NOC-AE-15003270, Response to Requests for Additional Information for the Review of the South Texas Project, Units 1 and 2, License Renewal Application - Set 28

Response to Requests for Additional Information for the Review of the South Texas Project, Units 1 and 2, License Renewal Application - Set 28

ML15197A029
Person / Time
Site:	South Texas
Issue date:	06/30/2015
From:	Gerry Powell South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-15003270, TAC ME4936, TAC ME4937
Download: ML15197A029 (75)

Text

Nuclear Operating Company South Texas Prolect Electric Generating Station P.O. Box 289 Wadsworth, Texas 77483 ,xIM .-

June 30, 2015 NOC-AE-1 5003270 10 CFR 54 STI: 34151048 File: G25 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Response to Requests for Additional Information for the Review of the South Texas Project, Units 1 and 2, License Renewal Application - Set 28 (TAC Nos. ME4936 and ME4937)

References:

1. Letter from G. T. Powell, STP, to NRC Document Control Desk, "License Renewal Application", dated October 25, 2010 (NOC-AE-1 0002607)

(ML103010257)

2. Letter from NRC to STP, "Requests for Additional Information for the Review of the South Texas Project, Units 1 and 2, License Renewal Application - Set 28", dated July 16, 2014, (TAC Nos. ME4936 and ME4937) (AE-NOC-14002550) (ML14183B719)

By Reference 1, STP Nuclear Operating Company (STP) submitted a License Renewal Application (LIRA) for South Texas Project Units 1 and 2.

By Reference 2, the NRC staff requested additional information for their review of the STP LIRA.

Requested information is related to STPs Aging Management of Reactor Vessel Internals, MRP-227A, and Reactor vessel clevis assemblies. STPNOC's response to the requests for additional information is provided in Enclosure 1 to this letter. Changes to LRA pages described in Enclosure 1 are depicted as line-in/line-out pages provided in Enclosure 2. provides a new LRA section Appendix C, Response to Applicant Action Items for Inspection and Evaluation Guidelines for PWR Internals and STP's Reactor Vessel Internals Program Inspection Plan.

There are no new regulatory commitments in this letter.

41(47:

NOC-AE-1 5003270 Page 2 of 3 Should you have any questions regarding this letter, please contact either Arden Aidridge, STP License Renewal Project Lead, at (361) 972-8243 or Rafael Gonzales, STP License Renewal Project regulatory point-of-contact, at (361) 972-4779.

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

Executed on ]-*e. 30) Zo*~

Date G. T. Powell Site Vice President RJG

Enclosures:

1. STPNOC Response to Requests for Additional Information

2. STPNOC LIRA Changes with Line-in/Line-out Annotations

3. New LRA Appendix C, Response to Applicant Action Items for Inspection and Evaluation Guidelines for PWR Internals and STP's Reactor Internals Program Inspection Plan

NOC-AE-1 5003270 Page 3 of 3 CC: (electronic copy)

(paper copy)

Morcian, Lewis & Bockius LLP Regional Administrator, Region IV Steve Frantz, Esquire U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard U.S. Nuclear Re~qulatory Commission Arlington, TX 76011-4511 Lisa M. Regner John W. Daily Tam Tran Lisa M. Regner Senior Project Manager NRG South Texas LP U.S. Nuclear Regulatory Commission John Ragan One White Flint North (O8H04) Chris O'Hara 11555 Rockville Pike Jim von Suskil Rockville, MD 20852 CPS Enercqy NRC Resident Inspector Kevin Polio U. S. Nuclear Regulatory Commission Cris Eugster P. O. Box 289, Mail Code: MNl16 L. D. Blaylock Wadsworth, TX 77483 John W. Daily Crain Caton & James, P.C.

License Renewal Project Manager Peter Nemeth (Safety)

U.S. Nuclear Regulatory Commission One White Flint North (MS 011-F1l) City of Austin Washington, DC 20555-0001 Cheryl Mele John Wester Tam Tran License Renewal Project Manager (Environmental) Texas Dept. of State Health Services U. S. Nuclear Regulatory Commission Richard A. Ratliff One White Flint North (MS 011 F01) Robert Free Washington, DC 20555-0001

Enclosure 1 NOC-AE-l15003270 Enclosure 1 STPNOC Response to Requests for Additional Information

Enclosure 1 NOC-AE-1 5003270 Page 1 of 8 SOUTH TEXAS PROJECT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION, SET 28 (TAC NOS. ME4936 AND ME4937)

RAI B2.1.35-1 - Aging Management of Reactor Vessel Internals and MRP-227A

Background:

The license renewal application (LIRA) for South Texas Project (STP), Units 1 and 2, proposed aging management for the reactor vessel internal (RVI) components based on a regulatory commitment in the LIRA's Updated Final Safety Analysis Report (UFSAR)

Supplement. The commitment stated that the applicant will develop an aging management program (AMP) and inspection plan based on augmented inspection activities for the components developed by the EPRI Materials Reliability Project (MRP),

and that the inspection plan will be submitted for NRC review and approval at least two (2) years prior to entering into the period of extended operation for STP, Units I and 2.

The NRC's recommended AMP for Pressured Water Reactor (PWR) RVIs in Revision 2 of NUREG-1801, "Generic Aging Lessons Learned (GALL) Report," is given in Section XI.M16A, "PWR Vessel Internals," which was issued in December 2010. On January 9, 2012, subsequent to the issuance of Revision 2 of the GALL Report, the EPRI MRP issued Technical Report No. 1022863, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A)," which included the NRC safety evaluation (SE) on the report's methodology dated December 16, 2011. On June 3, 2013, the staff revised AMP XI.M16A and the aging management review (AMR) items in the GALL Report for PWR RVI components to be consistent with the contents of the MRP-227-A report and issued them in License Renewal Interim Staff Guidance Document No. LR-ISG-201 1-04, "Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors."

On July 21, 2011, the NRC issued Regulatory Information Summary (RIS) 2011-07, "License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management," which provided updated NRC procedures for LIRA reviews of PWR RVI AMPs. This RIS identifies Category C plants as those plants that have an LIRA currently under review, and states that these applicants will be expected to revise their commitment for aging management of PWR vessel internals such that the information identified in the SE for MRP-227 would be submitted to the NRC for review and approval not later than two years after issuance of the renewed license or not later than two years before the plant enters the period of extended Operation, whichever comes first. STP, Units I and 2, are Category C plants in accordance with the RIS.

Issue:

The categorization of STP, Units I and 2, and other plants in Category C of RIS 2011-07 was based on an expectation that the LIRA would be reviewed and approved on a normal review schedule of 22 months, and that it would be an unreasonable burden to expect the applicant to address all aspects of the NRC's SE for MRP-227 within the LIRA review.

However, the applicant requested that the NRC place the LRA review on hold for the year 2013 to allow the applicant to address plant-specific technical issues. The staff noted that, as of the date of this RAI, the review of these technical issues is still on-going. Since the

Enclosure 1 NOC-AE-1 5003270 Page 2 of 8 resolution of these issues is not imminent and completion of the staff's review of the LRA is on-going, the staff has concluded that the applicant should provide an LRA update or amendment that includes updated AMP and AMR items for the RVI components, including responses to the applicable Applicant/License Actions Items identified in the staff's SE for MRP-227.

In addition to the issues identified above, the staff has noted that recent operating experience at one 4-loop PWR (in 2010) reported cracking and failures of some nickel alloy (Alloy X-750) clevis insert bolts in the clevis assemblies attached to the lower internal portion of its reactor vessel (see EPRI Technical Report No. 1022863, "Materials Reliability Program: Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines (MRP-227-A)," Appendix A). The staff further noted that some inspection routines (such as an ASME Section XI visual VT-3 inspection on a 10-year frequency) of the clevis insert assemblies may not be adequate to ensure the integrity of clevis insert assemblies during design basis events if multiple bolt failures occur prior to detection, and then the design basis event were to occur.

Request:

The staff requests that STPNOC provide the following:

1. MRP-227-A Applicant/Licensee Action Items (A/LAIs): Provide either an LRA amendment or update that includes an updated AMP, updated AMR items, and any applicable inspection plan(s) for the PWR RVI components at STP, Units 1 and 2, that are based on the guidance in LR-ISG-201 1-04, including responses to applicable A/LAIs identified in the staff's SE for MRP-227, dated December 16, 2011.

STPNOC Response:

STP has updated the PWR Reactor Internals Program to address MRP-227-A Applicant/Licensee Action Items (A/LAIs), update AMR tables, and Aging Management Program, and provided a program inspection plan for NRC review and approval, based on the guidance in LR-ISG-201 1-04 and the NRC staff's Safety Evaluation for MRP-227.

LRA Appendix B2.1 .35 and LR Basis Document AMP PWRI, PWR Reactor Internals program are revised to align the Scope of Program Element 1 component names with MRP-227-A and LR-ISG-2011-04 Table 3.1-1.

LIRA Table of Contents is revised to include Appendix C, Response to Applicant Action Items for Inspection and Evaluation Guidelines for PWR Internals.

LIRA Section 3.1.2.1.1 is revised to add Neutron Flux to the list of environments.

Enclosure 1 NOC-AE-1 5003270 Page 3 of 8 Further Evaluations LIRA Sections 3.1.2.2.6, 3.1.2.2.9, 3.1.2.2.12, and 3.1.2.2.15 are no longer applicable because STP PWR Reactor Internals program (B2.1 .35) relies on implementation of the inspection and evaluation guidelines in EPRI TR-1022863 (MRP-227-A) and EPRI TR-1 016609 (MRP-228) to manage the aging effects of the reactor vessel internal.

LIRA Table 3.1.1 is revised to include the verbiage from LR-ISG-201 1-04 for the Table 3.1.1 Item Numbers 3.1.1.22, 3.1.1.27, 3.1.1.30, 3.1.1.33, 3.1.1.37, 3.1.1.60, 3.1.1.63 and 3.1.1.80. The corresponding IDs from LR-ISG-2011-04 Table 3.1-1 are 59c, 59b, 53a, 59a, 53c, 54, 32, 53b.

The AMR lines from the original LIRA Table 3.1.2-1 pertaining to the reactor vessel internals are deleted and new lines are added using LR-ISG-201 1-04 Table IV.B2 Item numbers for the reactor vessel internals components applicable to STP.

Appendix C is added to provide the STP response to the Applicant Action Items for Inspection and Evaluation Guidelines for PWR Internals and STP's Reactor Vessel Internals Program Inspection Plan. provides the line-in/line-out revision to LRA Appendix B2.1 .35, LRA Table of Contents, LIRA Sections 3.1.2.1.1, 3.1.2.2.6, 3.1.2.2.9, 3.1.2.2.12, and 3.1.2.2.15 and Tables 3.1.1 and 3.1.2-1. provides a new LIRA section Appendix C, Response to Applicant Action Items for Inspection and Evaluation Guidelines for PWR Internals and STP's Reactor Vessel Internals Program Inspection Plan.

Request:

2. Reactor vessel clevis assemblies: Please address the clevis insert bolt operating experience issue cited above in reference to STP:

a) Describe the configuration of clevis insert assemblies at STP, including number of bolts in the assemblies. Specify the materials of fabrication, including any applicable heat treatments that were used for the design of the clevis insert bolts at STP.

b) Discuss and justify whether the operating experience associated with cracking of the clevis insert bolts is applicable to the clevis insert assembly designs at STP.

c) Describe the inspections that have been performed of the clevis insert bolts, including the type of inspection (e.g., VT-3). Discuss the visual inspection coverage that was achieved during these inspections. Clarify the ASME examination category that applies to inspections of the clevis insert bolts (and identify the applicable inspection method and frequency) and whether any past examinations have resulted in the detection of any indications of cracking or failures of the clevis insert bolts that are included in the clevis insert assembly designs. If so, provide the details of the inspection results and clarify the corrective actions that were taken at the facility to justify the structural integrity of the clevis insert assemblies and the intended safety function of the plant's core support structure and its components during plant operations.

d) Based on your responses to Parts (a) through (c) of this request, clarify whether the 10-year ISI basis (or the current approach used for STP) for the clevis insert bolts is sufficient to manage cracking and wear of the bolts during the period of extended operation. Justify your response to this request.

Enclosure 1 NOC-AE-1 5003270 Page 4 of 8 STPNOC Response:

(a) The clevis insert assemblies at South Texas Units 1 and 2 are comprised of eight clevis insert cap screws fabricated from Inconel X-750, SA-637, Grade 688, Type 2, two dowel pins fabricated from Alloy 600 and an Alloy 600 hard-faced U-shaped clevis insert that rests onto the clevis locations in the reactor vessel. In total there are six clevis locations (0, 60, 120, 180, 240 and 300 degrees). The clevis insert cap screws include a lock bar that is welded to the clevis insert face.

The clevis insert cap screws are attached to the interior face of the clevis insert assembly. See Figure 2-1 for a comparison of the clevis insert assemblies for STP and the reference plant that had clevis insert cap screw cracking.

The South Texas material used for the clevis insert cap screws is similar to that used by the operating experience referenced plant where cracking was observed.

The reference plant cap screw material heat treatment consisted of equalization heat treatment followed by a one-hour solution heat treatment. The precipitation hardening temperatures are the same, but the cooling time between temperatures for the reference plant is longer. For either type of heat treatment, susceptibility to primary water stress corrosion cracking is known to exist. STP specific heat treatment values for the above methods are Westinghouse proprietary information but can be made available upon request. The cap screws are of the same design, except that the South Texas cap screw shank length is slightly longer. The cap screws were installed with the same torque as that used for the referenced plant.

(b) The main function of the Lower Radial Support System (LRSS) is to prevent tangential or rotational motion of the lower internals assembly while permitting axial displacement and differential radial expansion. These supports are designed to prevent excessive lateral and rotational displacement of the lower internals during seismic and loss-of coolant accident (LOCA) conditions. The supports also limit displacements and misalignments in order to avoid overstressing the core barrel and to ensure that the control rods can be freely inserted. Therefore, assuming the clevis inserts remain in place as limited by the adjoining radial keys and support lugs, the design function of the LRSS will be maintained during seismic and LOCA conditions.

South Texas Units 1 and 2 have six radial supports spaced at 60-degree intervals around the circumference of the vessel. Although labeled as radial supports, the supports actually support the core barrel only in the tangential direction because the tangential clearances between the core barrel keys and the vessel clevis inserts are much smaller than the radial clearances. This basic arrangement is the same for South Texas units and the reference plant where clevis insert cap screw cracking was observed; however, the clevis insert designs are different. See Figure 2-1 for this comparison. The same number of eight cap screws is arranged in the same two vertical columns of four cap screws each. Two interference-fit dowel pins of the same size are located in-line with the cap screws in the same manner as the reference plant. The main design difference is that the South Texas reactor clevis insert is U-shaped, with the cap screws located inboard of the "U";

whereas the reference plant insert, while also being U-shaped, has flanges on either side where the cap screws are located. The tangential interference fit of the insert against the support lug is at the ends of these flanges for the reference plant design and on the sides of the "U" for the South Texas reactor design. Therefore, the tangential interference-fit compression stiffness of the two inserts is different.

Enclosure 1 NOC-AE-1 5003270 Page 5 of 8 The clevis insert cap screws for the South Texas units are of a similar design, of the same material with very similar heat treatment, torqued to the same degree, and operated at a slightly hotter TcoId inlet temperature as compared to the reference plant, it is possible that these cap screws can experience primary water stress corrosion cracking (PWSCC) similar to that of the reference plant.

Therefore, the operating experience relative to cracked clevis insert cap screws is applicable to South Texas units.

The structural evaluations performed by the reference plant to justify continued operation in the as-found condition demonstrated safe operation for an additional fuel cycle. The concern was possible long-term effects, such as the potential for vibratory loads to eventually cause loosening and wear of the insert and the subsequent increase in gaps between the insert, radial key and support lug.

A similar review of the structural adequacy of the South Texas clevis insert design was performed to determine if broken cap screws present a structural concern for safe operation. The structural aspects and loose parts assessment for the South Texas Units is discussed below:

Clevis Support Lug PrimaryStress The clevis insert, if completely loose to slide radially inward, is captured in a manner similar to the reference plant and is restrained by a similar radial gap before it contacts the radial key. With the clevis insert displaced fully inward, the primary stresses on the clevis support lugs remain acceptable relative to the reactor vessel original ASME 1971 Edition (through 1973 Summer Addenda) code of construction under plant-specific maximum upset and faulted condition loads due to seismic and LOCA conditions.

Clevis Insert PrimaryPlus Secondary Stress The increase in insert stress due to broken cap screws remains acceptable because the fatigue analysis maintains an acceptable usage factor less than 1.0.

Cap Screw Primary Plus Secondary Stress Where one column of cap screws is entirely broken, the resulting cap screw stress produced by this prying load on the insert is acceptable with four intact screws.

During heatup and steady-state, the clevis insert remains preloaded against the support lug, and this type of loading on the intact screws will not occur.

Clevis Insert Restraining Force (No Cap Screws)

If all of the cap screws are broken, and no restraint by the dowel pins is assumed, the clevis insert will maintain preload during steady-state operation and will maintain its ability to perform its intended function. Long-term wear between the insert and support lug is not expected or would be insignificant. The clevis insert is restrained tangentially by the support lugs and restrained radially by the limited clearance with the radial key. In addition, the insert has a thick upper flange that prevents it from falling downward, and the downward force from the downcomer flow will prevent it from working upward. The design installed at South Texas Units 1 and 2, which maintains a greater amount of preload between the insert and support lug as compared to the referenced plant, longer operation can be maintained before discernable degradation occurs.

Enclosure 1 NOC-AE-1 5003270 Page 6 of 8 During core barrel removal at cold conditions, the interference fit of the insert provides greater frictional force than the applied frictional force produced by the key sliding upward against the insert. The two dowel pins will also provide additional vertical constraint of the insert. Therefore, the clevis insert design prevents separation of the insert during core barrel removal operations if the cap screws (and dowel pins) are nonfunctional.

Loose PartsAssessment The insert cap screws have the same head design and locking device design as the reference plant. A lock bar is installed in a groove in the cap screw head, and the bar is welded to the insert counterbore where the cap screw is inserted. If a cap screw head should separate, the lock bar can wear and separate over time, causing the cap screw head to be loose in the counterbore recess. The maximum design gaps between the core barrel radial keys and the inserts are less than the height of the cap screw heads. Therefore, the cap screw heads remain captured, unless, over a long period of time, wear of the heads reduces the height of the heads by this amount. The cap screw head wear is expected to be small because the cap screw material is much harder than the clevis insert and radial key material.

The potential for loose parts due to wear-related degradation of the lock bars related to failed clevis insert cap screws at the degraded cap screw locations was evaluated. South Texas and the reference plant have different lower internal designs; however, the effects of where these loose parts would be captured or would impact against the lower internals are the same. Therefore, no significant degradation of mechanical components is expected as a result of the potential presence of loose parts from the lock bars in the primary system.

(c) STP performs remote visual examinations on 100% of accessible ASME Section Xl Category B-N-i, B-N-2 and B-N-3 examination areas in accordance with ASME Section Xl B-N-2 code categorization. As part of these examinations the clevis insert bolting, pins, and welded lock bars are verified intact. The examinations are performed with a submersible mini-submarine (mini-sub) with an attached color camera and associated support equipment. The most recent inspections were performed for Unit I in 2009 during 1RE015 and for Unit 2 in 2010 during 2RE014.

100% of the clevis insert bolt heads, pins, and lock bars were observed in conjunction with the B-N-2 inspection and no degradation or damage was identified. There are no STP inspection results that identified similar clevis insert bolt head detachment, as was identified at the referenced plant.

(d) Based on the structural evaluations above and operation with potential loose parts of the type and quantities that are no different than have already been evaluated, safe operation of the reactors and primary systems at South Texas Units 1 and 2 is assured. The ability of the Lower Radial Support System (LRSS) to perform its intended design function under seismic and LOCA condition loadings is unrelated to the integrity of the cap screws and dowel pins that are used to hold the clevis insert in place. Even if all of the cap screws and dowel pins separate, complete disengagement of one of the clevis inserts will not occur because of the small size of the gaps between the clevis inserts and radial keys. Wear or some degradation of a key might occur, but the key would still be expected to maintain functionality.

Taken as a whole, the core barrel and LRSS are expected to maintain their design function with degraded clevis insert bolts.

Enclosure 1 NOC-AE-1 5003270 Page 7 of 8 Augmentation of the reactor internals inspection program to include crack detection prior to cap screw failure is not required due to inherent design redundancy as discussed above. The effects of wear and/or looseness of the insert if the cap screws should become degraded, should not significantly affect the preload between the insert and support lug during steady-state operation. The Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A) categorization for wear-only is based on the primary concern for clevis insert looseness and wear of the clevis insert and radial key interfacing surfaces that could potentially lead to increased motion at the bottom end of the core barrel, rather than bolt material cracking. PWSCC was considered and screened in the Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191I). Actions to address PWSCC are included in MRP-227-A. Manifestation of cap screw cracking is identified as a result of the observation of wear (see note 2 of Table 4-9 in MRP-227-A).

Existing inspections are already in place to account for this concern. Qualified personnel at South Texas performing video camera inspections at 10-year intervals, as specified in ASME Code Section XI and MRP-227-A, are capable of identifying wear or dislodged components of the clevis insert cap screws or dowel pins at any location.

Visual inspection at 10-year intervals can also detect wear and displacement of the clevis insert. Inspection of the insert and key contact surfaces can detect wear in adjacent non-contact surfaces. If cap screw heads are observed to be loose, any displacement of the insert relative to the vessel support lug can be easily observed. During the last in-service inspections at the South Texas units in 2010, no indications of loosening or adverse wear were observed. Based on the above considerations and observations, it is concluded that the current in-service inspections to examine the clevis insert (and LRSS) with VT-3 is adequate without augmentation.

Enclosure 1 NOC-AE-1 5003270 Page 8 of 8 C ap Screw Centerline Radial Key Clevis i-Insert Referenced Plant South Texas Units 1 and 2 Figure 2-1: Lower Radial Support Comparison

Enclosure 2 NOC-AE-1 5003270 Enclosure 2 STPNOC LRA Changes with Line-in/Line-out Annotations

Enclosure 2 NOC-AE-1 5003270 Page 1 of 42 B2.1.35 PWR Reactor Internals Program Description The PWR Reactor Internals program manages cracking, loss of material, loss of fracture toughness, dimensional changes, and loss of preload for reactor vessel components that provide a core structural support intended function. The program implements the guidance of EPRI 1022863, PWR Internals Inspection and Evaluation Guideline (MRP-227-A, Rev .0) and EPRI 1016609, Inspection Standardfor PWR Internals (MRP-228, Rev. 0). The program manages aging consistent with the inspection guidance for Westinghouse designated primary components in Table 4-3 of MRP-227-A and Westinghouse designated expansion components in Table 4-6 of MRP-227-A, and the Westinghouse designated existing components in Table 4-9 of MRP-227-A. Primary components are expected to show the leading indications of the degradation effects. The expansion components are specified to expand the primary component sample should the indications of the sample be more severe than anticipated. The aging effects of a third set of MRP-227-A internals locations are deemed to be adequately managed by existing program components whose aging is managed consistent with ASME Section XI Table IWB-2500-1, Examination Category B-N-3.

Program examination methods include visual examination (VT-3), enhanced visual examination (EVT-1), volumetric examination, and physical measurements. Bolting ultrasonic examination technical justifications in MRP-228 have demonstrated the indication detection capability to detect loss of integrity of PWR internals bolts, pins, and fasteners, such as baffle-former bolting. For some components, the MRP-227-A methodology specifies a focused visual (VT-3) examination, similar to the current ASME Code,Section XI, Examination Category B-N-3 examinations, in order to determine the general mechanical and structural condition of the internals by (a) verifying parameters, such as clearances, settings, and physical displacements; and (b) detecting discontinuities and imperfections, such as loss of integrity at bolted or welded connections, loose or missing parts, debris, corrosion, wear, or erosion. In some cases, VT-3 visual methods are used for the detection of surface cracking when the component material has been shown to be tolerant of easily detected large flaws. In some cases, where even more stringent examinations are required, enhanced visual (EVT-1) examinations or ultrasonic methods of volumetric inspection, are specified for certain selected components and locations.

The program provides both examination acceptance criteria for conditions detected as a result of monitoring the primary components, as well as criteria for expanding examinations to the expansion components when warranted by the level of degradation detected in the primary components. Based on the identified aging effect, and supplemental examinations if required, the disposition process results in an evaluation and determination of whether to accept the condition until the next examination or implement corrective actions. Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection.

Enclosure 2 NOC-AE-1 5003270 Page 2 of 42 The PWR Vessel Internals program is a new program that has been implemented. The program will include future industry operating experience, as it is incorporated into the future revisions of MRP-227-A, to provide reasonable assurance for long-term integrity of the reactor internals. The reactor vessel internals included in the scope of the PWR Reactor Internals program are identified in Element 1. The scope of the program does not include welded attachments to the internal surface of the reactor vessel because these components are managed by the ASME Section Xl Inservice Inspection, Subsections IWB, IWO, and IWD program (B2.1.1) (exam category B-N-2) and/or the Nickel-Alloy Aging Management Program (B2.1.34). The scope of the program also does not include BMI flux thimble tubes which are managed by the Flux Thimble Tube Inspection program (B2.1.21).

Aging Management Program Elements The results of an evaluation of each element against the 10 elements described in Appendix A of NUREG-1 800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants are provided below.

Scope of Program- Element 1 The scope of the program applies the guidance in MRP-227-A which provides augmented inspection and flaw evaluation methodology for assuring the functional integrity of Westinghouse reactor vessel internals. The scope of the PWR Reactor Internals program includes components that provide a core structural support intended function and are managed by the Westinghouse designated primary components in Table 4-3 of MRP-227-A and Westinghouse designated expansion components in Table 4-6 of MRP-227-A and applicable MRP-227-A methodology license renewal applicant action items. MRP-227-A Table 4-9 also identifies existing program components whose aging is managed consistent with ASME Section Xl Table IWB-2500-1, Examination Category B-N-3.

Primary components are expected to show the leading indications of the degradation effects. The expansion components are specified to expand the primary component sample should the indications of the sample be more severe than anticipated. The aging effects of a third set of MRP-227-A internals locations are deemed to be adequately managed by existing program components whose aging is managed consistent with ASME Section Xl Table IWB-2500-1, Examination Category B-N-3.

The STP reactor vessel internals are divided into the following major component groups:

the lower core support assembly (including the entire core barrel assembly, baffle-former assembly, neutron shield panel, core support plate, and energy absorber assembly), the upper core support (UCS) assembly (including the upper support plate, support column, control rod guide tube assembly, upper core plate, and protective skirt), the incore instrumentation support structures (including the instrumentation columns (exit thermocouples), upper/lower tie plates, and instrumentation columns (BMI)), and miscellaneous alignment/interface components (including internals hold-down spring, upper core plate guide pins, and radial support keys including clevis inserts).

The following reactor vessel internals are included in the scope of the PWR Reactor Internals program:

1. Control rod guide tube assembly a,,-Bkh,*,

Enclosure 2 NOC-AE-1 5003270 Page 3 of 42

-Guide plate (cards) [Primary component]

-Lower flange welds and adjacent base metal^,,... (Adese ,n ^AMR by, Component" Typo.of "RV Control.,+.. Rod, G_,u;do Tube. AssembW*') [Primary component]

-Guide Tube Support Pins (Split Pins) (Addressed in ^AMR by, Component Type.of

,,RY! Cont,*+.,*rol,,R Guide__, Tube Bolting"),,,-,, [Existing programs component]

2. Core barrel assembly

-Upper core barrel flange weld and adjacent base metal (Addre"ssed in* AMR by, Component. Tye of.."RVI Core Barrel Assembl!*') [Primary component]

-Core barrel flange (Addressed in-^AMR by, Component. Tps o...f "R\/I Co*re' Barrel..

Asse*by~ [Expansion component and Existing programs component]

-Core barrel vertical axial welds and adjacent base metal [Expansion component]

-Core barrel circumferential girth welds and adjacent base metal [Primary component]

-Core barrel outlet nozzle welds and adjacent base metal [Expansion component]

-Lower core barrel flange weld and adjacent base metal Addressed in AMR by Component Types of "RVI Core Barrel Assembly") [Primary component]

3. Baffle-former assembly a*,-b,4 i,,i*,

-Baffle-former bolting [Primary component]

-Barrel to Former Boltincq [Expansion component]

-Baffle-former assembly baffle and former plates [Primary component]

4. Alignment and interfacing components

- Internals hold-down spring [Primary component]

-, Radial,* support. key _Clevis insert bolts [Existing programs component]

- Upper core plate g*i~ alicqnment pins [Existing programs component]

5. Bottom Mounted Instrumentation (BMI) Column assembly ..upport otr....tu....

- BMI Instrumentation columns bodies -BM [Expansion component]

6. Upper e~e-4s~ippe Internals assembly

Enclosure 2 NOC-AE-1 5003270 Page 4 of 42

- Upper core support pieteetPve-skirt [Existing programs component]

-Upper Core Plate [Expansion component]

7. Lower internal assembly Corc Support Structure

.. Lloe..r S e4*at -"*-*[Expansioncomponent]

The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are managed by the ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD program (B2.1.1) (exam category B-N-2) and/or the Nickel-Alloy Aging Management Program (B2.1.34). The scope of the program also does not include BMI flux thimble tubes which are managed by the Flux Thimble Tube Inspection program (B2.1.21).

The STP reactor vessel internals configuration does not include the lower internals assembly (lower support column bodies and lower core plate) noted in MRP-227-A.

The PWR Reactor Internals program is consistent with the following MRP-227-A assumptions (determination of applicability) which are based on PWR representative internals configurations and operational histories.

(1) STP has operated for less than 30 years of operation with high leakage core loading patterns. Operation with high leakage core loading was followed by implementation of a low leakage fuel management pattern for the remaining operating life.

(2) STP operates at fixed power levels and does not usually vary power based on calendar or load demand schedule.

(3) STP has not implemented any design changes beyond those identified in industry guidance or recommended by Westinghouse.

Preventive Actions - Element 2 The PWR Reactor internals program does not prevent degradation due to aging effects, but provides measures for monitoring to detect the degradation prior to loss of intended function. Preventive measures to mitigate aging effects such as loss of material and cracking include monitoring and maintaining reactor coolant water chemistry consistent with the guidelines of EPRI TR 1014986, PWR Primary Water Chemistry Guidelines, Volume 1. The primary water chemistry program is described separately in the Water Chemistry program (B2.1.2).

ParametersMonitored or Inspected - Element 3 The PWR Reactor Internals program monitors the following aging effects by inspection in accordance with the guidance of MRP-227-A or ASME Section XI Category B-N-3:

(1). Cracking Cracking is due to stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), irradiation assisted stress corrosion cracking (IASCC), or fatigue

Enclosure 2 NOC-AE-1 5003270 Page 5 of 42

/cyclical loading. Cracking is monitored with a visual inspection for evidence of surface breaking linear discontinuities or a volumetric examination. Surface examinations may also be used to supplement visual examinations for detection and sizing of surface-breaking discontinuities.

(2). Loss of Material Loss of Material is due to wear. Loss of material is monitored with a visual inspection for gross or abnormal surface conditions.

(3). Loss of Fracture Toughness Loss of Fracture Toughness is due to thermal aging or neutron irradiation embrittlement.

The impact of loss of fracture toughness is indirectly monitored by using visual or volumetric examination techniques to monitor for cracking and by applying applicable reduced fracture toughness properties in the flaw evaluations if cracking is detected and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation.

(4). Dimensional Changes Dimensional Changes are due to void swelling and irradiation growth, distortion or deflection. The program supplements visual inspection with physical measurements to monitor for any dimensional changes due to void swelling, irradiation growth, distortion, or deflection.

(5). Loss of Preload Loss of Preload is caused by thermal and irradiation-enhanced stress relaxation or creep.

Loss of preload is monitored with a visual inspection for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections.

The PWR Reactor Internals program manages the aging effects noted above consistent with the inspection guidance for Westinghouse designated primary components in Table 4-3 of MRP-227-A and Westinghouse designated expansion components in Table 4-6 of MRP-227-A. MRP-227-A also identifies Existing Program components whose aging is managed consistent with ASME Section Xl Table IWB-2500-1, Examination Category B-N-3. See the component list in element 1 to identify Primary, Expansion, and Existing components.

Detection of Aging Effects - Element 4 The PWR Reactor Internals program detects aging effects through the implementation of the parameters monitored or inspected criteria and bases for Westinghouse designated Primary Components in Table 4-3 of MRP-227-A and for Westinghouse designated Expansion Components in Table 4-6 of MRP-227-A. The aging effects of a third set of MRP-227-A internals locations identified in Table 4-9 of MRP-227-A are deemed to be adequately managed by existing program components whose aging is managed consistent with ASME Section XI Table IWB-2500-1, Examination Category B-N-3.

One hundred percent of the accessible volume/area of each component will be examined for the Primary and Expansion components inspection category components. The minimum examination coverage for primary and expansion inspection categories is 75 percent of the component's total (accessible plus inaccessible) inspection area/volume be examined. When addressing a set of like components (e.g. bolting), the minimum

Enclosure 2 NOC-AE-1 5003270 Page 6 of 42 examination coverage for primary and expansion inspection categories is 75 percent of the component's total population of like components (accessible plus inaccessible).

If defects are discovered during the examination, STP enters the information into the STP corrective action program and evaluates whether the results of the examination ensure that the component (or set of components) will continue to meet its intended function under all licensing basis conditions of operation until the next scheduled examination.

Engineering evaluations that demonstrate the acceptability of a detected condition will be performed consistent with WCAP-1 7096-NP.

Monitoring and Trending - Element 5 The program provides both examination acceptance criteria (See Element 6) for conditions detected as a result of monitoring the primary components as described in Element 4, as well as criteria for expanding examinations to the expansion components when warranted by the level of degradation detected in the primary components. Based on the identified aging effect, and supplemental examinations if required, the disposition process results in an evaluation and determination of whether to accept the condition until the next examination or implement corrective actions. Any detected conditions that do not satisfy the examination acceptance criteria (See Element 6) are required to be dispositioned through the corrective action program (See Element 7), which may require repair, replacement, or analytical evaluation for continued service until the next inspection.

Acceptance Criteria- Element 6 Examination acceptance for the primary and expansion component examinations are consistent with Section 5 of MRP-227-A. ASME Section Xl section IWB-3500 acceptance criteria apply to Existing Programs components. The following examination acceptance criteria apply to the STP reactor vessel internals:

Visual examination (VT-3) and enhanced visual examination (EVT-1)

For existing program components, the ASME Code Section Xl, Examination Category B-N-3 provides the following general relevant conditions for the visual (VT-3) examination of removable core support structures.

(1) Structural distortion or displacement of parts to the extent that component function may be impaired, (2) Loose, missing, cracked, or fractured parts, bolting, or fasteners, (3) Corrosion or erosion that reduces the nominal section thickness by more than 5 percent, (4) Wear of mating surfaces that may lead to loss of function; and (5) Structural degradation of interior attachments such that the original cross-sectional area is reduced more than 5 percent.

In addition, for the visual examinations (VT-3) of Primary and Expansion components, the PWR Reactor Internals program is consistent with the more specific descriptions of relevant conditions provided in Table 5-3 of MRP-227-A. EVT-1 examinations are used

Enclosure 2 NOC-AE-1 5003270 Page 7 of 42 for detecting small surface breaking cracks and surface crack length sizing when used in conjunction with sizing aids. EVT- 1 examination has been selected to be the appropriate NDE method for detection of cracking in plates or their welded joints. The relevant condition applied for EVT-1 examination is the same as found for cracking in ASME Section XI section 3500 which is crack-like surface breaking indications.

Volumetric examination Individual bolts are accepted (pass/fail acceptance) based on the detection of relevant indications established as part of the examination technical justification. When a relevant indication is detected in the cross-sectional area of the bolt, it is assumed to be non-functional and the indication is recorded. Bolted assemblies are evaluated for acceptance based on meeting a specified number and distribution of functional bolts. Acceptance criteria for volumetric examination of STP reactor internals bolting are consistent with Table 5-3 of MRP-227-A.

Physical Measurements measurement. The exami,,nation .'racceptance critorion for this............ is con...i;tent with Tab.le 5 3 of iMRPD22"7 A and,, requi,,res that* the rem,-a-ining, compr.ressible heig'-ht o,,f the spring,, shall p..o.vide hold down.. for..es within the,,*o*

plant peific design tolerance. Physical measurement of the internals hold down spring is not required because STP internals hold down spring is fabricated from 403 stainless steel.

Corrective Actions - Element 7 The following corrective actions are available for the disposition of detected conditions that exceed the examination acceptance criteria:

(1) Supplemental examinations to further characterize and potentially dispose of a detected condition consistent with Section 5.0 of MRP-227-A; (2) Engineering evaluation that demonstrates the acceptability of a detected condition consistent with WCAP-1 7096-NP; (3) Repair, in order to restore a component with a detected condition to acceptable status (ASME Section XI); or (4) Replacement of a component with an unacceptable detected condition (ASME Section xI)

(5) Other alternative corrective action bases if previously approved or endorsed by the NRC.

Relevant indications failing to meet applicable acceptance criteria are repaired or replaced in accordance with plant procedures. Appropriate codes and standards are specified in both the "ASME Section Xl Repair, Replacement, and Post-Maintenance Pressure Testing" procedure and in design drawings. Quality assurance requirements for repair and replacement activities are also included in the STP Operations Quality Assurance Plan.

Enclosure 2 NOC-AE-1 5003270 Page 8 of 42 STP site QA procedures, review and approval process, and administrative controls are implemented in accordance with the requirements of 10 CFR 50 Appendix B and are acceptable in addressing corrective actions. The QA program includes elements of corrective action, confirmation process and administrative controls, and is applicable to the safety-related and non-safety related systems, structures, and components that are subject to aging management review.

Confirmation Process- Element 8 STP site QA procedures, review and approval process, and administrative controls are implemented in accordance with the requirements of 10 CFR 50 Appendix B and are acceptable in addressing the confirmation process. The QA program includes elements of corrective action, confirmation process and administrative controls and is applicable to the safety-related and non-safety related systems, structures and components that are subject to aging management review.

Administrative Controls - Element 9 STP site QA procedures, review and approval process, and administrative controls are implemented in accordance with the requirements of 10 CFR 50. Appendix B and are acceptable in addressing administrative controls. The QA program includes elements of corrective action, confirmation process and administrative controls and is applicable to the safety-related and non-safety related systems, structures and components that are subject to aging management review.

Operating Experience - Element 10 Relatively few incidents of PWR internals aging degradation have been reported in operating U.S. commercial PWR plants. However, a considerable amount of PWR internals aging degradation has been observed in European PWRs, with emphasis on cracking of baffle-former bolting. The experience reviewed includes NRC Information Notice 84-18, Stress Corrosion Cracking (SCC) in PWR Systems and NRC Information Notice 98-11, Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants.

Most of the industry operating experience reviewed has involved cracking of austenitic stainless steel baffle-former bolts, or SCC of high-strength internals bolting. SCC of control rod guide tube split pins has also been reported.

Several other items with existing or suspected material degradation concerns that have been identified for PWR components are wear in thimble tubes and potentially in control guide cards and observed cracking in some high-strength bolting and in control rod guide tube alignment (split) pins. The latter are conditions that have been corrected primarily through bolt replacement with less susceptible material and improved control of pre-load.

Based on industry operating experience, STP replaced the Alloy-750 guide tube support pins (split pins) with strained hardened (cold worked) 316 stainless steel pins during Refueling Outage 1 RE1 2 (Spring 2005) for Unit 1 and Refueling Outage 2RE1 1 (Fall 2005) for Unit 2. The replacement was conducted to reduce the susceptibility for stress corrosion cracking in the split pins. There were no cracked Alloy X-750 pins discovered during the replacement process.

Enclosure 2 NOC-AE-1 5003270 Page 9 of 42 The ASME Code, Section Xl, Examination Category B-N-3 examinations of core support structures conducted during Refueling Outage 1RE15 (Fall 2009) for Unit 1, and Refueling Outage 2RE14 (Spring 2010) for Unit 2, did not identify any conditions that required repair, replacement or evaluation.

The ISI Program portion of the PWR Reactor Internals program at STP is updated to account for industry operating experience. ASME Section Xl is also revised every three years and addenda issued in the interim, which allows the code to be updated to reflect operating experience. The requirement to update the ISI Program to reference more recent editions of ASME Section XI at the end of each inspection interval ensures the ISI Program reflects enhancements due to operating experience that have been incorporated into ASME Section XI.

With exception of the ASME Section ISI portions, the PWR Reactor Internals program will be a new program and has no direct programmatic history. A key element of the MRP-227-A program is the reporting of aging of reactor vessel components. STP, through its participation in PWR Owners Group and EPRI-MRP activities, will continue to benefit from the reporting of inspection information and will share its own operating experience with the industry through those groups or INPO, as appropriate.

As additional Industry and applicable plant-specific operating experience become available, the OE will be evaluated and appropriately incorporated into the program through the STP Corrective Action and Operating Experience Programs. This ongoing review of OE will continue throughout the period of extended operation, and the results will be maintained on site. This process will confirm the effectiveness of this new license renewal aging management program by incorporating applicable OE and performing self assessments of the program.

Conclusion The implementation of the PWR Reactor Internals program provides reasonable assurance that aging effects will be adequately managed such that the systems and components within the scope of this program will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

Enclosure 2 NOC-AE-1 5003270 Page 10 of 42 LRA Table of Contents Appendix C No-Ue Response to Applicant Action Items for Inspection and Evaluation Guidelines for PWR Internals

Enclosure 2 NOC-AE-1 5003270 Page 11 of 42 3.1.2.1.1 Reactor Vessel and Internals Materials The materials of construction for the reactor vessel and internals component types are:

• Carbon Steel

• Carbon Steel with Stainless Steel Cladding

• High Strength Low Alloy Steel (Bolting)

• Nickel-Alloys

• Stainless Steel

• Stainless Steel Cast Austenitic (CASS)

Environment The reactor vessel and internals components are exposed to the following environments:

• Borated Water Leakage

• Reactor Coolant

• Neutron Flux Aging Effects Requiring Management The following reactor vessel and internals aging effects require management:

• Changes in dimensions

• Cracking

• Loss of fracture toughness

• Loss of material

• Loss of preload Aging Management Programs The following aging management programs manage the aging effects for the reactor vessel and internals component types:

• ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD (B2.1.1)

• Boric Acid Corrosion (B2.1.4)

• Flux Thimble Tube Inspection (B2.1.21)

• Nickel-Alloy Aging Management (B2.1 .34)

• Nickel-Alloy Penetration Nozzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors (B2.1.5)

Enclosure 2 NOC-AE-1 5003270 Page 12 of 42

• PWR Reactor Internals (B2.1 .35)

• Reactor Head Closure Studs (B2.1 .3)

• Reactor Vessel Surveillance (B2.1 .15)

• Water Chemistry (B2.1.2)

For Reactor Coolant System Nickel-Alloy Pressure Boundary Components, STP will:

(1) Implement applicable NRC Orders, Bulletins and Generic Letters associated with nickel-alloys; (2) implement staff-accepted industry guidelines, (3) participate in the industry initiatives, such as owners group programs and the EPRI Materials Reliability Program, for managing aging effects associated with nickel-alloys, and (4) upon completion of these programs, but not less than 24 months before entering the period of extended operation, STP will submit an inspection plan for reactor coolant system nickel-alloy pressure boundary components to the NRC for review and approval.

For Reactor Vessel Internals, STP will:

(1) Participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals. ; and (3) u.pon, completion of thcso, programs, but not""

inspection plan"for reactor internals to the NRC"fo re.ie. and... appro.a....

Enclosure 2 NOC-AE-1 5003270 Page 13 of 42 3.1.2.2.6 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement and Void Swelling Not applicable, the STP PWR Reactor Internals proglram (B2.1.35) relies on implementation of the inspection and evaluation .quidelines in EPRI TR-1 022863 (MRP-227-A) and EPRI TR-1 016609 to managqe the agqingq effects of the reactor vessel internal components.

Loss., of fracturev to

÷ughneSs

• duo to. neutron irradiation4..* .. mbrittlomen.. an void4

,,, well...ng, fo.

the plant speci;fic, PWR Reactor Internals program (B21o 5 based*

,,on*, the, guid..elines,*

prot',dcd in Er-I~ 1022863' (MRP 227 OA). Consitent *.A,;fh ER-D:I 1'022863 (IMRPD227 ,A\

RVI neutron shi;ld p.nel 3.1.2.2.9 Loss of Preload due to Stress Relaxation Not applicable, the STP PWR Reactor internals proglram (B2.1 .35) relies on implementation of the inspection and evaluation gquidelines in EPRI TR-1 022863 (MRP-227-A) and EPRI TR-1 016609 to managqe the aglingl effects of the reactor vessel internal components.

Less.*of,preload4 du' to stress relaxation. for. nickeUl*allo and . stainles steel reactor..,

interna*ls- components exposed to reactor coolant is- managed by* the p~lant specific-PWR Reactrf Inte;r nals p:\Irog *ram r(B2.. h)nasdn thl uieinlprvdd;nEPI1026 3.1.2.2.12 Cracking due to Stress Corrosion Cracking and Irradiation-Assisted Stress Corrosion Cracking (IASCC)

Not applicable, the STP PWR Reactor Internals progqram (B2.1.35) relies on implementation of the inspection and evaluation gquidelines in EPRI TR-1 022863 (MRP-227-A) and EPRI TR-1 016609 to manage the aglingl effects of the reactor vessel internal components.

Fo managing theIDO- agin effectf c.;hracing):*

due0*' tolstress coroio: cackn Dand: Irradatio assisted stescorso

• "* cracking of,,,stainless

, steel,,

reato internals9 components*'*'

exposed to~r rrrreatr coolnt* *÷,4 atr CIs-;n ' (;2. 1.2),'*,is augmented by;,*the plntspecific PnWR Reactor'/0 Ineal prga B..5 ae nte-udlnspoie nER 102863(MR 2271,A)., Consitn ih P;1283 MP27A, W ee

Enclosure 2 NOC-AE-1 5003270 Page 14 of 42 RV*\II Uppr Core Sulppor IUpper Suppo-,r,,,, .t Column,,

RVI\1 Uppe,-.,r Cre..-r, Suppr,-.t

• Upper,,r Suppo*rtv< Pl1,ate RV*\I IC'I Support Structures Exit Thermocouples

_RV\I ICI Support'*+%1 Structures UppeII~r/lowe^r Tic Plates 3.1.2.2.15 Changes in dimensions due to Void Swelling Not applicable, the STP PWR Reactor Internals progqram (B2.1.35) relies on implementation of the inspection and evaluation ,quidelines in EPRI TR-1022863 (MRP-227-A) and EPRI TR-1 016609 to manaqe the aglingl effects of the reactor vessel internal components.

..

Reactor Internals... prora ..... 1.5 baedo th guidelines _ro'-ded in EPI-- 1026

-RVI Control Rod Guide Tube ,Assembly

-RVI Control Rod Guide Tube Bolting

-RP5I Control Rod Guide Tube Gu-ide Plates RViI i Hol Doigwnl SpingII RV*\I Lower Cre,.+ Support Bolt-s o+ ,,+-,.

RVI Lower Cr,,,,e Suppor Clevi.s*.Iner Bo. I)**- ltng *"

• -RVI Radial Suppor't Keys and Cle'-is Inserts RI Upper* Cor Plat Gu Ide Pins R'iUpeCoeSpotrtet'eSkt RVI:\1 Uppe,,*r Co-re SupportH Uppe,,,r Co*re Plate RV*)/II Uppe~r Co*re Supportr' Upper Suppo~nrt Column~r

,Rv. Up~per Core Support. Up~per Support ColumnBs RV:)\II Uppe~r Core Supportr Uppe~r Supprtr P*lator

_RY:\/I Uppe~r Suppor Column~r BoIr ng,,rr 1:lf

Enclosure 2 NOC-AE-1 5003270 Page 15 of 42 Table 3.1.1 Summary of Aging Management Evaluations in Chapter IV of NUREG-1801 for Reactor Vessel, Internals, and Reactor Coolant System

,Item Component Type:* Aging Effect:/Mechanism  : .. Aging:Management:: FurherDicu..o Num ber:  ;. .. . ..  :::: *::. ..  : :*, i;: Program i;ii:::  :::*; valuation :*; .  :

.-- 22Stainlcss steci and Loss

_ .. toughness duc' of fracture*,* ,,TSAR., supplement*

.. No l Consistent-;with NUREG*

niocel alo",, reactor to...

neutro irradiatio...... cn to"""(1)*patiiptc181 ormacr 3..122SailesstelS, os f ratretognesu NWRC approel 21tmonths ReaCtorsi nternat with NU 35EG i includingq CASS, PH to thermal agqingq and neutron (B2.1 .35) 1801. Revision 2 and LR-ISG SS or martensitic SS) irradiation embrittlement, and 2011-04 or nickel alloy for CASS, martensitic SS, and Westinglhouse reactor PH 5S due to thermal aglingl vessel internal embrittlement; or changes in "Existingl Programs" dimensions due to void swellingl components or distortion: or loss of preload due to thermal and irradiation enhanced stress relaxation or creep: or loss of material due to wearr____________

1-- Stainless steel and Loss of preoaed due to stress FSRsplmn No Consistent with screws, bolts, tie* rods, programs (2) implement *'ff-ec t but;a

, diffren aging, For-NR approval... 21 on,..th..

m Reactor.. Internals (B2.1 .35) is before the extended,,, period an ,.dte,*,;, with excetio of,,

,*

ndustr' recomendatio...e.i isepprtboltng n V

Enclosure 2 NOC-AE-1 5003270 Page 16 of 42

Item Component Type; Aging Effect IMechaniSm i

Aging Maniagement, Further Discussion Num ber :  :  : :iz i,*i:i:  ::**i Program =..i:!:i !: ,EvaiUation::i

=Section 3.1 .2.2.9.

3.1.1.27 Stainless steel (SS, ILoss of fracture toutihness due PWR Vessel Internals N._o 'Consistent with includingq CASS, PH to neutron irradiation (B..5 NUREG-1801 Revision 2 and SS or martensitic SS) embrittlement and for CASS, ILR-ISG-201 1-04 or and nickel alloy martensitic SS, and PH SS due Westingqhouse reactor to thermal agqing embrittlement:

vessel internal or chanties in dimensions due "Expansion" to void swellingq or distortion: or components loss of preload due to thermal and irradiation enhanced stress relaxation or creep: or loss of material due to wear 4-3OStainless stool reactor ,Crackring due to stress Water*, Che.mistry (B2.1 .2) and. INe -onsistent,^,ith NU ,REG

..omponents (e.g,* issisted stes corrosio  ;,n commitmcnt to/ (1 par*tic.ipa.te environment, and agingg I pe.... =,*... ias rek~ig in industry R\/I ag*ing zffc,+but diffe;rent..' Ah/D are.

.... embly, RCCA guide4 programs (2) imp,,lemn

.... {tcd:llrl Water{* C'hcmr*;{

shroud assemblie, Controecmmndtin.CoSiEStent ith X nEPRIe

,,,, guide tube*(,, CRGPT) .Inspection for selected

~See further ev-aluJation in assembly,, Cre,,barrel Section*;3.1.2.2.12.

distributor assembly,

Enclosure 2 NOC-AE-1 5003270 Page 17 of 42 Item Component Type : Aging Effect!IMechanism Agigangeen Futhr isusio sn,, strumetatisn 3.1.1.30 Stainless steel or Crackingq due to stress Water Chemistry (B2.1 .2) and No__ Consistent with NUREG nickel alloy 'corrosion crackingq, irradiation- PWR Vessel Internals 1801 Revision 2 and LR-ISG-Westingqhouse reactor assisted stress corrosion (B2.1 .35) 2011-04.

vessel internal crackingq, or fatigque "Primary~components ______________ _____________ ____________

veslitrasn industory, RVIl agingt for this materia'l andw~

RVI inspection plan based on Section 3.1 .2.2.15.

industni recomendtion ______________

3.1.1.33 Stainless steel (SS. Loss of fracture toughness due PWR Vessel Internals No Consistent with NUREG including CASS, PH to neutron irradiation (B2.1 .35) 1801 Revision 2 and LR-ISG-SS or martensitic SS) embrittlement and for CASS. 2011-04 or nickel alloy martensitic SS, and PH SS due Westinghouse reactor to thermal agqingq embrittlement; vessel internal or changqes in dimensions due "Primary" components to void swellingq or distortion; or loss of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear vesselinternals .ater stres corrosion.. commitmen to*(1 pa* c;pate..... ron..... and aging copnet.(, cacig irradiati.*on assiste i*.nindustry RVagngo.ffec.t.*, but di.4ffrn

.... s are...

tube ssemblie, For NRC approval. 21* . months" by, the plant specific agin

Enclosure 2 NOC-AE-1 5003270 Page 18 of 42 Item= ": Component Type :': =Aging Effect IMechanism iNumber  ::Aging Management:::

Program  ; :i** EvalUation:::*

Further ::**1:*= :Discussion * :i: :

s-sembly,*h, "CEA shroud RVD\I inseton-,v÷,, p*lan based on Rea-cItor-f,, Inte*rn-als /*1B. .35) asoble, Corci,,ost- .... o....nda.tion C..onos-.stentwih EPRI distributor assembly) See f,-rher evaluation in

,,,Sectio3.1.2.2.17.

3.1.1.37 Stainless steel or Crackingq due to stress Water Chemistry (B2.1 .2) and N._o Consistent with nickel alloy ,corrosion crackingq, primary PWR Vessel Internals .NUREG-1801 Revision 2 and Westingqhouse reactor water stress corrosion (B2.1 .35) LR-ISG-201 1-04.

vessel internal :crackingl, irradiation-assisted "Existing Progqrams" stress corrosion cracking*, or components fat u 3.1.1 .60 Stainless steel flux Loss of material due to Wear Flux Thimble Tube Inspection No Consistent with thimble tubes (with or i(212)NUREG-1 801 NUREG-1 801 without chrome plating) Revision 2 and LR-ISG-2011-

"Existing Progqrams" 04__..

component_______

Steel*, reacto,-r vessel,

• A-1 Los

,of mate ria",-*;'l duh to We-alr Insr'v.c',e.,;, In*.,-spection (IIWA. Ne Cr-onosistent- w.,it-h NUREGD-t*

exposed to reactor guid tube boltingk,s- core.

.ssembly, core..supports for which the material,

• ' -*peamh -,,t,.,e,*i,4-,,*s environm

• FF,.,, nt agingh

,a .*nd

Enclosure 2 NOC-AE-1 5003270 Page 19 of 42

• Iltem i:; ComPonent Type Aging Effect IMechanism :i Aging Management  :* Furthe'r:* : ::::

i

:i;i*:**i=Evaluation=:,i *:Discuss~ion* i:*.i:

SNumber= ::Program

:: ........i: ::::i::::________,,_______________i ii  : Ri ecom m ended '*

3.1.1.63 Stainless steel, nickel Crackingq or Loss of material Inservice Inspection (IWB, N__o Consistent with NUREG-alloy or CASS reactor due to Wear IWC, and IWD) (B2.1 .1) or 1801 Revision 2 and LR-ISG-vessel internals, core PWR Vessel Internals 2011-04 support structure (B2.1 .35) components in MRP-227-A, exposed to reactor coolant and neutron flux 3*148 atasoii Lossof.f -rctuI, re toug...hnesdo* Thcrma*.'l Il, Agngn a"nd Neutrol*.--n Ne Except-,,*.Iionr .to, NUIE 1801.*_d*

o÷t'ainlc sool reacto*,,÷r ÷to, thc-rm'l agi*ng and* nuro,,,÷n llr-adiat;io; Ebitt';'lement* o-,f Aging ,eff..ct-÷ in* NURE 1801'_d vcssel internals/ (eg. rr,,dation ... brittlen

... ASS For, this material,, and, apfe,-i.Rtemel en'-ironment combination is internal assembly, CE, EP, 1016596... .

(R 27)..../

assembly, core support*

shilcd assembly, lower 3.1.1.80 Stainless steel Crackingq due to stress Water Chemistry (B2.1 .2) and N.oo Consistent with NUREG-Westingqhouse reactor corrosion crackingq, irradiation- PWR Vessel Internals 1801 Revision 2 and LR-ISG-vessel internal assisted stress corrosion (B2.1 .35) 2011-04 "Expansion" crackingq, or fatigque

_____corn ponents_________________________________________

Enclosure 2 NOC-AE-1 5003270 Page 20 of 42 Table 3.1.2-1 Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation - Reactor Vessel and Internals _______

Component Type I:ntendedl Material EnVironment: Aging Effect :Aging Management' NUREG,- :*,Table 1I, =Notes

___ __ __ __ i. __  : '* . ...  :: ... ::....:: M anagem ent 2...

Item....

. ...

RV Upper, PB Carbon Reactor Coolant Loss of material ASME Section Xl IV.A2-25 3.1.1.63 A Intermediate, Lower Steel with (Int) .Inservice Inspection, Shell and Welds Stainless Subsections IWB, IWC, Steel and IWD (B2.1.1)

Cladding_____________ ___ ___

RV!-Befffe-Edge 88 Stainless Reactor Cooelant Changes-in PWR Rcactor Internals l-\B2-4 &1A3 G-3 Bolting Steel (E- 4)

• ensieas B2o t -5)

D\V, Baffl,-,4d,. 88 Stainless. Reactor,. Coolant.,. "essef-prelead PWR Reactor Internals P42- &--4 -

R *lBfle~g 8=8 Stainless Reco oln Lo*ss of factu=re PWR Reactor Internals 4-VB2--6 &4-.A E--3 fRe-dg 88 Stainless Reactor-Coolant Cracking Wate *Ges~ t\.B2-10q &14a B\ . .....

RVl-BaffleFormer 88 SD Stainless Reactor Coolant *lt~ Time Liiedo Agngenl 4V-B2--4 1- -. A-

^B...ting S Steel (-Ext) fqmatiie-dag Anaysi. ealutedfo RV! £*f~e *ReaettheFormr , DFSLD, periodkinof extendedE,-,

RVI Baffle Former DF,-SLD, Stainless Reactor-Coolan Ghngs-inau IpwR Reactor Internals PAB2-1 -- 3 -

88..k,,S Steel (* teuisiens (B2.i* "5 RVI Baffle Former ,DF,-SLD, Stainless Reactor-Coolant Gia~4=akeing ater-GheImist l-V.B2 32 *-43 A-

^ .... il,, 88 Steel (-Ext) (212 n W

NOC-AE-1 5003270 Page 21 of 42

Enclosure 2 NOC-AE-1 5003270 Page 22 of 42 RVI Control Rod Reactor Coolant RVI Control Rod Reactor Coolant Guide,,4 Tu bc Gui,,do Ryl Control Rod Guide Tube Guide P4atee

Enclosure 2 NOC-AE-1 5003270 Page 23 of 42 Component Type Intended

,,Function M:l~ateriali

:: Environment;*<

, " Aginfg Effectl :=iiAging

,Requiring Management i:NUJREG- Table1  ::Notes

• Program ,:,i:1801 Vol. ,Item -

. . *:. i:==*

,> ...

; * :i: ... i=*,, i~ Man agem ent:** < 2 1tem

. ...

• '.

'= ..... " ... = :

RVI Core Barrel D.F, -SLD, S=tainless Rcco oln oso maera WeeG*~s~ l-V.B2-32 *---1~- A Asess8 Sel(22 RICrBrrl DF,-SLD, Stainless Reco oln ILos of material P-WR Rcactor Intcrnals 4/:B2-84 *146 Asemly5 Steel ,__._5_ ____B2._

RVI Core Barrel 88 Stainless Reco oln Ghanges-in PWR Reactor Internals lV*B- &1A* ..-

,sl-FSe Stieles { *PR eco !trt) C9-8 *E, RVI Corc Barrel 88 Stainless Raco Coln Less of-frcture PWR Reactor Internals 4V-B2-5 *142 semlFomrSteel C**(B.,.5 RVI Core Barrel 58 Stainless Reco oln Cracking Wae hmsr lV.B24!0 *-4 Assml Fre Steel (.E (B2.1.2)and PWR Belting Reactor Internals RVI Core Barrel 88 Stainless Reco Coln LGunmlaie lieLmted Aging tV.B2-~132 - A Assebl,-,FrDoer8 Steinless~ t "nalysi fReactor a-'"e eavaluat.ed f-,,or,, */o

• RVI Core Barre 8Z8 Stainless Reactor Coolant Los fmaeia pWR Rem-ctor~m tra~V.B2-2 1 33 A-*3 RVI Hold Down 88 Stils eco Coln Loss ofmteolnaerCemisry-V-B2-4-1 *44 RV,, Hold, Down 88 Stailes Recto Coln N.oneu... Non~e I* R\-4 34A- 3 I.T ISuabsections IWB, IWC,

NOC-AE-1 5003270 Page 24 of 42

Enclosure 2 NOC-AE-1 5003270 Page 25 of 42

.....

Component *Function Type intendled *'Matebrial;;.i EnVironment* :': Requiring

,.:Aging Effect: AgingProgrm.180 M~anagement,: NUREG Vo. Ie Tablem I Noe Str,-t'-es- Steel tE-x4) (2~

P4ates RVI ewr4radato 88 Stils Reactor Coolant Noekne ASM S.tinV.£kB-2 !5 4 O*-

Specimen Basket Steel InscIliIeIInspection, Subsections IWB, IWC, an pr~d fDextended RVI Loer Cor 88 Stiles ReactorCoolant None Nene 1qB2-45 &4-4*3 I*-4 RY oe oe 88 StinkiAless Reactor Coolant Naekn NSe SectonXI 4V.~B2- !5 -4*

RVI LowerCoe* 88 Stink IAless Raco Coln Cr~ack e Tm LmtdAing 1V.2 34-0 S~~e-Bels ftigu-damge nalyi e\aluate*0d for

Enclosure 2 NOC-AE-1 5003270 Page 26 of 42

.... Ppecte RVE Lower Core RVE Lower Core Absor-be-r Aoosembly

Enciosure 2 NOC-AE-1 5003270 Page 27 of 42 Component:

... Type: Intendedn..ti... Material: ..... ,:EnvirOnment: Aging Effect:

Requirin.... :AgingPrograme Management NUREG-1801 Vol. Table Ite I Notes::

':.. ::::::  : : ..  ::..  :: M anagem ent 2 Item _____:,___

RVI NeuJtron Shield SL-- St-aielese Reactor Coolant Neae Neae 4V* &1A-3 l-4 p-n*Steel _

RVI Ne'-tron Shield S4L--ifies RccorColn CrackingMEScto Xl l-*g- &---3 P~ane Steel {*4 Inser.... I""sp...tion RVI Ne'utron Shield SL-- Stahdles Reactor Cooante Neae NeVe .B2-9 &4-.-.--.-.--.-.-

RVI Neudtron Shield ,=L-D St*le Reactor Coolant Loss ofaterialWao Chmsr 4V.B2-32 *4--8 A RVI Radial Support -S8 Saniekeslos ReactorMColanttion NX! !V.~B2349 *--3 4G lse&Subsections IWB, IWC, RV ailSpot-8 Stainless Reactor Coolant Loss ofmaterial PWE Seaction Xnenls 1V.B2-34 *3---46a Keysate-Gie4Ae ,Steel {-E-x) Ineric Inpetin 1V4-*e45 Subsnlctions-NWBN WC, I\ q J34 RIRV dia Supprort 58 Ntiekl-Alye actorCoolant ¢'of,.;, maeilWte*Ge Loss ' B23 *A-8 A Plate Gueida Pis ____ rSeel (B2-).' 1-")* n W

~V-U~pe~Geie ta~iess 58 eactr Colan NO~ NReao PAntcma4,33Is Plate Gide

____.__ Stee Pin

Enclosure 2 NOC-AE-1 5003270 Page 28 of 42 Aging Effect*ii::

i,.:;Requiring Aging Management ii Program i Giat* Time Linitod Aging Cumulative fa'**;g*I ie- die'R~l Anaysi

,-v-a;luatcd fo÷rll

Enclosure 2 NOC-AE-1 5003270 Page 29 of 42 Aging =Effec~t

=:Requiring

!Aging:Program Management; Managements C,~-da'!*t,- ,A,p[!ys.s ewucted for Lhc period of extended GeIaerte W-Ghe~&t~y Neiie RVl-U~per-CeFe Sp~e~4ppe~

Suppe~t-Geh~R Beee

~VI4pe~-Ge~e

• uppeIt-LJVpei

Enclosure 2 NOC-AE-1 5003270 Page 30 of 42 Component Type I~ntended:l: Material :: :Environment i: ;,Aging Effect: Aging Manag~em~ent NUREG-: Table1I :iNotes:

Function: * = i =i iI= Requiring P..

.. rogram 1801ivVol.i! Item::: :,.........

..= :=**

. . . ...... I*: ... .. .. .. =:,i=  : M~anagem ent ......

• : 2..Item...._____......_

• \, ei... e,,, 58 S,.a,,de. Rcco oln Nene Nene RL-8B2-4-1~ 4.-

Cope'Jmjq e~rn Steel (*t Coumneft4nJ ,e8 Steel {Ext) ncicInpto, RVI, Uppcr Support SS StaPnless Reactor Coolant ofmtrAl

,-,Loss,- W te - ,,,e,-,t, 4-V7B2--40 *--8 RVI Uppfer Suporte S._S Stainless Reactor Coolant Nrckne NSME SetokX VB2.RP- 3.-1.1.63 A-Asemly Ln-ock Steel and euto) fu Inservice Inspection, 38 Device(Ext)Subsections IWB, IWC,

_____ _____ ____ ____ _____ ____ and IW D (B2.1.)_ _ _ _ _

RVl Baffle Former S._S Stainless Reactor Coolant CrLosofmtra ASME Section Xl IV.B2.RP- 3.1 .1.63 A Assembly Lock Steel and neutron flux *wa)Inservice Inspection. 382 Device (Ext) Subsections IWB, IWC, and IWO (B2.1.1)_____

RVI Baffle-Former IDSS D Stainless Reactor Coolant Loss of material WaSMer Shetiontry IV.B2-R32 13.1 .1.63 A Assembly (All S._S Steel (Ext) (B2.1 .2)

Components)

RVI Baffle-Former SS Stainless Reactor Coolant Chancges in PWR Reactor Internals IV.B2.RP- 13.1.1.33 A. 3 Assembly Baffle- Steel and neutron flux dimensions (B2.1 .35) 27_22 Former Boltinci ____ _ _ _ _ __ _ _ __ _ _ __ _ _ _ _ _ _ _ _

RVI Baffle-Former S._S Stainless Reactor Coolant Loss of fracture PWR Reactor Internals IV.B2.RP- 3.1 .1 .33 A, 3 Assembly Baffle- Steel and neutron flux tougihness (B2.1 .35)l 27__22 Former Boltingi _____E_____ ________ ________________ __________

Enclosure 2 NOC-AE-1 5003270 Page 31 of 42

• Component Type* IntendedFucto ,*Material ::::*Environment:: Aging Effect, 'i AgingProgram Requiring Management*,. NUREG-1t80Volvo. Table Item 1 I

• Notes RVI Baffle-Former SS Stainless Reactor Coolant Loss of preload PWR Reactor Internals IV.B2.RP- 3.1.1.33 A, 3 Assembly Baffle- Steel and neutron flux (B2.1 .35) 27:2 Former Boltingq (Ext___ ________

RVI Baffle-Former S._S Stainless Reactor Coolant C~rackingq Water Chemistry IV.B2.RP- 3.1.1.30 A, 3 A~ssembly Baffle- Steel and neutron flux (B2.1 .2) and PWR 271 Former Boltingq (Ext) Reactor Internals (B2.1 .35)

RVI Baffle-Former S._S Stainless Reactor Coolant Cumulative Time-Limited Agqingq IV.B2.RP- 3.1.1 .05 A Assembly Baffle- Steel and neutron flux fatigque damagqe Analysis evaluated for 30__33 Former Boltingq (Ext) the period of extended operation ______-_____

RVI Baffle-Former S._S Stainless Reactor Coolant Changqes in PWR Reactor Internals IV.B2.RP- 3.1.1.27 A,3 Assembly Barrel to Steel and neutron flux dimensions (B2.1 .35) 27_4 Former Boltingq(Et RVI Baffle-Former S._S Stainless Reactor Coolant Crackingq Water Chemistry IV.B2.RP- 3.1.1.80 A3 Assembly Barrel to Steel and neutron flux (B2.1 .2)and PWR 273 Former Boltingq (Ext) Reactor Internals (B2.1 .35)

RVI Baffle-Former S._S Stainless Reactor Coolant Cumulative Time-Limited Agqingq IV.B2.RP- 3.1 .1.05 A Assembly Barrel to Steel and neutron flux fatigue damagqe Analysis evaluated for 30_3 Former Boltingq (Ext) the period of extended

________ __________operation _____

R5VI Baffle-Former SS Stainless Reactor Coolant Loss of fracture PWR Reactor Internals IV.B2.RP- 3.1.1.27 A, 3

.A~ssembly Barrel to Steel and neutron flux tougqhness (B2.1 .35) 274 F~ormer Boltingq_____

RVI Baffle-Former - _SS Stainless Reactor Coolant Loss of preload PWR Reactor Internals IV.B2.RP- 3.1 .1.27 A, 3 Assembly Barrel to Steel and neutron flux (B2.1 .35) 274 Former Bolting *L RVI Baffle-Former OF, SLD, Stainless Reactor Coolant Changqes in PWR Reactor Internals IV.B2.RP- 3.1.1.33 A, 3 Assembly - baffle 5._S Steel and neutron flux dimensions (B2.1 .35) 270 and former Plates_____________________________________________

Enclosure 2 NOC-AE-1 5003270 Page 32 of 42

'. ...

Component .

Type= FUnctiOn' Intended

• Material  : ,;=  : i*Environment iEffect Requiring= :=Aging:

Aging  :: :P~rogram:::.: = I 1801 Man~agementi:

, NUREG-Vol.: ' : Table item*I::: Notes RVI Baffle-Former DF, ID' Stainless" Reactor Coolant Crackingq Water Chemistry IV.B2.RP- 3.1.1 .30 A, 3 Assembly-baffle S__S Steel and neutron flux (B2.1 .2) and PWR 270a and former plates .E(_Ex.) Reactor Internals

______________(B2.1 .35)

RVI Bottom SS Stainless Reactor Coolant Cumulative Time-Limited Agqingq IV.B2.RP- 3.1.1.05 A Mounted .Steel (Ext) fatigque damagqe Analysis evaluated for 30._33 Instrumentation the period of extended System operation RVI Bottom S__S Stainless Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1 .1 .83 A Mounted Steel and neutron flux (B2.1 .2) 24 Instrumentation (Ext)

System (All Components)

RVI Bottom S._S Stainless Reactor Coolant Crackingq Water Chemistry IV.B2.RP- 3.1 .1 .80 A, 3 Mounted Steel and neutron flux ('B2.1 .2) and PWR 293 Instrumentation (Ext) Reactor Internals System BMI (B2.1.35 Column Bodies _____ _____

RVI Bottom S.SS Stainless Reactor Coolant Loss of fracture PWR Reactor Internals IV.B2.RP- 3.1.1.27 A, 3 Mounted Steel and neutron flux tou~ghness (B2.1 .35) 292 Instrumentation (Ext)

System BMI Column Bodies

• RVI Bottom SS Stainless Reactor Coolant Loss of Material Flux Thimble Tube IV.B2.RP- 3.1 .1.60 A, 5 Mounted Steel and neutron flux (wear) Inspection (B2.1 .21) 284 Instrumentation (Ext)

System Flux Thimble Tubes__________________________________________________

Enclosure 2 NOC-AE-1 5003270 Page 33 of 42 Component Type Ilntended: Material *:Environment Aging Effect i Aging Management NUREG-* Table i'I Notes

.... : : ... Function I,,; =:,:*;: : i: ::* : *!: Requiring Program::

n 18 1 ol Ie RVI Bottom SS Stainless Reactor Coolant 'None None '"IV.B2.RP- NA _A Mounted Steel and neutron flux 265 Instrumentation (Ext)

System Column Collars, Extension Bars And Extension Tubes ____ _____

RVI Bottom SS Stainless Reactor Coolant None None IV.B2.RP- NA A Mounted Steel Cast and neutron flux .26._55 Instrumentation Austenitic (Ext)

System Column Cruciforms ________

RVI Bottom SS Stainless Reactor Coolant Crackingq ASME Section Xl IV.B2.RP- 3.1.1.63 A Mounted Steel and neutron flux Inservice Inspection, 382 Instrumentation (Ext) Subsections IWB, IWC, System Lockingq and IWD (B2.1.1)

Devices_______________

RVI Bottom SS Stainless Reactor Coolant Loss of material ASME Section Xl IV.B2.RP- 3.1.1.63 A Mounted Steel and neutron flux (wear) Inservice Inspection, 38._22 Instrumentation ('Ext) Subsections IWB, IWO, System Lockingq and IWO (B2.1.1)

Devices RVI Clevis Insert SS Nickel Alloys Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1.1 .83 A

('All Components) and neutron flux (B. 2 2_4 RVI Clevis Insert S.S Nickel Alloys Reactor Coolant Crackingq Water Chemistry IV.B2.RP- 3.1 .1 .37 A, 3 Bligand neutron flux (B2.1 .2) and PWR 39._.9 (Ext) Reactor Internals

________________ ~('B2.1 .35)_ _ _ _ __ _ _ _ _

RVI Clevis Insert SS Nickel Alloys Reactor Coolant Loss of material PWR Reactor Internals IV.B2.RP- 3.1 .1 .22 A, 3 Blngand neutron flux (wear) (B2.1.35) 285

Enclosure 2 NOC-AE-1 5003270 Page 34 of 42 Component Type :intended Function Materil .Envionmen Aging Effec~t :

Requi.ring Aging Management: !1801

!:"::Program' NUREG-i Vol. Tablel, Item ::,Notesi RVI Clevis Insert _SS Nickel Alloys Reactor Cooiant ILoss of preioad Water Chemistry lV.B2.RP- 3.1.1.22 A, 3 otngand neutron flux (B2.1 .2') and PWR 285 (Ext') Reactor Internals

_______ 8~~~(2.1.35)_____ ____

RVI Clevis Inserts SS Nickel Alloys Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1.1 .83 A and Lock Keys and neutron flux (82.1.2) 24.

RVI Control Rod S._S Stainless Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1.1.83 A_

Guide Tube Steel and neutron flux (B2.1.2') 24.

Assembly and Flow (Ext')

Downcommers (All Components)___________

RVI Control Rod S__S Stainless Reactor Coolant None None IV.B2.RP- NA A Guide Tube Steel and neutron flux 265 Assembly C Tubes ____xt)_

RVI Control Rod S5 Stainless Reactor Coolant Loss of material PWR Reactor Internals IV.B2.RP- 3.1.1.33 A, 3 Guide Tube Steel and neutron flux (wear') (B2.1 .35') 296 Assembly Guide (Ext')

Plates (Cards')_________________

RVI Control Rod S._S Stainless Reactor Coolant Crackinq 'Water Chemistry IV.B2.RP- 3.1.1.37 A, 3 Guide Tube Steel and neutron flux (B2.1 .2') and PWR 355 Assembly Guide (Ext') Reactor Internals Tube Support Pins ________(82.1 .35') _______________

RVI Control Rod 5.5S Stainless Reactor Coolant Loss of material PWR Reactor Internals IV.B2.RP- 3.1.1 .22 A, 3 Guide Tube Steel and neutron flux (wear') (B2.1 .35') 35._66 Assembly Guide (Ext')

Tube Support Pins____________________

RVI Control Rod SS Stainless Reactor Coolant Crackingq ASME Section Xl IV.B2.RP- 3.1.1 .63 A, Guide Tube Steel and neutron flux Inseryice Inspection, 382 Assembly Inserts (Ext') Subsections IWB, IWC, and Screw Lockinq and IWD (82.1.1')

Deyices ________ ___________ _____ ______ _____

Enclosure 2 NOC-AE-1 5003270 Page 35 of 42 Co pyn e nI t n e aeil E vr n e t A i g Ef c gn Ma a em n NUR G- Table~lI Notes S.-.Function. ..... :: .. Requirin....

,gem  :::i*P~rogram 1801 Vol. ~iItem ':

RVI Control Rod SS Stainless Reactor Coolant Loss of material ASME Section Xl IV.B2.RP- 3.1 .1 .63 A Guide Tube Steel and neutron flux (wear) Inservice Inspection, 382 Assembly Inserts (Ext) Subsections IWB, IWC, and Screw Lockingq and IWD (B2.1.1)

Devices (ISI Items No 10 RVI Control Rod S._S Stainless Reactor Coolant None None IV.B2.RP- NA A Guide Tube Steel and neutron flux 265 Assembly Sheaths _____ ______ xt)______________

RVI Control Rod SS Stainless Reactor Coolant Loss of fracture Water Chemistry IV.B2.RP- 3.1 .1.33 A, 3 Guide Tube Steel and neutron flux tougqhness (B2.1 .2) and PWR 29_77 Assembly Lower (Ext) ,Reactor Internals Flangqe Welds (B2.1 .35) ____

RVI Control Rod SS Stainless. Reactor Coolant Crackingq Water Chemistry IV.B2.RP- 3.1.1 .30 A, 3 Guide Tube Steel and neutron flux I(B2.1 .2) and PWR 298 Assembly Lower (Ext) Reactor Internals Flangqe Welds (B2.1 .35)

RVI Core Barrel S._S Stainless Reactor Coolant Cumulative Time-Limited Agqingq IV.B.RP- 3.1 .1 .05 A Assem~bly Steel and neutron flux fatigque damagqe Analysis evaluated for 303 (Ext) the period of extended operation RVI Core Barrel OF, SLD, Stainless *Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1.1.83 A Assembly (All S._S Steel and neutron flux (B2.1 .2) 24.

Components),(Et RVI Core Barrel DF SD Stainless Reactor Coolant Loss of material PWR Reactor Internals IV.B2.RP- 3.1.1.22 A,3 Assembly Core S._S Steel and neutron flux (wear) (B2.1 .35) 345 Barrel Flangqe ____xt)___

RVI Core Barrel DF SD Stainless Reactor Coolant Crackingq Water Chemistry IV.B2.RP- 3.1.1.80 A, 3 Assembly Core SSS Steel and neutron flux (B2.1 .2) and PWR 278 Barrel Outlet Nozzle (Ext) Reactor Internals Welds ________________________(B2.1 .35)

Enclosure 2 NOC-AE-1 5003270 Page 36 of 42

• Component ... : Type :Function Intended: Material ' :i Environimenti J:Aging Effect::i Aging ~Management: !NUREG-:I ReurngPorm 81Vo. Ie I

Tablel*i Notes:

, .. Managem.nt .. .. .. . .... .. 2....Item..

RVI oreBarel, Stainless Reactor Coolant Loss of fracture Water Chemistry IV.B2.RP-, 3.1.1.7 Assembly Core SS Steel and neutron flux tougqhness .(B2.1 .2) and PWR 278a Barrel Outlet Nozzle (Ext) Reactor Internals Welds __ _ _ _ ______ ________ ________(B2.1 .35) _ _ _ _

RVI Core Barrel IDSD Stainless Reactor Coolant Crackingq Water Chemistry IV.B2.RP- 13.1.1.30 A, 3 Assembly Lower SS Steel and neutron flux (B2.1 .2) and PWR 280 Core Barrel Flangqe (Ext) Reactor Internals Welds _____________(B2.1 .35)_____

RVI Core Barrel DF SD Stainless Reactor Coolant Crackingq Water Chemistry IV.B2.RP- 3.1 .1.30 A, 3 Assembly Upper SS Steel and neutron flux (B2.1 .2) and PWR 276 Core Barrel Flangqe (Ext) Reactor Internals Welds (B2.1 .35) ____

RVI Core Barrel S__S Stainless Reactor Coolant Crackingq Water Chemistry IV.B2.RP- .3.1 .1.30 A, 3 Assembly Upper Steel and neutron flux ('B2.1 .2) and PWR 387 Core Barrel and (Ext) Reactor Internals Lower Core Barrel ('B2.1 .35)

Circumferential (girth) Welds _____

RVI Core Barrel S._S Stainless Reactor Coolant Loss of fracture PWR Reactor Internals IV.B2.RP- 3.1.1.33 A, 3 Assembly Upper Steel and neutron flux tougqhness (B2.1 .35) 388 Core Barrel and (Ext) '

Lower Core Barrel Circumferential (gqirth) Welds ___________

RVI Core Barrel 55 Stainless Reactor Coolant ICrackingq Water Chemistry 'IV.B2.RP- 3.1.1.30 A, 3 Assembly Upper Steel and neutron flux (B2.1 .2) and PWR 387a Core Barrel and (Ext) Reactor Internals Lower Core Barrel (B2.1 .35)

Vertical (axial)

Welds_______________________________________________________

Enclosure 2 NOC-AE-1 5003270 Page 37 of 42 Component Type :Intended ::Material: Environment ;l:: Aging Effect ... :Aging Management NUREG,- TaleNoe RVI Core Barrel S._S Stainless Reactor Coolant iLoss of fracture PWR Reactor Internals I V.B2.RP- 3.1 .1 .27 A, 3 Assembly Upper Steel and neutron flux tougqhn*ess (B2.1.35) 388a Core Barrel and (Ext)

Lower Core Barrel Vertical (axial)

Welds ____ _____ ____

RVI Fuel Pin to SS Stainless Reactor Coolant 'Loss of material ASME Section Xl IV.B2.RP- 3.1.1.63 A Core Support Steel and neutron flux (wear) Inservice Inspection. 38.22 Lockingq Device (Ext) Subsections IWB, IWC, and IWD (B2.1.1)____

RVI Fuel Pin to SS Stainless Reactor Coolant Cracking ASME Section Xl IV.B2.RP- 3.1 .1.63 A Core Support Steel and neutron flux Inservice Inspection, 38.22 Lockingq Device (Ext) Subsections IWB, IWC,

_______________ ~~and IWD (B2.1.1) _____ ____

RVI Hold Down S._S Stainless Reactor Coolant None None IV.B2.RP- NA 1,4 SpigSteel and neutron flux 26_55 RVI Irradiation SS Stainless Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1.1.83 A Specimen Guides Steel and neutron flux (B. 2 2.4_

(All Components) ___Ext______

RVI Irradiation S._S Stainless Reactor Coolant Cracking ASME Section Xl IV.B2.RP- 3.1.1 .63 A Specimen Guides Steel and neutron flux Inservice Inspection. 382 Screw Locking (Ext) Subsections IWB, IWC, Device and Dowel and IWD ('B2.1.1)

Pins RVI Lower Internal DFS Stainless Reactor Coolant Crackingq Water Chemistry IV.B2.RP- =3.1.1.37 A, 3 Assembly-XL Lower Steel and neutron flux (B2.1 .2) and PWR 289 Core Plate (Ext) Reactor Internals (B2.1 .35) ____

RVI Lower Internal S._S Stainless Reactor Coolant Cumulative Time-Limited Aging IV.B2.RP- 3.1.1 .05 A Assembly-XL Lower Steel and neutron flux fatigue damage Analysis evaluated for 303, Core Plate (Ext) the period of extended

________

____ __ _______________ __ ___ ____ ___operation _ _ _ _ _ _ _ _ _ _ _ _____

Enclosure 2 NOC-AE-1 5003270 Page 38 of 42 Component Type Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Notes Function Requiring Program:: ... 1801 Vol. Item

________ _________ Management " , , ,2 Item RVI Lower Internal DF, S Stainless Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1.1.83 A Assembly-XL Lower Steel and neutron flux (B. 2 2.*4 Core Plate (Ext)_IVB2_RP-RVI Lower Internal DF S Stainless Reactor Coolant ILoss of material PWR Reactor Internals I.B2.R 3.1.1.27 A, 3 Assembly-XL Lower Steel and neutron flux (wear) (B2.1 .35) 288 Core Plate _Et RVI Lower Internal PF S Stainless Reactor Coolant Loss of fracture PWR Reactor Internals IV.B2.RP- 3.1.1.27 A, 3 Assembly-XL Lower Steel and neutron flux tougqhness (B2.1 .35) 288 Core Plate __ xt RVI Neutron Shield SLD Stainless Reactor Coolant Loss of material ASME Section Xl IV.B2.RP- 3.1.1.63 A Panel Screw Steel and neutron flux ('wear) Inservice Inspection. 38_22 Lockingq Devices (Ext) Subsections IWB, IWC, (ISI Item No 22) ___ ____ and IWD (B2.1.1)

RVI Neutron Shield SLO Stainless Reactor Coolant Crackingq ASME Section Xl IV.B2.RP- 3.1 .1 .63 A Panel Screw Steel and neutron flux Inservice Inspection, 38._22 Lockingq Devices (Ext) Subsections IWB, IWC,

____________

______ _ ___ ___ __ ___ ___ ___and IWD (B2.1..1)

RVl Radial Support S._S Nickel Alloys Reactor Coolant Crackingq ASME Section Xl IV.B2.RP- 3.1 .1.63 A

.Key Lock Keys and and neutron flux Inservice Inspection, 382 Clevis Insert Lock (Ext) Subsections IWB, IWC, Kesand IWD (B2.1.1)

RVI Radial Support SS Stainless Reactor Coolant Loss of material ASME Section Xl IV.B2.RP- 3.1 .1.63 A Key Lock Keys and Steel and neutron flux (wear) .Inservice Inspection. 382 Clevis Insert Lock (Ext) Subsections IWB, IWC, Kesand IWD (B2.1.1)

RVI Secondary S._S Stainless Reactor Coolant Cracking ASME Section Xl IV.B2.RP- 3.1.1.63 A Core Support Steel and neutron flux Inservice Inspection, 38.22 (SCS) Assembly (Ext) Subsections IWB, IWC, SCS locking Device and IWO (B2.1.1) ____

RVI Secondary SS Stainless Reactor Coolant Loss of material ASME Section Xl IV.B2.RP- 3.1.1.63 A Core Support Steel and neutron flux (wear._#.) .Inservice Inspection. 382 (SCS) Assembly (Ext) Subsections IWB, IWC, SCS locking Device ____________________and ,___ IWD (B2.1.1) ____ ____

Enclosure 2 NOC-AE-1 5003270 Page 39 of 42 Component Type Intended Material Environment Aging Effect Aging Management NUREG- Table I Notes Function .. ... Requiring Program ' 1801 Vol. Item Management .... .. 2 Item RVI Secondary SS Stainless Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1.1.83 A Core Support SCS) Steel and neutron flux (B. 2 2_4 Assembly (All (Ext)

C oinponents) ___ __ R e ac tor__ o_ t_ _Co _ _ __an _

RVI Shield SLD IStainless Reco, CoatLoss of material Water Chemistry IV.B2.RP- 3.1.1.83 A Assembly (All Steel and neutron flux (B. 2 24

,Components) -ti-RVI Upper Core SS Stainless Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1.1.83 _A Assembly - Upper Steel and neutron flux (B2.1 .2) 24 Support Column (Ext)

Assemblies (all components)

RVI Upper Core SS Stainless Reactor Coolant Cracking ASME Section Xl IV.B2.RP- 3.1 .1 .63 A, Assembly Steel and neutron flux Inservice Inspection. 382 Thermocouple (Ext) Subsections IWB, IWC, Clamps, Conduit and IWID ('B2.1.1)

Swagq ock Fittinqs, Bandins.qs and Tab Lock__s ____

RVI Upper Core S.__S Stainless Reactor Coolant Loss of material ASME Section Xl IV.B2.RP- 3.1 .1.63 A Assembly Steel and neutron flux (war Inservice Inspection, 38:2 Thermocouple (Ext) Subsections IWB, IWC, Clamps. Conduit and IWD (B2.1.1)

Swaglock Fittins.qs Banding.qs and Tab Locks RVI Upper Core SS Stainless Reactor Coolant Cumulative ,Time-Limited Ai ngq IV.B2.RP- 3.1.1.05 A Assembly -Upper Steel and neutron flux fatigue damage Analysis evaluated for 303 Core Plate (Ext) Ithe period of extended

________________ ~operation_____ ______

RVI Upper Core S_.S Stainless Reactor Coolant Loss of material PWR Reactor Internals IV.B2.RP- 3.1 .1.27 A, 3 Assembly -Upper Steel and neutron flux (wear)_&& (B2.1 .35) 290b Core Plate __ x___

Enclosure 2 NOC-AE-1 5003270 Page 40 of 42 Component Type Intended Material *Environment Aging Effect Aging Management: NUREG- Table I Notes

...Function  :  : .. Requiring Program

... *1801 Vol. ;Item ..

_____ _"__ ____Management .........  :, 21Item * * **" ..

RVI Upper Core SS Stainless Reactor Coolant Crackingq PWR Reactor Internals IV.B2.RP- 3.1.1.80 A, 3 Assembly-Upper Steel and neutron flux (B2.1 .35) 291 b Core Plate ,(Ext)_____________

RVI Upper Core SS Stainless Reactor Coolant Crackingq Water Chemistry IV.B2.RP- 3.1.1 .37 A, 3 Plate Aligqnment Steel and neutron flux (B2.1 .2) and PWR 30._1 Pins (Ext) Reactor Internals

________(B2.1 .35)

RVI Upper Core S.SS Stainless Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1.1,83 IA Plate Aliqnment Steel and neutron flux (B2.1 .2) 24 Pins _ *

,RVI Upper Core S._S Stainless Reactor Coolant Loss of material PWR Reactor Internals IV.B2.RP- 3.1.1 .22 A,3 Plate Aliqnment Steel and neutron flux (wear'l (B2.1 .35) 29_99 Pins _Et

.RVI Upper Core SS Stainless Reactor Coolant Cumulative Time-Limited Agi ng IV.B2.RP- 3.1 .1 .05 A Support-Upper Steel (Ext) fatigque damagqe Analysis evaluated for 30.33 Support Column .the period of extended

_________________operation_____

RVI Upper Core SS Stainless Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1.1.83 A Support-Upper Steel and neutron flux (B. 2 24.

Support Plate (Ext)

Assembly ('All Components)________

RVI Upper SS Stainless Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1 .1.83 A Instrument Conduit Steel and neutron flux (B.12 2.44 and Supports (all (Ext)

.components)

,RVI Upper SS Stainless Reactor Coolant Loss of material WtrCeityI.2R-3118 Instrumentation Steel and neutron flux (B. 2l24.

Conduit and (Ext)

Supports (All

.Corn ponents)

RVI Upper Support SS Stainless Reactor Coolant Loss of material Water Chemistry IV.B2.RP- 3.1 .1.83 A Plate Assembly (All Steel and neutron flux (B. 2 24, Components) E.______ __________ __________

Enclosure 2 NOC-AE-1 5003270 Page 41 of 42 Component Type Intended Material .Environment .Aging Effect Aging Management NUREG- Table I Notes

" Function o i' Requiring Program, 1t801 Vol. Item,

I. ....... * ,*M anagem ent . .... ::...2 Item

.RVI Upper Support 5.__S Stainless Reactor Coolant Crackingq Water Chemistry IV.B2.RP- 3.1.1.37 A, 3 Plate Assembly - Steel and neutron flux (B2.1 .2) and PWR 346

.Upper Support Skirt (Ext) Reactor Internals

_______________B2.1 .35)_ _ _ _

RVI Upper Support SS Stainless Reactor Coolant Cumulative Time-Limited Ai ngq IV.B2-303 3.1 .1 .05 A Plate Assembly- Steel and neutron flux fatigque damagqe Analysis evaluated for Upper Support Skirt (Ext) the period of extended

__________operation Seal Table P__B Stainless Borated Water None None 'IV.E-3 3.1.1.86 A

_____________________Steel Leakagqe (Ext)___________

Notes for Table 3.1 .2-1' Standard Notes:

A Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP is consistent with NUREG-1 801 AMP.

B Consistent with NUREG-1 801 item for component, material, environment, and aging effect. AMP takes some exceptions to NUREG-1 801 AMP.

C Component is different, but consistent with NUREG-1 801 item for material, environment, and aging effect. AMP is consistent with NUREG-1 801 AMP.

E Consistent with NUREG-1 801 for material, environment, and aging effect, but a different aging management program is credited or NUREG-1 801 identifies a plant-specific aging management program.

G Environment not in NUREG-1 801 for this component and material.

I Aging effect in NUREG-1 801 for this component, material and environment combination is not applicable.

Plant Specific Notes:

I Includes the plant specific Nickel-Alloy Aging Management Program (B2.1 .34) in addition to the programs identified in NUREG-1 801.

2 NUREG-1 801 does not address the aging effect of nickel-alloys in borated water leakage. Nickel-alloys subject to an air with borated water leakage environment are similar to stainless steel in a borated water leakage environment and do not experience aging effects due to borated water leakage.

3 Plant-specific aging management program PWR Reactor Internals (B2.1 .35), is credited to manage this aging effect.

4 Sec Fu-rther E'-aluJtion in Scction 3.1.2.2.15. STP uses 403 SS Hold Down Springqs which do not reguire agqingq managqement.

5 See Further Ewaluation in Section 3.1 .2.2.12. Agingq managqement progqram Flux Thimble Tube Inspection (B2.1 .21). is credited to manage this agqingq effect.

6 Scc Fu-rther Ev-alu-tion in Section 3.1.2.2.9.

7 See Further Evaluation in Section 3.1 .2.2.17.

8 Sce F'-rthor Evaluation in Section 3.1.2.2.6.

Enclosure 2 NOC-AE-1 5003270 Page 42 of 42 9- ConWtcnt with EPRI 1016596 (MRP 227), loss offractur t.. is not an' appliCablc**,

-oughness aging;,, effoct- requrin.. an g...... fo the, uppcr..

support column base.

Enclosure 3 NOC-AE-1 5003270 Enclosure 3 New LRA Appendix C, Response to Applicant Action Items for Inspection and Evaluation Guidelines for PWR Internals and STP's Reactor Internals Program Inspection Plan

Enclosure 3 NOC-AE-1 5003270 Page 1 of 19 STP License Renewal Application Technical Information Electric Power Research Institute (EPRI) has published the NRC-approved version of Materials Reliability Program (MRP) Report 1022863 (MRP-227-A), "Pressurized Water Reactor (PWR)

Internals Inspection and Evaluation Guidelines." This report was developed to provide inspection and evaluation guidelines as part of an aging management program for PWR reactor vessel internal components.

The NRC safety evaluation for MRP-227 is included in MRP-227-A, in which the NRC staff determined that MRP-227 is acceptable for referencing in license renewal applications for PWR internals inspection and evaluation. The safety evaluation includes eight plant-specific applicant action items. Included in these items is a request to provide a plant-specific reactor vessel internals inspection plan.

Appendix C includes the following:

1.0 STP conformance with MRP-227 Assumptions 2.0 Topical Report Conditions and Licensee Action Items 3.0 Plant-specific reactor vessel internals inspection plan.

Appendix C Response to Applicant Action Items for Inspection Page C-1 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 2 of 19 STP License Renewal Application Technical Information 1.0 Conformance with MRP-227 Assumptions 1.1 Assumption 1: Thirty years of operation with high leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management strategy for the remaining years of operation.

STPNOC Response: STPNOC implemented a full low-leakage fuel pattern during Cycle 6 for STP Unit I and Cycle 5 for STP Unit 2. The assumption of 30 years of operation with a high-leakage core is bounded for STP.

1.2 Assumption 2: Base load operation, i.e., typically operates at full power levels and does not usually vary power on a calendar or load demand schedule.

STPNOC Response: STP Unit 1 ahd Unit 2 are considered base load units.

1.3 Assumption 3: No design changes beyond those identified in general industry guidance or recommended by the original vendors.

STPNOC Response: .STP Unit I and Unit 2 have had no identified changes beyond those recommended by OEM.

2.0 Topical Report Conditions and Licensee Action Items The final NRC Safety Evaluation (SE) for MRP-227-A contains seven Topical Report Conditions and eight Applicant/Licensee Action Items. This section provides the STPNOC responses to the Topical Report Conditions and Applicant/Licensee Action Items that are incorporated in the inspection plan.

2.1 Topical Report Conditions 2.1.1 SE Section 4.1.1, Topical Report Condition 1: Moving components from "No Additional Measures" to "Expansion" category.

STPNOC Response. In accordance with SE Section 4.1.1, the (1) Upper Core Plate and (2) Lower Support Forging or Casting have been added to the STPNOC "Expansion"categoryand are contained in Table 4. 6. The components are linked to, and examination method is consistent with, the "Primary" components Control Rod Guide Tube Assembly Lower Flange Welds.

Appendix C Response to Applicant Action Items for Inspection Page C-2 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 3 of 19 STP License Renewal Application Technical information 2.1.2 SE Section 4.1.2, Topical Report Condition 2: Inspection of components subject to irradiation stress corrosion cracking.

STPNOC Response: In accordance with SE Section 4.1.2, the (I) Upper and Lower Core barrel Girth Welds and the (2) Lower Core Barrel Flange Welds have been added to the S TPNOC "Primary"Tnspectioncategory and are contained in table 4-3. The examination method is consistent with the MRP recommendations for these components, the exam coverage conforms to the criteria described in Section 3.3.1 of the NRC SE, and the re-examination period is on a 10-year interval consistent with other "Primary"inspection category components.

2.1.3 SE Section 4.1.3, Topical Report Condition 3: Inspection of high consequence components subject to multiple degradation mechanisms.

STPNOC Response: This condition and limitation is only applicable to B&W and CE designed reactors. NOT applicable to STPNOC (Westinghouse Design Plants).

2.1.4 SE Section 4.1.4, Topical Report Condition 4: Imposition of minimum examination coverage criteria for "Expansion" inspection category components.

STPNOC Response: One hundred percent of the volume/area of each accessible components will be examined. The minimum examination coverage for "Primary"and "Expansion"inspection categories is seventy-five percent of the component's total (accessible plus inaccessible) inspection volume/area be examined or, when addressing a set of like components, that the inspection examine a minimum sample size of seventy-five percent of the total population of like components. Defects shall be entered into the CorrectiveAction Program and evaluated whether the component, or set of like components, will continue to meet their intended function under all licensing basis conditions of operation until the next scheduled examination.

2.1.5 SE Section 4.1.5, Topical Report Condition 5: Examination frequencies for baffle former bolts and core shroud bolts.

STPNOC Response: The frequency of examination for baffle former bolts has been revised to a ten-year interval following the initial or baseline examination.

Appendix C Response to Applicant Action Items for Inspection Page C-3 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 4 of 19 STP License Renewal Application Technical Information 2.1.6 SE Section 4.1.6, Topical Report Condition 6: Periodicity of the re-examination of "Expansion" inspection category components.

STPNOC Response: In accordance with SE Section 4.1.6, Table 4-6 has been revised to require a 10-year re-examination frequency for all "Expansion" inspection category components once degradation is identified in the associated "Primary"inspection category component.

2.1.7 SE Section 4.1.7, Topical Report Condition 7: Updating of industry guideline MRP-227-A, Appendix A to include a reference to AMP XMI.M6A in NUREG 1801, Revision 2.

STPNOC Response: STPNOC conforms to the recommended program element criteria in AMP XI.MI6A, Revision 2. No plant specific actions are required.

2.2 Applicant/Licensee Action Items (A/LAI) 2.2.1 SE Section 4.2.1, Applicant/Licensee Action Item 1: Applicability of FMECA and Functionality Analysis Assumptions.

STPNOC Response: The STPNOC Units 1 and 2 Reactor Vessel Internals (RVI) components are reasonablyrepresented by the design and operating history assumptions regardingneutron fluence, temperature, materials, and stress values in the MRP-191 generic FailureModes, Effects and CriticalityAnalysis (FMECA) and in the MRP-232 functionality analysis. The STPNOC Units1I and 2 comply with A/LAI 1 of the NRC SE regarding MRP-22 7-A. Therefore the requirements are met for applicationof MRP-22 7-A as a strategy for managing age-related material degradationin the RVI components. (Ref. PWROG-15001-P) 2.2.2 SE Section 4.2.2, Applicant/Licensee Action item 2: PWR Vessel Internals Components within the Scope of License Renewal.

STPNOC Response: The generic scoping and screening of the RVI, as summarized in MRP-191 and MRP-232, to support the inspection sampling approach for aging management of the RVI specified in MRP-22 7-A are applicable to STPNOC Units 1 and 2 with no modifications for the components.

STPNOC Units 1 and 2 comply with A/LA I 2 of the NRC SE in MRP-22 7-A for all components. (Ref. PWROG-15001-P).

Appendix C Response to Applicant Action Items for Inspection Page 0-4 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 5 of 19 STP License Renewal Application Technical Information 2.2.3 SE Section 4.2.3, Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-Specific Existing Programs.

STPNOC Response: STPNOC replaced the X-750 guide tube supportpins (split pins) with cold worked (CW) 316 stainless steel during refueling outages 1RE12 and 2REll1(2005). The new split pins were qualified for a 60 year design objective at a 100% capacity factor (WCAP-16620-P, Rev.O). Potentialaging effects were evaluated including those identified in MRP-191 Table 5-1. No additionalinspection requirements were established for the con trol rod guide tube support pins in the design change packages that installed them based on the following:

• Cold-worked Type 316 SS split pins have been installed at other plants since 1997 and none of these plants have experienced any failures.

• Since other plants have installed split pins since 1997 and STPNOC did not install them until 2005 for Units 1 and 2, the other plants will provide a leading indicator.

At STPNOC the effects of aging on these components will be managed in the period of extended operation based on operating experience.

2.2.4 SE Section 4.2.4, Applicant/Licensee Action Item 4: B&W Core Support Structure Upper Flange Stress Relief.

STPNOC Response: This condition and limitation is only applicable to B&W designed reactors. NOT applicable to STPNOC (Westinghouse Design Plants).

2.2.5 SE Section 4.2.5, Applicant/Licensee Action Item 5: Application of Physical Measurements as Part of l&E Guidelines for B&W, CE, and Westinghouse RVI Components.

STPNOC Response: The reactorinternalshold down spring is fabricated with Type 403 stainless steel. The type 403 stainless steel reactorinternals hold down spring is screened as FMECA Group 1, Category A in MRP-191, Table 7-

2. The Category A assignment is applied to components that are deemed with "low probability of failure".

2.2.6 SE Section 4.2.6, Applicant/Licensee Action Item 6: Evaluation of Inaccessible and Non-Inspectable B&W Components.

STPNOC Response: This condition and limitation is only applicable to B&W designed reactors. NOT applicable to S TPNOC (Westinghouse Design Plants).

Appendix C Response to Applicant Action Items for Inspection Page 0-5 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 6 of 19 STP License Renewal Application Technical Information 2.2.7 SE Section 4.2.7, Applicant/Licensee Action Item 7: Plant Specific Evaluation of CASS Materials.

STPNOC Response: A/LAI 7, from the NRC final SE on MRP-227 [3], states that, for assessment of CASS materials, the applicant/licensee for license renewal may apply the criteria in the NRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, "Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components,""as the basis for determining whether the CASS materials are susceptible to the thermal aging mechanism. If the application of the applicable screening criteria for the components material demonstrates that the components are not susceptible to either TE or IE, or to the synergistic effects of TE and IE combined, then no other evaluation would be necessary.

Both the STP Units 1 and 2 upper internals assembly subcomponents a) upper support columns, and b) Column base, are comprised of CA SS, Grade CF8 material. Accounting for the elemental percentagesfrom the chemical data retrieved from the certified materialtest report (CMTRs), calculated percentage delta ferrite and potential for thermal embrittlement (TE), it is concluded that continued application of the MRP-22 7-A strategy will meet the requirement for managing age-relateddegradation of the STP Units 1 and 2 CASS reactor vessel internals components (Ref. PWROG- 1500 1-P).

2.2.8 SE Section 4.2.8, Applicant/Licensee Action Item 8: Submittal of Information for Staff Review and Approval.

STPNOC Response: In accordance with the NUREG-1801, Revision 2, the following information is provided for the following items (1) through (5) for staff review and approval.

1. An AMP for the facility that addresses the 10 program elements as defined in NUREG-1 801, Revision 2, AMP XI.M16A.

STPNOC Response: STPNOC AMP B2. 1.35, South Texas Project License Renewal ProgramEvaluation address the 10 program elements as defined NUREG-1801 (GALL Report), Revision 2, ChapterXL.MI6A, "PWR Vessel Internals" as modified by LR-/SG-201 1-04.

2. To ensure the MRP-227 program and plant-specific action items will be carried out by applicants/licensees, applicants/licensees are to submit an inspection plan which addresses the identified plant-specific action items for staff review and approval consistent with the licensing basis for the plant. If an applicant/licensee plans to implement an AMP which deviates from the guidance provided in MRP-227, as approved by the NRC, the applicant/licensee shall identify where their program deviates from the Appendix C Response to Applicant Action Items for Inspection Page C-6 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 7 of 19 STP License Renewal Application Technical information recommendations of MRP-227, as approved by the NRC, and shall provide a justification for any deviation which includes a consideration of how the deviation affects both "Primary" and "Expansion" inspection category components.

STPNOC Response: The PWR Vessel Internals program (B2. 1.35) is described in LRA Section A 1.35 and LRA Section B2. 1.35. The STPNOC Inspection Plan is included in LRA Appendix C and addressesplant specific action items and does NOT identify any deviations to MRP-22 7-A.

3. The regulation at 10 CFR 54.21(d) requires that an FSAR supplement for the facility contain a summary description of the programs and activities for managing the effects of aging and the evaluation of TLAAs for the period of extended operation. Those applicants for LR referencing MRP-227, as approyed by the NRC, for their RVI component AMP shall ensure that the programs and activities specified as necessary in MRP-227, as approved by the NRC, are summarily described in the FSAR supplement.

STPNOC Response: The UFSAR Supplement is included in LRA Appendix A, Section A 1.35 and includes a summary of the program and activities specified as necessary for the PWR Reactor Internals (B2. 1.35) program.

4. The regulation at 10 CFR 54.22 requires each applicant for LR to submit any TS changes (and the justification for the changes) that are necessary to manage the effects of aging during the period of extended operation as part of its LR application (LRA). For the plant CLBs that include mandated inspection or analysis requirements for RVI either in the operating license for the facility or in the facility TS, the applicant/licensee shall compare the mandated requirements with the recommendations in the NRC-approved version of MRP-227. If the mandated requirements differ from the recommended criteria in MRP-227, as approved by the NRC, the conditions in the applicable license conditions or TS requirements take precedence over the MRP recommendations and shall be complied with.

STPNOC Response: No changes to the Technical Specifications (TS) are required.

Appendix C Response to Applicant Action Items for Inspection Page C-7 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 8 of 19 STP License Renewal Application Technical Information

5. Pursuant to 10 CFR 54.21 (c)(1), the applicant is required to identify all analyses in the CLB for their RVI components that conform to the definition of a TLAA in 10 CFR 54.3 and shall identify these analyses as TLAAs for the application in accordance with the TLAA identification requirement in 10 CFR 54.21(c)(1). MRP-227 does not specifically address the resolution of TLAAs that may apply to applicant/licensee RVI components. Hence, applicants/licensees who implement MRP-227, as approved by the NRC, shall still evaluate the CLB for their facilities to determine if they have plant-specific TLAAs that shall be addressed. If so, the applicant's/licensee's TLAA shall be submitted for NRC review along with the applicant's/licensee's application to implement the NRC- approved version of MRP-227.

STPNOC Response: Reactor Internals TLAAs are addressed in LRA Section 4.3.3 and does not credit the PWR Reactor Vessel Internalsprogram (B2. 1.35).

Appendix C Response to Applicant Action Items for Inspection Page C-8 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 9 of 19 STP License Renewal Application Technical Information 3.0 PWR Vessel Internals Inspection Plan The PWR Vessel Internals Inspection Plan is provided in Tables A through E.

• Table A specifies the vessel internal components classified as Primary components and is based on MRP-227-A, Table 4.3.

• Table B specifies the vessel internal components classified as expansion components and is based on MRP-227-A, Table 4.6.

• Table C specifies the examination acceptance and expansion criteria and is based on MRP-227-A, Table 5.3.

• Table 0 specifies the components that are classified as Existing Program components.

• Table E provides the Examination Plan Summary Table item I ISG A: Westinghouse/STP Applicability.

Primary Effect::

Components Expansi:on EXamination

..  ::-: ...

_____________

,Examination

..:

.2011-04 * (Mechanism) Link (Note 1) Meth~odlFrequency:: Coverage  :

(Note 1) (see Addendum 5

• for figures)

Control Rod All plants Loss of Material None Visual (VT-3) 20% examination Guide Tube (Wear) examination no later of the number of Assembly than 2 refueling outages CRGT assemblies, Guide plates from the beginning of with all guide cards (cards) the license renewal within each period, and no earlier selected CRGT IV.B2.RP-296 than two refueling assembly outages prior to the start examined.

of the license renewal See Figure 4-20 period. Subsequent examinations are required on a ten-year

______________ _______________

________________interval. _________

Control Rod All plants Cracking (SCC, Bottom-mounted Enhanced visual (EV/T- 100% of outer Guide Tube Fatigue) instrumentation 1) examination to (accessible) CRGT Assembly Loss of Fracture (BMI) column determine the presence lower flange weld Lower flange Toughness (IE) bodies, of crack-like surface surfaces and welds flaws in flange welds no adjacent base Upper core plate later than 2 refueling metal on the IV.B2.RP-297 outages from the individual periphery IV.B2.RP-298 beginning of the license CRGT assemblies.

renewal period and (Note 5) subsequent examination See Figure 4-21.

on a ten-year interval. _________

Core Barrel All plants Cracking (SCC Core barrel outlet Periodic enhanced 100% of one side Assembly IASCC, Fatigue) nozzle welds visual (EVT-1) of the accessible Upper core examination, no later surfaces of the barrel flange than 2 refueling outages selected weld and weld from the beginning of adjacent base the license renewal metal (Note 3).

IV.B2.RP-276 period and subsequent See Figure 4-22.

examination on a ten-year interval.

Appendix C Response to Applicant Action for Items for Internals Inspection Page C-9 and Evaluation Guidelines PWR

Enclosure 3 NOC-AE-1 5003270 Page 10 of 19 STP License Renewal Application Technical Information Table A: WestinghoUselSTP Primary.components ::" , -' ** i,:

Item! ISG Applicability Effect!:= "ExpanSion Examination ... Exami nation 2011-04 (Mechanism) Link (Note 1) MethodiFrequency :Coverage (Note 1) (see Addendum 5 for figures)

Core Barrel All plants Cracking (SCC, Upper and lower Periodic enhanced Bolts and locking Assembly IASCC, Fatigue) core barrel vertical visual (EVT-1) devices on high Upper and lower Loss of Fracture axial welds examination, no later fluence core barrel Toughness (IE) than 2 refueling outages seams.100% of Circumferential from the beginning of one side of the girth welds the license renewal accessible period and subsequent surfaces of the IV,.B2.RP-387 examination on a ten- selected weld and IV.B2.RP-388 year interval, adjacent base metal (Note 3).

See Figure 4-22.

Core Barrel All plants Cracking (SCC, None Periodic enhanced 100% of one side Assembly Fatigue) visual (EVT-1) of the accessible Lower core examination, no later surfaces of the barrel flange than 2 refueling outages selected weld and weld (Note 4) from the beginning of adjacent base the license renewal metal (Note 3).

IV.B2.RP-280 period and subsequent examination on a ten-year interval.

Baffle-Former All plants with Assembly baffle-edge bolts Baffle-edge bolts N/A STP does not have baffle-edge bolts Baffle-Former All plants Cracking (SCC, Barrel-former bolts Baseline volumetric (UT) 100% of accessible Assembly IASCC, Fatigue) examination between 25 bolts (Note 2).

Baffle-former Loss of Fracture and 35 EFPY, with Heads accessible bolts Toughness (IE and subsequent examination from the core side.

Loss of Preload on a ten-year interval. UT accessibility IV.B2.RP-271 (iSR) may be affected by IV.B2.RP-272 complexity of head and locking device designs.

See Figures 4-23 and 4-24.

Appendix C Response to Applicant Action Items for Inspection Page C-10 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 11 of 19 STP License Renewal Application Technical Information Table A: Westinghous~e/STP Primary components ,i* .... .. :*"....::

"ltem I iSG Applicability: Effect=  : ExpanSion Examnination "  : EXamination 2011-04 , (MeChanism) Link (Note 1) Method/Frequency Coverage

"*(Note 1) (see Addendum 5

• for figures)

Baffle-Former All plants Change in None Visual (v-r-3) ... Core side surface Assembly Dimensions examination to check for as indicated.

Assembly (Distortion or Void evidence of distortion, See Figures 4-24, (Includes: Baffle Swelling), or with baseline 4-25, 4-26 and 4-plates, and Cracking (IASCC), examination between 20 27.

indirect effects of Fatigue) that and 40 EFPY and void swelling in results in subsequent former plates) a Abnormal examinations on a ten-interaction with fuel year interval.

IV.B2.RP-270 assemblies IV.B2.RP-270a e Gaps along high fiuence baffle joint

• Vertical displacement of baffle plates near high fiuence joint Alignment and All plants with 304 Interfacing stainless steel hold Components down springs Internals hold down spring STP utilizes 403 SS hold down IV.B2.RP-300 springs Thermal Shield All plants with N/A STP does not Assembly thermal shields utilize flexure Thermal shield design flexures Notes Table A:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3.

2. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined for inspection credit, 3, A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined from either the inner or outer diameter for inspection credit.

4. The lower core barrel flange weld may be alternatively designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

5. A minimum of 75% of the total identified sample population must be examined.

Appendix C Response to Applicant Action Items for Inspection Page C-11 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 12 of 19 STP License Renewal Application Technical Information Table B:.Westi nghouselSTP Expansion components :  :  :*  :*::;  : :

Item IiISG-2011' Applicability Effect 'Primary Link

• Examination' Examination 04 ....:(Mechanism) (Note 1) ;MethOd (Note Coverage

... 1) (see Addendum 5

....  ::  :  ::; ... for figures)

Upper Internals All plants Cracking (Fatigue), CRGT lower Enhanced visual 100% of accessible Assembly Loss of Material flange weld (EVT-1) surfaces (Note 2).

Upper core plate (Wear) examination.

Re-inspection every IV.B2.RP-290b 10 years following IV.B2.RP-291 b initial inspection.

Core Barrel All plants Assembly Lower support N/A STP has an forging or castings extended core does not have a Lower

________________Support Assembly _________

Baffle-Former All plants Cracking (IASCC, Baffle-former Volumetric (UT) 100% of accessible Assembly Fatigue) bolts examination, bolts. Accessibility Barrel to former Loss of Facture Re-inspection every may be limited by bolts Toughness "(IE), 10 years following presence of thermal Change in initial inspection, shields or neutron IVB2.RP-273 Dimensions pads (Note 2).

IV.B2.RP-274 (Distortion or Void See Figure 4-23.

Swelling) and Loss of Preload (ISR)

Lower Support All plants Assembly Lower support N/A STP has an column bolts extended core does not have a Lower Support Assembly Core Barrel All plants Cracking (SCC, Upper core barrel Enhanced visual 100% of one side of Assembly Fatigue) flange weld (EVT'-1) the accessible Core barrel outlet Facture Toughness examination, surfaces of the nozzle welds (IE of lower Re-inspection every selected weld and sections) 10 years following adjacent base IV.B2.RP-278 initial inspection. metal (Note 2).

IV.B2.RP-278a See Figure 4-22.

Core Barrel All plants Cracking (SCC, Upper and lower Enhanced visual 100% of one side of Assembly IASCC, Fatigue) core barrel (EVT-1) the accessible Upper and lower Facture Toughness Circumferential examination, surfaces of the core barrel vertical (IE) girth welds Re-inspection every selected weld and axial welds 10 years following adjacent base initial inspection, metal (Note 2).

IV.B2.RP-387a See Figure 4-22.

IV.B2.RP-388a Appendix C Response to Applicant Action Items for Inspection Page C-12 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 13 of 19 STP License Renewal Application Technical Information Table B: WestinghouselSTP Expansion components,....  : ,. .

item IISG-2011- Applicability Effect ... :Primary Link :Examination : Examination 04' =. . (Mechanism) (Note 1) Method (Note Cov/erage

= *1) (see Addendum 5

___________for figures)

Lower Support All plants Assembly Lower support N/A STP has an column bodies extended core does (non cast) not have Lower Support Assembly Lower Support All plants Assembly Lower support N/A STP has an column bodies extended core does (cast) not have a Lower Support Assembly Bottom Mounted All plants Cracking (Fatigue) Control rod guide Visual (VT'-3) 100% of BMI Instrumentation including the tube (CROT) examination of BMI column bodies for System detection of lower flanges column bodies as which difficulty is Bottom-mounted completely indicated by difficulty detected during flux instrumentation fractured column of thimble (BMI) column bodies bodies insertion/withdrawal insertion/withdrawal Facture Toughness of flux thimbles. .

IV.B2.RP-292 (IE) Re-inspection every See Figure 4-35.

IV.B2.RP-293 10 years following initial inspection.

Flux thimble insertion/withdrawal to be monitored at each inspection interval.__________

Notes to Table B:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3.

2. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

Appendix C Response to Applicant Action Items for Inspection Page C-13 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 14 of 19 STP License Renewal Application Technical Information Table C: westinghouse/STP examination acceptance and expansion criteria *"::

Item ... Applicability ,Examination Expansion Expansion Additional

• .. Acceptance : :Link(s) ..... Criteria Examination

= =Criteria (Note .,1) :: .; ,=._* ;Accetance_ .. *.

., =' ... . *=:; *,,  : =*: c riteria =

Control Rod All plants Visual (VT-3) None N/A N/A Guide Tube examination Assembly Guide plates The specific relevant (cards) condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.

Control Rod All plants Enhanced visual a. Bottom- a. Confirmation of a. For BMI column Guide Tube (EVT-1) examination mounted surface-breaking bodies, the specific Assembly instrumentation indications in two relevant condition Lower flange welds The specific relevant (BMI) column or more CRGT for the VT-3 condition is a bodies lower flange welds, examination is detectable crack-like b. upper core combined with flux completely surface indication, plate thimble fractured column insertion/withdrawal bodies.

difficulty, shall b. For cast lower require visual (VT- support column

3) examination of bodies, upper core BMI column bodies plate and lower by the completion support of the next forging/castings, refueling outage. the specific relevant

b. Confirmation of condition is a surface-breaking detectable crack-indications in two like surface or more CRGT indication.

lower flange welds shall require EVT-1 examination of s, upper core plate within three fuel cycles following the initial observation.

Appendix C Response to Applicant Action for Items for Internals Inspection Page 0-14 and Evaluation Guidelines PWR

Enclosure 3 NOC-AE-1 5003270 Page 15 of 19 STP License Renewal Application Technical Information T~hI~ (~ W~~fmnnhnIQ~I~TP a~iv~amin2tfln ~nt~nr~ ~nrI ~yn~incinn r~ri*r~rr~a Item  :-: Applicability: Examination :: ]Expansion i Expansion ... :Additional

• i " ii ,AccePtance =iLink(S) , Criteria

• I Examination :

i; *::;.,::'*;;;

=  : .... = : i :Criteria(Note 1): , * . i;; i;= ' : :':..= A cceptance -;

• Core Barrel All plants Periodic enhanced a. Core barrel a. The confirmed a and b. The Assembly visual (EVT-1) outlet nozzle detection and specific relevant Upper core barrel examination welds sizing of a surface- condition for the flange weld breaking indication expansion core The specific relevant with a length barrel outlet nozzle condition is a greater than two weld and lower detectable crack-like inches in the upper support column surface indication, core barrel flange body examination is weld shall require a detectable crack-that the EVT-1 like surface examination be indication.

I expanded-to include the core barrel outlet nozzle welds by the completion of the next refueling outage.

Core Barrel All plants Periodic enhanced None None None Assembly visual (EVT-1)

Lower core barrel examination flange weld (Note

2) The specific relevant condition is a detectable crack-like surface indication.

Core Barrel All plants Periodic enhanced Upper core barrel The confirmed The specific Assembly visual (EVT-1) vertical axial detection and relevant condition Upper core barrel examination welds sizing of a surface- for the expansion circumferential girth breaking indication upper core barrel welds The specific relevant with a length vertical axial weld condition is a greater than two examination is a detectable crack-like inches in the upper detectable crack-surface indication, core barrel like surface circumferential indication.

girth welds shall require that the EV-1'- examination be expanded to include the upper core barrel vertical axial welds by the completion of the next refueling outage.

Appendix C Response to Applicant Action Items for Inspection Page C-15 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 16 of 19 STP License Renewal Application Technical Information Table C: Westinghouse/STP examination acceptance and expansion criteria*==:,;*

Item Applicability Examination Expansion Expansion Additional SAcceptance :Link(s) , Criteria Examination

... . . , Criteria (Note 1) ,..... : Acceptance

___________ _______ _ :Criteria Core Barrel All plants Periodic enhanced Lower core barrel The confirmed The specific Assembly visual (EVT-1) vertical axial detection and relevant condition Lower core barrel examination welds sizing of a surface- for the expansion circumferential girth breaking indication lower core barrel welds The specific relevant with a length vertical axial weld condition is a greater than two examination is a detectable crack-like inches in the lower detectable crack-surface indication, core barrel like surface circumferential indication.

girth welds shall require that the EVT'-1 examination be expanded to include the lower core barrel vertical axial welds by the completion of the next refueling outage.

Baffle-Former All plants with Assembly baffle-edge bolts Baffle-edge bolts N/A STP does not have baffle-edge bolts Baffle-Former All plants Volumetric (UT) a. Barrel-former Confirmation that The examination Assembly examination bolts more than 5% of acceptance criteria Baffle-former bolts the baffle-former for the UT of the The examination bolts actually barrel-former bolts acceptance criteria for examined on the shall be established the UT of the baffle- four baffle plates at as part of the former bolts shall be the largest distance examination established as part of from the core technical the examination (presumed to be justification.

technical justification, the lowest dose locations) contain unacceptable indications shall require UT examination of the barrel-form er bolts bolts within the next three fuel cycles.

Appendix C Response to Applicant Action Items for Inspection Page C-16 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 17 of 19 STP License Renewal Application Technical Information Table C: Westinahouse/STP examination accentance and exnansion criteria Item Applicability Examination Expansion Expansion Additional  :

SAcceptance Link(s) Criteria': Examination

• : *:* Criteria (Note 1):i Acceptance ,

Baffle-Former All plants Visual (VT-3) None N/A N/A Assembly examination Assembly The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, vertical displacement of shroud plates near high fluence joints, and broken or damaged edge bolt locking systems along high fluence baffle plate joints.

Alignment and All plants with 304 None Interfacing stainless steel hold Components down springs Internals hold down spring STP utilizes 403 SS hold down springs Thermal Shield All plants with Assembly thermal shields Thermal shield flexures N/A STP does not utilize flexure

________________design___________ ______________________________

Notes to Table C:

1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).

2. The lower core barrel flange weld may alternatively be designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

Appendix C Response to Applicant Action items for inspection Page 0-17 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 18 of 19 STP License Renewal Application Technical Information Table D: Westinahouse/STP Existina Proarams comDonents item *: Applicability Effect Reference , Examination Examination Barll plants(Mechanism) _________ Method .. .Coverage ..

Core BreAlplnsLoss of material PWR Reactor Visual (VT-3) All accessible Assembly (Wear) Internals (B2.1.35) examination to surfaces at Core barrel flange IV.B2.RP-345 determine general specified frequency condition for excessive wear Upper Internals All plants Cracking (SCC, ASME Code Visual (VT-3) All accessible Assembly Fatigue) Section Xl Items examination surfaces at Upper support skirt Also includes bolts, 12 and 13 and specified frequency locking devices, PWR Reactor Internals (B2.1 .35)

IV.B2.RP-346 Lower Internals All plants Cracking (IASCC, PWR Reactor Visual (VTI-3) All accessible Assembly Fatigue) Internals (B2.1.35) examination of the surfaces at Lower core plate STP has the XL Facture Toughness lower core plates to specified frequency XL lower core plate lower core plate (IE) and Loss of iV.B2.RP-288 and detect evidence of (Note 1) Material (Wear) IV.B2.RP-289 distortion and/or (SR) loss of bolt integrity.

Bottom Mounted All plants Loss of material Flux Thimble Tube Surface (ET) Eddy current Instrumentation (Wear) inspection examination surface System (B2.1.21) examination as Flux thimble tubes defined in plant response to IEB 88-

______________09. (ref. 8.1.22)

Alignment and All plants Cracking (SCC, PWR Reactor Visual (VT-3) All accessible Interfacing lASCC, Fatigue Internals (B2.1.35) examination surfaces at Components Loss of material IV.B2.RP-399 and specified frequency Clevis insert bolts (Wear) Loss of IV.B2.RP-285 Preload (SR)

(Note 2)

Alignment and All plants Loss of material PWR Reactor Visual (VT-3) All accessible Interfacing (Wear) Internals (82.1.35) examination surfaces at Components Cracking (SCC) IV.B2.RP-299 and specified frequency Upper core plate IV.B2.RP-301 alignment pins RVI Control Rod All plants Loss of material PWR Reactor Visual (VT-3) All accessible Guide Tube (wear) internals (82.1.35) examination surfaces at Assembly Guide IV.B2.RP-356 specified frequency Tube Support Pins Notes to Table D:

1. XL = "Extra Long" referring to Westinghouse plants with 14-foot cores.

2. Bolt was screened in because of stress relaxation and associated cracking; however, wear of the clevis/insert is the issue.

Appendix C Response to Applicant Action Items for Inspection Page 0-18 and Evaluation Guidelines for PWR Internals

Enclosure 3 NOC-AE-1 5003270 Page 19 of 19 STP License Renewal Application Technical Information Table E: Examination Plan Summary Table Iseto dee..

Primary CmoetExpansion Links ,ISP0Cuil  :,tSPNCUt2 Co~mments Control Rod Guide See tables 4-3, 4-Tube Assembly 1 E66, and 5-3 for Guide plates (cards) None 20E26) 2RE25 primary/expansion (Spring 02) (Spring 2027) components, and acceptance criteria Control Rod Guide Bottom-mounted See tables 4-3, 4-Tube Assembly Instrumentation 6, and 5-3 for Lower flange welds (BMI) column 1RE26 2RE25 primary/expansion bodies, (Spring 2026) (Spring 2027) components, and Upper core plate acceptance criteria Core Barrel See tables 4-3, 4-Assembly Core barrel outlet 1 RE26 2RE25 6, and 5-3 for Upper core barrel nozzle welds (Spring 2026) (Spring 2027) primary/expansion flange weld components, and acceptance criteria Core Barrel See tables 4-3, 4-Assembly Upper and lower 2E56, and 5-3 for Upper and lower coebreetcl1 RE26 2E5primary/expansion core barrel axial welds (Spring 2026) (pig27) components, and Circumferential girth acceptance criteria welds Core Barrel See tables 4-3, 4-Assembly 6 n - o Lower core barrel None 1RE26 2RE25 6,imand/e5-3nfor flange weld (Spring 2026) (Spring 2027) components, and acceptance criteria Baffle-Former See tables 4-3, 4-Assembly 6 n - o Baffle-former bolts Barrel-former bolts 1RE26 2RE25 6,imand/e5-3nfor (Spring 2026) (Spring 2027) components, and acceptance criteria Baffle-Former See tables 4-3, 4-Assembly 6, and 5-3 for (Includes: Baffle primary/expansion plates, and indirect None 1RE26 2RE25 components, and effects of void (Spring 2026) (Spring 2027) acceptance criteria swelling in former plates) ____________________________ _________

Appendix C Response to Applicant Action Items for Inspection Page C-19 and Evaluation Guidelines for PWR Internals