ER 4.16.1 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 15 Commitment Number Commitment Implementation Schedule Related LRA Section Number culturally sensitive areas, and pi ctures of artifacts that are prevalent in the area of the Plant site. The revised procedures will also require training of the Site Environmental Coordinator and other personnel responsible for proper execution of excavation or other ground-disturbing activities. | include drawings and illustrations to assist users in identifying 8/9/2013 ER 4.16.1 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 15 Commitment Number Commitment Implementation Schedule Related LRA Section Number culturally sensitive areas, and pi ctures of artifacts that are prevalent in the area of the Plant site. The revised procedures will also require training of the Site Environmental Coordinator and other personnel responsible for proper execution of excavation or other ground-disturbing activities. | ||
[Added in ER revision submi tted in letter dated 3/4/2009] | |||
38 NSPM will conduct a Phase I Reconnaissance Field Survey of the disturbed areas within the Pl ant's boundaries. In addition, NSPM will conduct Phase I field surveys of areas of known archaeological sites to precisely determine their boundaries. | 38 NSPM will conduct a Phase I Reconnaissance Field Survey of the disturbed areas within the Pl ant's boundaries. In addition, NSPM will conduct Phase I field surveys of areas of known archaeological sites to precisely determine their boundaries. | ||
NSPM will use the results of these surveys to designate areas for archaeological protection. | NSPM will use the results of these surveys to designate areas for archaeological protection. | ||
[Added in ER revision submi tted in letter dated 3/4/2009] 8/9/2013 ER 4.16.2 39 NSPM will prepare, maintain and implement a Cultural Resources Management Plan (CRMP) to protect significant historical, archaeological, and cultural resources that may currently exist on the Plant site. In connection with the preparation of the CRMP, NSPM will conduct botanical surveys to identify culturally and medicinally important spec ies on the Plant site, and incorporate provisions to protect such plants into the | |||
CRMP. [Added in ER revision submi tted in letter dated 3/4/2009] 8/9/2013 ER 4.16.2 40 NSPM will consult with a qualif ied archaeologist prior to conducting any ground-disturbing activity in any area | CRMP. [Added in ER revision submi tted in letter dated 3/4/2009] 8/9/2013 ER 4.16.2 40 NSPM will consult with a qualif ied archaeologist prior to conducting any ground-disturbing activity in any area | ||
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designated as undisturbed and in any disturbed area that is | designated as undisturbed and in any disturbed area that is | ||
described as potentially containi ng archaeological resources (as determined by the Phase I Reconnaissance Field Survey 8/9/2013 ER 4.16.2 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 16 Commitment Number Commitment Implementation Schedule Related LRA Section Number discussed in Commitment Number 38). | described as potentially containi ng archaeological resources (as determined by the Phase I Reconnaissance Field Survey 8/9/2013 ER 4.16.2 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 16 Commitment Number Commitment Implementation Schedule Related LRA Section Number discussed in Commitment Number 38). | ||
[Added in ER revision submi tted in letter dated 3/4/2009] | |||
41 During the first refueling outage following refueling cavity leak repairs in each Unit (scheduled for refueling outages 1R26 and 2R26), concrete will be removed from the sump C pit to expose an area of the containment vessel bottom head. Visual | 41 During the first refueling outage following refueling cavity leak repairs in each Unit (scheduled for refueling outages 1R26 and 2R26), concrete will be removed from the sump C pit to expose an area of the containment vessel bottom head. Visual | ||
examination and ultrasonic thickness measurement will be performed on the portions of the containment vessels exposed by the excavations. An assessment of the condition of exposed concrete and rebar will also be performed. Degradation observed in the exposed containm ent vessel, concrete or rebar will be entered into the Co rrective Action Program and evaluated for impact on structural integrity and identification of additional actions that may be warranted. | examination and ultrasonic thickness measurement will be performed on the portions of the containment vessels exposed by the excavations. An assessment of the condition of exposed concrete and rebar will also be performed. Degradation observed in the exposed containm ent vessel, concrete or rebar will be entered into the Co rrective Action Program and evaluated for impact on structural integrity and identification of additional actions that may be warranted. | ||
[Added in letter dated 4/6/09 in response to Follow Up RAI | |||
B2.1.38] U1 - 8/9/2013 U2 - 10/29/2014 B2.1.38 42 During the two consecutive refueling outages following refueling cavity leak repairs in each Unit (scheduled for refueling outages | B2.1.38] U1 - 8/9/2013 U2 - 10/29/2014 B2.1.38 42 During the two consecutive refueling outages following refueling cavity leak repairs in each Unit (scheduled for refueling outages | ||
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1R26 and 2R26), visual inspecti ons will be performed of the areas where reactor cavity leakage had been observed | 1R26 and 2R26), visual inspecti ons will be performed of the areas where reactor cavity leakage had been observed | ||
previously to confirm that l eakage has been resolved. The inspection results will be documented. If refueling cavity leakage is again identified, the issue will be entered into the Corrective Action Program and evaluated for identification of additional actions to mitigate leakage and monitor the condition of the containment vessel and internal structures. | previously to confirm that l eakage has been resolved. The inspection results will be documented. If refueling cavity leakage is again identified, the issue will be entered into the Corrective Action Program and evaluated for identification of additional actions to mitigate leakage and monitor the condition of the containment vessel and internal structures. | ||
[Added in letter dated 4/6/09 in response to Follow Up RAI | |||
B2.1.38] U1 - 8/9/2013 U2 - 10/29/2014 B2.1.38 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 17 Commitment Number Commitment Implementation Schedule Related LRA Section Number 43 Preventive maintenance requirements will be implemented to require periodic replacement of rubber flexible hoses in the Diesel Generators and Support System that are exposed to fuel oil or lubricating oil internal environments. | B2.1.38] U1 - 8/9/2013 U2 - 10/29/2014 B2.1.38 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 17 Commitment Number Commitment Implementation Schedule Related LRA Section Number 43 Preventive maintenance requirements will be implemented to require periodic replacement of rubber flexible hoses in the Diesel Generators and Support System that are exposed to fuel oil or lubricating oil internal environments. |
Revision as of 23:39, 11 July 2019
ML091380115 | |
Person / Time | |
---|---|
Site: | Prairie Island |
Issue date: | 04/06/2009 |
From: | - No Known Affiliation |
To: | Office of Nuclear Reactor Regulation |
References | |
Download: ML091380115 (66) | |
Text
1 PrairieIslandNPEm Resource From: Travers, Greg [Greg.Travers@xenuclear.com]
Sent: Monday, April 06, 2009 5:57 PM To: Richard.Plasse@nrc.gov
Subject:
FW: PINGP Letter Responding to Follow Up RAI B2.1.38 Attachments:
20090406 Letter Responding to Follow Up RAI B2.1.38 et.al.pdf; 20090406 Letter Responding to Follow Up RAI B2.1.38 et.al.docSent again due to error message received upon first attempt. Please confirm receipt.
From: Travers, Greg Sent: Monday, April 06, 2009 4:51 PM To: 'Richard.Plasse@nrc.gov; 'Nathan.Goodman@nrc.gov' Cc: Eckholt, Gene F.; Davis, Marlys E.
Subject:
PINGP Letter Responding to Follow Up RAI B2.1.38 Attached are pdf and WORD versions of a letter responding to Follow Up RAI B2.1.38, RAI 3.3.2-8-1, and other follow up questions raised in a conference call on March 30, 2009.
Let me know if you have any questions.
Greg Travers 651-388-1121 Hearing Identifier: Prairie_Island_NonPublic Email Number: 1023 Mail Envelope Properties (9FA1D9F2F220C04F95D9394E3CF02DAB0163B2BE)
Subject:
FW: PINGP Letter Responding to Follow Up RAI B2.1.38 Sent Date: 4/6/2009 5:57:13 PM Received Date: 4/6/2009 5:57:22 PM From: Travers, Greg Created By: Greg.Travers@xenuclear.com Recipients: "Richard.Plasse@nrc.gov" <'Richard.Plasse@nrc.gov'>
Tracking Status: None
Post Office: enex02.ft.nmcco.net Files Size Date & Time MESSAGE 592 4/6/2009 5:57:22 PM 20090406 Letter Responding to Follow Up RAI B2.1.38 et.al.pdf 801210 20090406 Letter Responding to Follow Up RAI B2.1.38 et.al.doc 395328
Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date: Recipients Received:
1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121 April 6, 2009 L-PI-09-047 10 CFR 54 U S Nuclear Regulatory Commission
ATTN: Document Control Desk
Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306
License Nos. DPR-42 and DPR-60 Responses to NRC Requests for Additional Information Dated March 31, 2009 and Follow Up Questions Regarding Application for Renewed Operating Licenses By letter dated April 11, 2008, Northern States Power Company, a Minnesota Corporation, (NSPM) submitted an Application for Renewed Operating Licenses (LRA) for the Prairie Island Nuclear Generating Plant (PINGP) Units 1 and 2. In a letter dated November 5, 2008, the NRC transmitted Request for Additional Information (RAI)
AMP-B2.1.38-2 regarding that application. NSPM responded to that RAI in a letter
dated December 5, 2008, and provided additional information in a public meeting on March 2, 2009. Subsequently, in a letter dated March 31, 2009, the NRC provided a
follow up RAI related to the previous response and the meeting. Enclosure 1 of this
letter provides the NSPM response to that follow up RAI.
In the letter dated March 31, 2009, the NRC also transmitted RAI 3.3.2-8-1 related to Natural Rubber in Fuel Oil and Lubricating Oil. Enclosure 2 provides the NSPM response to RAI 3.3.2-8-1.
In a conference call on March 30, 2009, the NRC raised follow up questions about the LRA and certain previous RAI responses. Enclosure 3 of this letter provides a summary
of each follow up question followed by the NSPM response. provides the list of License Renewal Commitments, updated to reflect NSPM correspondence to date.
If there are any questions or if additional information is needed, please contact Mr. Eugene Eckholt, License Renewal Project Manager. Summary of Commitments This letter contains three new License Renewal commitments.
NSPM Response to Follow Up RAI B2.1.38 1Follow Up RAI B2.1.38 In letter L-PI-08-098, dated December 5, 2008, "Responses to NRC Requests for Additional Information Dated November 5, 2008 Regarding Application for Renewed
Operating Licenses," the applicant submitted responses to the staff's RAIs. In addition, the applicant provided information during a public meeting on March 2, 2009. The staff has reviewed the information provided, but needs additional information to complete its
review. Please provide the following information to supplement the initial response: a. Due to the leakage path of borated water along the bottom of the steel containment vessel, the vessel/concrete interface may remain wetted after
refueling outages. During the public meeting, the applicant discussed its plan to remove concrete from sump C, below the reactor vessel, to physically assess the condition of the steel vessel, concrete, and rebar. Please provide more detailed information about this plan including actions and schedule. Discuss what other actions will be taken, prior to the period of extended operation, to adequately
assess the condition of the steel vessel. b. In the absence of a commitment to fix the leakage prior to the period of extended operation, please explain how the IWE program, or a plant-specific program, will address the leakage and will ensure that these aging effects will be effectively
managed, especially in the inaccessible regions, during the period of extended
operation.c. Provide the chemical properties of the "white deposits" found on the concrete surfaces and address the possibility of calcium hydroxide Ca(OH) 2 leaching from concrete.NSPM Response to Follow Up RAI B2.1.38 Response to Part aLeak repairs for the Unit 1 refueling cavity are planned for refueling outage 1R26 (fall 2009). Leak repairs for the Unit 2 refueling cavity are planned for refueling outage 2R26 (spring 2010). During the next refueling outages following leak repairs (projected to be
spring 2011 and spring 2012, respectively), removal of concrete from each sump C pit will be performed to expose an area near the center of the inside surface of the containment vessel bottom head of each Unit.Site drawings indicate that there is
approximately a 16 inch thickness of reinforced concrete between the bottom of the sump C pit and the containment vessel. The method of concrete removal has not been determined, but will likely include core drilling or a similar process followed by manual chipping to avoid damage to the containment vessel. The concrete will be removed primarily to allow visual and UT thickness examination of a sample location in each
containment vessel bottom head. A secondary purpose is to provide access for an attempt to evacuate residual water that may be present after leak mitigation.A sample of the concrete excavated from close proximity to each containment vessel will be qualitatively assessed for strength (e.g., does not crush easily, no signs of NSPM Response to Follow Up RAI B2.1.38 2 porosity with fine aggregate particles easily dislodged, etc.) and will also be chemically
analyzed for changes caused by exposure to borated water leakage. Water that seeps into the exploratory opening, if any, will be sampled and analyzed for pH and ionic species. Reinforcing bar exposed by the excavation will be visually examined for indications of degradation. Visual and UT examinations will be performed on the portion of each containment vessel exposed by the excavation. Degradation observed in the
exposed containment vessel, concrete or rebar will be entered into the Corrective Action Program and evaluated for impact on structural integrity and identification of additional actions (e.g., inspection of additional locations) that may be warranted.
In addition to the physical actions described above, the following evaluations are planned:complete an evaluation of potential degradation to date of the containment
vessel, concrete and reinforcing steel perform a margin assessment of the containment vessel and interior structures, including the area around the transfer tube, to determine allowable degradation NSPM expects that the refueling cavity leak repairs in outages 1R26 and 2R26 will be successful. This will be confirmed during the two refueling outages following the repairs in each Unit by specifically inspecting the areas where reactor cavity leakage had been
observed previously. The inspection results will be documented. During subsequent outages, continued monitoring for leakage and degradation will be performed through the periodic examinations of the containment vessels and interior structures performed
in accordance with the ASME Section XI, Subsection IWE Program and the Structures
Monitoring Program. Following the leak repairs, if continued monitoring again identifies active refueling cavity leakage, further action will be taken as described in Part b below.
Response to Part bAs discussed in Part a above, NSPM expects to stop the refueling cavity leakage. However, if the repairs do not completely stop the leakage, NSPM will continue to take
actions to eliminate the leakage and will continue to monitor for degradation.
Any additional actions to be taken would be determined in accordance with the Corrective Action Program evaluation of the specific locations and quantities of
observed leakage. Actions may include removal of additional grout or concrete, visual
and UT thickness examinations from the interior of the containment vessel at exposed
locations, and UT thickness examinations from the annulus around the exterior of the
containment vessel. Examinations may be repeated at intervals commensurate with the
safety significance of any identified degradation and the extent of residual leakage, if any. Actions may also include testing of concrete and rebar in areas judged to have the potential for unacceptable degradation from borated water leakage. NSPM Response to Follow Up RAI B2.1.38 3Monitoring of the containment vessel and interior structures will be performed, at a minimum, in accordance with the ASME Section XI, Subsection IWE Program and the Structures Monitoring Program. Program requirements will be augmented as appropriate for the specific leakage or degradation observed, based on good
engineering practice.As discussed in LRA Section B2.1.4, the ASME Section XI, Subsection IWE Program provides for condition monitoring of Class MC pressure-retaining components by performing periodic visual examinations (general, VT-3, VT-1) of the Class MC components and their related items. Visual (VT-1) or volumetric examinations, as
applicable, are performed on components that require augmented examination.
Indications of leakage that affects Class MC pressure-retaining components would be
documented in the Corrective Action Program for evaluation and assignment of
appropriate corrective actions. When conditions are identified in accessible areas that
could indicate the presence of, or result in, degradation in inaccessible areas, the program requires an evaluation of the acceptability of affected inaccessible areas. Such
evaluations of inaccessible areas are required to be reported to the NRC in the ISI
summary report. In accordance with IWA-6000, the information to be reported includes a description of the type and estimated extent of degradation, and conditions that led to
degradation; an evaluation of each affected area and the result of the evaluation; and a
description of the necessary actions.As discussed in LRA Section B2.1.38, the Structures Monitoring Program is a condition monitoring program that manages aging effects in structures, supports and structural components within the scope of License Renewal. This includes containment interior
structures. The program performs periodic visual inspections to monitor the condition of structures, supports and components with respect to established acceptance criteria to ensure that degraded conditions are identified, evaluated, and, when necessary, corrected such that there is no loss of intended function. Inspections performed by the Structures Monitoring Program would be expected to identify conditions indicative of potential concrete or rebar degradation, including signs of leakage, as well as cracks, calcium deposits, spalling, and rust stains.
Response to Part c A task has been entered into the site Corrective Action Program to collect and chemically analyze deposits in areas showing signs of leakage in each Unit's
regenerative heat exchanger room and sump C, as applicable. The conditions
observed for each deposit, including photographs of typical leakage areas and an estimated volume of the deposit, will be documented in the Corrective Action Program in order to qualitatively assess the extent of the affected areas. The results of the analyses will also be documented in the Corrective Action Program. Records will be available on site for NRC review.
License Renewal CommitmentsThe response to this RAI contains two new License Renewal Commitments to be completed prior to entry of each Unit into the Period of Extended Operation. NSPM Response to Follow Up RAI B2.1.38 4New License Renewal Commitment Number 41 reads as follows: During the first refueling outage following refueling cavity leak repairs in each Unit (scheduled for refueling outages 1R26 and 2R26), concrete will be removed from the sump C pit to expose an area of the containment vessel bottom head. Visual
examination and ultrasonic thickness measurement will be performed on the
portions of the containment vessels exposed by the excavations. An assessment of
the condition of exposed concrete and rebar will also be performed. Degradation observed in the exposed containment vessel, concrete or rebar will be entered into the Corrective Action Program and evaluated for impact on structural integrity and
identification of additional actions that may be warranted.New License Renewal Commitment Number 42 reads as follows: During the two consecutive refueling outages following refueling cavity leak repairs in each Unit (scheduled for refueling outages 1R26 and 2R26), visual inspections will be performed of the areas where reactor cavity leakage had been observed previously to confirm that leakage has been resolved. The inspection results will be
documented. If refueling cavity leakage is again identified, the issue will be entered into the Corrective Action Program and evaluated for identification of additional actions to mitigate leakage and monitor the condition of the containment vessel and
internal structures.
NSPM Response to RAI 3.3.2-8-1 1 RAI 3.3.2-8-1 Natural Rubber in Fuel Oil and Lubricating OilIn AMR Table 3.3.2-8, there is an AMR line item for Flex Connections/Pressure Boundary/Natural Rubber/Fuel Oil (int) and Lubricating Oil (int)/none/none/ with plant
specific note G, 313.
P.A. Schweitzer, "Corrosion Resistance Tables - Metals, Nonmetals, Coatings, Mortars, Plastics, Elastomers and Linings, and Fabrics," Fourth Edition, Part B, Marcel Dekker, Copyright 1995 discuss natural rubber and EPDM and states that they are not resistant
to fuel oil or lubricating oil. Please provide a justification for using natural rubber in fuel oil and lubricating oil with no aging effects and no need for an aging management program.
NSPM Response to RAI 3.3.2-8-1 During the AMR process certain rubber hoses were conservatively identified as being made of Natural Rubber when the actual materials could not be readily identified.
Natural rubbers are not resistant to fuel oil or lubricating oil, and only qualified
elastomers, that are resistant, are used for fuel oil or lubricating oil service. In general, the rubber flexible connections in the Diesel Generators and Support System are constructed of nitrile rubber or are lined with teflon, both of which have outstanding resistance to oils. Regardless, based on vendor recommendations, NSPM has
determined that the rubber flexible hoses in the Diesel Generators and Support System
that are exposed to Fuel Oil or Lubricating Oil internal environments should be periodically replaced. Therefore, these hoses are not subject to aging management
review, and need not be addressed in the LRA.
Accordingly, a new License Renewal Commitment Number 43 is added to read as follows:Preventive maintenance requirements will be implemented to require periodic replacement of rubber flexible hoses in the Diesel Generators and Support System that are exposed to fuel oil or lubricating oil internal environments. Conforming changes are made to the LRA as follows:
In LRA Table 3.3.2-8, Auxiliary Systems - Diesel Generators and Support System - Summary of Aging Management Evaluation, on Page 3.3-156, the line items for Flex Connections / Natural Rubber / Fuel Oil (Int) and Lubricating Oil (Int), are deleted. Additionally, the flexible connections for the lubricating oil and fuel oil lines on the diesel driven cooling water pumps were determined to be constructed from stainless steel. To address aging management review of the stainless steel flexible connections on the
diesel driven cooling water pumps, the following LRA changes are made.
NSPM Response to RAI 3.3.2-8-1 2In LRA Table 3.3.2-8, Auxiliary Systems - Diesel Generators and Support System -
Summary of Aging Management Evaluation, on Page 3.3-153, the following line items
are added: Component Type IntendedFunction MaterialEnvironment Aging Effect RequiringManagement Aging Management Programs NUREG-1801 Volume 2 Line Item Table 1 Item Notes FlexConnections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material -Crevice Corrosion Fuel Oil Chemistry Program VII.H2-16 3.3.1-32 B FlexConnections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material -Crevice Corrosion One-Time Inspection Program VII.H2-16 3.3.1-32 A FlexConnections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material -
MIC Fuel Oil Chemistry Program VII.H2-16 3.3.1-32 B FlexConnections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material -
MIC One-Time Inspection Program VII.H2-16 3.3.1-32 A FlexConnections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material -
PittingCorrosion Fuel Oil Chemistry Program VII.H2-16 3.3.1-32 B FlexConnections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material -
PittingCorrosion One-Time Inspection Program VII.H2-16 3.3.1-32 A FlexConnections Pressure Boundary Stainless SteelLubricating Oil (Int)
Loss of Material -Crevice Corrosion Lubricating Oil Analysis Program VII.H2-17 3.3.1-33 A FlexConnections Pressure Boundary Stainless SteelLubricating Oil (Int)
Loss of Material -Crevice Corrosion One-Time Inspection Program VII.H2-17 3.3.1-33 A FlexConnections Pressure Boundary Stainless SteelLubricating Oil (Int)
Loss of Material -
PittingCorrosion Lubricating Oil Analysis Program VII.H2-17 3.3.1-33 A FlexConnections Pressure Boundary Stainless SteelLubricating Oil (Int)
Loss of Material -
PittingCorrosion One-Time Inspection Program VII.H2-17 3.3.1-33 A NSPM Responses to NRC Follow Up Questions 1B2.1.19 Fuel Oil Chemistry Program Follow Up Question In the 12/18/08 Response to NRC RAI B2.1.19-3, NSPM provided justification for the Exception to NUREG-1801 which states that day tanks and clean fuel oil leakage collection tanks are not subject to periodic draining, cleaning, and internal inspection.
The justification provided was based upon the high turnover rate for the fuel in the day
tanks and the close control of the fuel oil quality in the storage tanks that comprise the sources for the day and clean leakage collection tanks. In the 3/30/09 conference call, the NRC expressed concern that no monitoring of the tank condition was being performed, and requested additional explanation as to how aging effects will be adequately managed in the day tanks and clean fuel oil collection tanks.
The diesel fuel oil day tanks and clean fuel oil leakage collection tanks are accessible for external UT examination. In order to confirm the continued effectiveness of oil quality control and to verify the absence of aging effects, one-time inspections using
ultrasonic thickness measurements will be performed on selected day tanks and clean fuel oil leakage collection tanks prior to the period of extended operation as part of the
One-Time Inspection Program. Specific locations to be selected include: An external UT on select bottom locations on four of the seven diesel fuel oil day tanks. An external UT on select bottom locations of one of the two D1/D2 clean fuel oil leakage collection tanks. The inspection locations identified above are in addition to those previously listed in the NSPM Response to RAI B2.1.19-3 (12/18/08 letter).
The previous response incorrectly stated there were four Diesel Generator D1/D2 and four Diesel Generator D5/D6 fuel oil leakage collection tanks managed by the Fuel Oil Chemistry Program. However, as a result of recent plant walk downs, it has been determined that there are only two clean fuel oil leakage collection tanks, both associated with D1/D2. The Unit 1 D1/D2 Diesel Generator Clean Fuel Tanks are the
only fuel oil leakage collection tanks within the scope of the Fuel Oil Chemistry
Program. These Clean Fuel Tanks collect clean fuel from the injector drain headers and return the fuel to the day tank with the clean fuel return pump. The vented
collection header is continuously piped from each injector to the Clean Fuel Tank.
These fuel oil leakage collection tanks have a capacity of approximately 5 gallons each.
These leakage tanks are evaluated in PINGP LRA Table 3.3.2-8, Auxiliary Systems -
Diesel Generators and Support System - Summary of Aging Management Evaluation, on Page 3.3-179. The six remaining fuel oil leakage collection tanks, two associated with D1/D2 and four associated with D5/D6, are contaminated fuel oil leakage collection tanks that collect drops of fuel from pump and nozzle leaks. The four D5/D6 fuel oil leakage collection tanks in turn drain to Dirty Oil Tanks which also collect floor drains from the D5/D6 Fuel
Oil Day Tank and Lube Oil Tank rooms. These fuel oil leakage collection tanks and Dirty Oil Tanks are not pressurized.The tanks are periodically drained, and the NSPM Responses to NRC Follow Up Questions 2collected fluid is disposed of as waste.Due to the potential for contamination, the internal environment for these six fuel oil leakage collection tanks and the Dirty Oil Tanks are evaluated as Raw Water. The tanks are evaluated in LRA Table 3.3.2-8, Auxiliary Systems - Diesel Generators and Support System - Summary of Aging
Management Evaluation, on Page 3.3-181. Plant Specific Note 320 is applied to these
line items and states, "The environment for this component is evaluated as Raw Water
and is essentially waste water or a potential mixture of water and oil." Aging effects are
managed by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting
Components Program.
During the review of LRA Table 3.3.2-8, it was noted that the line items for Piping / Fittings and Valve Bodies did not fully account for the piping and valves that are used
for collection of dirty oil in the leakage collection tanks and are managed by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program. An entry for Sight Glasses in a Raw Water (Int) environment has also been
added. For completeness, LRA Table 3.3.2-8 is hereby revised to incorporate the
appropriate line items. In LRA Table 3.3.2-8, Auxiliary Systems - Diesel Generators and Support System - Summary of Aging Management Evaluation, the following line items are added: Component Type IntendedFunction MaterialEnvironment Aging Effect RequiringManagement Aging Management Programs NUREG-1801 Volume 2 Line Item Table 1 Item Notes Piping / Fittings Pressure Boundary Carbon SteelRaw Water (Int)Loss of Material -Crevice Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-19 3.3.1-76 E, 320 Piping / Fittings Pressure Boundary Carbon SteelRaw Water (Int)Loss of Material -Galvanic Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-19 3.3.1-76 E, 310, 320 Piping / Fittings Pressure Boundary Carbon SteelRaw Water (Int)Loss of Material -General Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-19 3.3.1-76 E, 320 Piping / Fittings Pressure Boundary Carbon SteelRaw Water (Int)Loss of Material -
MICInspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-19 3.3.1-76 E, 320 Piping / Fittings Pressure Boundary Carbon SteelRaw Water (Int)Loss of Material -
PittingCorrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-19 3.3.1-76 E, 320 NSPM Responses to NRC Follow Up Questions 3Component Type IntendedFunction Material Environment Aging Effect RequiringManagement Aging Management Programs NUREG-1801 Volume 2 Line Item Table 1 Item Notes Piping / Fittings Pressure Boundary Copper AlloyRaw Water (Int)Loss of Material -Crevice Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Piping / Fittings Pressure Boundary Copper AlloyRaw Water (Int)Loss of Material -
MICInspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Piping / Fittings Pressure Boundary Copper AlloyRaw Water (Int)Loss of Material -
PittingCorrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Piping / Fittings Pressure Boundary Copper AlloyRaw Water (Int)Loss of Material -
SelectiveLeaching Selective Leaching of Materials Program VII.H2-13 3.3.1-84 B, 320 SightGlasses Pressure Boundary GlassRaw Water (Int)None None VII.J-11 3.3.1-93 A, 320 Valve Bodies Pressure Boundary Brass Raw Water (Int)Loss of Material -Crevice Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Valve Bodies Pressure Boundary Brass Raw Water (Int)Loss of Material -
MICInspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Valve Bodies Pressure Boundary Brass Raw Water (Int)Loss of Material -
PittingCorrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Valve Bodies Pressure Boundary Brass Raw Water (Int)Loss of Material -
SelectiveLeaching Selective Leaching of Materials Program VII.H2-13 3.3.1-84 B, 320 Valve Bodies Pressure Boundary Bronze Raw Water (Int)Loss of Material -Crevice Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Valve Bodies Pressure Boundary Bronze Raw Water (Int)Loss of Material -
MICInspection of Internal Surfaces in Miscellaneous Piping and Ducting VII.C1-9 3.3.1-81 E, 320 NSPM Responses to NRC Follow Up Questions 4Component Type IntendedFunction Material Environment Aging Effect RequiringManagement Aging Management Programs NUREG-1801 Volume 2 Line Item Table 1 Item NotesComponents Program Valve Bodies Pressure Boundary Bronze Raw Water (Int)Loss of Material -
PittingCorrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320B2.1.19 Fuel Oil Analysis Follow Up Question In the 3/30/09 conference call, the NRC acknowledged PINGP's commitment to initiate annual testing of fuel oil for particulates, but the GALL recommends quarterly testing.
The reviewer indicated that annual particulate testing may be sufficient as long as
sediment and particulate testing results continue to meet acceptance standards. NSPM was requested to clarify whether the annual test frequency for particulates would be
increased in the event significant sediment or particulates are detected. As described in the Program Basis Document for the PINGP Fuel Oil Chemistry Program, when fuel oil analysis limits are not met, such as for sediment content or particulate contamination, the condition is entered into the Corrective Action Program
for evaluation and assignment of corrective actions. Actions may include replacement of diesel fuel oil, more frequent sampling, use of fuel oil additives such as biocides, or tank draining, cleaning and inspection, depending on the significance of the issue and
required actions to correct the condition. LRA Section 4.7.5 Turbine Missile Analysis TLAA Follow Up Question In the 3/30/09 conference call, the NRC questioned whether the analysis of a turbine rotor failure leading to turbine missiles is a TLAA as discussed in LRA Section 4.7.5.
This section primarily references a discussion in the USAR, but the USAR does not
make it clear that the analysis is part of the CLB and why it is considered a TLAA. The
reviewer requested that PINGP clarify whether the issue should be classified as a TLAA. If it is concluded to be a TLAA, specific CLB information which defines the safety
bases for turbine rotor failure (inspection frequency) should be provided.When the PINGP LRA was prepared, USAR Sections 11.2.3.2, Turbine Overspeed Control, and 12.2.7.1, Probability of Damage to Safeguard Equipment from Turbine
Missiles, were consulted. Section 12.2.7.1 contained a brief discussion that the probabilities of missile ejection at running speed and design overspeed meet the criteria
of the SRP "... through the remaining Operating License period in 2013 [Unit 2: 2014]."
Based on this statement, the analysis of turbine missiles was designated a TLAA, and a restatement of the USAR discussion was provided as a TLAA discussion in Section NSPM Responses to NRC Follow Up Questions 5 4.7.5 of the LRA. Upon further review of the reference documents underlying the USAR
discussion, NSPM has concluded that the analysis of the probability of damage to safeguard equipment from turbine missiles does not meet the criteria for a TLAA, and should be removed from the LRA.
The NRC approved the analysis methodology for determining the total probability of turbine missile ejection from an overspeed condition (WCAP-11525) in conjunction with PINGP License Amendments 86 (Unit 1) and 79 (Unit 2) issued by an NRC letter dated February 7, 1989. WCAP-11525, "Probabilistic Evaluation of Reduction in Turbine
Valve Test Frequency," considers the probabilities of rotor burst during design overspeed, intermediate overspeed, and destructive overspeed conditions. The total annual probability of turbine missile ejection is the sum of the annual probabilities of
failure at design overspeed, intermediate overspeed, and destructive overspeed conditions. The probability of failure at design and intermediate overspeed equals the
probability of turbine overspeed multiplied by the conditional probability of failure at overspeed.
The probability of turbine overspeed is a function of the reliability of the turbine control system and the turbine valves. The reliability of the control system and valves are a function of the turbine valve test interval and the design configuration. Valve operating
experience is a factor in determining this probability.The conditional probability of missile ejection given a design or intermediate overspeed condition is based on consideration of damage mechanisms (e.g., SCC and fatigue) and
the length of time the rotor is in service following its last inspection as discussed in
WCAP-11525, Section 8.0. The rotor inspection interval selected in WCAP-11525 for evaluation of the fully integral low pressure turbine rotor is 10 years (WCAP-11525, Section 8.2, Page 8-3), even though evaluations (e.g., Figure 12.2-38 of the PINGP
USAR from Reference 52 of the USAR) showed very low failure probabilities at 30
years with no inspection.
In accordance with Section 8.0 of WCAP-11525, neither the turbine valve test interval (e.g., 12 months) nor the rotor inspection interval (e.g.,10 years) involves a time-limited assumption defined by the current operating term of the plant. Therefore, the analysis
of turbine missiles discussed in USAR Section 12.2.7.1, including the effects of turbine
valve test frequency and turbine rotor inspection interval, are not TLAAs. Since this
issue does not meet the definition of a TLAA, it need not be discussed in the LRA.
Therefore, LRA Section 4.7.5 can be withdrawn. Conforming LRA changes to reflect withdrawal of Section 4.7.5 are as follows:
In LRA Section 3.4.2.3 on Page 3.4-21, the last bullet, Probability of Damage to Safeguards Equipment from Turbine Missiles, is deleted in its entirety.
In LRA Table 4.1-1 on Page 4.1-7, the last line item, Probability of Damage to Safeguards Equipment from Turbine Missiles, is deleted in its entirety. In LRA Section 4.7 on Page 4.7-1, the sentence is revised in its entirety to read: NSPM Responses to NRC Follow Up Questions 6Other PINGP-specific TLAAs include leak-before-break (LBB) analyses, underclad
cracking, RCP flywheel, and fatigue analysis of cranes. LRA Section 4.7.5 on Pages 4.7-4 through 4.7-6 is deleted in its entirety.
LRA Section A4.9 on Pages A-25 and A-26 is deleted in its entirety. License Renewal Commitment List 17 Pages Prairie Island Nuclear Generating Plant License Renewal Commitments The following table provides the list of commitments included in the Application for Renewed Operating Li c Island Nuclear Generating Plant (PINGP) Units 1 and 2, as updated in subsequent correspondence.
The commitments in this list are anticipated to be the final commitments which will be confirmed in the NR CReport (SER) for the renewed operating licenses. These commitments, as confirmed in the SER, will bec oissuance of the renewed licenses. In addition, as stated in the LRA, the final commitments will be incorpo r Safety Analysis Report (USAR).
Commitment Number Commitment Implementation Schedule S 1Each year, following the submittal of the PINGP License Renewal Application and at least three months before the scheduled completion of the NRC review, NMC will submit amendments to the PINGP application pursuant to 10 CFR
54.21(b). These revisions will identify any changes to the
Current Licensing Basis that materially affect the contents of
the License Renewal Application, including the USAR supplements.
12 months after LRA submittal
date and at least
3 months before
completion of
NRC review 2Following the issuance of the renewed operating license, the summary descriptions of aging management programs and
TLAAs provided in Appendix A, and the final list of License Renewal commitments, will be incorporated into the PINGP
USAR as part of a periodic USAR update in accordance with 10
CFR 50.71(e). Other changes to specific sections of the PINGP USAR necessary to reflect a renewed operating license will also
be addressed at that time.
First USAR update in accordance with 10 CFR 50.71(e) following issuance
of renewed
operating licenses 3An Aboveground Steel Tanks Program will be implemented.
Program features will be as described in LRA Section B2.1.2.
U1 - 8/9/2013 U2 - 10/29/2014 4Procedures for the conduct of inspections in the External Surfaces Monitoring Program, Structures Monitoring Program, Buried Piping and Tanks Inspection Program, and the RG
1.127 Inspection of Water-Control Structures Associated with U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 1
Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S Nuclear Power Plants Program will be enhanced to include
guidance for visual inspections of installed bolting. 5A Buried Piping and Tanks Inspection Program will be implemented. Program features will be as described in LRA Section B2.1.8.
U1 - 8/9/2013 U2 - 10/29/2014 6 The Closed-Cycle Cooling Water System Program will be enhanced to include periodic inspection of accessible surfaces
of components serviced by closed-cycle cooling water when the systems or components are opened during scheduled
maintenance or surveillance activities. Inspections are
performed to identify the presence of aging effects and to confirm the effectiveness of the chemistry controls. Visual
inspection of component internals will be used to detect loss of
material and heat transfer degradation. Enhanced visual or
volumetric examination techniques will be used to detect
cracking.[Revised in letter dated 1/20/2009 in response to RAI 3.3.2 01]U1 - 8/9/2013 U2 - 10/29/2014 7 The Compressed Air Monitoring Program will be enhanced as follows: Station and Instrument Air System air quality will be monitored and maintained in accordance with the instrument air quality guidance provided in ISA
S7.0.01-1996. Particulate testing will be revised
to use a particle size methodology as specified in
ISA S7.0.01. The program will incorporate on-line dew point
monitoring.
U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 2
Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S[Revised in letter dated 2/6/2009 in response to Region III
License Renewal Inspection] 8An Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program will be
completed. Program features will be as described in LRA Section B2.1.11.
U1 - 8/9/2013 U2 - 10/29/2014 9An Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program will be implemented. Program features will be as described in LRA Section B2.1.12.
U1 - 8/9/2013 U2 - 10/29/2014 10An Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in
Instrumentation Circuits Program will be implemented.
Program features will be as described in LRA Section B2.1.13.
U1 - 8/9/2013 U2 - 10/29/2014 11 The External Surfaces Monitoring Program will be enhanced as follows: The scope of the program will be expanded as necessary
to include all metallic and non-metallic components within
the scope of License Renewal that require aging
management in accordance with this program. The program will ensure that surfaces that are
inaccessible or not readily visible during plant operations will be inspected during refueling outages. The program will ensure that surfaces that are inaccessible or not readily visible during both plant operations and refueling outages will be inspected at
intervals that provide reasonable assurance that aging U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 3
Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule Seffects are managed such that the applicable components will perform their intended function during
the period of extended operation. The program will apply physical manipulation techniques, in addition to visual inspection, to detect aging effects in
elastomers and plastics. The program will include acceptance criteria (e.g.,
threshold values for identified aging effects) to ensure
that the need for corrective actions will be identified
before a loss of intended functions. The program will ensure that program documentation
such as walkdown records, inspection results, and other records of monitoring and trending activities are auditable
and retrievable.[Revised in letter dated 2/6/2009 in response to RAI B2.1.14-1 Follow Up question] 12The Fire Protection Program will be enhanced to require periodic visual inspection of the fire barrier walls, ceilings, and floors to be performed during walkdowns at least once every
refueling cycle. [Revised in letter dated 12/5/2008 in response to RAI B2.1.15-3]
U1 - 8/9/2013 U2 - 10/29/2014 13 The Fire Water System Program will be enhanced as follows: The program will be expanded to include eight additional yard fire hydrants in the scope of the annual visual inspection and flushing activities.
U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 4
Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S The program will require that sprinkler heads that have been in place for 50 years will be replaced or a
representative sample of sprinkler heads will be tested using the guidance of NFPA 25, "Inspection, Testing and
Maintenance of Water-Based Fire Protection Systems" (2002 Edition, Section 5.3.1.1.1). Sample testing, if
performed, will continue at a 10-year interval following the initial testing.
14The Flux Thimble Tube Inspection Program will be enhanced as follows: The program will require that the interval between inspections be established such that no flux thimble tube is predicted to incur wear that exceeds the established
acceptance criteria before the next inspection. The program will require that re-baselining of the examination frequency be justified using plant-specific
wear rate data unless prior plant-specific NRC
acceptance for the re-baselining was received. If design changes are made to use more wear-resistant thimble tube materials, sufficient inspections will be conducted at an adequate inspection frequency for the new materials. The program will require that flux thimble tubes that cannot be inspected must be removed from service.
U1 - 8/9/2013 U2 - 10/29/2014 15The Fuel Oil Chemistry Program will be enhanced as follows: Particulate contamination testing of fuel oil in the eleven fuel oil storage tanks in scope of License Renewal will be
performed, in accordance with ASTM D 6217, on an U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 5
Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S annual basis. One-time ultrasonic thickness measurements will be performed at selected tank bottom and piping locations prior to the period of extended operation. 16A Fuse Holders Program will be implemented. Program features will be as described in LRA Section B2.1.20.
U1 - 8/9/2013 U2 - 10/29/2014 17 An Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program will be implemented. Program features will be as described in LRA Section B2.1.21 U1 - 8/9/2013 U2 - 10/29/2014 18 An Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program will be implemented. Program features will be as described in LRA section B2.1.22.
Inspections for stress corrosion cracking will be performed by
visual examination with a magnified resolution as described in 10 CFR 50.55a(b)(2)(xxi)(A) or with ultrasonic methods.
[Revised in letter dated 2/6/2009 in response to RAI B2.1.22-1 Follow Up question]
U1 - 8/9/2013 U2 - 10/29/2014 19 The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program will be enhanced as follows: Program implementing procedures will be revised to ensure the components and structures subject to inspection are clearly identified. Program inspection procedures will be enhanced to include the parameters corrosion and wear where U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 6
Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S omitted.20A Metal-Enclosed Bus Program will be implemented. Program features will be as described in LRA Section B2.1.26.
U1 - 8/9/2013 U2 - 10/29/2014 21 Number Not Used
[Revised in letter dated 3/27/2009]
22The Nickel-Alloy Penetration Nozzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors Program will be enhanced as follows: The program will require that any deviations from implementing the appropriate required inspection methods of the NRC First Revised Order EA-03-009, "Issue of Order Establishing Interim Inspection
Requirements for Reactor Pressure Vessel Heads at
Pressurized Water Reactors," dated February 20, 2004 (Order), as amended, will be submitted for NRC review and approval in accordance with the Order, as amended. The program will require that any deviations from implementing the required inspection frequencies mandated by the Order, as amended, will be submitted for NRC review and approval in accordance with the Order, as amended. The program will require that relevant flaw indications detected during the augmented inspections of the upper vessel head penetration nozzles will be evaluated in
accordance with the criteria provided in the letter from Mr. Richard Barrett, NRC, Office of Nuclear Reactor U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 7
Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S Regulation (NRR), Division of Engineering to Alex
Marion, Nuclear Energy Institute (NEI), dated April 11, 2003, or in accordance with NRC-approved Code Cases that incorporate the flaw evaluation procedures and criteria of the NRC's April 11, 2003, letter to NEI. The program will require that, if leakage or evidence of cracking in the vessel head penetration nozzles (including associated J-groove welds) is detected while ranked in the "Low," "Moderate," or "Replaced"
susceptibility category, the nozzles are to be immediately
reclassified to the "High" susceptibility category and the required augmented inspections for the "High" susceptibility category are to be implemented during the same outage the leakage or cracking is detected.
23 A One-Time Inspection Program will be completed. Program features will be as described in LRA Section B2.1.29.
U1 - 8/9/2013 U2 - 10/29/2014 24A One-Time Inspection of ASME Code Class 1 Small-Bore Piping Program will be completed. Program features will be as described in LRA Section B2.1.30.
U1 - 8/9/2013 U2 - 10/29/2014 25For the PWR Vessel Internals Program, PINGP commits to the following activities for managing the aging of reactor vessel internals components: Participate in the industry programs for investigating and managing aging effects on reactor internals; Evaluate and implement the results of the industry programs as applicable to the reactor internals; and Upon completion of these programs, but not less than 24 U1 - 8/9/2011 U2 - 10/29/2012 Updated through 4/6/2009 8
Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule Smonths before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC
for review and approval.
26The Reactor Head Closure Studs Program will be enhanced to incorporate controls that ensure that any future procurement of
reactor head closure studs will be in accordance with the material and inspection guidance provided in NRC Regulatory
Guide 1.65.
U1 - 8/9/2013 U2 - 10/29/2014 27The Reactor Vessel Surveillance Program will be enhanced as follows: A requirement will be added to ensure that all withdrawn and tested surveillance capsules, not discarded as of
August 31, 2000, are placed in storage for possible future reconstitution and use. A requirement will be added to ensure that in the event spare capsules are withdrawn, the untested capsules are placed in storage and maintained for future insertion.
U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 9
Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S 28 The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants Program will be enhanced as follows: The program will include inspections of concrete and steel components that are below the water line at the Screenhouse and Intake Canal. The scope will also
require inspections of the Approach Canal, Intake Canal, Emergency Cooling Water Intake, and Screenhouse
immediately following extreme environmental conditions
or natural phenomena including an earthquake, flood, tornado, severe thunderstorm, or high winds. The program parameters to be inspected will include an inspection of water-control concrete components that are below the water line for cavitation and erosion degradation. The program will visually inspect for damage such as cracking, settlement, movement, broken bolted and
welded connections, buckling, and other degraded conditions following extreme environmental conditions or natural phenomena.
U1 - 8/9/2013 U2 - 10/29/2014 29 A Selective Leaching of Materials Program will be completed.Program features will be as described in LRA B2.1.36.
U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 10 Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S 30 The Structures Monitoring Program will be enhanced as follows: The following structures, components, and component supports will be added to the scope of the inspections:
o Approach Canal o Fuel Oil Transfer House o Old Administration Building and Administration Building Addition o Component supports for cable tray, conduit, cable, tubing tray, tubing, non-ASME vessels, exchangers, pumps, valves, piping, mirror
insulation, non-ASME valves, cabinets, panels, racks, equipment enclosures, junction boxes, bus
ducts, breakers, transformers, instruments, diesel equipment, housings for HVAC fans, louvers, and
dampers, HVAC ducts, vibration isolation
elements for diesel equipment, and miscellaneous electrical and mechanical equipment items o Miscellaneous electrical equipment and instrumentation enclosures including cable tray, conduit, wireway, tube tray, cabinets, panels, racks, equipment enclosures, junction boxes, breaker housings, transformer housings, lighting
fixtures, and metal bus enclosure assemblies o Miscellaneous mechanical equipment enclosures including housings for HVAC fans, louvers, and
dampers o SBO Yard Structures and components including SBO cable vault and bus duct enclosures.
U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 11 Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S o Fire Protection System hydrant houses o Caulking, sealant and elastomer materials o Non-safety related masonry walls that support equipment relied upon to perform a function that
demonstrates compliance with a regulated
event(s). The program will be enhanced to include additional inspection parameters. The program will require an inspection frequency of once every five (5) years for structures and structural components within the scope of the program. The
frequency of inspections can be adjusted, if necessary, to
allow for early detection and timely correction of negative trends. The program will require periodic sampling of groundwater and river water chemistries to ensure they remain non-aggressive.
31A Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) Program will be implemented. Program features will be as described in LRA Section B2.1.39.
U1 - 8/9/2013 U2 - 10/29/2014 32 The Water Chemistry Program will be enhanced as follows: The program will require increased sampling to be performed as needed to confirm the effectiveness of
corrective actions taken to address an abnormal chemistry condition. The program will require Reactor Coolant System dissolved oxygen Action Level limits to be consistent with U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 12 Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S the limits established in the EPRI PWR Primary Water Chemistry Guidelines." [Revised in letter dated 12/5/2008 in response to RAI B2.1.40-3]
33 The Metal Fatigue of Reactor Coolant Pressure Boundary Program will be enhanced as follows: The program will monitor the six component locations identified in NUREG/CR-6260 for older vintage Westinghouse plants, either by tracking the cumulative number of imposed stress cycles using cycle counting, or
by tracking the cumulative fatigue usage, including the
effects of coolant environment. The following locations
will be monitored:
o Reactor Vessel Inlet and Outlet Nozzles o Reactor Pressure Vessel Shell to Lower Head o RCS Hot Leg Surge Line Nozzle o RCS Cold Leg Charging Nozzle o RCS Cold Leg Safety Injection Accumulator Nozzle o RHR-to-Accumulator Piping Tee Program acceptance criteria will be clarified to require corrective action to be taken before a cumulative fatigue usage factor exceeds 1.0 or a design basis transient
cycle limit is exceeded. [Revised in letter dated 1/9/2009 in response to RAI 4.3.1.1-1]
U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 13 Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S 34 Reactor internals baffle bolt fatigue transient limits of 1835 cycles of plant loading at 5% per minute and 1835 cycles of plant unloading at 5% per minute will be incorporated into the
Metal Fatigue of Reactor Coolant Pressure Boundary Program and USAR Table 4.1-8.
U1 - 8/9/2013 U2 - 10/29/2014 35NSPM will perform an ASME Section III fatigue evaluation of the lower head of the pressurizer to account for effects of insurge/outsurge transients. The evaluation will determine the
cumulative fatigue usage of limiting pressurizer component(s)
through the period of extended operation. The analyses will
account for periods of both "Water Solid" and "Standard
Steam Bubble" operating strategies. Analysis results will be
incorporated, as applicable, into the Metal Fatigue of Reactor
Coolant Pressure Boundary Program. [Revised in letter dated 1/9/2009 in response to RAI 4.3.1.1-1]
U1 - 8/9/2013 U2 - 10/29/2014 36 NSPM will complete fatigue calculations for the pressurizer surge line hot leg nozzle and the charging nozzle using the methodology of the ASME Code (Subsection NB) and will
report the revised CUFs and CUFs adjusted for environmental effects at these locations as an amendment to the PINGP LRA. Conforming changes to LRA Section 4.3.3, "PINGP EAF Results," will also be included in that amendment to reflect
analysis results and remove references to stress-based
fatigue monitoring.
[Added in letter dated 1/9/2009 in response to RAI 4.3.1.1-1]
April 30, 2009 37NSPM will revise procedures for excavation and trenching controls and archaeological, cultural and historic resource
protection to identify sensitive areas and provide guidance for ground-disturbing activities. The procedures will be revised to
include drawings and illustrations to assist users in identifying 8/9/2013 Updated through 4/6/2009 14 Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S culturally sensitive areas, and pictures of artifacts that are
prevalent in the area of the Plant site. The revised procedures
will also require training of the Site Environmental Coordinator
and other personnel responsible for proper execution of excavation or other ground-disturbing activities. [Added in ER revision submitted in letter dated 3/4/2009]
38NSPM will conduct a Phase I Reconnaissance Field Survey of the disturbed areas within the Plant's boundaries. In addition, NSPM will conduct Phase I field surveys of areas of known archaeological sites to precisely determine their boundaries.
NSPM will use the results of these surveys to designate areas
for archaeological protection. [Added in ER revision submitted in letter dated 3/4/2009]
8/9/2013 39NSPM will prepare, maintain and implement a Cultural Resources Management Plan (CRMP) to protect
significant historical, archaeological, and cultural resources that may currently exist on the Plant site. In
connection with the preparation of the CRMP, NSPM will
conduct botanical surveys to identify culturally and medicinally important species on the Plant site, and
incorporate provisions to protect such plants into the
CRMP.[Added in ER revision submitted in letter dated 3/4/2009]
8/9/2013 40NSPM will consult with a qualified archaeologist prior to conducting any ground-disturbing activity in any area
designated as undisturbed and in any disturbed area that is described as potentially containing archaeological resources (as
determined by the Phase I Reconnaissance Field Survey 8/9/2013 Updated through 4/6/2009 15 Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S discussed in Commitment Number 38). [Added in ER revision submitted in letter dated 3/4/2009]
41During the first refueling outage following refueling cavity leak repairs in each Unit (scheduled for refueling outages 1R26 and
2R26), concrete will be removed from the sump C pit to expose
an area of the containment vessel bottom head. Visual examination and ultrasonic thickness measurement will be performed on the portions of the containment vessels exposed
by the excavations. An assessment of the condition of exposed
concrete and rebar will also be performed. Degradation
observed in the exposed containment vessel, concrete or rebar
will be entered into the Corrective Action Program and evaluated for impact on structural integrity and identification of
additional actions that may be warranted.
[Added in letter dated 4/6/09 in response to Follow Up RAI B2.1.38]U1 - 8/9/2013 U2 - 10/29/2014 42 During the two consecutive refueling outages following refueling cavity leak repairs in each Unit (scheduled for refueling outages
1R26 and 2R26), visual inspections will be performed of the
areas where reactor cavity leakage had been observed
previously to confirm that leakage has been resolved. The inspection results will be documented. If refueling cavity leakage is again identified, the issue will be entered into the Corrective Action Program and evaluated for identification of additional actions to mitigate leakage and monitor the condition of the containment vessel and internal structures.
[Added in letter dated 4/6/09 in response to Follow Up RAI B2.1.38]U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 16 Prairie Island Nuclear Generating Plant License Renewal Commitments Commitment Number Commitment Implementation Schedule S 43Preventive maintenance requirements will be implemented to require periodic replacement of rubber flexible hoses in the
Diesel Generators and Support System that are exposed to fuel oil or lubricating oil internal environments.
[Added in letter dated 4/6/09 in response to RAI 3.3.2-8-1]
U1 - 8/9/2013 U2 - 10/29/2014 Updated through 4/6/2009 17 1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121 April 6, 2009 L-PI-09-047 10 CFR 54 U S Nuclear Regulatory Commission
ATTN: Document Control Desk
Washington, DC 20555-0001
Prairie Island Nuclear Generating Plant Units 1 and 2
Dockets 50-282 and 50-306
License Nos. DPR-42 and DPR-60
Responses to NRC Requests for Additional Information Dated March 31, 2009 and Follow Up Questions Regar ding Application for Renewed Operating Licenses
By letter dated April 11, 2008, Norther n States Power Company, a Minnesota Corporation, (NSPM) submitted an Application for Renewed Operating Licenses (LRA) for the Prairie Island Nuclear Generating Plan t (PINGP) Units 1 and 2. In a letter dated November 5, 2008, the NRC transmitted Request for Additional Information (RAI)
AMP-B2.1.38-2 regarding that application. NSPM responded to that RAI in a letter dated December 5, 2008, and provided additional informati on in a public meeting on March 2, 2009. Subsequently, in a letter dated March 31, 2009, the NRC provided a follow up RAI related to the previous respon se and the meeting. Enclosure 1 of this letter provides the NSPM response to that follow up RAI.
In the letter dated March 31, 2009, the NRC al so transmitted RAI 3.3.2-8-1 related to Natural Rubber in Fuel Oil and Lubricating Oil. Enclosure 2 provides the NSPM response to RAI 3.3.2-8-1.
In a conference call on March 30, 2009, the NRC raised follow up questions about the LRA and certain previous RAI responses. Enclos ure 3 of this letter provides a summary of each follow up question followed by the NSPM response.
provides the list of Licens e Renewal Commitments, updated to reflect NSPM correspondence to date.
If there are any questions or if additional information is needed, please contact Mr. Eugene Eckholt, License Renewal Project Manager. Summary of Commitments This letter contains three new License Renewal commitments.
Document Control Desk Page 2 2New License Renewal Commitment Number 41 reads as follows:
During the first refueling outage following refueling cavity leak repairs in each Unit (scheduled for refueling outages 1R26 and 2R26), concrete will be removed from the sump C pit to expose an area of the c ontainment vessel bottom head. Visual examination and ultrasonic thickness measurement will be performed on the portions of the containment vessels expos ed by the excavations. An assessment of the condition of exposed concrete and rebar will also be performed. Degradation observed in the exposed containment vessel, concrete or rebar will be entered into the Corrective Action Program and evaluated fo r impact on structural integrity and identification of additional actions that may be warranted.
New License Renewal Commitment Number 42 reads as follows:
During the two consecutive refueling outages following refueling cavity leak repairs in each Unit (scheduled for refueling outages 1R26 and 2R26), visual inspections will be performed of the ar eas where reactor cavity leakage had been observed previously to confirm that leakage has been resolved. The inspection results will be documented. If refueling cavity leakage is again identified, the issue will be entered into the Corrective Action Program and eval uated for identificati on of additional actions to mitigate leakage and monitor the condition of the containment vessel and
internal structures.
New License Renewal Commitment Number 43 reads as follows:
Preventive maintenance requirements will be implemented to require periodic replacement of rubber flexib le hoses in the Diesel G enerators and Support System that are exposed to fuel oil or lubricating oil internal environments.
These actions will be completed in Unit 1 by August 9, 2013, and in Unit 2 by October 29, 2014, the dates on which the Units will enter the Period of Extended Operat ion. I declare under penalty of perjury that the foregoing is true and correct.
Executed on April 6, 2009.
/S/ Michael D. Wadley
Michael D. Wadley
Site Vice President, Prairie Island Nu clear Generating Plant Units 1 and 2 Northern States Power Company - Minnesota
Enclosures (4)
cc:
Administrator, Region III, USNRC License Renewal Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC
Prairie Island Indian Communi ty ATTN: Phil Mahowald Minnesota Department of Commerce NSPM Response to Follow Up RAI B2.1.38 1Follow Up RAI B2.1.38 In letter L-PI-08-098, dated December 5, 2008, "Responses to NRC Requests for Additional Information Dated November 5, 2008 Regarding Application for Renewed Operating Licenses," the applic ant submitted responses to th e staff's RAIs. In addition, the applicant provided information during a publ ic meeting on March 2, 2009. The staff has reviewed the information provided, but needs additional information to complete its review. Please provide the following informa tion to supplement the initial response:
- a. Due to the leakage path of borated water along the bottom of the steel containment vessel, the vessel/concrete interface may remain wetted after refueling outages. During the public meet ing, the applicant discussed its plan to remove concrete from sump C, below the reactor vessel, to physically assess the condition of the steel vessel, concrete , and rebar. Please provide more detailed information about this plan including acti ons and schedule. Discuss what other actions will be taken, prior to the period of extended operation, to adequately assess the condition of the steel vessel.
- b. In the absence of a commitment to fix the leakage prior to the period of extended operation, please explain how the IWE progr am, or a plant-spec ific program, will address the leakage and will ensure that t hese aging effects will be effectively managed, especially in the inaccessible r egions, during the period of extended operation.
- c. Provide the chemical properties of the "white deposits" found on the concrete surfaces and address the possibilit y of calcium hydroxide Ca(OH) 2 leaching from concrete.
NSPM Response to Follow Up RAI B2.1.38 Response to Part a
Leak repairs for the Unit 1 refueling cavi ty are planned for refueling outage 1R26 (fall 2009). Leak repairs for the Unit 2 refueling cavity are planned for refueling outage 2R26 (spring 2010). During the next refueling outages following leak repairs (projected to be spring 2011 and spring 2012, respectively), re moval of concrete from each sump C pit will be performed to expose an area near the center of the inside surface of the containment vessel bottom head of each Unit. Site drawings indicate that there is approximately a 16 inch thickness of rein forced concrete between the bottom of the sump C pit and the containment vessel. The method of concrete removal has not been determined, but will likely include core drilling or a similar process followed by manual chipping to avoid damage to the containmen t vessel. The concrete will be removed primarily to allow visual and UT thickness examination of a sample location in each containment vessel bottom head. A secondary purpose is to provide access for an attempt to evacuate residual water that ma y be present after leak mitigation.
A sample of the concrete excavated from close proximity to each containment vessel will be qualitatively assessed for strength (e.g., does not crush easily, no signs of NSPM Response to Follow Up RAI B2.1.38 2 porosity with fine aggregate particles easily dislodged, etc.) and will also be chemically analyzed for changes caused by exposure to bor ated water leakage. Water that seeps into the exploratory opening, if any, will be sampled and analyzed for pH and ionic species. Reinforcing bar exposed by t he excavation will be visually examined for indications of degradation. Visual and UT examinations will be per formed on the portion of each containment vessel exposed by the excavation. Degradation observed in the exposed containment vessel, concrete or rebar will be entered into the Corrective Action Program and evaluated for impact on structural integrity and identific ation of additional actions (e.g., inspection of additional locations) that may be warranted.
In addition to the physical actions descr ibed above, the following evaluations are planned:
complete an evaluation of potential degradation to date of the containment vessel, concrete and reinforcing steel perform a margin assessment of the containment vessel and interior structures, including the area around the transfer t ube, to determine allowable degradation
NSPM expects that the refue ling cavity leak repairs in outages 1R26 and 2R26 will be successful. This will be confirmed during the two refueling outages following the repairs in each Unit by specifically inspecting t he areas where reactor cavity leakage had been observed previously. The inspection results will be documented. During subsequent outages, continued monitoring for leakage and degradation will be performed through the periodic examinations of the containment vessels and interior structures performed in accordance with the ASME Section XI, S ubsection IWE Program and the Structures Monitoring Program.
Following the leak repairs, if continued monitoring again identifies active refueling cavity leakage, further action will be taken as described in Part b below.
Response to Part b
As discussed in Part a above, NSPM expects to stop the refueling cavity leakage. However, if the repairs do not completely stop the leakage, NSPM will continue to take actions to eliminate the leakage and will c ontinue to monitor for degradation.
Any additional actions to be taken woul d be determined in accordance with the Corrective Action Program evaluation of t he specific locations and quantities of observed leakage. Actions may include remova l of additional grout or concrete, visual and UT thickness examinations from the inte rior of the containment vessel at exposed locations, and UT thickness examinations from the annulus around t he exterior of the containment vessel. Examinations may be repeated at intervals commensurate with the safety significance of any identified degradat ion and the extent of residual leakage, if any. Actions may also include testing of c oncrete and rebar in areas judged to have the potential for unacceptable degradation from borated water leakage.
NSPM Response to Follow Up RAI B2.1.38 3Monitoring of the containment vessel and interior structures will be performed, at a minimum, in accordance with the ASME Section XI, Subsection IWE Program and the Structures Monitoring Pr ogram. Program requirements will be augmented as appropriate for the specific leakage or degradation observed, based on good engineering practice.
As discussed in LRA Section B2.1.4, the ASME Section XI, Subsection IWE Program provides for condition monitoring of Class MC pressure-retaining components by performing periodic visual ex aminations (general, VT-3 , VT-1) of the Class MC components and their related item
- s. Visual (VT-1) or vo lumetric examinations, as applicable, are performed on components that require augm ented examination. Indications of leakage that affects Class MC pressure-retaining components would be documented in the Corrective Action Progr am for evaluation and assignment of appropriate corrective actions. When conditions are identified in accessible areas that could indicate the presence of, or result in, degradation in ina ccessible areas, the program requires an evaluation of the acceptability of affected inaccessible areas. Such evaluations of inaccessible areas are required to be repo rted to the NRC in the ISI summary report. In accordance with IWA-6000, the information to be reported includes a description of the type and estimated extent of degradation, and conditions that led to degradation; an evaluation of eac h affected area and the result of the evaluation; and a description of the necessary actions.
As discussed in LRA Section B2.1.38, the St ructures Monitoring Pr ogram is a condition monitoring program that manages aging effects in structures, supports and structural components within the scope of License Renewal.
This includes containment interior structures. The program performs periodic visual inspections to monitor the condition of structures, supports and components with respect to established acceptance criteria to ensure that degraded conditions are identified, evaluated, and, when necessary, corrected such that there is no loss of int ended function. Inspections performed by the Structures Monitoring Program would be expected to identify conditions indicative of potential concrete or rebar degr adation, including signs of leakage, as well as cracks, calcium deposits, spalli ng, and rust stains.
Response to Part c
A task has been entered into the site Corre ctive Action Program to collect and chemically analyze deposits in areas showing signs of leakage in each Unit's regenerative heat exchanger room and sump C, as applicable. The conditions observed for each deposit, including photographs of typical leakage areas and an
estimated volume of the deposit, will be doc umented in the Correct ive Action Program in order to qualitatively asse ss the extent of the affected areas. The results of the analyses will also be documented in the Corrective Action Program. Records will be available on site for NRC review.
License Renewal Commitments
The response to this RAI contains tw o new License Renewal Commitments to be completed prior to entry of each Unit in to the Period of Ext ended Operation. NSPM Response to Follow Up RAI B2.1.38 4
New License Renewal Commitment Number 41 reads as follows:
During the first refueling outage following refueling cavity leak repairs in each Unit (scheduled for refueling outages 1R26 and 2R26), concrete will be removed from the sump C pit to expose an area of the c ontainment vessel bottom head. Visual examination and ultrasonic thickness measurement will be performed on the portions of the containment vessels expos ed by the excavations. An assessment of the condition of exposed concrete and rebar will also be performed. Degradation observed in the exposed containment vessel, concrete or rebar will be entered into the Corrective Action Program and evaluated fo r impact on structural integrity and identification of additional actions that may be warranted.
New License Renewal Commitment Number 42 reads as follows:
During the two consecutive refueling outages following refueling cavity leak repairs in each Unit (scheduled for refueling outages 1R26 and 2R26), visual inspections will be performed of the ar eas where reactor cavity leakage had been observed previously to confirm that leakage has been resolved. The inspection results will be documented. If refueling cavity leakage is again identified, the issue will be entered into the Corrective Action Program and eval uated for identificati on of additional actions to mitigate leakage and monitor the condition of the containment vessel and
internal structures.
NSPM Response to RAI 3.3.2-8-1 1 RAI 3.3.2-8-1 Natural Rubber in Fuel Oil and Lubricating Oil
In AMR Table 3.3.2-8, there is an AMR line item for Flex Connections/Pressure Boundary/Natural Rubber/Fuel Oil (int) and Lubricating Oil (int)/none/none/ with plant specific note G, 313.
P.A. Schweitzer, "Corrosion Resistance Tables
- Metals, Nonmetals, Coatings, Mortars, Plastics, Elastomers and Linings, and Fabrics," Fourth Edition, Part B, Marcel Dekker, Copyright 1995 discuss natural rubber and EPDM and states that they are not resistant to fuel oil or lubricating oil.
Please provide a justification for using natural rubber in fuel oil and lubricating oil with no aging effects and no need for an aging management program.
NSPM Response to RAI 3.3.2-8-1 During the AMR process certain rubber hoses we re conservatively identified as being made of Natural Rubber when the actual mate rials could not be readily identified.
Natural rubbers are not resist ant to fuel oil or lubric ating oil, and only qualified elastomers, that are resistant, are used for fuel oil or lubricating oil service. In general, the rubber flexible connections in the Diesel Generat ors and Support System are constructed of nitrile rubber or are lined wi th teflon, both of which have outstanding resistance to oils. Regardless, based on vendor recommendations, NSPM has
determined that the rubber fl exible hoses in the Diesel Generators and S upport System that are exposed to Fuel Oil or Lubricat ing Oil internal environments should be periodically replaced. Ther efore, these hoses are not subject to aging management review, and need not be addressed in the LRA.
Accordingly, a new License Renewal Co mmitment Number 43 is added to read as follows:
Preventive maintenance requirements will be implemented to require periodic replacement of rubber flexib le hoses in the Diesel G enerators and Support System that are exposed to fuel oil or lubricating oil internal environments.
Conforming changes are made to the LRA as follows:
In LRA Table 3.3.2-8, Auxiliary Systems - Diesel Generators and Support System - Summary of Aging Management Evaluation, on Page 3.3-156, the line items for Flex Connections / Natural Rubber / Fuel Oil (Int) and Lubricating Oil (Int), are deleted.
Additionally, the flexible connec tions for the lubricating oil and fuel oil lines on the diesel driven cooling water pumps were determined to be constructed from stainless steel. To address aging management review of the stainless steel flexible connections on the diesel driven cooling water pumps , the following LRA changes are made.
NSPM Response to RAI 3.3.2-8-1 2 In LRA Table 3.3.2-8, Auxiliary Systems - Diesel Generators and Support System - Summary of Aging Management Evaluation, on Page 3.3-153, the following line items are added:
Component Type Intended Function Material Environment Aging Effect Requiring ManagementAging Management Programs NUREG-1801 Volume 2 Line Item Table 1 ItemNotes Flex Connections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material - Crevice Corrosion Fuel Oil Chemistry Program VII.H2-16 3.3.1-32 B Flex Connections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material - Crevice Corrosion One-Time Inspection Program VII.H2-16 3.3.1-32 A Flex Connections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material -
MIC Fuel Oil Chemistry Program VII.H2-16 3.3.1-32 B Flex Connections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material -
MIC One-Time Inspection Program VII.H2-16 3.3.1-32 A Flex Connections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material -
Pitting Corrosion Fuel Oil Chemistry Program VII.H2-16 3.3.1-32 B Flex Connections Pressure Boundary Stainless Steel Fuel Oil (Int)
Loss of Material -
Pitting Corrosion One-Time Inspection Program VII.H2-16 3.3.1-32 A Flex Connections Pressure Boundary Stainless Steel Lubricating Oil (Int)
Loss of Material - Crevice Corrosion Lubricating Oil Analysis Program VII.H2-17 3.3.1-33 A Flex Connections Pressure Boundary Stainless Steel Lubricating Oil (Int)
Loss of Material - Crevice Corrosion One-Time Inspection Program VII.H2-17 3.3.1-33 A Flex Connections Pressure Boundary Stainless Steel Lubricating Oil (Int)
Loss of Material -
Pitting Corrosion Lubricating Oil Analysis Program VII.H2-17 3.3.1-33 A Flex Connections Pressure Boundary Stainless Steel Lubricating Oil (Int)
Loss of Material -
Pitting Corrosion One-Time Inspection Program VII.H2-17 3.3.1-33 A NSPM Responses to NRC Follow Up Questions 1B2.1.19 Fuel Oil Chemistry Program Follow Up Question In the 12/18/08 Response to NRC RAI B2.1.
19-3, NSPM provided ju stification for the Exception to NUREG-1801 which states that day tanks and clean fuel oil leakage collection tanks are not subjec t to periodic draining, cleaning , and internal inspection.
The justification provided was based upon the high turnover rate for the fuel in the day tanks and the close control of t he fuel oil quality in the stor age tanks that comprise the sources for the day and clean leakage collection tanks. In the 3/30/
09 conference call, the NRC expressed concern that no monito ring of the tank condition was being performed, and requested additional explanation as to how aging effects will be adequately managed in the day tanks and cl ean fuel oil collection tanks.
The diesel fuel oil day tanks and clean fuel oil leakage collection tanks are accessible for external UT examination. In order to confirm the continued effectiveness of oil quality control and to verify the absence of aging effects, one-time inspections using ultrasonic thickness measurements will be pe rformed on selected day tanks and clean fuel oil leakage collection tanks prior to the period of extended operation as part of the One-Time Inspection Program. Specific locations to be selected include:
An external UT on select bottom locations on four of the seven diesel fuel oil day tanks.
An external UT on select bottom loca tions of one of the two D1/D2 clean fuel oil leakage collection tanks.
The inspection locations identified above are in addition to those previously listed in the NSPM Response to RAI B2.1.19-3 (12/18/08 letter).
The previous response incorrectly stated t here were four Diesel Generator D1/D2 and four Diesel Generator D5/D6 fuel oil leakage collection t anks managed by the Fuel Oil Chemistry Program. However, as a result of recent plant walk downs, it has been determined that ther e are only two clean fuel oil l eakage collection tanks, both associated with D1/D2. The Unit 1 D1/D2 Diesel Generator Clean Fuel Tanks are the only fuel oil leakage collection tanks withi n the scope of the Fuel Oil Chemistry Program. These Clean Fuel Tanks collect cl ean fuel from the injector drain headers and return the fuel to the day tank with the clean fuel return pump. The vented collection header is continuous ly piped from each injector to the Clean Fuel Tank.
These fuel oil leakage collecti on tanks have a capacity of appr oximately 5 gallons each. These leakage tanks are evaluated in PINGP LRA Table 3.3.2-8, Auxiliary Systems -
Diesel Generators and Support System - Summary of Aging Management Evaluation, on Page 3.3-179.
The six remaining fuel oil l eakage collection tanks, two a ssociated with D1/D2 and four associated with D5/D6, are cont aminated fuel oil leakage collection tanks that collect drops of fuel from pump and nozzle leaks. The four D5/D 6 fuel oil leakage collection tanks in turn drain to Dirty Oil Tanks which al so collect floor drains from the D5/D6 Fuel Oil Day Tank and Lube Oil Tank rooms. These fuel oil leakage collection tanks and Dirty Oil Tanks are not pressurized.
The tanks are periodically drained, and the NSPM Responses to NRC Follow Up Questions 2 collected fluid is disposed of as waste.
Due to the potential for contamination, the internal environment for these six fuel oil leakage collection tanks and the Dirty Oil Tanks are evaluated as Raw Water. The tanks are evaluated in LRA Table 3.3.2-8, Auxiliary Systems - Diesel Generators and Support System - Summary of Aging Management Evaluation, on Page 3.3-181. Plant Specific No te 320 is applied to these line items and states, "The environment for th is component is evaluated as Raw Water and is essentially waste water or a potential mixture of wate r and oil." Aging effects are managed by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program.
During the review of LRA Table 3.3.2-8, it was noted that the line items for Piping / Fittings and Valve Bodies did not fully account for the piping and valves that are used for collection of dirty oil in the l eakage collection tanks and are managed by the Inspection of Internal Surfaces in Mi scellaneous Piping and Du cting Components Program. An entry for Sight Glasses in a Raw Water (Int) environment has also been added. For completeness, LRA Table 3.3.2-8 is hereby revised to incorporate the appropriate line items.
In LRA Table 3.3.2-8, Auxiliary Systems - Diesel Generators and Support System -
Summary of Aging Management Evaluation, the following line items are added:
Component Type Intended Function Material Environment Aging Effect Requiring ManagementAging Management Programs NUREG-1801 Volume 2 Line Item Table 1 ItemNotes Piping / Fittings Pressure Boundary Carbon Steel Raw Water (Int) Loss of Material - Crevice Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-19 3.3.1-76 E, 320 Piping / Fittings Pressure Boundary Carbon Steel Raw Water (Int) Loss of Material - Galvanic Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-19 3.3.1-76 E, 310, 320 Piping / Fittings Pressure Boundary Carbon Steel Raw Water (Int) Loss of Material - General Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-19 3.3.1-76 E, 320 Piping / Fittings Pressure Boundary Carbon Steel Raw Water (Int) Loss of Material -
MIC Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-19 3.3.1-76 E, 320 Piping / Fittings Pressure Boundary Carbon Steel Raw Water (Int) Loss of Material -
Pitting Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-19 3.3.1-76 E, 320 NSPM Responses to NRC Follow Up Questions 3Component Type Intended Function Material Environment Aging Effect Requiring ManagementAging Management Programs NUREG-1801 Volume 2 Line Item Table 1 ItemNotes Piping / Fittings Pressure Boundary Copper Alloy Raw Water (Int) Loss of Material - Crevice Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Piping / Fittings Pressure Boundary Copper Alloy Raw Water (Int) Loss of Material -
MIC Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Piping / Fittings Pressure Boundary Copper Alloy Raw Water (Int) Loss of Material -
Pitting Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Piping / Fittings Pressure Boundary Copper Alloy Raw Water (Int) Loss of Material -
Selective Leaching Selective Leaching of Materials Program VII.H2-13 3.3.1-84 B, 320 Sight Glasses Pressure Boundary Glass Raw Water (Int) None None VII.J-11 3.3.1-93 A, 320 Valve Bodies Pressure Boundary Brass Raw Water (Int) Loss of Material - Crevice Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Valve Bodies Pressure Boundary Brass Raw Water (Int) Loss of Material -
MIC Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Valve Bodies Pressure Boundary Brass Raw Water (Int) Loss of Material -
Pitting Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Valve Bodies Pressure Boundary Brass Raw Water (Int) Loss of Material -
Selective Leaching Selective Leaching of Materials Program VII.H2-13 3.3.1-84 B, 320 Valve Bodies Pressure Boundary Bronze Raw Water (Int) Loss of Material - Crevice Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 Valve Bodies Pressure Boundary Bronze Raw Water (Int) Loss of Material -
MIC Inspection of Internal Surfaces in Miscellaneous Piping and Ducting VII.C1-9 3.3.1-81 E, 320 NSPM Responses to NRC Follow Up Questions 4Component Type Intended Function Material Environment Aging Effect Requiring ManagementAging Management Programs NUREG-1801 Volume 2 Line Item Table 1 ItemNotesComponents Program Valve Bodies Pressure Boundary Bronze Raw Water (Int) Loss of Material -
Pitting Corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program VII.C1-9 3.3.1-81 E, 320 B2.1.19 Fuel Oil Analysis Follow Up Question In the 3/30/09 conference call, the NRC ack nowledged PINGP's commitment to initiate annual testing of fuel oil for particulates, but the GALL recommends quarterly testing.
The reviewer indicated that annual particulate testing may be sufficient as long as sediment and particulate testing results c ontinue to meet acceptance standards. NSPM was requested to clarify whether the annua l test frequency for particulates would be increased in the event significant sedi ment or particulates are detected.
As described in the Program Basis Docum ent for the PINGP Fuel Oil Chemistry Program, when fuel oil analysis limits are no t met, such as for sediment content or particulate contamination, the condition is entered into the Corrective Action Program for evaluation and assignment of corrective actions. Actions may include replacement of diesel fuel oil, more frequent sampling, use of fuel oil additives such as biocides, or tank draining, cleaning and inspection, depending on the significance of the issue and required actions to correct the condition.
LRA Section 4.7.5 Turbine Missile Analysis TLAA Follow Up Question In the 3/30/09 conference call, the NRC questioned whether the analysis of a turbine rotor failure leading to turbine missiles is a TLAA as discussed in LRA Section 4.7.5. This section primarily refer ences a discussion in the USAR, but the USAR does not make it clear that the analysis is part of the CLB and why it is considered a TLAA. The reviewer requested that PINGP clarify whether the issue should be classified as a TLAA. If it is concluded to be a TLAA, specif ic CLB information which defines the safety bases for turbine rotor failure (inspection frequency) should be provided.
When the PINGP LRA was prepared, USAR Sections 11.2.
3.2, Turbine Overspeed Control, and 12.2.7.1, Probability of Damage to Saf eguard Equipment from Turbine Missiles, were consulted. Section 12.2.7.1 contained a brie f discussion that the probabilities of missile ejecti on at running speed and design ov erspeed meet the criteria of the SRP "... through the remaining Operati ng License period in 2013 [Unit 2: 2014]."
Based on this statement, the analysis of turbine missiles was designated a TLAA, and a restatement of the USAR discussion was provided as a TLAA discussion in Section NSPM Responses to NRC Follow Up Questions 5 4.7.5 of the LRA. Upon furt her review of the reference documents underlying the USAR discussion, NSPM has concluded that the analysis of the probability of damage to safeguard equipment from turbi ne missiles does not meet the criteria for a TLAA, and should be removed from the LRA.
The NRC approved the analysis methodology fo r determining the total probability of turbine missile ejection from an overspeed condition (WCAP-11525) in conjunction with PINGP License Amendments 86 (Unit 1) and 79 (Unit 2) issued by an NRC letter dated February 7, 1989. WCAP-11525, "Probabilistic Evaluation of Reduction in Turbine Valve Test Frequency," considers the pr obabilities of rotor burst during design overspeed, intermediate overspeed, and destr uctive overspeed conditions. The total annual probability of turbine miss ile ejection is the sum of the annual probabilities of failure at design overspeed, intermediat e overspeed, and destructive overspeed conditions. The probability of failure at design and intermediate overspeed equals the probability of turbine overspeed multiplied by the conditional probability of failure at overspeed.
The probability of turbine overspeed is a function of the reliability of the turbine control system and the turbine valves. The reliability of the control system and valves are a function of the turbine valve test interval and the design configuration. Valve operating experience is a factor in determining this probability.
The conditional probability of missile ejecti on given a design or intermediate overspeed condition is based on consideration of damage mechanisms (e.g., SCC and fatigue) and the length of time the rotor is in service following its last inspection as discussed in WCAP-11525, Section 8.0. The rotor ins pection interval selected in WCAP-11525 for evaluation of the fully integr al low pressure turbine rotor is 10 years (WCAP-11525, Section 8.2, Page 8-3), even though evaluati ons (e.g., Figure 12.2-38 of the PINGP USAR from Reference 52 of the USAR) s howed very low failure probabilities at 30 years with no inspection.
In accordance with Section 8.0 of WCAP-11525, neither the turbine valve test interval (e.g., 12 months) nor the rotor inspection interval (e.g.,10 year s) involves a time-limited assumption defined by the current operating te rm of the plant. Ther efore, the analysis of turbine missiles discussed in USAR Section 12.2.7.1, including t he effects of turbine valve test frequency and turbine rotor inspection interval, are not TLAAs. Since this issue does not meet the defin ition of a TLAA, it need not be discussed in the LRA.
Therefore, LRA Section 4.7.5 can be withdrawn.
Conforming LRA changes to reflect withdr awal of Section 4.7.5 are as follows:
In LRA Section 3.4.2.3 on Page 3.4-21, the last bullet, Probability of Damage to Safeguards Equipment from Turbine Missiles, is deleted in its entirety.
In LRA Table 4.1-1 on Page 4.1-7, the la st line item, Probabi lity of Damage to Safeguards Equipment from Turbine Missiles, is deleted in its entirety.
In LRA Section 4.7 on Page 4.7-1, the senten ce is revised in its entirety to read: NSPM Responses to NRC Follow Up Questions 6 Other PINGP-specific TLAAs include leak-before-break (LBB) analyses, underclad cracking, RCP flywheel, and fatigue analysis of cranes.
LRA Section 4.7.5 on Pages 4.7-4 through 4.7-6 is deleted in its entirety.
LRA Section A4.9 on Pages A-25 and A-26 is deleted in its entirety.
Enclosure 4 License Renewal Commitment List
17 Pages Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 1 The following table provides the list of commitments included in the Application for Renewed Operating Licenses (LRA) for Prair ie Island Nuclear Generating Plant (PINGP) Units 1 and 2, as updated in subsequent correspondence.
The commitments in this list are anticipat ed to be the final commitments which will be confirmed in the NRC's Safety Evaluation Report (SER) for the renewed operating licen ses. These commitments, as confirmed in the SER, will become effective upon NRC issuance of the renewed licenses. In addition, as stated in the LRA, the final commi tments will be incorporated into the Updat ed Safety Analysis Report (USAR).
Commitment Number Commitment Implementation Schedule Related LRA Section Number 1 Each year, following the submittal of the PINGP License Renewal Application and at l east three months before the scheduled completion of the NRC review, NMC will submit
amendments to the PINGP application pursuant to 10 CFR 54.21(b). These revisions will identify any changes to the Current Licensing Basis that materially affect the contents of the License Renewal Application, including the USAR supplements.
12 months after LRA submittal
date and at least
3 months before
completion of NRC review 1.4 2 Following the issuance of the renewed operating license, the summary descriptions of aging management programs and TLAAs provided in Appendix A, a nd the final list of License Renewal commitments, will be in corporated into the PINGP USAR as part of a periodic USAR update in accordance with 10
CFR 50.71(e). Other changes to specific sections of the PINGP USAR necessary to reflect a renewed operating license will also be addressed at that time.
First USAR update in accordance with 10 CFR 50.71(e) following issuance of renewed operating licenses A1.0 3 An Aboveground Steel Tanks Program will be implemented.
Program features will be as descr ibed in LRA Section B2.1.2.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.2 4 Procedures for the conduct of inspections in the External Surfaces Monitoring Program, St ructures Monitoring Program, Buried Piping and Tanks Ins pection Program, and the RG 1.127 Inspection of Water-Control Structures Associated with U1 - 8/9/2013 U2 - 10/29/2014 B2.1.6 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 2 Commitment Number Commitment Implementation Schedule Related LRA Section Number Nuclear Power Plants Program will be enhanced to include guidance for visual inspections of installed bolting. 5 A Buried Piping and Tanks Inspection Program will be implemented. Program features will be as described in LRA Section B2.1.8.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.8 6 The Closed-Cycle Cooling Water System Program will be enhanced to include periodic inspection of accessible surfaces
of components serviced by closed-cycle cooling water when the systems or components are opened during scheduled maintenance or surveillance ac tivities. Inspections are performed to identify the pres ence of aging effects and to confirm the effectiveness of the chemistry controls. Visual inspection of component internals will be used to detect loss of
material and heat transfer degr adation. Enhanced visual or volumetric examination tec hniques will be used to detect cracking.
[Revised in letter dated 1/20/2009 in response to RAI 3.3.2 01] U1 - 8/9/2013 U2 - 10/29/2014 B2.1.9 7 The Compressed Air Monitoring Program will be enhanced as follows:
Station and Instrument Ai r System air quality will be monitored and maintained in accordance with
the instrument air quality guidance provided in ISA S7.0.01-1996. Particulat e testing will be revised to use a particle size methodology as specified in
ISA S7.0.01.
The program will incorpor ate on-line dew point monitoring.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.10 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 3 Commitment Number Commitment Implementation Schedule Related LRA Section Number [Revised in letter dated 2/6/2009 in response to Region III
License Renewal Inspection] 8 An Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program will be completed. Program features will be as described in LRA Section B2.1.11.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.11 9 An Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program will be implemented. Program features will be as described in LRA Section B2.1.12.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.12 10 An Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualific ation Requirements Used in Instrumentation Circuits Program will be implemented.
Program features will be as descr ibed in LRA Section B2.1.13.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.13 11 The External Surfaces Monitoring Program will be enhanced as follows: The scope of the program will be expanded as necessary to include all metallic and non-metallic components within
the scope of License Renewal that require aging
management in accordance with this program.
The program will ensure that surfaces that are inaccessible or not readily visible during plant operations will be inspected durin g refueling outages.
The program will ensure that surfaces that are inaccessible or not readily visible during both plant operations and refueling outages will be inspected at intervals that provide reasonable assurance that aging U1 - 8/9/2013 U2 - 10/29/2014 B2.1.14 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 4 Commitment Number Commitment Implementation Schedule Related LRA Section Number effects are managed such that the applicable components will perform their intended function during
the period of ex tended operation.
The program will apply physical manipulation techniques, in addition to visual inspection, to detect aging effects in elastomers and plastics.
The program will include acc eptance criteria (e.g., threshold values for identif ied aging effects) to ensure that the need for corrective actions will be identified
before a loss of intended functions.
The program will ensure that program documentation such as walkdown records, inspection results, and other
records of monitoring and trendi ng activities are auditable and retrievable. [Revised in letter dated 2/6/2009 in response to RAI B2.1.14-1 Follow Up question] 12 The Fire Protection Program will be enhanced to require periodic visual inspection of the fire barrier walls, ceilings, and
floors to be performed during walkdowns at least once every refueling cycle. [Revised in letter dated 12/5/2008 in response to RAI B2.1.15-3]
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.15 13 The Fire Water System Program will be enhanced as follows: The program will be expanded to include eight additional yard fire hydrants in the scope of the annual visual inspection and flushing activities.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.16 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 5 Commitment Number Commitment Implementation Schedule Related LRA Section Number The program will require that sprinkler heads that have been in place for 50 years will be replaced or a representative sample of sprinkler heads will be tested using the guidance of NFPA 25, "Inspection, Testing and Maintenance of Water-Based Fire Protection Systems" (2002 Edition, Section 5.3.1.
1.1). Sample testing, if performed, will continue at a 10-year interval following the initial testing.
14 The Flux Thimble Tube Ins pection Program will be enhanced as follows: The program will require t hat the interval between inspections be established such that no flux thimble tube is predicted to incur wear that exceeds the established acceptance criteria before the next inspection. The program will require that re-baselining of the examination frequency be justif ied using plant-specific wear rate data unless prior plant-specific NRC acceptance for the re-baselining was received. If design changes are made to use more wear-resistant thimble tube materials, sufficient inspections will be conducted at an adequate inspection frequency for the new materials. The program will require that flux thimble tubes that cannot be inspected must be removed from service.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.18 15 The Fuel Oil Chemistry Progr am will be enhanced as follows: Particulate contamination test ing of fuel oil in the eleven fuel oil storage tanks in scope of License Renewal will be performed, in accordance with ASTM D 6217, on an U1 - 8/9/2013 U2 - 10/29/2014 B2.1.19 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 6 Commitment Number Commitment Implementation Schedule Related LRA Section Number annual basis. One-time ultrasonic thickness measurements will be performed at selected tank bottom and piping locations prior to the period of extended operation. 16 A Fuse Holders Program will be implemented. Program features will be as described in LRA Section B2.1.20.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.20 17 An Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program will be implemented. Program features will be as described in LRA Section B2.1.21 U1 - 8/9/2013 U2 - 10/29/2014 B2.1.21 18 An Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program will be implemented. Program features will be as described in LRA section B2.1.22. Inspections for stress corrosion cracking will be performed by visual examination with a magni fied resolution as described in 10 CFR 50.55a(b)(2)(xxi)(A) or with ultrasonic methods. [Revised in letter dated 2/6/2009 in response to RAI B2.1.22-1 Follow Up question]
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.22 19 The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Program will be enhanced as follows: Program implementing procedures will be revised to ensure the components and structures subject to inspection are clearly identified. Program inspection procedures will be enhanced to include the parameters corrosion and wear where U1 - 8/9/2013 U2 - 10/29/2014 B2.1.23 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 7 Commitment Number Commitment Implementation Schedule Related LRA Section Number omitted. 20 A Metal-Enclosed Bus Program will be implemented. Program features will be as described in LRA Section B2.1.26.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.26 21 Number Not Used
[Revised in letter dated 3/27/2009]
22 The Nickel-Alloy Penetration No zzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors Program will be enhanced as follows: The program will require that any deviations from implementing the appropria te required inspection methods of the NRC First Revised Order EA-03-009, "Issue of Order Establishing Interim Inspection
Requirements for Reactor Pressure Vessel Heads at
Pressurized Water Reactors," dated February 20, 2004 (Order), as amended, will be submitted for NRC review and approval in accordance with the Order, as amended. The program will require that any deviations from implementing the required inspection frequencies mandated by the Order, as amended, will be submitted for NRC review and approval in accordance with the Order, as amended. The program will require that relevant flaw indications detected during the augmented inspections of the upper vessel head penetration nozzles will be evaluated in accordance with the criteria provided in the letter from Mr. Richard Barrett, NRC, Offi ce of Nuclear Reactor U1 - 8/9/2013 U2 - 10/29/2014 B2.1.28 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 8 Commitment Number Commitment Implementation Schedule Related LRA Section Number Regulation (NRR), Division of Engineering to Alex
Marion, Nuclear Energy Inst itute (NEI), dated April 11, 2003, or in accordance with NRC-approved Code Cases that incorporate the flaw evaluation procedures and criteria of the NRC's April 11, 2003, letter to NEI. The program will require that , if leakage or evidence of cracking in the vessel head penetration nozzles (including associated J-groove welds) is detected while ranked in the "Low," "Moderate," or "Replaced" susceptibility category, the no zzles are to be immediately reclassified to the "High" susceptibility category and the required augmented inspec tions for the "High" susceptibility category are to be implemented during the same outage the leakage or cracking is detected.
23 A One-Time Inspection Program will be completed. Program features will be as described in LRA Section B2.1.29.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.29 24 A One-Time Inspection of ASME Code Class 1 Small-Bore Piping Program will be completed.
Program featur es will be as described in LRA Section B2.1.30.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.30 25 For the PWR Vessel Internals Program, PINGP commits to the following activities for managing the aging of reactor vessel internals components: Participate in the industry programs for investigating and managing aging effects on reactor internals; Evaluate and implement the results of the industry programs as applicable to the reactor internals; and Upon completion of these pr ograms, but not less than 24 U1 - 8/9/2011 U2 - 10/29/2012 B2.1.32 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 9 Commitment Number Commitment Implementation Schedule Related LRA Section Number months before entering the per iod of extend ed operation, submit an inspection plan for reactor internals to the NRC for review and approval.
26 The Reactor Head Closure Studs Program will be enhanced to incorporate controls that ensure that any future procurement of reactor head closure studs will be in accordance with the material and inspection guidance provided in NRC Regulatory Guide 1.65.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.33 27 The Reactor Vessel Surveillance Program will be enhanced as follows: A requirement will be added to ensure that all withdrawn and tested surveillance capsules, not discarded as of
August 31, 2000, are placed in storage for possible future reconstitution and use. A requirement will be added to ensure that in the event spare capsules are withdrawn, the untested capsules are placed in storage and maintained for future insertion.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.34 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 10 Commitment Number Commitment Implementation Schedule Related LRA Section Number 28 The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants Program will be enhanced as follows: The program will include inspections of concrete and steel components that are bel ow the water line at the Screenhouse and Intake Canal. The scope will also require inspections of the Approach Canal, Intake Canal, Emergency Cooling Water Intake, and Screenhouse
immediately following extreme environmental conditions or natural phenomena includi ng an earthquake, flood, tornado, severe thunderstorm, or high winds. The program parameters to be inspected will include an inspection of water-control concrete components that are below the water line for cavitation and erosion degradation. The program will visually inspect for damage such as cracking, settlement, mo vement, broken bolted and welded connections, bu ckling, and other degraded conditions following extreme environmental conditions or natural phenomena.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.35 29 A Selective Leaching of Material s Program will be completed.
Program features will be as described in LRA B2.1.36.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.36 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 11 Commitment Number Commitment Implementation Schedule Related LRA Section Number 30 The Structures Monitoring Program will be enhanced as follows: The following structures, components, and component supports will be added to the scope of the inspections:
o Approach Canal o Fuel Oil Transfer House o Old Administration Building and Administration Building Addition o Component supports for cable tray, conduit, cable, tubing tray, tubing, non-ASME vessels, exchangers, pumps, valves, piping, mirror
insulation, non-ASME valv es, cabinets, panels, racks, equipment enclosures, junction boxes, bus
ducts, breakers, transformers, instruments, diesel equipment, housings for HVAC fans, louvers, and
dampers, HVAC ducts, vibration isolation elements for diesel equipment, and miscellaneous
electrical and mechanical equipment items o Miscellaneous electrical equipment and instrumentation enclosures including cable tray, conduit, wireway, tube tray, cabinets, panels, racks, equipment enclosures, junction boxes, breaker housings, transformer housings, lighting
fixtures, and metal bus enclosure assemblies o Miscellaneous mechanical equipment enclosures including housings for HVAC fans, louvers, and dampers o SBO Yard Structures and components including SBO cable vault and bus duct enclosures.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.38 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 12 Commitment Number Commitment Implementation Schedule Related LRA Section Number o Fire Protection System hydrant houses o Caulking, sealant and elastomer materials o Non-safety related masonry walls that support equipment relied upon to perform a function that demonstrates compliance with a regulated
event(s). The program will be enhanced to include additional inspection parameters. The program will require an inspection frequency of once every five (5) years for structures and structural components within the scope of the program. The
frequency of inspections can be adjusted, if necessary, to
allow for early detection and timely correction of negative trends. The program will require periodic sampling of groundwater and river water c hemistries to ensure they remain non-aggressive.
31 A Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) Program will be impl emented. Program features will be as described in LRA Section B2.1.39.
U1 - 8/9/2013 U2 - 10/29/2014 B2.1.39 32 The Water Chemistry Program will be enhanced as follows: The program will require increased sampling to be performed as needed to confi rm the effectiveness of corrective actions taken to address an abnormal chemistry condition. The program will require R eactor Coolant System dissolved oxygen Action Level limits to be consistent with U1 - 8/9/2013 U2 - 10/29/2014 B2.1.40 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 13 Commitment Number Commitment Implementation Schedule Related LRA Section Number the limits established in the EPRI PWR Primary Water Chemistry Guidelines." [Revised in letter dated 12/5/2008 in response to RAI B2.1.40-3]
33 The Metal Fatigue of Reactor Coolant Pressure Boundary Program will be enhanced as follows:
The program will monitor t he six component locations identified in NUREG/CR-6260 for older vintage Westinghouse plants, either by tracking the cumulative number of imposed stress cycles using cycle counting, or
by tracking the cumulative fatigue usage, including the
effects of coolant environmen
- t. The following locations will be monitored:
o Reactor Vessel Inlet and Outlet Nozzles o Reactor Pressure Vessel Shell to Lower Head o RCS Hot Leg Surge Line Nozzle o RCS Cold Leg Charging Nozzle o RCS Cold Leg Safety Injection Accumulator Nozzle o RHR-to-Accumulator Piping Tee Program acceptance criteria will be clarified to require corrective action to be taken before a cumulative fatigue usage factor exceeds 1.0 or a design basis transient cycle limit is exceeded. [Revised in letter dated 1/9/2009 in response to RAI 4.3.1.1-1]
U1 - 8/9/2013 U2 - 10/29/2014 B3.2 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 14 Commitment Number Commitment Implementation Schedule Related LRA Section Number 34 Reactor internals baffle bolt fatigue transient limits of 1835 cycles of plant loading at 5%
per minute and 1835 cycles of plant unloading at 5% per minut e will be incorpor ated into the Metal Fatigue of Reactor Coolant Pressure Boundary Program and USAR Table 4.1-8.
U1 - 8/9/2013 U2 - 10/29/2014 B3.2 35 NSPM will perform an ASME Sect ion III fatigue evaluation of the lower head of the pressurize r to account for effects of insurge/outsurge transients.
The evaluation will determine the cumulative fatigue usage of limiting pressurizer component(s)
through the period of extended operation. The analyses will account for periods of both "Water Solid" and "Standard Steam Bubble" operating strat egies. Analysis results will be incorporated, as applicable, into the Metal Fatigue of Reactor
Coolant Pressure Boundary Program. [Revised in letter dated 1/9/2009 in response to RAI 4.3.1.1-1]
U1 - 8/9/2013 U2 - 10/29/2014 4.3.1.3 36 NSPM will complete fatigue calc ulations for the pressurizer surge line hot leg nozzle and the charging nozzle using the methodology of the ASME Code (Subsection NB) and will
report the revised CUFs and CUFs adjusted for environmental effects at these locations as an amendment to the PINGP LRA. Conforming changes to LR A Section 4.3.3, "PINGP EAF Results," will also be included in that amendment to reflect analysis results and remove references to stress-based
fatigue monitoring.
[Added in letter dated 1/9/2009 in response to RAI 4.3.1.1-1] April 30, 2009 4.3.3 37 NSPM will revise procedures for excavation and trenching controls and archaeological, cult ural and historic resource protection to identify sensitiv e areas and provide guidance for ground-disturbing activities. The procedures will be revised to
include drawings and illustrations to assist users in identifying 8/9/2013 ER 4.16.1 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 15 Commitment Number Commitment Implementation Schedule Related LRA Section Number culturally sensitive areas, and pi ctures of artifacts that are prevalent in the area of the Plant site. The revised procedures will also require training of the Site Environmental Coordinator and other personnel responsible for proper execution of excavation or other ground-disturbing activities.
[Added in ER revision submi tted in letter dated 3/4/2009]
38 NSPM will conduct a Phase I Reconnaissance Field Survey of the disturbed areas within the Pl ant's boundaries. In addition, NSPM will conduct Phase I field surveys of areas of known archaeological sites to precisely determine their boundaries.
NSPM will use the results of these surveys to designate areas for archaeological protection.
[Added in ER revision submi tted in letter dated 3/4/2009] 8/9/2013 ER 4.16.2 39 NSPM will prepare, maintain and implement a Cultural Resources Management Plan (CRMP) to protect significant historical, archaeological, and cultural resources that may currently exist on the Plant site. In connection with the preparation of the CRMP, NSPM will conduct botanical surveys to identify culturally and medicinally important spec ies on the Plant site, and incorporate provisions to protect such plants into the
CRMP. [Added in ER revision submi tted in letter dated 3/4/2009] 8/9/2013 ER 4.16.2 40 NSPM will consult with a qualif ied archaeologist prior to conducting any ground-disturbing activity in any area
designated as undisturbed and in any disturbed area that is
described as potentially containi ng archaeological resources (as determined by the Phase I Reconnaissance Field Survey 8/9/2013 ER 4.16.2 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 16 Commitment Number Commitment Implementation Schedule Related LRA Section Number discussed in Commitment Number 38).
[Added in ER revision submi tted in letter dated 3/4/2009]
41 During the first refueling outage following refueling cavity leak repairs in each Unit (scheduled for refueling outages 1R26 and 2R26), concrete will be removed from the sump C pit to expose an area of the containment vessel bottom head. Visual
examination and ultrasonic thickness measurement will be performed on the portions of the containment vessels exposed by the excavations. An assessment of the condition of exposed concrete and rebar will also be performed. Degradation observed in the exposed containm ent vessel, concrete or rebar will be entered into the Co rrective Action Program and evaluated for impact on structural integrity and identification of additional actions that may be warranted.
[Added in letter dated 4/6/09 in response to Follow Up RAI
B2.1.38] U1 - 8/9/2013 U2 - 10/29/2014 B2.1.38 42 During the two consecutive refueling outages following refueling cavity leak repairs in each Unit (scheduled for refueling outages
1R26 and 2R26), visual inspecti ons will be performed of the areas where reactor cavity leakage had been observed
previously to confirm that l eakage has been resolved. The inspection results will be documented. If refueling cavity leakage is again identified, the issue will be entered into the Corrective Action Program and evaluated for identification of additional actions to mitigate leakage and monitor the condition of the containment vessel and internal structures.
[Added in letter dated 4/6/09 in response to Follow Up RAI
B2.1.38] U1 - 8/9/2013 U2 - 10/29/2014 B2.1.38 Prairie Island Nuclear Generating Plant License Renewal Commitments Updated through 4/6/2009 17 Commitment Number Commitment Implementation Schedule Related LRA Section Number 43 Preventive maintenance requirements will be implemented to require periodic replacement of rubber flexible hoses in the Diesel Generators and Support System that are exposed to fuel oil or lubricating oil internal environments.
[Added in letter dated 4/6/09 in response to RAI 3.3.2-8-1]
U1 - 8/9/2013 U2 - 10/29/2014 Table 3.3.2-8