ML090930096

2009/03/27 PINGP Lr - PINGP Letter Providing Supplemental Information Regarding Contention 8

ML090930096
Person / Time
Site:	Prairie Island
Issue date:	03/27/2009
From:	- No Known Affiliation
To:	Division of License Renewal
References
Download: ML090930096 (62)

Text

1 PrairieIslandNPEm Resource From: Vincent, Robert [Robert.V incent@xenuclear.com]

Sent: Friday, March 27, 2009 2:13 PM To: Richard Plasse; Nathan Goodman Cc: Eckholt, Gene F.; Davis, Marlys E.

Subject:

PINGP Letter Providing Supplemental Information Regarding Contention 8 Attachments:

20090327 Supplemental Letter to Resolve Cont ention 8 .pdf; 20090327 Supplemental Letter to Resolve Contention 8 .docAttached are pdf and WORD versions of a letter supplementing the LRA to resolve Contention 8. The letter adds a plant-specific Nickel-Alloy Nozzles and Penetrations Program to the LRA to replace the existing commitment which is specified in GALL. Let me know if you have any questions.

Bob Vincent Licensing Lead, License Renewal Project 651-388-1121 X7259 (office)

651-267-7207 (fax)

Hearing Identifier: Prairie_Island_NonPublic Email Number: 974 Mail Envelope Properties (9FA1D9F2F220C04F95D9394E3CF02DAB013A723A)

Subject:

PINGP Letter Providing Supplemental Information Regarding Contention 8 Sent Date: 3/27/2009 2:12:36 PM Received Date: 3/27/2009 2:12:47 PM From: Vincent, Robert Created By: Robert.Vincent@xenuclear.com Recipients: "Eckholt, Gene F." <Gene.Eckholt@xenuclear.com>

Tracking Status: None "Davis, Marlys E." <Marlys.Davis@xenuclear.com>

Tracking Status: None "Richard Plasse" <Richard.Plasse@nrc.gov> Tracking Status: None "Nathan Goodman" <Nathan.Goodman@nrc.gov> Tracking Status: None

Post Office: enex02.ft.nmcco.net Files Size Date & Time MESSAGE 417 3/27/2009 2:12:47 PM 20090327 Supplemental Letter to Resolve Contention 8 .pdf 1079505 20090327 Supplemental Letter to Resolve Contention 8 .doc 322112

Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date: Recipients Received:

Document Control Desk Page 2 Enclosures (2) cc: Administrator, Region III, USNRC License Renewal Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC Prairie Island Indian Community ATTN: Phil Mahowald Minnesota Department of Commerce Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 1 LRA Sections A2.27 and B2.1.27, and selected information in LRA Sections 2, 3, B1.1, B1.5, and B2.0, are hereby revised to provide updated information regarding the Nickel-

Alloy Nozzles and Penetrations Program. The updated information designates the

program as a plant-specific program, expands the level of detail in the LRA, and reflects the incorporation of new NRC requirements from 10 CFR 50.55a. The LRA changes are as follows: In LRA Section 2.1.1.3, Interim Staff Guidance (ISG) Discussion, on Page 2.1-5, the second paragraph of the discussion for LR-ISG-19B is deleted and replaced with the

following:The PINGP Nickel-Alloy Penetration Nozzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors Program manages aging effects on the reactor head penetrations. With respect to the management of nickel-alloy nozzles and penetrations not associated with the reactor vessel head penetration nozzles, a plant-specific program, the Nickel-Alloy Nozzles and Penetrations

Program, is credited with managing the effects of cracking due to PWSCC. As this issue evolves under the existing regulatory process, these programs will be modified

as necessary in response to industry initiatives and NRC guidance and

requirements.In LRA Section 3.1.2.2.13, Cracking due to Primary Water Stress Corrosion Cracking (PWSCC), on Pages 3.1-14 and 3.1-15, the eighth sentence "For the Nickel-Alloy

Nozzles and Penetrations Program ... and staff-accepted industry guidelines." is deleted and replaced with the following:

The Nickel-Alloy Nozzles and Penetrations Program ranks the Alloy 600/82/182 locations for primary water stress corrosion cracking susceptibility and then utilizes inspections, mitigation techniques, and repair/replacement activities to manage the

effects of cracking due to primary water stress corrosion cracking. The Nickel-Alloy

Nozzles and Penetrations Program implements the inspection of the Alloy 600/82/182 materials through the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program.

In LRA Table 3.1.2-1, Reactor Vessel, Internals, and Reactor Coolant System - Pressurizer System - Summary of Aging Management Evaluation, on Page 3.1-51, in

the line item for Surge Nozzle being managed by the Nickel-Alloy Nozzles and Penetrations Program, the Notes entry is changed from A, 111 to E, 111.

In LRA Table 3.1.2-4, Reactor Vessel, Internals, and Reactor Coolant System - Reactor Vessel System - Summary of Aging Management Evaluation, on Page 3.1-97, in the line item for Core Support Pads being managed by the Nickel-Alloy Nozzles and

Penetrations Program, the Notes entry is changed from A to E.

Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 2 In LRA Table 3.1.2-4, Reactor Vessel, Internals, and Reactor Coolant System - Reactor Vessel System - Summary of Aging Management Evaluation, on Page 3.1-100, in the line item for Instrumentation Tube Penetrations (Bottom Head) being managed by the Nickel-Alloy Nozzles and Penetrations Program, the Notes entry is changed from A to E.LRA Section A2.27 on Pages A-11 and A-12 is revised in its entirety to read as follows: A2.27 Nickel-Alloy Nozzles and Penetrations Program The Nickel-Alloy Nozzles and Penetrations Program manages the aging effect of cracking due to primary water stress corrosion cracking (PWSCC) of nickel-alloy pressure boundary and structural components exposed to primary coolant. The

Alloy 600/82/182 locations are ranked for PWSCC susceptibility. The program manages these components for cracking due to PWSCC utilizing inspections, mitigation techniques, and repair/replacement activities. The program implements

the inspection of the Alloy 600/82/182 materials through the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program. The program complies with applicable NRC Orders, and implements applicable NRC Bulletins, Generic Letters, and staff-accepted industry guidelines. In LRA Section B1.1, Overview, as revised in Enclosure 1 of NSPM letter dated March 12, 2009, the second paragraph, "Two of the programs consist of commitments

...NUREG-1801 Chapter IV." is deleted in its entirety and replaced with the following: In certain cases a program may consist solely of a commitment if the associated NUREG-1801 program description is only comprised of specified commitment language in NUREG-1801 Chapter IV (e.g., NUREG-1801 Programs XI.M11 and XI.M16). Where this appendix provides a full ten element description of a program

that NUREG-1801 only defines as a commitment, the program is designated a plant-specific program. Plant-specific programs are described in terms of their consistency with the ten generic program elements defined in Appendix A.1, Section

A.1.2.3 of NUREG-1800.In LRA Section B1.5, Aging Management Programs, on Page B-4, the introductory paragraph is revised to read as follows:

The AMPs credited with managing the effects of aging at PINGP are described in the following sections. Where these programs are addressed in NUREG-1801, they have been evaluated as being either fully consistent with, or, with some exceptions and/or enhancements, consistent with the NUREG-1801 programs.

Where plant-specific AMPs are credited, they have been evaluated for Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 3consistency with the ten elements of an acceptable aging management program defined in Appendix A.1, Section A.1.2.3 of NUREG-1800. In LRA Section B2.0, Aging Management Programs Correlation, on Page B-7, line item XI.M11 of the NUREG-1801 program correlation table is revised to appear as follows: NUREG-1801 IDNUREG-1801 ProgramPINGP Program NUREG-1801 ComparisonXI.M11Nickel-AlloyNozzles and

PenetrationsNickel-Alloy Nozzles and Penetrations Program

[Section B2.1.27]

Existing Plant-Specific Program In LRA Section B2.0 on Page B-12, Note 1 is revised in its entirety to read, "Not Used." LRA Section B2.1.27 on Page B-58 is revised in its entirety to read as follows: B2.1.27 Nickel-Alloy Nozzles and Penetrations Program Program Description The Nickel-Alloy Nozzles and Penetrations Program manages the aging effect of cracking due to primary water stress corrosion cracking (PWSCC) of nickel-alloy pressure boundary and structural components exposed to primary coolant. The

Alloy 600/82/182 locations are ranked for PWSCC susceptibility. The program manages these components for cracking due to PWSCC utilizing inspections, mitigation techniques, and repair/replacement activities. The Nickel-Alloy Nozzles and Penetrations Program implements the inspection of the Alloy 600/82/182 materials through the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program. The program complies with applicable NRC Orders, and implements applicable NRC Bulletins, Generic Letters, and staff-accepted industry guidelines.

NUREG-1800 Consistency The Prairie Island Nuclear Generating Plant Nickel-Alloy Nozzles and Penetrations Program is an existing plant-specific program. The program consists of the ten elements of an acceptable AMP as described in NUREG-1800 Appendix A.1, Section A.1.2.3 and Table A.1-1.

Exceptions to NUREG-1800 or NUREG-1801 None Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 4 Enhancements None Aging Management Program Elements The elements of the Nickel-Alloy Nozzles and Penetrations Program are described below. The results of an evaluation of each element with respect to the NUREG-

1800, Appendix A.1, Section A.1.2.3, "Aging Management Program Elements" and Table A.1-1, "Elements of an Aging Management Program for License Renewal," are

also provided.

Scope of Program The Nickel-Alloy Nozzles and Penetrations Program manages the aging effect of cracking due to primary water stress corrosion cracking (PWSCC) of pressure

boundary and structural components constructed of Alloy 600 and welds

constructed of the associated Alloy 82/182 filler metals exposed to primary coolant. The scope of the Nickel-Alloy Nozzles and Penetrations Program includes the Reactor Vessel System instrumentation tube penetrations (bottom head), thirty-six in each Unit, constructed of Alloy 600; the Reactor Vessel System core support pads (Alloy 600), four in each Unit; and the Unit 2

pressurizer surge nozzle-to-safe end dissimilar metal weld (Alloy 82).

The program is based upon the industry guidance provided in EPRI MRP-126, "Generic Guidance for Alloy 600 Management" (1009561), November 2004 and

NEI 03-08, "Guideline for the Management of Materials Issues," May 2003. The

program complies with applicable NRC Orders, and implements applicable NRC Bulletins, Generic Letters, and staff-accepted industry guidelines.

This AMP consists of PINGP activities that manage aging effects for components of the following systems: Pressurizer (PS) System Reactor Vessel (RV) System This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.1, Scope of Program.

Preventive Actions The program considers various mitigative and repair options to ensure that nickel-alloy components continue to perform their intended functions during the

period of extended operation. Selection of a mitigation strategy may be based upon available technology, ASME Code requirements, radiological conditions, and economic factors. Most mitigative actions implemented by the industry since the mid-1990s have utilized PWSCC resistant Alloy 690/52/152 materials.

Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 5To mitigate the effects of PWSCC on the Unit 2 pressurizer surge nozzle weld (Alloy 82), a full structural weld overlay (FSWOL) on the pressurizer surge

nozzle-to-safe end dissimilar metal and safe end-to-reducer stainless steel butt

welds was installed in October 2008 during the PINGP Unit 2 refueling outage (2R25). The FSWOL was installed using Alloy 52M weld material. Preventive measures to mitigate PWSCC are in accordance with the PINGP Water Chemistry Program. The Water Chemistry Program manages aging

effects by controlling concentrations of known detrimental chemical species such as chlorides, fluorides, sulfates and dissolved oxygen below the levels known to cause degradation. The program includes specifications for chemical species, sampling and analysis frequencies, and corrective actions for control of water chemistry. This program conforms to the EPRI PWR Primary Water Chemistry Guidelines. The PINGP Water Chemistry Program is further described in LRA Section B2.1.40, Water Chemistry Program. This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.2, Preventive Actions. Parameters Monitored/Inspected The PINGP Nickel-Alloy Nozzles and Penetrations Program monitors for cracking due to PWSCC of Alloy 600/82/182 materials exposed to primary coolant. The program performs condition monitoring examinations of the lower reactor vessel

head surface and each bottom-mounted instrumentation tube penetration. These examinations monitor for through-wall cracks that may exist in the nozzles or

their associated partial penetration J-groove welds. For in-scope pressure boundary components, the program monitors for evidence of reactor coolant leakage which may manifest itself in the form of boric acid residues or corrosion

products. The core support pads and the Unit 2 pressurizer surge nozzle-to-safe

end weld are monitored for evidence of cracking. The Nickel-Alloy Nozzles and Penetrations Program implements the inspection of the Alloy 600/82/182 materials through the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program. This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.3, Parameters Monitored or Inspected.

Detection of Aging Effects The program utilizes visual and volumetric examination techniques to detect cracking in Alloy 600/82/182 materials. 10 CFR 50.55a requires that all power

reactors maintain an Inservice Inspection Program in accordance with the ASME

Boiler and Pressure Vessel Code,Section XI. The Nickel-Alloy Nozzles and Penetrations Program implements the inspection of the Alloy 600/82/182 Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 6materials through the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program.For the reactor vessel core support pads, the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program conducts a visual VT-1 examination of the accessible interior attachment welds per Table IWB-2500-1, Examination Category B-N-2, once per Inservice Inspection interval.The FSWOL on the pressurizer surge nozzle-to-safe end dissimilar metal weld (Alloy 82) and safe end-to-reducer stainless steel butt weld was installed at PINGP Unit 2 with NRC approval. Inservice examinations are conducted ultrasonically with the examination volume as defined in ASME Section XI, Nonmandatory Appendix Q, Figure Q-4300-1. Inservice examinations as

described in Q-4300 are performed in accordance with the requirements of MRP-139, "Primary System Piping Butt Weld Inspection and Evaluation Guidelines,"

with the additional requirement of at least one ultrasonic examination within ten years of the FSWOL application. The MRP-139 guidance for ISI goes beyond the current ASME Code inspection requirements for PINGP Unit 2. With the installation of the FSWOL, the Unit 2 pressurizer surge nozzle-to-safe end weld reverts back to Category "B" per MRP-139 which will require examinations in accordance with the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program, with specified limitations, modifications and NRC-approved alternatives.

Reactor pressure vessel bottom head bare metal visual examinations are performed by removing insulation sections and/or examining under the insulation

using remote viewing equipment that provides a high degree of resolution in order to identify very small volumes of boric acid that may result from Alloy 600 PWSCC. In response to NRC Bulletin 2003-02, PINGP committed to 100% bare

metal visual inspection of the lower reactor pressure vessel dome up to and including each bottom mounted instrumentation (BMI) penetration to RPV junction every other refueling outage. Fully qualified and certified VT-2 examiners perform the inspections of the bottom head penetrations. The inspections are in compliance with ASME Code Case N-722, "Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials," as required by and modified by 10 CFR 50.55a(g)(6)(ii)(E). 10 CFR 50.55a(g)(6)(ii)(E) requires non-visual

inspection for items where leakage is identified to determine whether

circumferential cracking is present in the flawed material and if multiple

circumferential flaws have initiated.These inspections of nickel-alloy components ensure timely detection of cracking due to PWSCC prior to a loss of component intended function. The program complies with applicable NRC Orders, and implements applicable NRC Bulletins, Generic Letters, and staff-accepted industry guidelines.

Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 7This element is consistent with the recommendations of NUREG-1800, Element

A.1.2.3.4, Detection of Aging Effects.

Monitoring and Trending The Nickel-Alloy Nozzles and Penetrations Program ranked the Alloy 600/82/182 locations based on four main criteria: PWSCC susceptibility (e.g., operational time and temperature), failure consequence, leakage detection margin, and radiation dose rates. Additionally, material heat susceptibility and other industry

experience were also considered.

The program incorporates the inspection schedules and frequencies for the nickel-alloy components in accordance with the PINGP ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program and, where

applicable, ASME Code Case N-722, subject to the conditions specified in 10 CFR 50.55a(g)(6)(ii)(E). Flaw indications detected during the required examinations are dispositioned in accordance with the Acceptance Criteria and

Corrective Actions program elements discussed below.

In accordance with ASME Code Case N-722, visual examinations of highly susceptible Alloy 600/82/182 pressure retaining components are required during each refueling outage. Other Alloy 600/82/182 pressure retaining components

that are considered less susceptible to PWSCC cracking are required to be

examined by visual examinations every other refueling outage or once per

interval. The PINGP Nickel-Alloy Nozzles and Penetrations Program ranks the

reactor pressure vessel bottom head penetrations as moderate for their lower

susceptibility to PWSCC given the cooler temperature environment, good volumetric examination experience, and the medium-to-high failure consequence. ASME Code Case N-722 requires inspection of the reactor

pressure vessel bottom head penetrations every other refueling outage.

The reactor vessel core support pads are structural, non-pressure retaining, components and consequently are not included in ASME Code Case N-722. In accordance with 10 CFR 50.55a(g)(6)(ii)(E)(1), the inspection requirements of ASME Code Case N-722 do not apply to components with pressure retaining

welds fabricated with Alloy 600/82/182 materials that have been mitigated by weld overlay. Therefore, the Unit 2 pressurizer surge nozzle-to-safe end weld is

not covered under ASME Code Case N-722. This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.5, Monitoring and Trending.

Acceptance Criteria The implementing procedure or work order specifies examination requirements and acceptance criteria in accordance with the applicable regulatory

requirements or industry guidelines. The acceptance criteria, against which the Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 8need for corrective actions are evaluated, ensure that the component intended functions are maintained under all current licensing basis design conditions

during the period of extended operation.

For the reactor vessel core support pads, the PINGP ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program conducts visual VT-1 examination of the accessible welds. The PINGP ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program requires that indications

and relevant conditions detected during examination be evaluated in accordance

with ASME Section XI, Paragraph IWB-3520.1. The Inservice Inspection requirements for the Unit 2 pressurizer surge nozzle-to-safe end weld for the extended period of operation will be in accordance with the

PINGP ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program, with specified limitations, modifications and NRC-approved alternatives. Indications and relevant conditions detected during examination are required to be evaluated in accordance with ASME Section XI, Article IWB-3500.

The weld is also subject to visual examination (VT-2) during pressure testing in accordance with ASME Section XI, Table IWB-2500-1, Examination Category B-P. The examination results are evaluated in accordance with the acceptance criteria of ASME Section XI, Paragraph IWB-3522.

Reactor pressure vessel bottom head bare metal visual examinations are performed in order to identify very small volumes of boric acid that may result

from Alloy 600 PWSCC. The acceptance criteria for this examination is the lack of any relevant indication, namely evidence of any leakage arising from the

penetration to head interface, and the lack of any boric acid accumulations on the carbon steel head surfaces that may result in corrosion. The acceptance standards are in accordance with ASME Section XI, Paragraph IWB-3522 per ASME Code Case N-722, subject to the conditions specified in 10 CFR 50.55a(g)(6)(ii)(E). The acceptance standards of ASME Section XI, Paragraph IWB-3522 are also applied to relevant indications identified during system pressure testing performed in accordance with ASME Section XI, Table IWB-

2500-1, Examination Category B-P. This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.6, Acceptance Criteria.

Corrective Actions Indications are evaluated per the acceptance criteria, which determine relevant flaw indications that are unacceptable for further service. Relevant flaw indications are corrected through implementation of appropriate

repair/replacement activities.If visual examination of the reactor vessel instrumentation tube penetrations (bottom head) in accordance with ASME Code Case N-722 identifies leakage or Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 9evidence of cracking, additional actions shall be performed as specified in paragraphs 10 CFR 50.55a(g)(6)(ii)(E)(2) through (4). If PWSCC related flaws

are detected in the pressurizer surge nozzle FSWOL, the repair/replacement activity will include removal of the weld overlay and the original dissimilar metal

weld.Repair/replacement activities comply with ASME Section XI as invoked by 10 CFR 50.55a or approved ASME Code Cases as referenced in the latest version of NRC Regulatory Guide 1.147. Proposed alternative repair/replacement activities, if any, will be submitted to the NRC for review and approval in accordance with 10 CFR 50.55a(a)(3)(i) or 10 CFR 50.55a(a)(3)(ii).

Identified flaws are entered into the PINGP Corrective Action Program for appropriate disposition. A repair, replacement or an evaluation is performed for

all flaws that exceed the acceptance standards.

See LRA Section B1.3 for further discussion of this element.

This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.7, Corrective Actions.

Confirmation Process Corrective action effectiveness is part of the PINGP Corrective Action Program.

See LRA Section B1.3 for further discussion of this element.

This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.8, Confirmation Process.

Administrative Controls See LRA Section B1.3for the discussion of this element. This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.9, Administrative Controls.

Operating Experience A review of operating experience for the PINGP Nickel-Alloy Nozzles and Penetrations Program identified no adverse trends or issues with program performance. The review of operating experience, as discussed below, indicates that the Nickel-Alloy Nozzles and Penetrations Program is effective in utilizing inspections, mitigation techniques, and repair/replacement activities.

A visual VT-1 examination of the accessible welds of the reactor vessel core support pads was conducted in October 2004 for Unit 1 and in May 2005 for Unit Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 102. No recordable indications on the core support pads were detected in either

Unit.The PlNGP Unit 2 pressurizer surge nozzle-to-safe end weld was ultrasonically examined in November 2006. The examination met the ASME Section XI and EPRI MRP-139 requirements for examination coverage. No PWSCC indications were detected. Ultrasonic examinations of the Unit 2 surge nozzle-to-safe end

dissimilar metal weld were conducted again in September 2008, prior to

installation of the full structural weld overlay (FSWOL). No reportable indications were identified. Then in October 2008, following installation of the FSWOL, ultrasonic examinations (UT) were performed of the new overlay weld and the

nozzle-to-safe end dissimilar metal weld. 100 percent of the Code required

volume was achieved during the examinations. The UT exams resulted in no recordable indications. PINGP conducted bare metal visual examinations of the reactor vessel instrumentation tube penetrations (bottom head) in May 2006 for Unit 1 and April

2005 for Unit 2. No indications were observed.

Industry operating experience is monitored to incorporate any necessary changes to the program. PINGP participates in industry activities such as the

Owners Group meetings and the EPRI Material Reliability Program initiatives to monitor or inspect for cracking due to PWSCC. The PINGP Nickel-Alloy Nozzles and Penetrations Program effectively monitors the condition of components within the license renewal boundary and ensures

aging effects are acceptably managed. This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.10, Operating Experience.

ConclusionThe PINGP Nickel-Alloy Nozzles and Penetration Program is an existing program that manages the effects of cracking due to primary water stress corrosion cracking of the nickel-alloy pressure boundary and structural components exposed to primary coolant. The program has been effective in monitoring the nickel-alloy components and no adverse trends or significant conditions related to these components have

been identified.

Implementation of the Nickel-Alloy Nozzles and Penetrations Program provides reasonable assurance that aging effects will be managed such that structures, systems, and components within the scope of this program will continue to perform

their intended function(s) during the period of extended operation.

Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 11In the list of Preliminary License Renewal Commitments, Commitment 21 is deleted in its entirety. A revised preliminary commitment list which reflects this change is provided

as Enclosure 2 to this letter. Updated Preliminary License Renewal Commitment List 16 Pages Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 1The following table provides the list of preliminary commitments included in the Application for Renewe Licenses (LRA) for Prairie Island Nuclear Generating Plant (PINGP) Units 1 and 2. These commitmen t contents of the LRA as submitted, and any updates provided in subsequent correspondence, but are c opreliminary in that the specific wording of some commitments may change, and additional commitment s during the NRC review of the LRA. The final commitments as submitted by NMC, and accepted by NRC, are expected to be confirmed in t Evaluation Report (SER) for the renewed operating licenses. The final commitments, as confirmed in t h become effective upon NRC issuance of the renewed op erating licenses. In addition, as stated in the Lcommitments will be incorporated into the Updated Safety Analysis Report (USAR).

Commitment Number Commitment Implementation Schedule 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 2 Following 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 3 An 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 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 2 Commitment Number Commitment Implementation Schedule4 Procedures 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

Nuclear Power Plants Program will be enhanced to include

guidance for visual inspections of installed bolting.

U1 - 8/9/2013 U2 - 10/29/2014 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 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 U1 - 8/9/2013 U2 - 10/29/2014 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 3 Commitment Number Commitment Implementation Schedule to use a particle size methodology as specified in

ISA S7.0.01. The program will incorporate on-line dew point

monitoring.[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 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 10 An 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 U1 - 8/9/2013 U2 - 10/29/2014 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 4 Commitment Number Commitment Implementation Schedulewill 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 effects 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] 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.

U1 - 8/9/2013 U2 - 10/29/2014 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 5 Commitment Number Commitment Implementation Schedule[Revised in letter dated 12/5/2008 in response to RAI B2.1.15-3]

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. 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.

U1 - 8/9/2013 U2 - 10/29/2014 14 The 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.

U1 - 8/9/2013 U2 - 10/29/2014 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 6 Commitment Number Commitment Implementation Schedule The program will require that flux thimble tubes that cannot be inspected must be removed from service. 15 The 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 annual basis. One-time ultrasonic thickness measurements will be performed at selected tank bottom and piping locations prior to the period of extended operation.

U1 - 8/9/2013 U2 - 10/29/2014 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 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 U1 - 8/9/2013 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 7 Commitment Number Commitment Implementation Schedule (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 omitted.U2 - 10/29/2014 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

[Withdrawn 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 U1 - 8/9/2013 U2 - 10/29/2014 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 8 Commitment Number Commitment Implementation Scheduleimplementing 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

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 U1 - 8/9/2013 U2 - 10/29/2014 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 9 Commitment Number Commitment Implementation Scheduledescribed in LRA Section B2.1.30.

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 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

U1 - 8/9/2011 U2 - 10/29/2012 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 U1 - 8/9/2013 U2 - 10/29/2014 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 10 Commitment Number Commitment Implementation Scheduleplaced in storage and maintained for future insertion.

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 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 11 Commitment Number Commitment Implementation Schedule 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 U1 - 8/9/2013 U2 - 10/29/2014 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 12 Commitment Number Commitment Implementation Schedule SBO cable vault and bus duct enclosures.

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.

U1 - 8/9/2013 U2 - 10/29/2014 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 13 Commitment Number Commitment Implementation Schedule The program will require Reactor Coolant System dissolved oxygen Action Level limits to be consistent with

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 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 14 Commitment Number Commitment Implementation Schedule 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 36NSPM 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 8/9/2013 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 15 Commitment Number Commitment Implementation Scheduleground-disturbing activities. The procedures will be revised to

include drawings and illustrations to assist users in identifying 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 8/9/2013 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 16 Commitment Number Commitment Implementation Schedule 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

discussed in Commitment Number 38).

[Added in ER revision submitted in letter dated 3/4/2009]

1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121 March 27, 2009 L-PI-09-042 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

Supplemental Information Regarding 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 Pl ant (PINGP) Units 1 and 2. This letter amends the LRA to provide supplemental information addressing certain issues that have been raised as contentions in this License Renewal proceeding.

contains revisions to LRA Se ctions A2.27 and B2.1.27 to incorporate information regarding the plant-specific Nickel-Alloy Nozzles and Penetrations Program.

Conforming changes are also provided for LR A Sections 2, 3, B1.1, B1.5, and B2.0.

provides an updated version of the Preliminary License Renewal Commitments list contained in the LRA transmitta l letter. This updated list reflects the commitment changes made to date in NSPM correspondence.

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 no new commitment

s. Commitment No. 21 in the list of Preliminary License Renewal Commitm ents contained in the LRA transmittal letter dated April 11, 2008, is withdrawn.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on March 27, 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 Document Control Desk Page 2 Enclosures (2)

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

Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 1 LRA Sections A2.27 and B2.1.27, and selected information in LRA Sections 2, 3, B1.1, B1.5, and B2.0, are hereby revised to provide updated information regarding the Nickel-

Alloy Nozzles and Penetrations Program. The updated information designates the program as a plant-specific pr ogram, expands the level of det ail in the LRA, and reflects the incorporation of new NRC requirements from 10 CFR 50.55a. The LRA changes are as follows:

In LRA Section 2.1.1.3, Interim Staff Guidance (ISG) Discu ssion, on Page 2.1-5, the second paragraph of the discussion for LR-ISG

-19B is deleted and replaced with the following:

The PINGP Nickel-Alloy Penetration Nozzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reac tors Program manages aging effects on the reactor head penetrations. With respec t to the management of nickel-alloy nozzles and penetrations not associated wit h the reactor vessel head penetration nozzles, a plant-specific program, the Nickel-Alloy Nozzles and Penetrations Program, is credited with managing the effect s of cracking due to PWSCC. As this issue evolves under the existing regulatory process, these programs will be modified as necessary in response to industry initiatives and NRC guidance and

requirements.

In LRA Section 3.1.2.2.13, Cracking due to Primary Water Stress Corrosion Cracking (PWSCC), on Pages 3.1-14 and 3.1-15, the eighth sentence "For the Nickel-Alloy Nozzles and Penetrations Program ... and staff-accepted industry guidelines." is deleted and replaced with the following:

The Nickel-Alloy Nozzles and Penetrati ons Program ranks the Alloy 600/82/182 locations for primary water stress corrosi on cracking susceptibi lity and then utilizes inspections, mitigation techniques, and repai r/replacement activities to manage the effects of cracking due to primary water stress corrosion cracking. The Nickel-Alloy Nozzles and Penetrations Program impl ements the inspection of the Alloy 600/82/182 materials through the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program.

In LRA Table 3.1.2-1, Reactor Vessel, Internals, and Reacto r Coolant System -

Pressurizer System - Summary of Aging Management Evaluation, on Page 3.1-51, in the line item for Surge Nozzle being m anaged by the Nickel-Alloy Nozzles and Penetrations Program, the Notes entry is changed fr om A, 111 to E, 111.

In LRA Table 3.1.2-4, Reactor Vessel, Inte rnals, and Reactor Cool ant System - Reactor Vessel System - Summary of Aging Managem ent Evaluation, on Page 3.1-97, in the line item for Core Support Pads being managed by the Nickel-Alloy Nozzles and Penetrations Program, the Notes entry is changed from A to E.

Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 2 In LRA Table 3.1.2-4, Reactor Vessel, Inte rnals, and Reactor Cool ant System - Reactor Vessel System - Summary of Aging Managem ent Evaluation, on Page 3.1-100, in the line item for Instrumentation Tube Penetra tions (Bottom Head) bei ng managed by the Nickel-Alloy Nozzles and Penetrations Program, the Notes entry is changed from A to E.

LRA Section A2.27 on Pages A-11 and A-12 is re vised in its entirety to read as follows:

A2.27 Nickel-Alloy Nozzles and Penetrations Program The Nickel-Alloy Nozzles and Penetrations Program manages the aging effect of cracking due to primary water stress corro sion cracking (PWSCC) of nickel-alloy pressure boundary and structural componen ts exposed to primary coolant. The Alloy 600/82/182 locations are ranked fo r PWSCC susceptibility. The program manages these components for cracking due to PWSCC utilizing inspections, mitigation techniques, and repair/replacemen t activities. The program implements the inspection of the Alloy 600/82/182 ma terials through t he ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program.

The program complies with applicable NRC Orders, and implements applicable NRC Bulletins, Generic Letters, and st aff-accepted industry guidelines.

In LRA Section B1.1, Overview, as revised in Enclosure 1 of NSPM letter dated March 12, 2009, the second paragraph, "Two of the programs consist of commitments ...NUREG-1801 Chapter IV." is deleted in its entirety and replaced with the following:

In certain cases a program may consist so lely of a commitment if the associated NUREG-1801 program description is only comprised of specified commitment language in NUREG-1801 Chapter IV (e.g., NUREG-1801 Programs XI.M11 and XI.M16). Where this appendix provides a full ten element description of a program that NUREG-1801 only defines as a commitm ent, the program is designated a plant-specific program. Plant-specific pr ograms are described in terms of their consistency with the ten generic program elements defined in Appendix A.1, Section A.1.2.3 of NUREG-1800.

In LRA Section B1.5, Aging Management Progr ams, on Page B-4, the introductory paragraph is revised to read as follows:

The AMPs credited with managing the effe cts of aging at PINGP are described in the following sections. Where thes e programs are addressed in NUREG-1801, they have been evaluated as being either fully consistent wit h, or, with some exceptions and/or enhancement s, consistent with the NUREG-1801 programs.

Where plant-specific AMPs are cr edited, they have been evaluated for Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 3consistency with the ten elements of an acceptable aging ma nagement program defined in Appendix A.1, Sect ion A.1.2.3 of NUREG-1800.

In LRA Section B2.0, Aging Management Programs Correlation, on Page B-7, line item XI.M11 of the NUREG-1801 program correlation table is revi sed to appear as follows:

NUREG-1801 ID NUREG-1801 Program PINGP Program NUREG-1801 Comparison XI.M11 Nickel-Alloy Nozzles and

Penetrations Nickel-Alloy Nozzles and

Penetrations Program

[Section B2.1.27]

Existing Plant-Specific Program

In LRA Section B2.0 on Page B-12, Note 1 is revised in its entirety to read, "Not Used."

LRA Section B2.1.27 on Page B-58 is revi sed in its entirety to read as follows:

B2.1.27 Nickel-Alloy Nozzles and Penetrations Program Program Description The Nickel-Alloy Nozzles and Penetrations Program manages the aging effect of cracking due to primary water stress corro sion cracking (PWSCC) of nickel-alloy pressure boundary and structural componen ts exposed to primary coolant. The Alloy 600/82/182 locations are ranked fo r PWSCC susceptibility. The program manages these components for cracking due to PWSCC utilizing inspections, mitigation techniques, and repair/replacement activities. The Nickel-Alloy Nozzles and Penetrations Program implements the inspection of the Alloy 600/82/182 materials through the ASME Section XI Inserv ice Inspection, Subsections IWB, IWC, and IWD Program.

The program complies with applicable NRC Or ders, and implements applicable NRC Bulletins, Generic Letters, and st aff-accepted industry guidelines.

NUREG-1800 Consistency The Prairie Island Nuclear Generating Plan t Nickel-Alloy Nozzles and Penetrations Program is an existing plant-specific program. The progr am consists of the ten elements of an acceptable AMP as described in NUREG-1800 Appendix A.1, Section A.1.2.3 and Table A.1-1.

Exceptions to NURE G-1800 or NUREG-1801 None

Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 4 Enhancements None Aging Management Program Elements

The elements of the Nickel-Alloy Nozzles and Penetrations Program are described below. The results of an evaluation of each element with respect to the NUREG-1800, Appendix A.1, Secti on A.1.2.3, "Aging Managemen t Program Elements" and Table A.1-1, "Elements of an Aging Manage ment Program for License Renewal," are also provided.

Scope of Program The Nickel-Alloy Nozzles and Penetrations Program manages the aging effect of cracking due to primary water stress corrosion cracking (PWSCC) of pressure boundary and structural components c onstructed of Alloy 600 and welds constructed of the associated Alloy 82/

182 filler metals exposed to primary coolant. The scope of the Nickel-Alloy Nozzles and Penetrations Program includes the Reactor Vessel System in strumentation tube penetrations (bottom head), thirty-six in each Unit, construc ted of Alloy 600; the Reactor Vessel System core support pads (Alloy 600), four in each Unit; and the Unit 2 pressurizer surge nozzle-to-safe end dissi milar metal weld (Alloy 82).

The program is based upon the industry guidance provided in EPRI MRP-126, "Generic Guidance for Alloy 600 Managem ent" (1009561), November 2004 and NEI 03-08, "Guideline for the Management of Materials Issues," May 2003. The program complies with applicable NRC Orders, and implements applicable NRC Bulletins, Generic Letters, and st aff-accepted industry guidelines.

This AMP consists of PINGP activities that manage aging e ffects for components of the following systems:

• Pressurizer (PS) System

• Reactor Vessel (RV) System This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.1, Sc ope of Program.

Preventive Actions The program considers various mitigativ e and repair options to ensure that nickel-alloy components continue to perfo rm their intended functions during the period of extended oper ation. Selection of a mitigation strategy may be based upon available technology, ASME Code requirements, radiological conditions, and economic factors. Most mitigative actions implemented by the industry since the mid-1990s have utilized PWSCC resistant Alloy 690/52/152 materials.

Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 5 To mitigate the effects of PWSCC on the Unit 2 pressurizer surge nozzle weld (Alloy 82), a full structural weld over lay (FSWOL) on the pressurizer surge nozzle-to-safe end dissimilar metal and safe end-to-reducer stainless steel butt welds was installed in October 2008 duri ng the PINGP Unit 2 refueling outage (2R25). The FSWOL was installed using Alloy 52M weld material.

Preventive measures to mitigate PWSCC are in accordance with the PINGP Water Chemistry Program. The Wa ter Chemistry Program manages aging effects by controlling concentrations of k nown detrimental chemical species such as chlorides, fluorides, sulfates and dissolved oxygen below the levels known to cause degradation. The program includes specifications for chemical species, sampling and analysis frequencies, and corrective actions for control of water

chemistry. This program conforms to the EPRI PW R Primary Water Chemistry Guidelines. The PINGP Water Chemistry Program is further described in LRA Section B2.1.40, Wate r Chemistry Program.

This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.2, Preventive Actions.

Parameters Monitored/Inspected The PINGP Nickel-Alloy Nozzles and Penetra tions Program monitors for cracking due to PWSCC of Alloy 600/82/182 material s exposed to primary coolant. The program performs condition monitoring exam inations of the lower reactor vessel head surface and each bottom-mounted instrume ntation tube penetration. These examinations monitor for through-wall cracks that may exist in the nozzles or their associated partial penetration J-gr oove welds. For in-scope pressure boundary components, the program monitors for evidence of reactor coolant leakage which may manifest it self in the form of boric acid residues or corrosion products. The core support pads and the Un it 2 pressurizer surge nozzle-to-safe end weld are monitored for evidence of cr acking. The Nickel-Alloy Nozzles and Penetrations Program implements the inspection of the Alloy 600/82/182 materials through the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program.

This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.3, Parameters M onitored or Inspected.

Detection of Aging Effects The program utilizes visual and volumetric examination techniques to detect cracking in Alloy 600/82/182 materials. 10 CFR 50.55a requires that all power reactors maintain an Inservice Inspecti on Program in accordance with the ASME Boiler and Pressure Vessel Code, Sect ion XI. The Nickel-Alloy Nozzles and Penetrations Program implements the inspection of the Alloy 600/82/182 Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 6materials through the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program.

For the reactor vessel core support pads, the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program conducts a visual VT-1 examination of the accessi ble interior attachment welds per Table IWB-2500-1, Examination Category B-N-2, once per Inservice Inspection interval.

The FSWOL on the pressurizer surge nozzle-to-safe end dissimilar metal weld (Alloy 82) and safe end-to-reducer stainless steel butt weld was installed at PINGP Unit 2 with NRC approval. Inservice examinations are conducted ultrasonically with the ex amination volume as defined in ASME Section XI, Nonmandatory Appendix Q, Figure Q-4300-1.

Inservice examinations as described in Q-4300 are performed in acco rdance with the requi rements of MRP-139, "Primary System Piping Butt Weld Inspection and Evaluation Guidelines," with the additional requirement of at least one ultrasonic examination within ten years of the FSWOL application. T he MRP-139 guidance for ISI goes beyond the current ASME Code inspection requirements for PINGP Unit 2.

With the installation of the FSWOL, the Unit 2 pressurizer surge nozzle-to-safe end weld reverts back to Category "B" per MRP-139 which will require

examinations in accordanc e with the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Pr ogram, with specif ied limitations, modifications and NRC-approved alternatives.

Reactor pressure vessel bottom head bar e metal visual examinations are performed by removing insulation sections and/or examining under the insulation using remote viewing equipment that pr ovides a high degree of resolution in order to identify very small volumes of bor ic acid that may result from Alloy 600 PWSCC. In response to NRC Bulletin 2003-02, PINGP commi tted to 100% bare metal visual inspection of the lower reactor pressure vessel dome up to and including each bottom mounted instru mentation (BMI) penet ration to RPV junction every other refueling outage. Fully qualified and certified VT-2 examiners perform the inspections of the bottom head penetrations. The inspections are in compliance with ASME Code Case N-722, "Additional Examinations for PWR Pressure Re taining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials," as required by and modified by 10 CFR 50.55a(g)(6)(ii)(E). 10 CFR 50.55a(g)(6)(ii)(E) requires non-visual inspection for items where leakage is identified to determine whether circumferential cracking is present in the flawed material and if multiple circumferential flaws have initiated.

These inspections of nickel-alloy component s ensure timely detection of cracking due to PWSCC prior to a loss of component intended function. The program complies with applicable NRC Orders, and implements applicable NRC Bulletins, Generic Letters, and staff-accepted industry guidelines.

Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 7This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.4, Detection of Aging Effects.

Monitoring and Trending

The Nickel-Alloy Nozzles and Penetrati ons Program ranked the Alloy 600/82/182 locations based on four main criteria

PWSCC susceptibility (e.g., operational time and temperature), failure consequence, leakage detection margin, and radiation dose rates. Additionally, material heat suscept ibility and other industry experience were also considered.

The program incorporates the ins pection schedules and frequencies for the nickel-alloy components in accordance with the PINGP ASME Section XI Inservice Inspection, Subsections IW B, IWC, and IWD Pr ogram and, where applicable, ASME Code Case N-722, subjec t to the conditions specified in 10 CFR 50.55a(g)(6)(ii)(E). Flaw indi cations detected during the required examinations are disposit ioned in accordance with the Acceptance Criteria and Corrective Actions program elements discussed below.

In accordance with ASME Code Case N-722, visual examinations of highly susceptible Alloy 600/82/182 pressure re taining components are required during each refueling outage. Other Alloy 600/82/182 pressure retaining components that are considered less susceptible to PWSCC cracking are required to be examined by visual examinations ever y other refueling outage or once per interval. The PINGP Nickel-Alloy Nozzl es and Penetrations Program ranks the reactor pressure vessel bottom head penet rations as moderate for their lower susceptibility to PWSCC given the cooler temperature environment, good volumetric examination experience, and the medium-to-high failure consequence. ASME Code Case N-722 requires inspection of the reactor

pressure vessel bottom head penetrations every other refueling outage.

The reactor vessel core support pads are structural, non-pressure retaining, components and consequently are not included in ASME Code Case N-722. In accordance with 10 CFR 50.55a(g)(6)(ii)(E)(1), the inspection requirements of ASME Code Case N-722 do not apply to components with pressure retaining welds fabricated with Alloy 600/82/182 materials that have been mitigated by weld overlay. Therefore, the Unit 2 pressurizer surge nozzle-to-safe end weld is not covered under ASME Code Case N-722.

This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.5, Moni toring and Trending.

Acceptance Criteria The implementing procedure or work order specifies examination requirements and acceptance criteria in accordance with the applicable regulatory requirements or industry guidelines. The acceptance criteria, against which the Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 8need for corrective actions are evaluated, ensure that the component intended functions are maintained under all current licensing basis design conditions during the period of ex tended operation.

For the reactor vessel core support pads , the PINGP ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program conducts visual VT-1 examination of the accessible welds. The PINGP ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program requires that indications and relevant conditions detec ted during examination be evaluated in accordance with ASME Section XI, Paragraph IWB-3520.1.

The Inservice Inspection requirements for the Unit 2 pressurizer surge nozzle-to-safe end weld for the extended period of operation will be in accordance with the PINGP ASME Section XI Inservice Inspec tion, Subsections IWB, IWC, and IWD Program, with specified limitations, modifications and NRC-approved alternatives. Indications and relevant c onditions detected during examination are required to be evaluated in accordance with ASME Section XI, Article IWB-3500. The weld is also subject to visual exam ination (VT-2) during pressure testing in accordance with ASME Section XI, Table IWB-2500-1, Examination Category B-P. The examination results are evaluated in accordance with the acceptance criteria of ASME Sect ion XI, Paragraph IWB-3522.

Reactor pressure vessel bottom head bar e metal visual examinations are performed in order to identify very small volumes of boric acid that may result from Alloy 600 PWSCC. The acceptance crit eria for this examination is the lack of any relevant indication, namely ev idence of any leakage arising from the penetration to head interface, and the lack of any boric acid accumulations on the carbon steel head surfaces that may result in corrosion. The acceptance standards are in accordance with ASME Section XI, Paragraph IWB-3522 per ASME Code Case N-722, subject to t he conditions specified in 10 CFR 50.55a(g)(6)(ii)(E). The acceptance st andards of ASME Section XI, Paragraph IWB-3522 are also applied to relevant indications identified during system pressure testing performed in accordance with ASME Section XI, Table IWB-2500-1, Examinati on Category B-P.

This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.6, Acceptance Criteria.

Corrective Actions Indications are evaluated per the acceptance criteria, which determine relevant flaw indications that are unacceptable fo r further service. Relevant flaw indications are corrected through implementation of appropriate repair/replacement activities.

If visual examination of the reactor vessel instrumentation tube penetrations (bottom head) in accordance with ASME Code Case N-722 identifies leakage or Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 9evidence of cracking, additional actions shall be performed as specified in paragraphs 10 CFR 50.55a(g)(6)(ii)(E)(2) th rough (4). If PWSCC related flaws are detected in the pressurizer surge nozzle FSWOL, the repair/replacement activity will include removal of the weld overlay and the original dissimilar metal weld.

Repair/replacement activities comply wit h ASME Section XI as invoked by 10 CFR 50.55a or approved ASME Code Cases as referenced in the latest version of NRC Regulatory Guide 1.147. Proposed alternative repair/replacement activities, if any, will be submitted to the NRC for review and approval in accordance with 10 CFR 50.55a(a)(3)(i) or 10 CFR 50.55a(a)(3)(ii).

Identified flaws are entered into the PINGP Corrective Action Program for appropriate disposition. A repair, replacement or an evaluation is performed for all flaws that exceed the acceptance standards.

See LRA Section B1.3 for furt her discussion of this element.

This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.7, Corrective Actions.

Confirmation Process

Corrective action effectiveness is part of the PINGP Correctiv e Action Program.

See LRA Section B1.3 for furt her discussion of this element.

This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.8, Confirmation Process.

Administrative Controls See LRA Section B1.3 for the discussion of this element.

This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.9, Administrative Controls.

Operating Experience A review of operating experience for the PINGP Nickel-Alloy Nozzles and Penetrations Program identified no adv erse trends or issues with program performance. The review of operating experience, as discussed below, indicates that the Nickel-Alloy Nozzles and Penetrations Program is effective in utilizing inspections, mitigation techniques, and repair/replacement activities.

A visual VT-1 examination of the acce ssible welds of the reactor vessel core support pads was conducted in October 2004 for Unit 1 and in May 2005 for Unit Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 102. No recordable indications on the core support pads were detected in either Unit.

The PlNGP Unit 2 pressurizer surge nozzl e-to-safe end weld was ultrasonically examined in November 2006. The examination met the ASME Section XI and EPRI MRP-139 requirements fo r examination coverage. No PWSCC indications were detected. Ultrasonic examinations of the Unit 2 surge nozzle-to-safe end dissimilar metal weld were conduct ed again in September 2008, prior to installation of the full structural weld ov erlay (FSWOL). No reportable indications were identified. Then in October 2008, following installation of the FSWOL, ultrasonic examinations (UT) were perfo rmed of the new overlay weld and the nozzle-to-safe end dissimilar metal weld. 100 percent of the Code required volume was achieved during the examinations. The UT exams resulted in no recordable indications.

PINGP conducted bare metal visual examinations of the reactor vessel instrumentation tube penetrati ons (bottom head) in May 2006 for Unit 1 and April 2005 for Unit 2. No indications were observed.

Industry operating experience is moni tored to incorporate any necessary changes to the program. PINGP participates in industry activities such as the Owners Group meetings and the EPRI Material Reliability Program initiatives to monitor or inspect for cracking due to PWSCC.

The PINGP Nickel-Alloy Nozzles and Penetra tions Program effectively monitors the condition of components within the license renewal boundary and ensures aging effects are acceptably managed.

This element is consistent with the recommendations of NUREG-1800, Element A.1.2.3.10, Operating Experience.

Conclusion The PINGP Nickel-Alloy Nozzles and Penetra tion Program is an existing program that manages the effects of cracking due to primary water stress corrosion cracking of the nickel-alloy pressure boundary and st ructural components exposed to primary coolant. The program has been effective in monitoring the nickel-alloy components and no adverse trends or significant condi tions related to these components have been identified.

Implementation of the Nicke l-Alloy Nozzles and Penetra tions Program provides reasonable assurance that aging effects will be managed such that structures, systems, and components within the scope of this program will continue to perform their intended function(s) during the period of extended operation.

Revisions to LRA Sections 2 and 3 and Appendices A and B Regarding the Nickel-Alloy Nozzles and Penetrations Program 11In the list of Preliminary License Renewa l Commitments, Commitment 21 is deleted in its entirety. A revised preliminary commitment list which reflects this change is provided as Enclosure 2 to this letter.

Enclosure 2 Updated Preliminary License Renewal Commitment List

16 Pages Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 1 The following table provides the list of preliminary commitments included in the Application for Renewed Operating Licenses (LRA) for Prairie Island Nucl ear Generating Plant (PINGP) Units 1 and

2. These commitments reflect the contents of the LRA as submitted, and any updates provided in subsequent correspondence, but are considered preliminary in that the s pecific wording of some commitments may change, and additional commitments may be made, during the NRC review of the LRA. The final commitments as submitted by NMC, and accepted by NRC, are expected to be confirmed in the NRC's Safety Evaluation Report (SER) for the renewed operating licenses. The final commitm ents, as confirmed in the SER, will become effective upon NRC issuance of the renewed operating licenses. In addition, as stated in the LRA, the final commitments will be incorporated into the Updated 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 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 2Commitment Number Commitment Implementation Schedule Related LRA Section Number 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 Nuclear Power Plants Program will be enhanced to include guidance for visual inspections of installed bolting.

U1 - 8/9/2013 U2 - 10/29/2014 B2.1.6 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 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 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 U1 - 8/9/2013 U2 - 10/29/2014 B2.1.10 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 3Commitment Number Commitment Implementation Schedule Related LRA Section Number to use a particle size methodology as specified in

ISA S7.0.01.

The program will incorpor ate on-line dew point monitoring. [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 U1 - 8/9/2013 U2 - 10/29/2014 B2.1.14 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 4Commitment Number Commitment Implementation Schedule Related LRA Section Number 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 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 trending 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.

U1 - 8/9/2013 U2 - 10/29/2014 B2.1.15 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 5Commitment Number Commitment Implementation Schedule Related LRA Section Number [Revised in letter dated 12/5/2008 in response to RAI B2.1.15-3]

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. 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.

U1 - 8/9/2013 U2 - 10/29/2014 B2.1.16 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.

U1 - 8/9/2013 U2 - 10/29/2014 B2.1.18 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 6Commitment Number Commitment Implementation Schedule Related LRA Section Number The program will require that flux thimble tubes that cannot be inspected must be removed from service. 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 annual basis. One-time ultrasonic thickness measurements will be performed at selected tank bottom and piping locations prior to the period of extended operation.

U1 - 8/9/2013 U2 - 10/29/2014 B2.1.19 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 C ables 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 corrosi on cracking will be performed by visual examination with a magn ified 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 U1 - 8/9/2013 B2.1.23 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 7Commitment Number Commitment Implementation Schedule Related LRA Section Number (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 omitted. U2 - 10/29/2014 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

[Withdrawn 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 U1 - 8/9/2013 U2 - 10/29/2014 B2.1.28 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 8Commitment Number Commitment Implementation Schedule Related LRA Section Number 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 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 U1 - 8/9/2013 U2 - 10/29/2014 B2.1.30 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 9Commitment Number Commitment Implementation Schedule Related LRA Section Number described in LRA Section 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 months before entering the per iod of extend ed operation, submit an inspection plan for reactor internals to the NRC for review and approval.

U1 - 8/9/2011 U2 - 10/29/2012 B2.1.32 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 U1 - 8/9/2013 U2 - 10/29/2014 B2.1.34 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 10Commitment Number Commitment Implementation Schedule Related LRA Section Number placed in storage and maintained for future insertion.

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 Preliminary License Renewal Commitments Updated through 3/27/2009 11Commitment 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 c able tray, conduit, cable, tubing tray, tubing, non-ASME vessels, exchangers, pumps, va lves, 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 U1 - 8/9/2013 U2 - 10/29/2014 B2.1.38 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 12Commitment Number Commitment Implementation Schedule Related LRA Section Number SBO cable vault and bus duct enclosures.

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.

U1 - 8/9/2013 U2 - 10/29/2014 B2.1.40 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 13Commitment Number Commitment Implementation Schedule Related LRA Section Number The program will require R eactor Coolant System dissolved oxygen Action Level limits to be consistent with 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 Preliminary License Renewal Commitments Updated through 3/27/2009 14Commitment 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 Bo undary 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 8/9/2013 ER 4.16.1 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 15Commitment Number Commitment Implementation Schedule Related LRA Section Number ground-disturbing activities. The procedures will be revised to

include drawings and illustrations to assist users in identifying 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 8/9/2013 ER 4.16.2 Prairie Island Nuclear Generating Plant Preliminary License Renewal Commitments Updated through 3/27/2009 16Commitment Number Commitment Implementation Schedule Related LRA Section Number 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

discussed in Commitment Number 38).

[Added in ER revision submi tted in letter dated 3/4/2009]