W3F1-2013-0070, Submittal of Reactor Vessel Internals Aging Management Program Consistent with MRP-227-A

Submittal of Reactor Vessel Internals Aging Management Program Consistent with MRP-227-A

ML13352A041
Person / Time
Site:	Waterford
Issue date:	12/16/2013
From:	Jarrell J Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2013-0070
Download: ML13352A041 (81)

Text

Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093 Tel 504 739 6685 Fax 504 739 6698

'""Entergy jjarrel@entergy.com John P Jarrell III Manager, Regulatory Assurance Waterford 3 W3F1-2013-0070 December 16, 2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Submittal of Reactor Vessel Internals Aging Management Program Consistent with MRP-227-A Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

REFERENCES:

1. Entergy Letter dated February 5, 2005, "Supplement to Amendment Request NPF-38-249 Extended Power Uprate" [Adams Accession No. ML050400463]

2. Entergy Letter dated April 7, 2010, "Commitment Change Associated with Reactor Vessel Internals Degradation Management Program"

[Adams Accession No. ML100990355]

3. NRC letter dated June 12, 2009, "Waterford Steam Electric Station, Unit 3 - Request for Alternative W3-ISI-006 for the Second 10-Year Inservice Inspection Interval (TAC No. MD9671)" [Adams Accession No. ML091210375]

4. TR MRP-227 SER [Adams Accession No. ML111600498]

5. NRC Regulatory Issue Summary 2011-07 License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management (Dated July 21, 2011) [Adams Accession No. ML111990086]

6. Entergy Letter dated December 19, 2011, "Commitment Change for Reactor Vessel Internals Degradation Management Program" [Adams Accession No. ML11356A083]

7. NRC letter dated December 16, 2011, "Revision 1 to the Final Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Program (MRP) Report 1016596 (MRP-227),

Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines" (TAC NO. ME0680)" [Adams Accession No. ML11308A770]

8. Entergy Letter dated February 27, 2012, "Commitment Date Change for Submittal of Reactor Vessel Internals Degradation Management Program Consistent with MRP-227-A" [Adams Accession No. ML12059A077]

W3F1-2013-0070 Page 2

Dear Sir or Madam:

This letter is to transmit to the NRC for your review and approval the Waterford Steam Electric Station, Unit 3 (WF3) Reactor Vessel Internals Aging Management Program (AMP) developed to implement MRP-227-A, Rev 0.

The Waterford 3 Reactor Vessel Internals AMP meets a "Needed" element of MRP-227-A and is a description of the program, including the inspection plan.

This also complies with a committed action from previous Entergy letters, specifically Entergy Letter dated February 5, 2005 (Reference 1) as changed by Entergy Letter dated April 7, 2010 (Reference 2), Entergy Letter dated December 19, 2011 (Reference 6) and Entergy Letter dated February 27, 2012 (Reference 8).

Discussion:

Entergy Letter dated February 5, 2005 (Reference 1), Entergy Operations, Inc. (Entergy) made the following commitment:

Entergy Operations, Inc (Entergy) is currently an active participant in the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) research initiatives on aging related degradation of reactor vessel internal components. Entergy commits to:

a. continue its active participation in the MRP initiative to determine appropriate reactor vessel internals degradation management programs,

b. evaluate the recommendations resulting from this initiative and implement a reactor vessel internals degradation management program applicable to Waterford 3,

c. incorporate the resulting reactor vessel internals inspections into the Waterford 3 augmented inspection plan as appropriate.

In addition, as requested by the NRC, a description of the program, including the inspection plan, will be submitted to the NRC for review and approval. The submittal date will be within 24 months after the final EPRI MRP recommendations are issued or within five years from the date of issuance of the uprated license, whichever comes first.

Entergy Letter dated April 7, 2010 (Reference 2) notified the NRC of a change to the commitment of record made in Reference 1. The fundamental change affected the schedule for submitting a description of a reactor vessel internals degradation management program to the NRC for review and approval. The revised commitment required submittal of the plan by December 31, 2011.

Entergy Letter dated December 19, 2011 (Reference 6) notified the NRC of a change to the commitment of record made in Reference 2. The fundamental change affects the schedule for submitting a description of a reactor vessel internals degradation management program, also

W3F1 -2013-0070 Page 3 referred to as an AMP, to the NRC for review and approval. The commitment of record required submittal of the plan by December 31, 2011. The revised commitment required submittal of the plan 24 months prior to entering the period of extended operation associated with its License Renewal Application as provided for in NRC Regulatory Issue Summary 2011-07 License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management (Reference 5).

Entergy Letter dated February 27, 2012 (Reference 8) notified the NRC of a change to the commitment made in Reference 6. The fundamental change affects the schedule for submitting a description of a Management Program (AMP), to the NRC for review and approval. The commitment of record required submittal of the plan 24 months prior to entering the period of extended operation associated with its License Renewal Application as provided for in NRC Regulatory Issue Summary 2011-07 License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management (Reference 5). The revised commitment required submittal of the plan to the NRC for review and approval within twenty-four months following issuance of MRP-227-A (that is, no later than December 16, 2013) which is in accordance with the EPRI/MRP Initiative Needed requirement (Reference 7).

This letter contains no new commitments.

If you have any questions or require additional information, please contact John Jarrell, Regulatory Assurance Manager at (504) 739-6685.

Sincerely JPJ/RJP

)rd 3 Reactor Vessel Internals Aging Management Program

W3F1-2013-0070 Page 4 cc: Mr. Marc L. Dapas RidsRgn4MailCenter@nrc.gov Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Marlone.Davis@nrc.gov Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Kaly.Kalyanam@nrc.gov Attn: Mr. N. Kalyanam Mail Stop O-07D1 Washington, DC 20555-0001

Attachment To W3FI-2013-0070 Waterford 3 Reactor Vessel Internals Aging Management Program

Report No. 1001328.401 Revision 1 Project No. 1001328 December 2013 Waterford Steam Electric Station Unit 3 Reactor Vessel Internals Aging Management Document Preparedfor."

Entergy OperationsInc.

ContractOrder No. 10309874 Preparedby:

StructuralIntegrity Associates, Inc.

San Jose, California Prepared by: Date: 12/12/2013 Chris S. Lohse, P.E.

Reviewed by: Date: 12/12/2013 Timothy J. Griesbach Approved by: Date: 12/12/2013 Timothy J. Griesbach

REVISION CONTROL SHEET Document Number: 1001328.401

Title:

Waterford Steam Electric Station Unit 3 Reactor Vessel Internals Aging Management Document Client:

SI Project Number: 1001328 Quality Program: [ Nuclear LI Commercial Section Pages Revision Date Comments 1 8-12 0 8/31/2011 Initial Issue 2 13- 16 3 17-30 4 31-33 5 34-36 6 37 7 38-39 8 40 1 9- 13 1 12/12/2013 Revised to incorporate the changes due to 2 14-17 MRP-227-A 3 18-31 4 32-34 5 35-37 6 38 7 39-41 8 42-43 9 44

WATERFORD STEAM AND ELECTRIC STATION Reactor Vessel Internals Aging Management Document December 2013 Document XXXXX Revision 1 Quality Class III 4Z-/f -/3 Date (3

Dte1 Date Reactor Coolant Sy m (RCS)

Materials Degrad ion Management Program (MDMP) Manager

/4

Record of Revisions Rev. Date Description/Affected Pages 0 8/31/2011 Initial Issue 1 12/12/2013 Revised to incorporate the changes due to MRP-227-A

Table of Contents SECTION PAGE LIST OF ACRONYMS ........................................................................................................ 7

1.0 INTRODUCTION

..................................................................................................... 9 1.1 O bjective ....................................................................................................................... 9 1.2 B ackground ................................................................................................................... 9 1.3 Responsib ilities ........................................................................................................... 12 2.0 DISCUSSION ................................................................................................................ 14 2.1 Mechanisms of Age-Related Degradation in PWR Internals ................................ 14 2.2 Aging Management Strategy ................................................................................. 16 3.0 WSES REACTOR VESSEL INTERNALS DESIGN [6] .................................... 18 3.1 Upper Internals Assembly ..................................................................................... 18 3.2 C ore Support B arrel ................................................................................................. 18 3.3 Low er Support A ssem bly ........................................................................................ 19 3.4 Core Shroud Assembly .......................................................................................... 19 3.5 Control Element Assembly Shroud Assemblies .................................................... 19 3.6 In-Core Instrumentation Support System ............................................................... 19 3.7 D esign Modifications ............................................................................................ 29 3.8 Description of Existing Aging Management Documents ....................................... 29 4.0 PROGRAM DESCRIPTION ................................................................................. 32 4.1 Preventive A ctions ................................................................................................. 32 4.2 O perational Experience .......................................................................................... 32 4.3 Component Inspection and Evaluation Overview .................................................. 32 4.4 Inspection and Evaluation Requirements for Primary Components ...................... 33 4.6 Inspection of Existing Plant Components ............................................................ 34 4.7 Examination Systems (MRP-227-A Section 7.4) .................................................. 34 4.8 Inspection Schedule .............................................................................................. 34 5.0 EXAMINATION ACCEPTANCE AND EXPANSION CRITERIA .................. 35 5.1 Examination Acceptance Criteria ........................................................................... 35 5.2 EXPANSION CRITERIA ....................................................................................... 36 5.3 EVALUATION, REPAIR AND REPLACEMENT STRATEGY (MRP-227-A SECTIONS 7.5, 7.6, AND 7.7) ............................................................................... 36 6.0 OPERATING EXPERIENCE AND ADDITIONAL CONSIDERATIONS ........... 38 6.1 Internal and External .............................................................................................. 38 7.0 RESPONSES TO THE NRC SAFETY EVALUATION REPORT APPLICANT/LICENSEE ACTION ITEMS ....................................................... 39 7.1 SER Section 4.2.1, Applicant/Licensee Action Item 1: ........................................ 39 7.2 SER Section 4.2.2, Applicant/Licensee Action Item 2: ........................................ 40

7.3 SER Section 4.2.3, Applicant/Licensee Action Item 3: ........................................ 40 7.4 SER Section 4.2.4, Applicant/Licensee Action Item 4: ........................................ 40 7.5 SER Section 4.2.5, Applicant/Licensee Action Item 5: ........................................ 40 7.6 SER Section 4.2.6, Applicant/Licensee Action Item 6: ........................................ 41 7.7 SER Section 4.2.7, Applicant/Licensee Action Item 7: ........................................ 41 7.8 SER Section 4.2.8, Applicant/Licensee Action Item 8: ........................................ 41

8.0 REFERENCES

........................................................................................................ 42 9.0 LIST OF DRAW INGS ............................................................................................. 44 ATTACHMENT A SIGNIFICANT INTERNAL AND EXTERNAL OPERATING EXPERIENCE REVIEW AND EVALUATION ................................................................ 72 ATTACHMENT B OPEN ACTION TRACKING LOG ..................................................... 73 ATTACHMENT C WSES REACTOR VESSEL INTERNALS DRAWINGS ................ 74

List of Tables TABLE NO. PAGE Table 1. Key Elements of the Reactor Vessel Internals Aging Management Document: ............ 11 Table 2. C-E Plants Primary Components Applicable to WSES [3] ...................................... 45 Table 3. C-E Plants Expansion Components Applicable to WSES [3] .................................. 52 Table 4. C-E Plants Existing Program Components Applicable to WSES [3] ....................... 56 Table 5. Inspection Schedule for WSES Primary Components per MRP-227-A ................... 57 Table 6. Inspection Schedule for WSES Existing Program Components Listed in M RP -227-A ............................................................................................................................ 58 Table 7.Section XI 10 Year ISI Examinations of BN2 and BN3 Internals Components for W S E S [14] ............................................................................................................................. 59 Table 8. C-E Plants Examination Acceptance and Expansion Criteria Applicable to W SE S [3] ............................................................................................................................... 63 Table 9. WSES Inspection Plan Summary Table ................................................................... 67

List of Figures FIGURE NO. PAGE Figure 1. Combustion Engineering Vessel and Internals Arrangement .................................. 21 Figure 2. Overview of Typical C-E Internals .......................................................................... 22 Figure 3. Core Shroud A ssem bly ............................................................................................ 23 Figure 4. Welded Core Shroud with Potential Crack and Void Swelling Locations ............... 24 Figure 5. Locations of Potential Separation between Core Shroud Sections Caused by Swelling Induced Warping of Thick Flange Plates in C-E Welded Core Shroud A ssem bled in Stacked Sections ........................................................................................ 25 Figure 6. Typical C-E Core Support Barrel Structure ........................................................... 26 Figure 7. Combustion Engineering Instrument Tube and Connected CEA Shroud Tube ..... 27 Figure 8. Lower Core Support Structure .................................................................................. 28 Figure 9. WSES Reactor Vessel Internals Inspection Plan ..................................................... 71

LIST OF ACRONYMS AMD Aging Management Document AMP Aging Management Program ARDM Age-related degradation mechanism ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel CASS Cast austenitic stainless steel C-E Combustion Engineering CEOG Combustion Engineering Owners Group CFR Code of Federal Regulations CLB Current licensing basis EFPY Effective full power years EPRI Electric Power Research Institute EVT Enhanced visual testing (visual NDE method indicated as EVT-I)

FMECA Failure modes, effects, and criticality analysis GALL Generic Aging Lessons Learned I&E Inspection and Evaluation IASCC Irradiation Assisted Stress Corrosion Cracking ICI In-Core Instrumentation ISI Inservice Inspection LRA License Renewal Application MRP Materials Reliability Program NDE Nondestructive Examination NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OE Operating Experience PWR Pressurized Water Reactor PWROG Pressurized Water Reactor Owners Group RCA Reactor Coolant System RFO Refueling Outage RV Reactor Vessel RVI Reactor Vessel Internals SCC Stress Corrosion Cracking SER Safety Evaluation Report SS Stainless Steel TLAA Time-limited Aging Analysis TS Technical Specifications UT Ultrasonic Testing UGS Upper Guide Structure

VT Visual Testing WSES Waterford Steam Electric Station

1.0 INTRODUCTION

1.1 Objective This program document describes the potential aging concerns in the reactor vessel internals (RVI) and implements the industry recommended guidance for managing these aging concerns at the Waterford Steam and Electric Station (WSES). This program document coordinates with the existing ASME Section XI inservice inspection (ISO program and supplements that program with augmented examinations for managing the potential aging effects. This program document establishes appropriate monitoring and inspection programs to maintain the reactor vessel internals functionality; the strategy is to assure nuclear safety and plant reliability. This document will provide assurance that WSES operations will continue to be conducted in accordance with the current licensing bases for the reactor vessel internals, and it will provide the technical basis for managing the time-limited aging concerns for the duration of plant life. This document identifies the internals components that must be considered for aging management review. The program plan supports the NEI 03-08 Materials Initiative Process [1], the NEI 03-08 Guideline for the Management of Materials Issues [2], and the EPRI Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A) [3]. Later revisions of the MRP-227 guideline will also be incorporated if they affect this program.

1.2 Background The reactor vessel internals were constructed in accordance with the ASME Boiler and Pressure Vessel Code, Sections I1, III, and XI, as applicable. The WSES design Code is the 1971 version of the ASME Section III with Addenda through Summer of 1971. The reactor internals assembly is a part of the reactor coolant system (RCS). The reactor internals are long-lived passive structural components designed to support the functions of RCS core cooling, control element assembly (CEA) insertion, and integrity of the fuel and pressure vessel boundary. The core support structures provide support and restraint of the core. Static (deadweight and mechanical) loads from the assembled components, fuel assemblies, and dynamic loads (hydraulic flow, flow-induced vibration, thermal expansion, and seismic) and LOCA loads are all carried by the core support assembly. In addition to core support, the internals assemblies provide a flaw boundary to direct coolant flow from the cold leg inlet nozzles, down the annulus to the lower plenum, past the core, into the upper plenum region above the core, and out the outlet nozzles to the hot leg piping.

Industry experience and research has shown that active degradation mechanisms may be present that could affect the ability of the internals components to perform their design functions.

Because of this, industry groups such as EPRI and other PWR Owners Groups began an effort to investigate these aging mechanisms, examine the materials of construction, consider the individual plant designs and operating conditions, and determine the internals components that may be susceptible to degradation and could potentially lead to loss of function.

To manage these aging concerns, the EPRI Materials Reliability Program (MRP) first published the MRP-227 guidelines document in December 2008 with an NRC approved version, MRP-227-A, issued in December 2011, which contained "Mandatory" and "Needed" actions under the NEI 03-08 Materials Initiative [2]. Implementation of this Reactor Vessel Internals Aging Management Document fulfills MRP-227-A Section 7.2 [3] requirements for WSES. In addition, Entergy actively participates in the PWR Owners Group Materials Subcommittee and the EPRI MRP. These industry groups actively manage generic work with a focus on improving plant performance and providing an effective interface with the NRC. Best practices and lessons learned are shared and discussed among members. Entergy will maintain active participation in these industry groups.

The following table (Table 1) outlines key elements of the Reactor Vessel Internals Aging Management Document, and provides reference to where additional information can be found.

Table 1. Key Elements of the Reactor Vessel Internals Aging Management Document:

Plan Attribute Approach and supplemental information Scope of Program The Reactor Vessel Internals Aging Management Document Manager is responsible for implementation of this program. Supplemental inspections of RV internals are described in N4RP-227-A [311. Additional actions and long range plans for aging management of internals are described within this document.

2 Preventive Measures Preventive measures are described in Section 4. 1.

3 Parameters Monitored Additional monitoring parameters may be needed, such as cycle counting, to assure that the design basis usage factor is not exceeded for core support structures.

4 Inspection Plan for The WSES ASME Section XI [4] ISI program for B-N-2 and B-N-3 internals Detection of Aging components, and the additional locations identified in MRP-227-A [3], form Effects the inspection plan for detection and monitoring of aging effects in the RV internals.

5 Inspection Program for This program, in combination with the ASME Section XI [4] ISI program, Monitoring and Trending provides direction for inspections required to support continued RV internals component reliability.

6 Acceptance Criteria Acceptance criteria used in the RV Internals Aging Management Document shall be based on the most appropriate ASME Section XI [4] criteria as described in Section 5.1. Where specific industry criteria are developed, those criteria will be incorporated into this program document. Reference 5 should be considered whenever developing plant specific Acceptance Criteria.

7 Corrective Actions Components with identified relevant conditions shall be dispositioned as described in Section 5.3. The disposition can include a supplementary examination to further characterize the relevant condition, an engineering evaluation to show that the component is capable of continued operation with a known relevant condition until the next planned inspection, or repair/replacement to remediate the relevant condition. Additional inspections of expansion category components may also be required.

8 Confirmation Process The RV Internals Aging Management Document Manager shall perform a and Self Assessment program assessment in accordance with EN-DC-202, Rev. 5 [1].

9 Administrative Controls This program is a support program of EN-DC-202, Rev. 5 [1].

10 Operating Experience Operating experience gained through professional contacts and Industry groups such as the EPRI MRP, the PWROG, INPO, WANO and International Sites shall be incorporated into this program document in a timeframe consistent with the significance.

1.3 Responsibilities The Reactor Vessel Internals Program Manager has overall responsibility for the development and implementation of the Reactor Vessel Internals aging management plan. The responsibilities for implementing the NEI 03-08 Materials Initiative Process are described in Reference 1.

The Reactor Vessel Internals Program Manager is responsible for:

• Overall development of the RVI aging management plan,

" Administering and overseeing the implementation of the RVI aging management plan,

• Ensuring that regulatory requirements related to inspection activities, if any, are met and incorporated into the plan,

• Communicating with senior management on periodic updates to the plan,

" Planning control and implementation of the RVI aging management plan,

• Maintaining the RVI aging management plan to incorporate changes and updates based on new knowledge and experience gained,

• Reviewing and approving industry and vendor programs related to RVI aging management activities,

• Processing of any deviations taken from IP guidelines in accordance with NEI 03-08

[2] requirements,

" Ensure prompt notification of the RCS Materials Degradation Management Program Manager whenever an issue or indication of potential generic Industry significance is identified,

" Participating in the planning and implementation of inspections of the internals, and

• Participating in the industry groups such as the PWROG, MRP-ITG, etc.

The ISI Engineer is responsible for:

" Planning and implementing inspections required by Section XI B-N-3 [4], the supplemental inspections identified in this program plan, and any other plant-specific commitments for inspection for managing aging of the internals,

• Providing the NDE services,

• Reviewing and approving the vendor NDE procedures and personnel qualifications,

" Providing direction and oversight of contracted NDE activities,

" Participating in industry groups such as PDI, EPRI Inspection Working Group, etc.

2.0 Discussion 2.1 Mechanisms of Age-Related Degradation in PWR Internals The EPRI MRP program considered all the potential aging mechanisms that could affect PWR internals for the long term. Of particular concern are those aging mechanisms that could have an impact on component functionality. The age-related degradation mechanisms used for the screening of the PWR internals for susceptibility were as follows:

2.1.1 Stress Corrosion Cracking Stress Corrosion Cracking (SCC) refers to local, non-ductile cracking of a material due to a combination of tensile stress, environment, and metallurgical properties. The actual mechanism that causes SCC involves a complex interaction of environmental and metallurgical factors. The aging effect is cracking.

2.1.2 Irradiation-Assisted Stress Corrosion Cracking Irradiation-assisted stress corrosion cracking (IASCC) is a unique form of SCC that occurs only in highly-irradiated components. The aging effect is cracking.

2.1.3 Wear Wear is cause by the relative motion between adjacent surfaces, with the extent determined by the relative properties of the adjacent materials and their surface condition. The aging effect is loss of material.

2.1.4 Fatigue Fatigue is defined as the structural deterioration that can occur as the result of repeated stress/strain cycles caused by fluctuating loads and temperatures. After repeated cyclic loading of sufficient magnitude, microstructural damage can accumulate, leading to macroscopic crack initiation at the most highly affected locations. Subsequent mechanical or thermal cyclic loading can lead to growth of the initiated crack. Corrosion fatigue is included in the degradation description.

Low-cycle fatigue is defined as cyclic loads that cause significant plastic strain in the highly stressed regions, where the number of applied cycles is increased to the point where the crack eventually initiates. When the cyclic loads are such that significant plastic deformation does not occur in the highly stressed regions, but the loads are of such increased frequency that a fatigue crack eventually initiates, the damage accumulated is said to have been caused by high-cycle fatigue. The aging effects of low-cycle fatigue and high-cycle fatigue are additive.

Fatigue crack initiation and growth resistance is governed by a number of material, structural and environmental factors, such as stress range, loading frequency, surface condition and presence of

deleterious chemical species. Cracks typically initiate at local geometric stress concentrations, such as notches, surface defects, and structural discontinuities. The aging effect is cracking.

2.1.5 Thermal Aging Embrittlement Thermal aging embrittlement is the exposure of delta ferrite within cast austenitic stainless steel (CASS) and precipitation-hardenable (PH) stainless steel to high inservice temperatures, which can result in an increase in tensile strength, a decrease in ductility, and a loss of fracture toughness. Some degree of thermal aging embrittlement can also occur at normal operating temperatures for CASS and PH stainless steel internals. CASS components have a duplex microstructure and are particularly susceptible to this mechanism. While the initial aging effect is loss of ductility and toughness, unstable crack extension is the eventual aging effect if a crack is present and the local applied stress intensity exceeds the reduced firacture toughness.

2.1.6 Irradiation Embrittlement Irradiation embrittlement is also referred to as neutron embrittlement. When exposed to high energy neutrons, the mechanical properties of stainless steel and nickel-base alloys can be changed. Such changes in mechanical properties include increasing yield strength, increasing ultimate strength, decreasing ductility, and a loss of fracture toughness. The irradiation embrittlement aging mechanism is a function of both temperature and neutron fluence. While the initial aging effect is loss of ductility and toughness, unstable crack extension is the eventual aging effect if a crack is present and the local applied stress intensity factor exceeds the reduced fracture toughness.

2.1.7 Void Swelling and Irradiation Growth Void swelling is defined as a gradual increase in the volume of a component caused by formation of microscopic cavities in the material. These cavities result from the nucleation and growth of clusters of irradiation produced vacancies. Helium produced by nuclear transmutations can have a significant impact on the nucleation and growth of cavities in the material. Void swelling may produce dimensional changes that exceed the tolerances on a component. Strain gradients produced by differential swelling in the system may produce significant stresses. Severe swelling (>5% by volume) has been correlated with extremely low fracture toughness values.

Also included in this mechanism is irradiation growth of anisotropic materials, which is known to cause significant dimensional changes in in-core instrumentation tubes fabricated from zirconium alloys. While the initial aging effect is dimensional change and distortion, severe void swelling may result in cracking under stress.

2.1.8 Thermal and Irradiation-Enhanced Stress Relaxation or Creep The loss of preload aging effect can be caused by the aging mechanisms of stress relaxation or creep. Thermal stress relaxation (or, primary creep) is defined as the unloading of preloaded components due to long-term exposure to elevated temperatures, such as seen in PWR internals.

Stress relaxation occurs under conditions of constant strain where part of the elastic strain is replaced with plastic strain. Available data show that thermal stress relaxation appears to reach saturation in a short time (<100 hours) at PWR internals temperatures.

Creep (or more precisely, secondary creep) is a slow, time and temperature dependent, plastic defornation of materials that can occur when subjected to stress levels below the yield strength (elastic limit). Creep occurs at elevated temperatures where continuous deformation takes place under constant strain. Secondary creep in austenitic stainless steels is associated with temperatures higher than those relevant to PWR internals even after taking into account gamma heating. However, irradiation-enhanced creep (or more simply, irradiation creep) or irradiation-enhanced stress relaxation (ISR) is a thermal process that depends on the neutron fluence and stress; and, it can also be affected by void swelling, should it occur. The aging effect is a loss of mechanical closure integrity (or, preload) that can lead to unanticipated loading which, in turn, may eventually cause subsequent degradation by fatigue or wear and result in cracking.

2.2 Aging Management Strategy The MRP-227-A [3] guidelines define a supplemental inspection program for managing aging effects and to develop this aging management document for WSES. The EPRI MRP Reactor Internals Focus Group developed the MRP-227-A Guidelines to support the demonstration of continued functionality, with requirements for inspections to detect the effects of aging degradation with requirements for the evaluation of those aging effects. The aging management strategy used to develop MRP-227-A combined the results of functionality assessment with component accessibility, operating experience, existing evaluations, and prior examination results to determine the appropriate aging management methodology, baseline examination timing, and the need and timing of subsequent inspections and identified the components and locations for supplemental examination by categorization. A description of the categorization process used to develop this program document is given below.

In accordance with the MRP-227-A I&E Guidelines [3], this inspection strategy consists of the following:

• Selection of items for inspection,

• Selection of the type of examination appropriate for each degradation mechanism, Specification of the required level of examination qualification, Schedule of first inspection and frequency of subsequent inspections, Requirements for sampling and coverage,

• Requirements for expansion of scope if unanticipated indications are found,

• Inspection acceptance criteria,

• Methods for evaluating examination results not meeting the acceptance criteria,

• Updating the program based on industry-wide results; and

• Contingency measures to repair, replace, or mitigate.

The specifics of the WSES reactor vessel internals design are described in Section 3.0

3.0 WSES Reactor Vessel Internals Design [61 The WSES Unit 3 was designed by Combustion Engineering (C-E) with a welded core shroud assembly and is made up of two vertical sections that are connected with a circumferential weld.

The components of the reactor vessel internals are divided into smaller sub-assemblies consisting of the upper guide structure, CEA shroud assemblies, core support barrel assembly, core shroud assembly, lower internals assembly, and in-core instrumentation system. The arrangement of a typical C-E design vessel and internals package is shown in Figure 1.

The C-E designed PWR internals consist of three major structural assemblies, plus three other sets of major components. The three major assemblies are the: (1) upper internals assembly, (2) core support barrel assembly, and (3) lower internals assembly. In addition, the three other sets if major components are in the control element assembly shroud assemblies, core shroud assembly, and in-core instrumentation support system. The overview of the C-E designed PWR internals is shown in Figure 2.

3.1 Upper Internals Assembly The upper internals assembly is located above the reactor core, within the core support barrel assembly, and is removed during refueling as a single component in order to provide access to the fuel assemblies. The upper internals assembly consists of the upper guide structure support plate, the fuel assembly alignment plate, the control element assembly shroud assemblies, the upper guide structure grid assembly, the upper guide structure cylinder, the in-core instrumentation support system and the hold-down ring (or expansion compensation ring). The functions of the upper internals assembly are to provide alignment and support to the fuel assemblies, to maintain control element assembly shroud spacing, to prevent movement of the fuel assemblies in the case of a severe accident condition, and to protect the control rods from cross-flow effects in the upper plenum. The flange on the upper end of the upper internals assembly rests on the core support barrel.

3.2 Core Support Barrel The core support barrel assembly consists of the core support barrel, the core support barrel upper flange, core support barrel alignment keys, and the core support barrel snubbers. The core support barrel is a cylinder which contains the core and other internals. Its function is to resist static loads from the fuel assemblies and other internals, and dynamic loads from normal operating hydraulic flaw, seismic events, and loss-of-coolant-accident (LOCA) events. The core support barrel also supports the lower internals assembly and its core support plate, upon which the fuel assemblies rest. The core support barrel upper flange is a thick ring that supports and suspends the core support barrel from a ledge on the reactor vessel.

3.3 Lower Support Assembly The lower internals assembly consists of the core support plate, the fuel alignment pins, the core support columns, the in-core instrumentation (ICI) support system, and the lower support structure beam assemblies. The core support plate functions are to position and support the reactor core, and to provide control of reactor coolant flow into each fuel assembly. The core support plate transmits the weight of the core to the core support barrel by means of the vertical core support columns, an annular skirt, and the lower support beams.

3.4 Core Shroud Assembly The core shroud assembly is located within the core support barrel and directly below the upper internals assembly. The core shroud assembly is attached to the core support barrel by threaded fasteners for those internals with a bolted core shroud and top-mounted ICI. The core shroud assembly is attached to the core support plate - an element of the lower internals assembly - by welds. The shroud assembly is attached to the lower internals assembly cylinder by welding.

The core shroud assembly functions are to provide a boundary between reactor coolant flow on the outside of the core support barrel and the reactor coolant flow through the fuel assemblies, to limit the amount of coolant bypass flow, and to reduce the lateral motion of the fuel assemblies.

3.5 Control Element Assembly Shroud Assemblies The control element assembly shroud assemblies consist of control element assembly shrouds, the control element assembly shroud bolts, and the control element assembly shroud extension shaft guides. The shroud tubes protect the control rods from cross-flow effects in the upper plenum. The bottom part of the shrouds is bolted at their lower end to the fuel assembly alignment plate. The extension shaft guides also protect the control rods from cross-flow effects in the upper plenum, and provide lateral support and alignment of the control element assembly extension shafts during refueling operations. The control element drive mechanisms are positioned on the reactor vessel closure head and are coupled to the control element assemblies by the control element assembly extension shafts. Control element assembly shroud assemblies are attached to the upper guide structure support plate by tie rods.

3.6 In-Core Instrumentation Support System The in-core instrumentation support system consists of in-core instrumentation guide tubes and components which provide support to the in-core instrumentation. The in-core instrumentation is inserted through the reactor vessel head through a nozzle into a guide tube. The guide tubes interface with the thimble support plate, which is perforated to fit over the control element assembly extension shaft guides, with a connection to the upper guide structure support plate.

ICI thimble tube assemblies extend downward from a flanged connection at the thimble support

• plate (in the original design) through the fuel alignment plate and into the reactor core. The upper portion of the ICI thimble tube exists between the thimble support plate and fuel alignment

plate, while the lower ICI thimble tube is the zirconium alloy portion that extends into the fuel assemblies.

The vessel internals drawings for WSES are provided in Attachment C. The WSES drawings used in this aging management document are listed as Drawings I through 4 in Section 9.0.

CEDM NOZZLE INSTRUMENTATIO NOZZLE CONTROL ELEMENT ASSEMBLY FULLY WITHDRAWN HOLODOWN RING ALIGNMENT KEY UPPER GUIDE STRUCTURE 30" I.. INLET NOZZLE FUEL ALIGNMENT 42" I.D. OUTLET PLATE NOZZLE CORE SUPPORT BARREL FUEL ASSEMBLY CORE SHROUD SURVEILANCE HOLDER CORE SUPPORT PLATE LOWER SUPPORT STRUCTURE SNUBBER FLOW SKIRT CORE STOP Figure 1. Combustion Engineering Vessel and Internals Arrangement

CORE SUPPORT BARREL CORE SHROWU LOWR SUPPORT ASSEB4LY Figure 2. Overview of Typical C-E Internals

Flanges -

Circumferential Weld Horizontal Stiffeners Figure 3. Core Shroud Assembly

Weld hxedau pmftkvydya1cted by swdhl"nhaduma stilbw I

aboe IAC dvewM Wd bdam patWý dfetd by sWaftkin bwxtxatt sEffwn Figure 4. Welded Core Shroud with Potential Crack and Void Swelling Locations

Figure 5. Locations of Potential Separation between Core Shroud Sections Caused by Swelling Induced Warping of Thick Flange Plates in C-E Welded Core Shroud Assembled in Stacked Sections

Flange Weld 0 Axial Weld Urper Core Barrel to Lower Core Barrel Circumferential Weld Lower Barrel Axial Weld Lower Barrel Circumferential Weld Lower Barrel Axial Weld Core Darrel to 3upport Plate Weld Figure 6. Typical C-E Core Support Barrel Structure

HOLDOOW4J RING

-i VGS SUIPPORT ASS~MBL' CEA SHROUDS 70 FUEL AIJQNME1JT-_

PLATE Figure 7. Combustion Engineering Instrument Tube and Connected CEA Shroud Tube

Figure 8. Lower Core Support Structure 3.7 Design Modifications 3.7.1 Instrument Tube Assemblies Design modifications were performed at WSES to reduce the flow induced wear and fretting of the Instrument Tube Assemblies. Operating experience has shown that the Instrument Tube Assemblies may be susceptible to flow induced vibration causing wear of the instrument tubes.

In addition, the fixed incore instrument (ICI) thimble tubes were observed to have irradiation growth of the zirconium materials, and these had to be replaced with shorter designed tubes. The Instrument Tube Assemblies were replaced in 2006 with a modified design using shortened assemblies to offset growth over time [7]. These assemblies are inspected periodically to monitor for wear and irradiation growth of the components (see Table 9).

3.8 Description of Existing Aging Management Documents The overall strategy for managing the effects of aging in the reactor vessel internals components at WSES is supported by the following existing programs:

• Reactor Vessel Internals Inspection Program per ASME Section XI [4]

• Water Chemistry Program [9] as described in Reference [1]

• Industry Programs for Managing Aging of Internals These are established programs that support the aging management of RCS components in addition to the RVI components. Although affiliated with and supporting the aging management of reactor vessel internals, these programs will continue to be managed under the existing structure.

3.8.1 ASME Section XI Inservice Inspection Program of Vessel Internals The ASME Section XI [4] Inservice Inspection Program is an existing program that facilitates inspections to identify and correct degradation in Class 1, 2 and 3 piping components, their supports and integral attachments, including welds, pump casings, valve bodies, pressure retaining bolting, piping/component supports and reactor head closure studs. These are identified in ASME Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," [4] or commitments requiring augmented inservice inspections. This program is in accordance with 10CFR50.55a [8]. The original Section XI inspection plan was based on knowledge at the time of original license and an expected service life of 40-years.

MRP-227-A [3] is designed to supplement original inspection requirements to address aging beyond the original design life.

The categories applying to the vessel internals include: 1) the interior attachments beyond the beltline (B-N-2) and (2) core support structures (B-N-3). The core support structures shall be removed from the reactor vessel for examination during the vessel ISI examination.

3.8.2 Water Chemistry Program The water chemistry program is credited for managing aging effects by controlling the environment to which internal surfaces of systems and components are exposed. Such effects include the following:

• Loss of material due to general, pitting and crevice corrosion, 0 Cracking due to SCC, Other materials degradation effects (e.g., steam generator tube degradation caused by denting, intergranular attack, and outer diameter stress corrosion cracking)

The aging effects are minimized by controlling the chemical species that cause the underlying mechanisms that produce them. The water chemistry program provides assurance that an elevated level of contaminants and, where applicable, oxygen does not exist in the system and components covered by the program, thus minimizing the occurrence of aging effects, and maintaining each component's ability to perform the intended functions.

Waterford has recently installed zinc injection as part of the RCS water chemistry program. This is monitored by the WSES Water Chemistry Program [9] in accordance to the EPRI PWR Primary Water Chemistry Guidelines [13].

3.8.3 Changes in Plant Operation Entergy Operations, Inc. submitted a request to the US Nuclear Regulatory Commission, a request for changes to the Waterford Steam Electric Station, Unit 3, Operating License and Technical Specifications (TSs). The requested changes were pertaining to increase the power level from 3390 Megawatts thermal (MWt) to 3441 MWt, an approximately 1.5% increase. This increase was based on the installation of a leading edge flow meter system in the feed water pipe from the main feed water header, which reduces the flow and temperature uncertainties, and the revision of Appendix K to Title 10, Code of Federal Regulations, part 50 (10 CFR Part 50),

which no longer requires a 2% flow uncertainty for the loss-of-coolant accident (LOCA) analysis.

In a letter dated March 29, 2002, the Staff approved the amendment changes to the WSES operating license and TS associated with the increase in the power level from 3390 MWt to 3441 MWt.

3.8.4 Industry Programs Entergy actively participates in the EPRI Materials Reliability Program and the PWR Owners Group that provides information on specific issues related to degradation of C-E designed reactor vessel internals.

4.0 Program Description Management of component aging effects includes actions to prevent or control degradation due to aging effects, review of operational experience to better understand the potential for degradation to occur, inspections to detect the onset of aging effects in susceptible components, protocols for evaluation and remediation of degradation due to aging, and procedures to ensure component aging is managed in a coordinated program.

4.1 Preventive Actions WSES is currently managing water chemistry to mitigate SCC initiation in nickel alloys. This is addressed by the WSES Water Chemistry Program [9].

4.2 Operational Experience Operational experience related to degradation of reactor internal components covered in this aging management document will be reviewed on a periodic basis. This review should include both domestic and international experience. A periodic review of significant OE review is performed and documented (see Attachment A) including reference to any consequential actions.

Worldwide operation experience through 2009 is summarized in Reference 10. Results of reactor internal components inspected in accordance with MRP-227-A will be summarized in the biannual MRP Inspection Data Survey, MRP-219 [11].

4.3 Component Inspection and Evaluation Overview A description of Aging Management Document categorization and the steps used to develop this program document are given below.

This program summarizes the guidance of the MRP I&E guidelines necessary to understand implementation but does not duplicate the full discussion of the technical bases. MRP-227-A [3]

and its supporting documents should be consulted for a complete description of the technical bases of the program.

MRP-227-A [3] establishes four groups of reactor internals components with respect to inspection requirements: Primary, Expansion, Existing Programs, and No Additional Measures, as summarized below.

Primary: Those PWR internals components that are highly susceptible to the effects of at least one of the eight aging mechanisms were placed in the Primary group. The aging management requirements are described in the I&E Guidelines and are needed to ensure functionality of Primary components. The Primary group also includes components which have low or moderate susceptibility to a specific aging degradation effect, but for which no highly susceptible component exists or for which no highly susceptible component is accessible.

" Expansion: Those PWR internals components that are highly or moderately susceptible to the effects of at least one of the eight aging mechanisms, but for which functionality assessment has shown a degree of tolerance to those effects, were placed in the Expansion group. The schedule for implementation of aging management requirements for Expansion components will depend on the findings from the examinations of the Primary components at individual plants.

" Existing Programs: Those PWR internals components that are susceptible to the effects of at least one of the eight aging mechanisms and for which existing program elements are capable of managing those effects, were placed in the Existing Programs group.

" No Additional Measures: Those PWR internals components for which the effects of all eight aging mechanisms are below the screening criteria, and which were placed in Category A by the initial screening step were placed in the No Additional Measures group. Through the functionality assessment process, some of the PWR internals components other than Category A components were also placed in No Additional Measures. No further action is required for managing the aging of the No Additional Measures components, other than the continuation of any existing plant requirements that apply to these components. Many of the No Additional Measures components are not core support structures, and therefore may not be covered by a program element such as the ASME B&PV Code, or Section XI periodic in-service examination [4].

The inspections required for Primary and Expansion components were selected from existing, visual, surface, and volumetric examination methodologies that are applicable and appropriate for the expected degradation effect (e.g., cracking caused by particular mechanisms, loss of material caused by wear). The inspection methodologies include: Visual (VT-3) examinations, Visual (VT-I) examinations, surface examinations, volumetric (specifically, UT) examinations, and physical measurements. MRP-227-A provides detailed justification for the components selected for inspection and the specific examination methodologies selected for each. The MRP-228 report, PWR Internals Inspection Standards [12], provides detailed examination requirements for the components listed.

4.4 Inspection and Evaluation Requirements for Primary Components The inspection requirements for Primary Components of C-E designed plants applicable to WSES from MRP-227-A [3] are provided in Table 2.

4.5 Inspection and Evaluation Requirements for Expansion Components The inspection requirements for Expansion Components of C-E designed plants applicable to WSES from MRP-227-A [3] are provided in Table 3.

4.6 Inspection of Existing Plant Components The list of Existing Plant Components of C-E designed plants applicable to WSES from MRP-227-A [3] are provided in Table 4. This list of components in the current Section XI ISI program for WSES designated as B-N-2 and B-N-3 locations are shown in Table 3 [4].

The reactor vessel inspection plan for WSES is provided in Figure 9. The current ISI program considering existing inspections will be implemented for each inspection interval [14].

Supplemental inspections in accordance with the requirements of NRP-227-A will be scheduled and implemented in accordance with future license renewal commitments. These additional examinations, the methods to be used, and the acceptance and expansion criteria are described below.

The ASME Section XI ISI Inspections and additional augmented inspections [14] identified in this Reactor Vessel Internals (RVI) Aging Management Document will be performed in accordance with the required inspection interval.

4.7 Examination Systems (MRP-227-A Section 7.4)

Equipment, techniques, procedures and personnel used to perform examinations required under this program shall be consistent with the requirements of MRP-228 Section 7.2 [12]. Indications detected during these examinations shall be characterized and reported in accordance with the requirements of MRP-228, Sections 7.3 and 7.4.

4.8 Inspection Schedule The inspection schedule for the WSES RVI primary components is provided in Table 5. The inspection schedule for the existing program components addressed in MRP-227-A is listed in Table 6. The inspection plan summary table for WSES augmented exams per MRP-227-A is given in Table 9.

5.0 Examination Acceptance and Expansion Criteria 5.1 Examination Acceptance Criteria 5.1.1 Visual (VT-3) Examination Visual (VT-3) examination has been determined to be an appropriate NDE method for the detection of general degradation conditions in many of the susceptible components. The ASME Code Section XI, Examination Category B-N-3 [4], provides a set of relevant conditions for the visual (VT-3) examination of removable core support structures in IWB-3520.2. These are:

I. 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%;

4. Wear or mating surface 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%.

For components in the Existing Programs group, these general relevant conditions are sufficient.

However, for components where visual (VT-3) is specified in the Primary or the Expansion group, more specific descriptions of the relevant conditions are provided in Table 8. Typical examples are "fractured material" and "completely separated material." One or more of these specific relevant condition descriptions may be applicable to the Primary and Expansion components listed in Tables 2 and 3. The examination acceptance criteria for components requiring visual (VT-3) examinations is thus the absence of the relevant condition(s) specified in Table 8. The disposition can include a supplementary examination to further characterize the relevant condition, an engineering evaluation to show that the component is capable of continued operation with a known relevant condition, or repair/replacement to remediate the relevant condition.

5.1.2 Visual (VT-1) Examination Visual (VT-1) examination is defined in the ASME Code Section XI as an examination "conducted to detect discontinuities and imperfections on the surface of components, including such conditions as cracks, wear, corrosion, or erosion." For these guidelines VT-I has only been selected to detect distortions as evidenced by small gaps between the upper-to-lower mating surfaces of C-E welded core shroud assembled in two vertical sections. The examination acceptance criterion is thus the absence of the relevant condition of gaps that would be indicative of distortion from void swelling.

5.1.3 Surface Examination Surface ET (eddy current) examinations are specified as an alternative or as a supplement to visual examinations. No specific acceptance criteria for surface (ET) examination of PWR internals locations are provided in the ASME Code Section XI. Since surface ET is employed as a signal-based examination, a technical justification per the Inspection Standard, MRP-228 [12]

provides the basis for detection and length sizing of surface-breaking or near-surface cracks.

The signal-based relevant indication for surface (ET) is thus the same as the relevant condition for enhanced visual (EVT-1) examination. The acceptance criteria for enhanced visual (EVT-1) examinations are therefore applied when this method is used as an alternative or supplement to visual examination.

5.1.4 Volumetric Examination There are no required volumetric examinations required for WSES vessel internals.

Locations for augmented MRP-227-A inspections for the WSES reactor vessel internals are identified in Figure 9.

5.2 Expansion Criteria The criteria for expanding the scope of examination from the Primary components to their linked Expansion components are contained in Table 8.

5.3 Evaluation, Repair and Replacement Strategy (MRP-227-A Sections 7.5, 7.6, and 7.7)

Any condition detected during examinations that do not satisfy the examination acceptance criteria of Section 5.1 shall be entered and dispositioned in the Corrective Action Program.

The options listed below will be considered for disposition of such conditions. Selection of the most appropriate option(s) will be dependent on the nature and location of the indication detected.

1. Supplemental examinations, such as surface examination to supplement a visual (VT-1) examination to further characterize and potentially dispose of a detected condition

2. Engineering evaluation that demonstrate the acceptability of a detected condition;

3. Repair to restore a component with a detected condition to acceptable status; or

4. Replacement of a component.

The methodology used to perform Engineering Evaluations to determine the acceptability of a detected condition (item 2 above) shall be conducted in accordance with NRC approved evaluation methodology. WCAP-17096, Reactor Internals Acceptance Criteria Methodology and Data Requirements [5] is currently under NRC review for this purpose.

5.3.1 Reporting Reporting and documentation of relevant conditions and disposition of findings will be performed consistent with WSES Quality Assurance policies and procedures. A summary report shall be provided to the EPRI MRP Program Manager of all inspections and monitoring, items requiring evaluation, and new repairs. This report shall be provided within 120 days of the completion of the outage during which the activities occur. The MRP reporting template should be used for the report.

Inspection results having potential Industry significance shall be expeditiously reported to the RCS Materials Degradation Program manager for consideration of reporting under the NEI 03-08, Emergent Issue Protocol [2].

5.3.2 Trending and Monitoring Inspection results that exceed recording criteria should be quantified to the extent possible and monitored for changes as determined by the Corrective Action program. Such monitoring actions should be incorporated into inspection procedures, or separately tracked in Attachment B.

6.0 Operating Experience and Additional Considerations 6.1 internal and External Operating Experience should be periodically reviewed and evaluated for applicability to this program document. Evaluation of internal observations and significant external events should be periodically documented in Attachment A.

7.0 Responses to the NRC Safety Evaluation Report Applicant/Licensee Action Items As part of the NRC Final Safety Evaluation of MRP-227 [3], a number of action items and conditions were specified by the staff. Wherever possible, these items have been addressed in the appropriate sections of this document. All NRC action items and conditions not addressed elsewhere in this document are discussed in this section.

7.1 SER Section 4.2.1, Applicant/Licensee Action Item 1:

WSES has assessed its plant design and operating history and has determined that MRP-227-A [3] is applicable to the facility. The assumptions regarding plant design and operating history made in MRP-191 [15] are appropriate for WSES and there are no differences in component inspection at WSES. WSES operated the first 22 effective full power years (EFPY) of operation with a high leakage core loading pattern. The FMECA and functionality analyses were based on the assumption of 30 years of operation with high leakage core loading patterns; therefore, WSES is bounded by the assumption in MRP-191 [15].

Operations at WSES conform to the assumptions in Section 2.4 of MRP-227-A [3].

• WSES operated for 22 effective full power years (EFPY) with high-leakage core patterns, followed by implementation of a low-leakage fuel management strategy for the remaining years of operation;

• WSES operates as a base load unit, and

• No design changes were implemented beyond those identified in general industry guidance or recommended by the vendor (C-E or Westinghouse)

During the review of MRP-227, Rev. 0, the NRC staff questioned the basis for the assumptions used during the scoping, screening, and functionality analyses used to develop the I&E Guidelines. In January and March 2013, meetings were held between EPRI, Westinghouse and NRC to address these concerns for "bounding" assumptions on a fleet basis. Following those meetings, Westinghouse provided the NRC with a Technical Basis Document supporting the assumptions used to bound the fleet in MRP-191 and MRP-227-A.

MRP 2013-025 [16] contains the approach for C-E plants to address the plant applicability for specific concerns by the NRC. The attachment to the letter discusses the generic evaluations that Westinghouse provided to the NRC to address the issue generically for the fleet. The document that Westinghouse provided to the NRC is WCAP-17780-P and contains the information to demonstrate plant-specific applicability of MRP-227-A. For example, a plant-specific determination of the applicability of the assumptions used in developing the sampling inspection strategies in MRP-227-A are to verify that the neutron fluence and heat generation rates are within the limiting threshold values:

0 Active core power density < 110 Watts/cm 3 for C-E designed plants, and V Heat generation figure of merit, F < 68 Watts/cm 3 for C-E designed plants.

A validation of the bounding assumptions for WSES will be confirmed.

7.2 SER Section 4.2.2, Applicant/Licensee Action Item 2:

This licensee action item will be addressed if WSES chooses to pursue license renewal.

7.3 SER Section 4.2.3, Applicant/Licensee Action Item 3:

The SE for MRP-227 [3] requires C-E plants to evaluate whether existing plant-specific programs are adequate to manage the aging effects of(i) thermal shield positioning pins and (2) in-core instrument thimble tubes. WSES does not have a thermal shield, so inspections of the positioning pins are not applicable. The ICI thimble tubes are managed In accordance with Design Change 020701067-6 and 051001333-2 as discussed in Table 9.

7.4 SER Section 4.2.4, Applicant/Licensee Action Item 4:

This action does not apply to C-E designed units.

7.5 SER Section 4.2.5, Applicant/Licensee Action Item 5:

Per the SE for MRP-227 [3], C-E designed plants are required to provide plant-specific acceptance criteria to be applied when performing physical measurements for measuring distortion in the gap between the top and bottom core shroud segments in units with core barrel shrouds assembled in two vertical sections. Figure 5 illustrates the location at the plane where the flanges of the top and bottom core shroud segments meet, and where the potential for flange separation caused by void swelling could occur. The examination requirement for this primary location is visual (VT-1) inspection to detect the relevant condition, which is visible flange separation. No physical measurements are needed unless the relevant condition is detected. If the relevant condition is detected, Table 4-2 of MRP-227-A requires three to five measurements of the extent of that separation from the core side at the core shroud re-entrant comers, along with an evaluation to determine the frequency and method to be used for any additional examinations. A/LAI 5 requires that acceptance criteria for any measured separation be provided on a plant-specific basis. Since the functionality analyses used to identify the effects of void swelling for core shrouds welded from two vertical sections are known to be very conservative, and since those effects after 60 years of conservative operation were shown to be very locally concentrated in the re-entrant comer regions, the acceptance criteria are conditional based upon the results of VT-I examinations during the license renewal period. Therefore, the satisfaction of this licensee action item will be addressed if WSES chooses to pursue license renewal.

7.6 SER Section 4.2.6, Applicant/Licensee Action Item 6:

This action does not apply to the C-E designed units.

7.7 SER Section 4.2.7, Applicant/Licensee Action Item 7:

The SE for MIRP-227 [3] requires the applicants/licensees of C-E reactors to develop plant-specific analyses to be applied for their facilities to demonstrate that the C-E lower support columns will maintain their functionality during the period of extended operation or for additional RVI components that may be fabricated from CASS, martensitic stainless steel or precipitation hardened stainless steel materials. WSES does have CASS materials in the lower support structure, specifically the lower support columns. This issue will be addressed when WSES considers application for license renewal.

7.8 SER Section 4.2.8, Applicant/Licensee Action Item 8:

As the submittal of the program for staff review is driven by license renewal commitments, WSES will determine whether to submit the document at the time of license renewal.

8.0 References

1. Entergy Document No. EN-DC-202, Rev. 5, "NEI 03-08 Materials Initiative Process,"

Entergy Nuclear Management Manual, 5/18/11 or later applicable revision. (SI File No.

1001328.201).

2. Nuclear Energy Institute, "Revision 2 to NEI 03-08, Guideline for the Management of Materials Issues," dated January, 2010. (SI File No. 100 1328.202).

3. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), EPRI, Palo Alto, CA: 2011. 1022863. (SI File No.

1001328.203).

4. ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2001 Edition, 2003 Addenda.

5. WCAP-17096-NP, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 2 or later revision, December 2009. (SI File No. 1001328.204).

6. Waterford Steam Electric Station Training Material, "Reactor Vessel Internals," SD-RVI, Rev. 8. (SI File No. 100 1328.205).

7. Entergy Nuclear Report, "Recommendations for Replacement ICI Thimble Measurement," WSES-ME-08-0005-000, Revision 0. (SI File No. 1001328.21 1P).

8. U.S. Code of Federal Regulations, "Title 10, Energy, Part 50, "Domestic Licensing of Production and Utilization Facilities," 50.55a, "Codes and Standards."

9. Entergy Procedure Document, "Maintaining Reactor Coolant Chemistry," CE-002-006, Revision 311, September 2013. (S1 File No. 1001328.213).

10. EPRI Letter MRP 2010-025, "Summary of Operating Experience with Pressurized Water Reactor Internals through 2009," March 30, 2010 or later revision. (SI File No.

1001328.206P). EPRI PROPRIETARY MA TERIAL.

11. EPRI Report MRP-219, "Materials Reliability Program: Pressurized Water Reactor Inspection Data Survey," Latest Revision. (SI File No. 1001328.207P). EPRI PROPRIETARYMATERIAL.

12. EPRI Report MRP-228, "Materials Reliability Program: Inspection Standard for Reactor Internals," Latest Revision. (SI File No. 100 1328.208).

13. "Pressurized Water Reactor Primary Water Chemistry Guidelines," Volumes I and 2, Revision 6, Electric Power Research Institute, Palo Alto, CA: 2007, 1014986.

14. Entergy Program Document, "Program Section for ASME Section XI, Division I Inservice Inspection Program," Program Section No: SEP-ISI-104, Revision 1, April 2012. (SI File No. 1001328.216).

15. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191). EPRI, Palo Alto CA: 2006. 1013234. (SI File No. 1001328.215P). EPRI PROPRIETARY MA TERIAL.

16. Materials Reliability Program Letter No. MRP 2013-025, "

Subject:

MRP-227-A Applicability Template Guideline," October 14, 2013. (SI File No. 1001328.217).

9.0 List of Drawings

1. Combustion Engineering Drawing E-9270-164-325, Revision No. 3, Sheets 1 and 2 of 2, "Core Shroud Assembly As-Built," SI File No. 1001328.209.

2. Combustion Engineering Drawing E-9270-164-312, Revision No. 1, Sheets 1 and 2 of 2, "Core Plate and Lower Support Assembly As Built," SI File No. 100 1328.209.

3. Combustion Engineering Drawing E-9270-164-303, Revision 5, "Reactor Internals Assembly," SI File No. 1001328.209.

4. Combustion Engineering Drawing E-9270-164-331, Revision 5, Sheet 4 of 5, "Upper Guide Structure Assy As-Built," SI File No. 1001328.209.

Table 2. C-E Plants Primary Components Applicable to WSES 1-3]

Item Applicability Effect Expansion Examination Examination (Mechanism) Link (Note 1) Method/Frequency Coverage (Note 1)

Core Shroud Assembly Plant designs with Cracking Remaining Enhanced visual Axial and horizontal core shrouds (IASCC) axial welds (EVT-1) examination weld seams at the (Welded) assembled in two no later than 2 core shroud re-entrant vertical sections Aging refueling outages comers as visible Core shroud plate-former Management from the beginning of from the core side of plate welds (IE) the license renewal the shroud, within six period and inches of central subsequent flange and horizontal examination on a ten- stiffeners.

year interval.

See Figures 3, 4, and 5.

Item Applicability Effect Expansion Examination Examination (Mechanism) Link (Note 1) Method/Frequency Coverage (Note 1)

Core Shroud Assembly Plant designs with Distortion None Visual (VT-1) If a gap exists, make core shrouds examination no later three to five (Welded) assembled in two (Void than 2 refueling measurements of gap vertical sections Swelling), as outages from the openings from the Assembly evidenced by beginning of the core side at the (Horizontal interface gap separation license renewal shroud re-entrant between upper and lower between the period. comers. Then, core shroud sections) upper and evaluate the swelling lower core Subsequent on a plant-specific shroud examinations on a basis to determine segments ten-year interval. frequency and method for additional Aging examinations.

Management (IE) See Figures 3, 4, and 5.

Item Applicability Effect Expansion Examination Examination (Mechanism) Link (Note 1) Method/Frequency Coverage (Note 1)

Core Support Barrel All plants Cracking Lower core Enhanced visual 100% of the Assembly (SCC) support beams (EVT-1) examination accessible surfaces of no later than 2 the upper flange weld Upper (core support Core support refueling outages to include a minimum barrel) flange weld barrel from the beginning of of 75% of the total assembly the license renewal weld length from upper cylinder period. Subsequent either side (inner or Upper core examinations on a outer diameter).

barrel flange ten-year interval.

See Figure 6.

Core Support Barrel All plants Cracking Lower Enhanced visual 100% of the Assembly (SCC, IASCC) cylinder axial (EVT-1) examination accessible surfaces of welds no later than 2 the lower cylinder Lower cylinder girth Aging refueling outages welds to include a welds Management from the beginning of minimum of 75% of (IE) the license renewal the total weld length period. Subsequent from either side examinations on a (inner or outer ten-year interval, diameter).

See Figure 6.

Item Applicability Effect Expansion Examination Examination (Mechanism) Link (Note 1) Method/Frequency (Note 1) Coverage Lower Support Structure All plants Cracking None Visual (VT-3) 100% of the (SCC, IASCC, examination no later accessible surfaces of Core support column Fatigue than 2 refueling the core support welds including outages from the column welds to damaged or beginning of the include a minimum fractured license renewal of 75% of the total material) period. Subsequent population of core examinations on a support column Aging ten-year interval. welds.

Management (IE, TE) See Figure 8.

Item Applicability Effect Expansion Examination Examination (Mechanism) Link (Note 1) Method/Frequency Coverage (Note 1)

Core Support Barrel All plants Cracking None If fatigue life cannot Examination Assembly (Fatigue) be demonstrated by coverage to be time-limited aging defined by evaluation Lower flange weld analysis (TLAA), to determine the enhanced visual potential location and (EVT-1) extent of fatigue examination, no later cracking.

than 2 refueling outages from the beginning of the See Figure 6.

license renewal period. Subsequent examination on a ten-year interval.

____________________________ a _________________ ________________ L J

Item Applicability Effect Expansion Examination Examination (Mechanism) Link (Note 1) Method/Frequency (Note 1) Coverage Lower Support Structure All plants with a Cracking None If fatigue life cannot Examination core support plate (Fatigue) be demonstrated by coverage to be Core support plate time-limited aging defined by evaluation Aging analysis (TLAA), to determine the Management enhanced visual potential location and (IE) (EVT-1) extent of fatigue examination, no later cracking.

than 2 refueling outages from the See Figure 8.

beginning of the license renewal period. Subsequent examination on a ten-year interval.

Item Applicability Effect Expansion Examination Examination (Mechanism) Link (Note 1) Method/Frequency Coverage (Note 1)

Control Element All plants with Cracking Remaining Visual (VT-3) 100% of tubes in Assembly instrument guide (SCC, Fatigue) instrument examination no later peripheral CEA tubes in the CEA that results in guide tubes than 2 refueling shroud assemblies Instrument guide tubes shroud assembly missing within the outages from the (i.e., those adjacent to supports or CEA shroud beginning of the the perimeter of the separation at assemblies license renewal fuel alignment plate).

the welded period. Subsequent joint between examination on a ten- See Figure 7.

the tubes and year.

supports Plant specific component integrity assessments may be required if degradation is detected and remedial action is needed.

Note:

1. Examination acceptance criteria and expansion criteria for C-E components are in Table 8.

Table 3. C-E Plants Expansion Components Applicable to WSES [3]

Effect Expansion Examination Examination Item Applicability (Mechanism) Link (Note 1) Method/Frequency Coverage (Note 1)

Core Shroud Plant designs with Cracking Core shroud Enhanced visual Axial weld seams Assembly (Welded) core shrouds (IASCC) plate-former (EVT-1) other than the core assembled in two plate weld examination, shroud re-entrant Remaining Axial vertical sections Aging comer welds at the Welds Management Re-inspection core mid-plane.

(IE) every 10 years following initial See Figures 3, 4, inspection, and 5.

Core Support Barrel All plants Cracking (SCC, Upper (core Enhanced visual 100% of accessible Assembly Fatigue) support (EVT- 1) welds and adjacent barrel) flange examination, base metal. (Note 2)

Lower core barrel weld flange Re-inspection every 10 years following initial See Figure 6.

inspection.

Effect Expansion Examination Examination Item Applicability (Mechanism) Link (Note 1) Method/Frequency Coverage (Note 1)

Core Support Barrel All plants Cracking (SCC) Upper (core Enhanced visual 100% of accessible Assembly support (EVT-1) surfaces of the Aging barrel) flange examination, welds and adjacent Upperweld base metal. (Note 2)

(IE) Re-inspection (including welds) every 10 years following initial inspection. See Figure 6.

Core Support Barrel All plants Cracking (SCC) Upper (core Enhanced visual 100% of accessible Assembly support (EVT- 1) bottom surface of barrel) flange examination, the flange. (Note 2)

Upper core barrel weld flange Re-inspection every 10 years following initial See Figure 6.

inspection.

Examination Examination Item Applicability Effect Expansion Method/Frequency Coverage Iy(Mechanism) Link(Note 1) (Note 1)

Core Support Barrel All plants Cracking (SCC) Core barrel Enhanced visual 100% of one side of Assembly assembly (EVT-1) the accessible weld girth welds examination, with and adjacent base Core barrel assembly initial and metal surfaces for axial welds subsequent the weld with the examinations highest calculated dependent on the operating stress.

results of core barrel assembly girth weld girthweldSee Figure 6.

examinations.

Lower Support All plants except Cracking (SCC, Upper (core Enhanced visual 100% of accessible Structure those with core Fatigue) support (EVT-1) surfaces (Note 2).

shrouds assembled including barrel) flange examination.

Lower core support with full-height damaged or weld beams shroud plates fractured evRe-inspection material eey1 er e iue8 following initial inspection.

Effect Expansion Examination Examination Item Applicability Effectaim Expnsi(one1 Method/Frequency Coverage Iy(Mechanism) Link(Note 1) (Note 1)

Control Element All plants with Cracking (SCC, Peripheral Visual (VT-3) 100% of tubes in Assembly instrument guide Fatigue) that instrument examination. CEA shroud tubes in the CEA results in guide tubes assemblies (Note 2).

Remaining shroud assembly missing within the Re-inspection instrument guide supports of CEA shroud every 10 years See Figure 7.

tubes separation at the assemblies following initial welded joint inspection.

between the tubes and supports.

Note:

1. Examination acceptance criteria and expansion criteria for CE components are in Table 8.

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).

Table 4. C-E Plants Existing Program Components Applicable to WSES [3]

ItemApplcabiity Effect Item Applicability (Mechanism) Primary Link Examination Method Examination Coverage Core Shroud Assembly All plants Loss of material ASME Code Visual (VT-3) First 10-year ISI after 40 (wear) Section XI examination, general years of operation, and Guide lugs condition examination at each subsequent Guide lug inserts and bolts for detection of interval.

excessive or asymmetrical wear. Accessible surfaces at specified frequency.

Lower Support Structure All plants Loss of material ASME Code Visual (VT-3) Accessible surfaces at with core (wear) Section XI examination, specified frequency.

Fuel alignment pins shrouds assembled in Aging two vertical Management (IE sections and ISR)

Core Barrel Assembly All plants Loss of material ASME Code Visual (VT-3) Area of the upper flange (wear) Section XI examination, potentially susceptible to Upper flange wear.

Table 5. Inspection Schedule for WSES Primary Components pe MRP-227-A Year 2014 20M5 2017 2018 202 2021 202 2024 Outage R19 R20 R21 R22 Rn3 R24 26 R2 Sbeaso S F S F S F S F Nominal Cycle Laengan 18 is8 18 18 18 1. 18 WIPY 25.4 269 2.4 29.9 31.4 32.9 X4 35,9

___________ __________ ____________Core Barrel Out Componn ISm-start Nrt-uency ee**

l Melhod thi'an**n Items (Table 3)

Core ShroudAssembly Core houdplate- NLT2 RFO from LR 10-year Interval Cracing (IASCC) Enhanced visual (EVT-1) Remainingaxial welds (Welded) trerplate welds Distortion (Void _________ I______

Core Shroud Assembly Swelling), as ewednced (Welded) Asseitly NILT2 RFO from LR l0-year Interval by seperaticn betwee Visual (VT-1) None Wheupper and ower core shroud sedimentrs Lower coe support beans Care Supprat Bate Llpe (care seppa Core sugput berroi esserrtl upper corebSupport B Lipper(core weldot *INLT 2 RFO from LR 10-year Interval Cracking(SCC) Enhanced usual (EVT-I) cyie r Asemnbly barrel)lange weld clne Upper core l large Core Support Barel Lower cylinder gith NLT 2 RFO from LR 10-yearInterval Cracking(=C, IASCC) Enhanced usual (EVT1) Lower cginderaxiel welds Asseml weds Craking (I , SCC.i core Swpon Barrel Core suppolt column NLT2 RFO from LR 10-yearInterval FatigaeIncluding r3 No Asembtly welds dmged artectured Material)

Core uppot B"Enhanced'Asuall (EVT-1) if CeonireuoBrt r Low atnge weld NLT 2 RFO from LR 10-yearIntermal Cracking(Feliaue) fatiguelife carnot be None Assembly demonstrated by TLAA.

Enhnced Vsual(EVT-1)if Low Support Structure Core supipot plate NLT 2 RFO from LR 10-year Interval Cracklng (Fatigue) ge life cannot be Note decanstlrted by TLhAA II Cracking (SCC, Fatigue) Visual (VT-3), Plant-specific Chat results i n missi ng I ntegrfty assessments may ControlySomen Instrument guide NLT2RFO from LR 10-yearinterval sutpprts orse araton be requiredRemaininginstrument guide tubes Asembly tubes atthe welded joint berqie ferddnwithin CbAshroudassemblies and remedial between the tubesand is detected action is needed support

Table 6. Insoection Schedule for WSES Existing Proaram Comnonents Listed in MRP-227-A Year 2014 2015 2017 2018 20=0 2021 2023 2024 Outag R19 R20 R2.1 R2 R23 R24 R25 R26 Seamer S F S F S F S F Nordnf~al CydoLelU. 18 18 is 18 is 18 18 1s mOt vroquaancw agur rIl Mln Met lbi iEFPY WA*

-a-- *m.......

.m 25,41 2 2 9,9 3L4. 32.9~ 3.4 35.9 First lOYear ISI after LRand at Core Shroud Assembly Guidelugs ASMESection Xl each subsequent Lossof Material (Wear) Visual (VT-3) AM ofshluppfliogse potutirly dI to w ear.

Auscept.n inspection interval Accessible Core Shroud Assembly l s and ASME Section Xl surface at Cracking (SCC, IASCC, Visual (VT-3)

Corehrou Assmbly bolts k,:cssrle surfatcenornp ecl terf specified Fatigue) ed nqncy. 1 frequency Accessible surface at Lower Support Structure Fuel alignment pins ASME Section Xl Loss of Material (Wear) IVisual (VT-3) specified frequency rusttO-year[SIa1t 40yearso'operution, Area of the odoatacuhsubsequent intava[

upper flange Core Barrel Assembly JUpperflange ASME Section Xl potentially LossofMaterial (Wear) Visual (VT-3) susceptible to wear

Table 7.Section XI 10 Year ISI Examinations of BN2 and BN3 Internals Components for WSES [14 Last 3rd Interval Completed Scheduled ComponentID Description ISONumber ASMECat ASMEItem 2 Interval Next Insp RF17 R.V.INTERIOR - AS 1-1200 / 1564-921; Spring RF20 Fall 01-054 ACCESSIBLE 1177 B-N-i B13.10 2011 2015 Spring 1-1200/1564-505, 2008 RF20 Fall 01-039 R.V. SNUBBER LUG AT O0 1564-63 B-N-2 B13.60 RF15 2015 Spring 1-1200 / 1564-505, 2008 RF20 Fall 01-040 R.V. SNUBBER LUG AT 600 1564-63 B-N-2 B13.60 RF15 2015 Spring R.V. SNUBBER LUG AT 1-1200 / 1564-505, 2008 RF20 Fall 01-041 1200 1564-63 B-N-2 B 13.60 RF15 2015 Spring R.V. SNUBBER LUG AT 1-1200 / 1564-505, 2008 RF20 Fall 01-042 1800 1564-63 B-N-2 B 13.60 RF15 2015 Spring R.V. SNUBBER LUG AT 1-1200 / 1564-505, 2008 RF20 Fall 01-043 2400 1564-63 B-N-2 B 13.60 RF15 2015

Last 3rd Interval Completed Scheduled ComponentID Description ISONumber ASMECat ASMEItem 2 Interval Next Insp Spring R.V. SNUBBER LUG AT 1-1200 / 1564-505, 2008 RF20 Fall 01-044 3000 1564-63 B-N-2 B 13.60 RF15 2015 Spring R.V. CORE STOP LUG AT 1-1200 / 1564-62, 2008 RF20 Fall 01-045 100 1564-63 B-N-2 B 13.60 RF15 2015 Spring R.V. CORE STOP LUG AT 1-1200 / 1564-62, 2008 RF20 Fall 01-046 400 1564-63 B-N-2 B 13.60 RF15 2015 Spring R.V. CORE STOP LUG AT 1-1200 / 1564-62, 2008 RF20 Fall 01-047 850 1564-63 B-N-2 B 13.60 RF15 2015 Spring R.V. CORE STOP LUG AT 1-1200 / 1564-62, 2008 RF20 Fall 01-048 1300 1564-63 B-N-2 B 13.60 RF15 2015 Spring R.V. CORE STOP LUG AT 1-1200 / 1564-62, 2008 RF20 Fall 01-049 1600 1564-63 B-N-2 B 13.60 RF15 2015

Last 3rd Interval Completed Scheduled Component_ID Description ISO_Number ASMECat ASMEItem 2 Interval Next Insp Spring R.V. CORE STOP LUG AT 1-1200 / 1564-62, 2008 RF20 Fall 01-050 2050 1564-63 B-N-2 B13.60 RF15 2015 Spring R.V. CORE STOP LUG AT 1-1200 / 1564-62, 2008 RF20 Fall 01-051 2500 1564-63 B-N-2 B 13.60 RF15 2015 Spring R.V. CORE STOP LUG AT 1-1200 / 1564-62, 2008 RF20 Fall 01-052 2800 1564-63 B-N-2 B113.60 RF15 2015 Spring R.V. CORE STOP LUG AT 1-1200 / 1564-62, 2008 RF20 Fall 01-053 3250 1564-63 B-N-2 B13.60 RF15 2015 Spring SURVEILLANCE CAPSULE 1-1200 / 1564-62, 2008 RF20 Fall 01-056 HOLDER AT 104 DEG 1564-63 B-N-2 B13.50 RF15 2015 Spring SURVEILLANCE CAPSULE 1-1200 / 1564-62, 2008 RF20 Fall 01-057 HOLDER AT 97 DEG 1564-63 B-N-2 B 13.50 RF15 2015

Last 3rd Interval Completed Scheduled Component_ID Description ISO_Number ASMECat ASMEItem 2 Interval Next Insp Spring SURVEILLANCE CAPSULE 1-1200 / 1564-62, 2008 RF20 Fall 01-058 HOLDER AT 85 DEG 1564-63 B-N-2 B13.50 RF15 2015 Spring SURVEILLANCE CAPSULE 1-1200 / 1564-62, 2008 RF20 Fall 01-059 HOLDER AT 263 DEG 1564-63 B-N-2 B 13.50 RF15 2015 Spring SURVEILLANCE CAPSULE 1-1200/ 1564-62, 2008 RF20 Fall 01-060 HOLDER AT 277 DEG 1564-63 B-N-2 B 13.50 RF15 2015 Spring SURVEILLANCE CAPSULE 1-1200 / 1564-62, 2008 RF20 Fall 01-061 HOLDER AT 284 DEG 1564-63 B-N-2 B13.50 RF15 2015 Spring 1-1200/ 1564-871, 2008 RF20 Fall 01-062 FLOW BAFFLE 1564-63 B-N-2 B 13.60 RF15 2015 Spring R.V.INTERIOR & CSB - CSB 1-1200 / 1564-62, 2008 RF20 Fall 01-055 REMOVED 1564-63 B-N-3 B 13.70 RF15 2015

Table 8. C-E Plants Examination Acceptance and Expansion Criteria Applicable to WSES [3]

Exam ination I .. Ex a so Ad iin a Ex m n t o Item Applicability Acceptance (Note 1) Criteria Expansion Link(s) Expansion Criteria Accetance Acceptance Critia Criteria Core Shroud Plant designs Enhanced Visual Remaining Confirmation that a The specific relevant Assembly with core (EVT-1) axial welds surface-breaking condition is a (Welded) shrouds examination. indication > 2 inches in detectable crack-like Core shroud plate- assembled in length has been surface indication.

former plate weld two vertical The specific relevant detected and sized in sections condition is a the core shroud plate-detectable crack-like former plate weld at surface indication. the core shroud re-entrant comers (as visible from the core side of the shroud),

within 6 inches of the central flange and horizontal stiffeners, shall require EVT-1 examination of all remaining axial welds by the completion of the next refueling outage.

Examination Expansion Additional Examination Item Applicability Acceptance Criteria Link(s) Expansion Criteria Acceptance Criteria (Note 1) Acceptance Criteria Core Shroud Plant designs Visual (VT- 1) None N/A N/A Assembly with core examination.

(Welded) shrouds Assembly assembled in The specific relevant (Horizontal interface two vertical condition is gap between upper sections evidence of physical and lower core separation between shroud sections) the upper and lower core shroud sections.

Core Support Barrel All plants Enhanced Visual Lower core Confirmation that a The specific relevant Assembly (EVT-1) support beams surface-breaking condition is a Upper (core support examination. Upper core indication >2 inches in detectable crack-like barrel) flange weld barrel cylinder length has been surface indication.

The specific relevant (including detected and sized in condition is a welds) the upper flange weld detectable crack-like Upper core shall require that an surface indication, barrel flange EVT-1 examination of the lower support beams, upper core barrel cylinder and upper core barrel flange be performed by the completion of the next refueling outage.

Examination Expansion Additional Examination Item Applicability Acceptance Criteria Expansion Criteria Acceptance Criteria (Note 1) Link(s)

Core Support Barrel All plants Enhanced Visual Lower Confirmation that a The specific relevant Assembly (EVT-1) cylinder axial surface-breaking condition for the Lower cylinder girth examination, welds indication >2 inches in expansion lower welds length has been cylinder axial welds is The specific relevant detected and sized in a detectable crack-like condition is a the lower cylinder girth surface indication.

detectable crack-like weld shall require an surface indication. EVT-1 examination of all accessible lower cylinder axial welds by the completion of the next refueling outage.

Lower Support All plants Visual (VT-3) None None N/A Structure examination.

Core support column The specific relevant welds condition is missing or separated welds.

Core Support Barrel All plants Visual (EVT-1) None N/A N/A Assembly examination.

Lower flange weld The specific relevant condition is a detectable crack-like indication.

Examination Expansion Additional Examination Item Applicability Acceptance Criteria Link(s) Expansion Criteria Acceptance Criteria (Note 1) AcceptanceCriteria Lower Support All plants with Enhanced Visual None N/A N/A Structure a core support (EVT- 1)

Core support plate plate examination.

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

Control Element All plants with Visual (VT-3) Remaining Confirmed evidence of The specific relevant Assembly instruments examination, instrument missing supports or conditions are missing Instrument guide tubes in the tubes within separation at the supports and separation tubes CEA shroud The specific relevant the CEA welded joint between at the welded joint assembly conditions are shroud the tubes and supports between the tubes and missing supports and assemblies, shall require the visual the supports.

separation at the (VT-3) examination to welded joint be expanded to the between the tubes remaining instrument and the supports. tubes within the CEA shroud assemblies by completion of the next

____ _ _refueling outage.

Note:

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

Table 9. WSES Inspection Plan Summary Table Primary Component Expansion Links Inspection Type and Inspection Schedule Coverage Core Shroud Remaining axial Enhanced visual TBD based on Assembly welds (EVT-I) future license examination, renewal Core shroud Coverage: Axial and commitments.

plate-former plate horizontal weld weld seams at the core shroud re-entrant comers as visible from the core side of the shroud, within six inches of central flange and horizontal stiffeners.

Core Shroud None Visual (VT-I) TBD based on Assembly examination. If gap future license exists, 3 to 5 renewal measurements of commitments.

Assembly gap openings from (Horizontal the core side at the interface gap core shroud re-etatcres between tipper and entrant comers.

lower core shroud Then, evaluate the sections) swelling on a plant-specific basis to determine frequency and method of additional examinations.

Primary Component Expansion Links Inspection Type and Inspection Schedule Coverage Core Support Barrel Lower core support Enhanced visual TBD based on Assembly beams (EVT-I) future license examination. renewal Core support barrel Coverage: 100% of commitments.

Upper (core support assembly upper the accessible barrel) flange weld cylinder surfaces of the Upper core barrel upper flange weld to include a minimum flange of 75% of the total weld length from either side (inner or outer diameter).

Core Support Barrel Lower cylinder Enhanced visual TBD based on Assembly axial welds (EVT-1) future license examination, renewal Lower cylinder girth Coverage: 100% of commitments.

welds the accessible surfaces of the lower cylinder welds to include a minimum of 75% of the total weld length from either side (inner or outer diameter).

Lower Support None Visual (VT-3). TBD based on Structure Coverage: 100% of future license the accessible renewal Core support surfaces of the core commitments.

column welds support column welds to include a minimum of 75% of the total population of core support column welds.

Primary Component Expansion Links Inspection Type and Inspection Schedule Coverage Core Support Barrel None Enhanced visual TBD based on Assembly (EVT-1) future license examination, renewal Coverage: Defined commitments. The Lower flange weld by evaluation to need for determine the examination can be potential location determined by and extent of fatigue results of plant-cracking. specific fatigue analysis.

Lower Support None Enhanced visual TBD based on Structure (EVT-1) future license examination, renewal Coverage: Defined commitments. The Core support plate by evaluation to need for determine the examination can be potential location determined by and extent of fatigue results of plant-cracking. specific fatigue analysis.2 Control Element Remaining VT-3 examination. Examination no Assembly instrument guide Coverage: 100% of later than 2 tubes within the tubes in peripheral refueling outages CEA shroud CEA shroud from the beginning Instrument guide assemblies assemblies (i.e., of the license tubes those adjacent to the renewal period and perimeter of the fuel subsequent alignment plate). examination on a ten-year interval.'

Primary Component Expansion Links Inspection Type and Inspection Schedule Coverage Other Supplemental Examinations External to MRP-227-A and the ISI Program ICI Thimbles None Physically measure In accordance with length of specified Design Change thimbles to monitor 020701067-6 and growth project time 051001333-2 before contact with guide tube bottom Notes:

1. Inspections will depend on future schedule for vessel 10-year ISI exams when the core barrel is removed.

2. Cycle counting of design transients can be performed to demonstrate that cumulative fatigue usage factor (CUF) is less than 1.0.

Figure 9. WSES Reactor Vessel Internals Inspection Plan Attachment A Significant Internal and External Operating Experience Review and Evaluation Event ID or Event Applicability to Determination of Target Entergy Description Waterford 3 Required Actions Completion Date Reference Number Zircaloy section Applicable Site Design change Design Change of Incore OE implemented to completed. Low Instrument replace thimbles frequency length thimbles with reduced monitoring is exhibited late length Zircaloy required blooming growth to allow for effects which additional resulted in rowth cycle thimble contact g y with fuel guide tube OE Core Barrel Possibly PWROG is 2015 Alignment Key applicable evaluating a (Clevis) bolting project to address failures alignment key performance criteria OE Excessive Guide Not applicable, None None Card wear at a Westinghouse WSES CEA PWR shroud design does not make use of guide card like features OE (various Baffle and Not Applicable None None events) former bolting failures WSES shroud is all welded design. There have been no events involving CEA shroud bolting failures

Attachment B Open Action Tracking Log Item Action Description of Action Planned Comments Tracking Completion Reference Date 2

3 4

5 6

7

Attachment C WSES Reactor Vessel Internals Drawings Page Entergy Title C-E Drawing #

Drawing #

C-2 Reactor Internals Assembly E-9270-164-303, Rev. 5 C-3 Core Shroud Assembly "As Builts" E-9270-164-325, Sheet 1, Rev. 1 C-4 Core Shroud Assembly "As Builts", Core Shroud E-9270-164-325, Sheet Segment 2, Rev. 1 C-5 Core Plate and Lower Support Assembly "As E-9270-164-312, Sheet Builts" 1, Rev. 1 C-6 Core Plate and Lower Support Assembly "As E-9270-164-312, Sheet Builts" 2, Rev. 1 C-7 Core Support Barrel "As Builts" E-9270-164-313, Rev. 1 C-8 Upper Guide Structure Assembly "As Builts" E-9270-164-331, Sheet 4, Rev. 5 WSES specific drawings included in this report are proprietary to Westinghouse and Entergy. Reactor internals management engineering program documents should generally include drawings with sufficient detail to unambiguously show the components and/or locations that will be inspected. As such, as-built drawings are preferred, if available.