ML15160A193

NYS000496 - Dacimo, Fred, Entergy, Letter to Document Control Desk, USNRC, License Renewal Application - Revised Reactor Vessel Internals Program and Inspection Plan, NL-12-037 (February 17, 2012) (ML12060A312)

ML15160A193
Person / Time
Site:	Indian Point
Issue date:	02/17/2012
From:	State of NY, Office of the Attorney General
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 27908, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15160A193 (79)

Text

NYS000496 Submitted: June 9, 2015 w

K-Enterqy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 SwEntergy Buchanan, N.Y. 10511-0249 Tel (914) 254-2055 Fred Dacimo Vice President NL-12-037 Operations License Renewal February 17, 2012 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

License Renewal Application - Revised Reactor Vessel Internals Program and Inspection Plan Compliant with MRP-227-A Indian Point Nuclear Generating Unit Nos' 2 and 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64

REFERENCES:

1. Entergy Letter dated April 23, 2007, Fred Dacimo to Document Control Desk, "License Renewal Application" (NL-07-039)

2. Entergy Letter dated July 14, 2010, Fred Dacimo to Document Control Desk, "Amendment 9 to License Renewal Application (LAR) - Reactor Vessel Internals Program" (NL-10-063).

3. Entergy Letter dated September 28, 2011, Fred Dacimo to Document Control Desk, "Completion of Commitment #30 Regarding the Reactor Vessel Internals Inspection Plan" (NL-1 1-107).

4. EPRI, Materials Reliability Program (MRP), Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227).

5. NRC, "Final Safety Evaluation of EPRI Report, Materials Reliability Program Report 1016596, Revision 0, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines" dated June 22, 2011.

6. NRC Letter dated January 27, 2012, "Summary of Conference Call Concerning the Reactor Vessel Internals Program at Indian Point Nuclear Generating Station, Units 2 and 3".

Dear Sir or Madam:

Entergy Nuclear Operations, Inc. applied for renewal of the Indian Point Energy Center (IPEC) operating license by the reference 1 letter which included a list of regulatory

ý,j 2,

NL-12-037 Docket Nos. 50-247 and 50-286 Page 2 of 2 commitments. The commitment list contained commitment # 30 for submitting an inspection plan for reactor vessel internals. Reference 2 provided the Indian Point Nuclear Generating Unit Nos. 2 & 3 Reactor Vessel Internals Program Plan. Reference 3 provided information supporting the completion of commitment # 30 to the License Renewal. The Application regarding the Aging Management Programs for Reactor Vessel Internals and the IPEC Reactor Vessel Internals Inspection Plan were developed to meet MRP-227 (reference 4) and addresses all action items and conditions as a result of the NRC Final Safety Evaluation of MRP-227 (Reference 5).

As discussed in Reference 6, Entergy is revising Reference 2 and affected portions of Reference 3 (Attachments 1 and 2, respectively) to reflect the issuance of the final approved MRP-227-A.

There are no new commitments identified in this submittal. If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-254-6710.

Sincer FD/cbr Attachments: 1. Indian Point Energy Center Revised Reactor Vessel Internals Program Compliant with MRP-227-A

2. Indian Point Energy Center Revised Reactor Vessel Internals Inspection Plan Compliant with MRP-227-A cc: Mr. William Dean, Regional Administrator, NRC Region I Mr. J. Boska, Senior Project Manager, NRC, NRR, DORL Mr. Robert F. Kuntz, NRC Sr. Project Manager, Division of License Renewal Mr. David Wrona, NRC Branch Chief, Engineering Review Branch I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel SNRC Resident Inspectors Office, Indian Point Ms. Bridget Frymire, NYS Dept. of Public Service Mr. Francis J. Murray, Jr., President and CEO, NYSERDA

ATTACHMENT 1 TO NL-12-037 Indian Point Energy Center Revised Reactor Vessel Internals Program Compliant with MRP-227-A Additions Underlined Deletions Lined Out ENTERGY NUCLEAR OPERATIONS, INC INDIAN POINT NUCLEAR GENERATING UNITS 2 AND 3 DOCKET NOS. 50-247 & 50-286

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 1 Page 1 of 10 A.2.1.41 Reactor Vessel Internals Aging Management Activities The Reactor Vessel Internals (RVI) Program is a new plant specific program to manage aging effects of reactor vessel internals using the guidance from the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The MRP inspection and evaluation (I&E) guidelines for managing the effects of aging on pressurized water reactor vessel internals are presented in MRP-227-A, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The MRP also developed inspection requirements specific to the inspection methods delineated in MRP-227-A, as well as requirements for qualification of the nondestructive examination (NDE) systems used to perform those inspections. These inspection requirements are presented in MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals."

MRP-227-A and MRP-228 provide the basis of the IPEC Reactor Vessel Internals (RVI)

Program. ,Rcvision.s to MRP 227 and MRP 228, including any "hangs rsuiting fromFthe , NRC rview of the documents (issued as MRP 227 A and MRP 228 A), will be iRo*r*pratod i*t* tho IPEC RVI Prg*ram. The RVI Program will monitor the effects of aging degradation mechanisms on the intended functions of the internals through periodic and conditional examinations. The RVI Program will detect and evaluate cracking, loss of material, reduction of fracture toughness, loss of preload and dimensional changes of vessel internals components in accordance with MRP-227-A inspection requirements and evaluation acceptance criteria.

The IPEC RVI Program will be implemented and maintained in accordance with the guidance in NEI 03-08 [Addenda], Addendum A, "RCS Materials Degradation Management Program Guidelines." Any deviations from mandatory, needed, or good practice implementation requirements established in MRP-227-A or MRP-228, will be dispesitie4 resolved in accordance with the NEI 03-08 implementation protocol. The RVI Program will be implemented prior to the period of extended operation.

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 1 Page 2 of 10 A.3.1.41 Reactor Vessel Internals Aging Management Activities The Reactor Vessel Internals (RVI) Program is a new plant specific program to manage aging effects of reactor vessel internals using the guidance from the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The MRP inspection and evaluation (I&E) guidelines for managing the effects of aging on pressurized water reactor vessel internals are presented in MRP-227-A, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The MRP also developed inspection requirements specific to the inspection methods delineated in MRP-227-A, as well as requirements for qualification of the nondestructive examination (NDE) systems used to perform those inspections. These inspection requirements are presented in MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals."

MRP-227-A and MRP-228 provide the basis of the IPEC Reactor Vessel Internals (RVI)

Program. R nto MRP 227 and MIRP 228, including any changc Fr.culting from the NRCP rcviow of the documontS (issued as MRP 227 A and MRP 228 A), will bo incorporatcd into the IPEC RVI Po*gra*m. The RVI Program will monitor the effects of aging degradation mechanisms on the intended function of the internals through periodic and conditional examinations. The RVI Program will detect and evaluate cracking, loss of material, reduction of fracture toughness, loss of preload and dimensional changes of vessel internals components in accordance with MRP-227-A inspection requirements and evaluation acceptance criteria.

The IPEC RVI Program will be implemented and maintained in accordance with the guidance in NEI 03-08 [Addenda], Addendum A, "RCS Materials Degradation Management Program Guidelines." Any deviations from mandatory, needed, or good practice implementation requirements established in MRP-227-A or MRP-228, will be dispesitieFled resolved in accordance with the NEI 03-08 implementation protocol. The RVI Program will be implemented prior to the period of extended operation.

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 1 Page 3 of 10 B.1.42 Reactor Vessel Internals Program Program Description The Reactor Vessel Internals Program is a new plant-specific program. Revision 1 of NUREG-1801 includes no aging management program description for PWR reactor vessel internals.

NUREG-1801,Section XI.M16, PWR Vessel Internals, instead defers to the guidance provided in Chapter IV line items as appropriate. The Chapter IV line item guidance recommends actions to:

.(1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval."

The industry programs for investigating and managing aging effects on reactor internals are part of the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The MRP developed inspection and evaluation (I&E) guidelines for managing the effects of aging on pressurized water reactor vessel internals. These guidelines, as reviewed and accepted by the NRC, are presented in MRP-227-A, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The I&E guidelines include:

• summary descriptions of PWR internals and functions;

• summary of the categorization and aging management strategy development of potentially susceptible locations, based on the safety and economic consequences of aging degradation;

• direction for methods, extent, and frequency of one-time, periodic, and conditional examinations and other aging management methodologies;

• acceptance criteria for the one-time, periodic, and conditional examinations and other aging management methodologies; and

• methods for evaluation of aging effectS that exceed conditions that fail to meet the examination acceptance criteria.

The MRP also developed inspection procedure requirements specific to the inspection methods delineated in MRP-227-A, as well as requirements for qualification of the nondestructive examination (NDE) systems used to perform those inspections. These inspection procedure requirements are presented in MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals."

MRP-227-A and MRP-228 provide the basis of the IPEC Reactor Vessel Internals (RVI)

Program. Revisions to MRP-227-A and MRP-228, including anY changes ... ulting from, the NRCC re"..w of the d' .... n u.d as MRP 227 A and MRP 228 A), will be incorporated into the IPEC RVI Program.

The RVI Program will monitor the effects of aging on the intended function of the reactor vessel internals through periodic and conditional examinations. The RVI Program will detect and evaluate cracking, loss of material, reduction of fracture toughness, loss of preload and

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 1 Page 4 of 10 dimensional changes of vessel internals components in accordance with MRP-227-A inspection recommendations and evaluation acceptance criteria.

IPEC will implement and maintain the RVI Program in accordance with the guidance in NEI 03-08 [Addenda], Addendum A, "RCS Materials Degradation Management Program Guidelines."

Any deviations from mandatory, needed, or good practice implementation activities established in MRP-227-A or MRP-228, will be managed in accordance with the NEI 03-08 implementation protocol.

The Reactor Vessel Internals Program is implemented through the Indian Point Energy Center Reactor Vessel Internals Inspection Plan (Reference B.2-3). The inspection plan provides additional details, including:

" Identification of items for inspection,

" Specification of the type of examination appropriate for each degradation mechanism,

" Specification of the required level of examination qualification,

" Schedule of initial inspection, schedule and frequency of subsequent inspections,

• Criteria for sampling and coverage,

• Criteria for expansion of scope if unacceptable indications are found,

• Inspection acceptance criteria,

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

• Provisions for updating the program based on industry-wide results, and

" Contingency measures to repair, replace or mitigate unacceptable examination results.

The Indian Point Energy Center Reactor Vessel Internals Inspection Plan also includes responses to applicable license renewal applicant action items identified in the NRC's safety evaluation of MRP-227 (incorporated in MRP-227-A).

Evaluation

1. Scope of Program MRP-227-A guidelines are applicable to reactor internal structural components. The scope does not include consumable items such as fuel assemblies and reactivity control assemblies which are periodically replaced based on neutron flux exposure. The scope does not include welded attachments to the reactor vessel which are considered part of the vessel, or nuclear instrumentation (flux thimble tubes) which forms part of the reactor coolant pressure boundary. Other programs manage the effects of aging on these components.

MRP-227-A separates PWR internals components into four groups depending on (1) their susceptibility to and tolerance of aging effects, and (2) the existence of programs that manage the effects of aging. These groupings include:

• Primary - those internals components that are highly susceptible to the effects of at least one aging mechanism (identified in Table 4-3 of MRP-227A.);

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 1 Page 5 of 10 Expansion - those internals components that are highly or moderately susceptible to the effects of at least one aging mechanism, but for which functionality assessment has shown a degree of tolerance to those effects (identified in Table 4-6 of MRP-227-A);

• Existing Programs - those internals components that are susceptible to the effects of at least one aging mechanism and for which generic and plant-specific existing AMP elements are capable of managing those effects (identified in Table 4-9 of MRP-227-A); and

• No Additional Measures - those internals components for which the effects of aging mechanisms are below the MRP-227-A screening criteria (internals components not included in Tables 4-3, 4-6 or 4-9 of MRP-227-A.).

The categorization of internals components for Westinghouse PWRs, as presented in MRP-227-A, applies to IPEC Unit 2 and Unit 3 vessel internals. The component inspections identified in MRP-227-A, Tables 4-3 and 4-6 for primary and expansion group components, define the scope of the IPEC RVI Program inspections. Those components subject to aging management by existing programs, as delineated in MRP-227-A, Table 4-9, are included in the scope of those programs, and are not part of the RVI Program inspections.

Components that are not included in Tables 4-3, 4-6 or 4-9 are considered to be within the scope of the program, but require no specific inspections.

2. Preventive Actions The Reactor Vessel Internals Program is a condition monitoring program that does not include preventive actions. However, primary water chemistry is maintained in accordance with EPRI guidelines by the Water Chemistry Control - Primary and Secondary Program, which minimizes the potential for loss of material, stress corrosion cracking (SCC), primarv water stress corrosion cracking (PWSCC), and irradiation assisted stress corrosion cracking (IASCC).

Plant operations also influence aging of the vessel internals. The general assumptions about plant operations used in the development of the MRP-227-A guidelines are applicable to the IPEC units. Both units are base loaded and both implemented low leakage core loading patterns within the first 30 years of operation. IPEC has implemented no design changes to reactor vessel internals beyond those identified in general industry guidance or recommended by Westinghouse.

3. Parameters Monitored or Inspected The RVI Program will monitor the effects of aging on the intended function of the internals through periodic and conditional examinations and other aging management methods, as required. As described in MRP-227-A, the program contains elements that will monitor and inspect for the parameters that indicate the progress of each of these effects. The component inspections identified in MRP-227-A, Tables 4-3 and 4-6 for primary and expansion group components respectively, set fourth the parameters monitored by the IPEC RVI Proaram inspections.

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 1 Page 6 of 10 The program will use NDE techniques to detect loss of material through wear, identify changes in dimension due to void swelling and irradiation growth, distortion, or deflection distotion of components, and locate cracks induced by SCC, PWSCC, IASCC, or fatique/cyclical loading. Loss of preload, caused by thermal and irradiation-enhanced stress relaxation or creep, is indirectly monitored by inspecting for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections.

The reduction of fracture toughness, induced by either thermal a-ging or neutron irradiation embrittlement, is indirectly monitored by using visual or volumetric examination techniques to monitor for cracking in the components and by applying applicable reduced fracture toughness properties in flaw evaluations where warranted.

Visual examinations (VT-3) will be used to detect wear. Visual examinations (VT-3) will also detect distortion or cracking through indications such as gaps or displacement along component joints and broken or damaged bolt locking systems. Direct measurements of spring height will be used to detect distortion of the internals hold down spring. Visual examinations (EVT-1) will be used to detect broken components and crack-like surface flaws of components and welds. Volumetric (ultrasonic) examinations will be used to locate cracking of bolting. (MRP, 227*,Tabl*os, 4 3 nd 6)

4. Detection of Aging Effects The RVI Program will detect cracking, loss of material, reduction of fracture toughness, loss of preload and dimensional changes (distortion) of vessel internals components in accordance with the specific provisions of MRP-227-A. The NDE systems (i.e., the combinations of equipment, procedure, and personnel) used to detect these aging effects will be qualified in accordance with MRP-228. The RVI Program will conduct inspections of primary group components as follows delineated in MRP-227-A, Table 4-3):

Ps~e~wdfio-visuaI examinations (VT 3) will detoct loss of matorial duo to woar fromn control rod guide tubo guide plates and thclrmal shield flexuro plates-.

Pio;dic visual examinations (VT 3) of the baffle formo assembly . plates and edgo bolts will detect symptoms of distortion due to Yead swolling or cracking fromR IASCO. The .

symptoms include abnremal interactions with fuel assemblies, gapS Or displacemonet along component joints, brokon or damagod bolt locking systems6, and failed Or missing bolts.

Diroct m.easuremont. of "Gprng height will detect distortion of the internals hold down spi*ng due to a less of cutnffess. Measurements wil be takon porloJ1alcay, as necaoa to aeteFrmi 0 the life of theo *pr*Rg.

Peri*odic*i,*ual -xaminations (EVT- 1) will doteI t GrFak like surface flaws of the *otr*l roed guide tube assembly 1owor flange welds and the upper core barrel to flIange weld-.

Volum~etric (UJT) examinations will locate cracking of baffle formerF bolting. B3aselin~e and subsequent mneasurements will be used to confirmA the stability of the_ bolting pattern.

Indications from EVT-1 or UT inspections may result in additional inspections of expansion group components, as determined by expansion criteria delineated in MRP-227-A, Table 5-

3. The relationships between primary group component inspection findings and additional inspections of expansion group components are as fellews described in MRP-227A, Table 4-6.

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 1 Page 7 of 10 Indications from the EVT 1 in-pection- of tho control rod guide tube assembly l.w.r flango wc..,, may re.ult in E'J 1 inspections of the lower support column bodies and V, 3 inspections of boftom moeunted instrumcentatien columnR bodicc to detect cracking-.

.ndications fro the. EVT 1 in.pectio. Of the upper coro barrol to flange weld may result in EVT 1 inSPections of the remaini*g core barrel welds Indications fromA the UJT inspections of baffle forerFA boltin ma rslt in UJT inspecstiens of the lower SUPPort column bolts and the barrel formeir bolts foer crmackinmg.-

5. Monitoring and Trending The RVI Program uses the inspection guidelines for PWR internals in MRP-227-A.

Inspections in accordance with these guidelines will provide timely detection of aging effects. In addition to the inspections of primary group components, expansion group components have been defined should the scope of examination and re-examination require expansion beyond the primary group. Records of inspection results are maintained allowing for comparison with subsequent inspection results.

.IPEC will share inSPection results with the industr' in In accordance with the geed,4.. +,,

recommendation of MRP-227-A, IPEC will provide a summary report of all inspections and monitoring, items requiring evaluation, and new repairs to the MRP Program Manager. The IPEC-spec.ific results will be incorporated into an overall industry report that will track industry progress and will aid in evaluation of potentially significant issues, identification of fleet trends, and determination of any needed revisions to the MRP-227-A guidelines.

6. Acceptance Criteria The RVI Program acceptance criteria are 49441 provided in Section 5 of MRP-227-A. Table 5-3 and Sections 5.1 through 5.3 of MRP-227-A provides the acceptance criteria for inspections of the IPEC primary and expansion group components. The criteria for expanding the examinations from the primary group components to include the expansion group components are also delineated in MRP-227-A, Table 5-3. The examination acceptance criteria include: (i) specific, descriptive relevant conditions for the visual (VT-3) examinations; (ii) requirements for recording and dispositioning surface breaking indications that are detected and sized for length by the visual (EVT-1) examinations; a, -(iii) requirements for system-level assessment of bolted assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits and (iv) requirements for fit up limits on physical measurements of the hold down springs.

7. Corrective Action Conditions adverse to quality, such as failures, malfunctions, deviations, defective material or equipment, and nonconformances, are promptly identified and corrected. In the case of significant conditions adverse to quality, measures are implemented to ensure that the cause of the nonconformance is determined and that corrective action is taken to preclude recurrence. In addition, the cause of the significant condition adverse to quality and the corrective action implemented is documented and reported to appropriate levels of management. The Entergy (10 CFR Part 50, Appendix B) Quality Assurance Program, including relevant corrective action controls, applies to the RVI Program.

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 1 Page 8 of 10 Any detected condition that dee6 n*t'*atiy fails to meet the examination acceptance criteria must be processed through the corrective action program. Example methods for analytical disposition of unacceptable conditions are discussed or referenced in Section 6 of MRP-227-A. The evaluation methods include recommendations for flaw depth sizing and for crack growth determinations as well for performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications. These methods or other NRC-approved evaluation d*omnstrated and voifiod aterativo methods may be used. The alternative of component repair and replacement of PWR vessel internals is subject to the applicable requirements of the ASME Code Section XI.

8. Confirmation Process This attribute is discussed in Section B.0.3.

9. Administrative Controls This attribute is discussed in Section B.0.3.

10. Operating Experience Relatively few incidents of PWR internals aging degradation have been reported in operating U.S. commercial PWR plants. However, PWR internals aging degradation has been observed in European PWRs, specifically with regard to cracking of baffle-former bolting. For this reason, the U.S. PWR owners and operators created a program to inspect the baffle-former bolting to determine whether similar aging degradation might be expected to occur in U.S. plants. A benefit of this decision was the experience gained with the UT examination techniques used in the inspections.

In addition, the industry undertook laboratory testing projects to gather the materials data necessary to support future inspections and evaluations. Other confirmed or suspected material degradation concerns that the industry has identified for PWR components are wear in thimble tubes, potential wear in control rod guide tube guide plates, and cracking in some high-strength bolting. The industry has addressed the last concern primarily through replacement of high-strength bolting with bolt material that is less susceptible to cracking and by improved control of pre-load.

The RVI Program established in accordance with the MRP-227-A guidelines is a new program. Accordingly, there is no direct programmatic history for IPEC. However, program inspections will use qualified techniques similar to those successfully used at IPEC and throughout the industry for ASME Section Xl Code inspections. Internals inspections (VT-3) have been conducted at IPEC in accordance with ASME Section Xl Code requirements, with no indications of component degradation. IPEC has appropriately responded to industry operating experience for reactor vessel internals. For example, guide tube support pins (split pins) have been replaced in both units on the basis of industry experience. As with other U.S. commercial PWR plants, cracking of baffle former bolts is recognized as a potential issue for the IPEC units. As a result, IPEC has monitored industry developments and recommendations regarding these components.

Development of the MRP-227-A guidelines is based upon industry operating experience, research data, and vendor evaluations. Reactor vessels internals aging degradation

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 1 Page 9 of 10 incidents in both U.S. and foreign plants were considered in the development of the MRP-227-A guidelines. As implemented, this program will account for applicable future operating experience during the period of extended operation.

Conclusion The RVI Program will be effective at managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls in accordance with MRP-227-A and MRP-228 guidelines and current IPEC programs. The RVI Program will provide reasonable assurance that the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 1 Page 10 of 10 B.2 REFERENCES B.2-1 NUREG-1 800, StandardReview Plan for Review of License Renewal Applications for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, September 2005.

B.2-2 NUREG-1 801, Generic Aging Lessons Learned (GALL) Report, U.S. Nuclear Regulatory Commission, September 2005.

B.2-3 Indian Point Energy Center Reactor Vessel Internals Inspection Plan.

ATTACHMENT 2 TO NL-12-037 Indian Point Energy Center Revised Reactor Vessel Internals Inspection Plan Compliant with MRP-227-A Additions Underlined Deletions Lined Out ENTERGY NUCLEAR OPERATIONS, INC INDIAN POINT NUCLEAR GENERATING UNITS 2 AND 3 DOCKET NOS. 50-247 & 50-286

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan I

INTRODUCTION 1.1 Aging Management Program Inspection Plan The EPRI MRP guidelines define a supplemental inspection program for managing aging effects on the reactor vessel internals and were used to develop this inspection plan for IPEC Units 2 and 3. The EPRI MRP Reactor Internals Focus Group developed the MRP-227-A Gguidelines to support the demonstration of continued functionality, with requirements for inspections to detect the effects of aging along with requirements for the evaluation of detected aging effects, if any. The development of MRP-227-A combined the results of component functionality assessments with component accessibility, operating experience, existing evaluations and prior examination results to determine the appropriate aging management methods, initial examination timing and the need and timing for f-subsequent inspections and identified the components and locations for supplemental examination.

in a...r.dan wi-th MRP 227, tThis inspection plan includes:

" Identification of items for inspection,

" Specification of the type of examination appropriate for each degradation mechanism,

" Specification of the required level of examination qualification,

• Schedule of initial inspection schedule and frequency of subsequent inspections,

" Criteria for sampling and coverage,

• Criteria for expansion of scope if ....an.ieia.*ed unacceptable indications are found,

" Inspection acceptance criteria,

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

• Provisions for ulpdating the program based on industry-wide results; and

" Contingency measures to repair, replace or mitigate unacceptable examination results.

Page 1

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan 2

BACKGROUND OF IPEC REACTOR VESSEL INTERNALS DESIGN This section provides a summary of the design characteristics for the IPEC Westinghouse PWR internals.

2.1 Westinghouse Internals Design Characteristics A schematic view of a typical set of Westinghouse-designed PWR internals is Figure 2-1. More detailed views of selected internals components are Figures 2-2 through 2-16 at the end of this section. These figures are typical and are not an exact representation of the IPEC internals.

To help in the categorization of IPEC internals design characteristics as discussed in MRP-227-A Section 3.1.3, the following information is provided. IPEC Units 2 and 3 are Westinghouse four loop plants with a downflow baffle-barrel region flow design, and a top hat design upper support plate. Unit 2 had an original thermal output rating of 2758 MWth and Unit 3 had an original thermal output rating of 3025 MWth. Unit 2 has a current thermal output rating of 3216 MWth and Unit 3 has a current thermal output rating of 3188 MWth.

Page 2

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian PointEnergy Center Reactor Vessel Internals Inspection Plan ROD TRAVEL HOUSING INSTRUMENTATION PORTS THERMAL SLEEVE LIFTING LUG CLOSURE HEAD ASSEMBLY HOLD-DOWN SPRING CONTROL ROD GUIDE TUBE CONTROL ROD DRIVE SHAFT INLET NOZZLE BAFFLE RADIAL CONTROL ROD SUPPORT CLUSTER (W-IHDRAWI BAFFLE CORE SUPPORT ACCESS PORT COLUMNS INSTRUMENTATION REACTOR VESSEL THIMBLE GUIDES RADIAL SUPPORT CORE SUPPORT LOWER CORE PLATE Figure 2-1 Overview of typical Westinghouse internals Page 3

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Westinghouse internals consist of two basic assemblies: an upper internals assembly that is removed during each refueling operation to obtain access to the reactor core and a lower internals assembly that can be removed following a complete core off-load.

The reactor core is positioned and supported by the upper internals and lower internals assemblies. The individual fuel assemblies are positioned by fuel alignment pins in the upper core plate and the lower core plate. These pins control the orientation of the core with respect to the upper and lower internals assemblies. The lower internals are aligned with the upper internals by the upper core plate alignment pins and secondarily by the head/vessel alignment pins. The lower internals are aligned to the vessel by the lower radial support/clevis assemblies and by the head/vessel alignment pins. Thus, the core is aligned with the vessel by a number of interfacing components.

The lower internals assembly is supported in the vessel by clamping to a ledge below the vessel-head mating surface and is closely guided at the bottom by radial support/clevis assemblies. The upper internals assembly is clamped at this same ledge by the reactor vessel head. The bottom of the upper internals assembly is closely guided by the core barrel alignment pins of the lower internals assembly.

Upper Internals Assembly The major sub-assemblies that constitute the upper internals assembly are the: (1) upper core plate (UCP); (2) upper support column assemblies; (3) control rod guide tube assemblies; and (4)

.upper support plate (USP).

During reactor operation, the upper internals assembly is preloaded against the fuel assembly springs and the internals hold down spring by the reactor vessel head pressing down on the outside edge of the USP. The USP acts as the divider between the upper plenum and the reactor vessel head and as a relatively stiff base for the rest of the upper internals. The upper support columns and the control rod guide tubes are attached to the USP. The UCP, in turn, is attached to the upper support columns. The USP design at IPEC is designated as a top hat design.

The UCP is perforated to permit coolant to pass from the core below into the upper plenum between the USP and the UCP. The coolant then exits through the outlet nozzles in the core barrel. The UCP positions and laterally supports the core by fuel alignment pins extending below the plate. The UCP contacts and preloads the fuel assembly springs and thus maintains contact of the fuel assemblies with the lower core plate (LCP) during reactor operation.

The upper support columns vertically position the UCP and are designed to take the uplifting hydraulic flow loads and fuel spring loads on the UCP. The control rod guide tubes are bolted to the USP and pinned at the UCP so they can be easily removed if replacement is desired. The control rod guide tubes are designed to guide the control rods in and out of the fuel assemblies to control power generation. Guide tube cards are located within each control rod guide tube Page 4

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan to guide the absorber rods. The control rod guide tubes are also slotted in their lower sectiors to allow coolant exiting the core to flow into the upper plenum.

The upper instrumentation columns are bolted to the USP. These columns support the thermocouple guide tubes that lead the thermocouples from the reactor head through the upper plenum to just above the UCP.

The UCP alignment pins locate the UCP laterally with respect to the lower internals assembly.

The pins must laterally support the UCP so that the plate is free to expand radially and move axially during differential thermal expansion between the upper internals and the core barrel. The UCP alignment pins are the interfacing components between the UCP and the core barrel.

Lower Internals Assembly The fuel assemblies are supported inside the lower internals assembly on top of the LCP. The functions of the LCP are to position and support the core and provide a metered control of reactor coolant flow into each fuel assembly. The LCP is elevated above the lower support casting by support columns and bolted to a ring support attached to the inside diameter of the core barrel. The support columns transmit vertical fuel assembly loads from the LCP to the much thicker lower support casting. The function of the lower support casting is to provide support for the core. The lower support casting is welded to and supported by the core barrel, which transmits vertical loads to the vessel through the core barrel flange.

The primary function of the core barrel is to support the core. A large number of components are

.attached to the core barrel, including the baffle/former assembly, the core barrel outlet nozzles, the thermal shields, the alignment pins that engage the UCP, the lower support casting, and the LCP. The lower radial support/clevis assemblies restrain large transverse motions of the core barrel but at the same time allow unrestricted radial and axial thermal expansion.

The baffle and former assembly consists of vertical plates called baffles and horizontal support plates called formers. The baffle plates are bolted to the formers by the baffle/former bolts, and the formers are attached to the core barrel inside diameter by the barrel/former bolts. Baffle plates are secured to each other at selected corners by edge bolts. In addition, at IPEC, corner brackets are installed behind and bolted to the baffle plates. The baffle/former assembly forms the interface between the core and the core barrel. The baffles provide a barrier between the core and the former region so that a high concentration of flow in the core region can be maintained.

A secondary benefit is to reduce the neutron flux on the vessel.

The function of the core barrel outlet nozzles is to direct the reactor coolant, after it leaves the core, radially outward through the reactor vessel outlet nozzles. The core barrel outlet nozzles are located in the upper portion of the core barrel directly below the flange and are attached to the core barrel, each in line with a vessel outlet nozzle.

Page 5

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Additional neutron shielding of the reactor vessel is provided in the active core region by thermal shields attached to the outside of the core barrel.

A flux thimble is a long, slender stainless steel tube that passes from an external seal table, through a bottom mounted nozzle penetration, through the lower internals assembly, and finally extends to the top of a fuel assembly. The flux thimble provides a path for a neutron flux detector into the core and is subjected to reactor coolant pressure and temperature on the outside surface and to atmospheric conditions on the inside. The flux thimble path from the seal table to the bottom mounted nozzles is defined by flux thimble guide tubes, which are part of the primary pressure boundary and not part of the internals. The bottom-mounted instrumentation (BMI) columns provide a path for the flux thimbles from the bottom of the vessel into the core. The BMI columns align the flux thimble paths with instrumentation thimbles in the fuel assembly.

In the upper internals assembly, the upper support plate, the upper support columns, and the upper core plate are considered core support structures. In the lower internals assembly the lower core plate, the lower support casting, the lower support columns, the core barrel including the core barrel flange, the radial support/clevis assemblies, the baffle plates, and the former plates are classified as core support structures.

Wear Area Figure 2-2 Typical Weatinghouse control rod guide card (17x17 fuel assembly)

Page 6

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan PE UP-I 3 e9 pe qi1 1p5 q Lower Flange Weld CRGT Split J

Pins Figure 2-3 Typical Westinghouse control rod guide tube assembly Page 7

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Axial Weld Flange Weld Upper Core Barrel to Lower Core Barrel Circumferential Weld n

Low*er Barrel..

Circumferential Weld Lower Barrel Axial Weld Core Barrel to Support, Plate Weld Figure 2-4 Major fabrication welds in typical Westinghouse core barrel Page 8

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 IndianPoint Energy Center Reactor Vessel InternalsInspection Plan RAFFLE TO FORMER BOLT (LONG & (OfT)

COAER EDGE BRACKET BAFFLE TO FORMER BOLT Figure 2-5 Bolt locations in typical Westinghouse baffle-former-barrel structure.

Page 9

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan w )

0 Figure 2-6 Baffle-edge bolt and baffle-former bolt locations at high fluence seams in bolted baffle-former assembly Page 10

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan H

High Fluence Seams Figure 2-7 High fluence seam locations in Westinghouse baffle-former assembly Page 11

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Potential Gaps Baffle-Former Levels Figure 2-8 Exaggerated view of void swelling induced distortion in Westinghouse baffle-former assembly.

Page 12

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Vertical Displacement Figure 2-9 Vertical displacement of Westinghouse baffle plates caused by void swelling.

Page 13

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan TOP SUPPORT PLATEr CORE BARREL Figure 2-10 Schematic cross-sections of the Westinghouse hold-down springs Page 14

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Core Barrel Thermal Shield Thermal Shield Flexure Core Support Figure 2-11 Location of Westinghouse thermal shield flexures Page 15

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Lower Core Plate Lower Core Support Structure Core Support Plate (Forging)

Figure 2-12 Schematic indicating location of Westinghouse lower core support structure. Additional details shown in Figure 23-13 LOWER CORE PLATE DIFFUSER PLATE SCORE SUPPORT PLATE/FORGING BOTTOM MOUNTED INSTRUMENTATION COLUMN Figure 2-13 Westinghouse lower core support structure and bottom mounted instrumentation columns. Core support column bolts fasten the core support columns to the lower core plate Page 16

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan 9

Figure 2-14 Typical Westinghouse core support column. Core support column bolts fasten the top of the support column to the lower core plate Figure 2-15 Examples of Westinghouse bottom mounted instrumentation column designs Page 17

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Figure 2-16 Typical Westinghouse thermal shield flexure Page 18

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan 3

INSPECTION PLAN

SUMMARY

Management of component aging effects includes actions to prevent or control aging effects, review of operating experience to better understand the potential for aging effects to occur, inspections to detect the onset of aging effects in susceptible components, and protocols for evaluation and remediation of the effects of aging and pr..ed.res to en.....u r.cm.. ntgi effets are managed-in a coor.dinated pr..gram..

3.1 Component inspection and Evaluation Overview This discussion summarizes the guidance of the MRP Inspection & Evaluation (I&E) guidelines necessary to understand implementation but does not duplicate the full discussion of the technical bases. MRP-227-A and its supporting documents provide further information on the technical bases of the program.

MRP-227-A establishes four groups of reactor internals components with respect to inspections:

Primary, Expansion, Existing Programs and No Additional Measures, as summarized below.

" Primary: Those PWR internals 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 that are needed to ensure functionality of Primary components are described in the I&E guidelines. The Primary group also includes components which have shown a degree of tolerance 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 that are highly or moderately susceptible to the effects of at least one of the eight aging mechanisms, but for which a 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.

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

" No Additional Measures: Those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria were placed in the No Additional Measures group. Items categorized as Category A in MRP- 191 are those for which aging effects are below the screening criteria, so that aging degradation significance is minimal. Primary, expansion, and Page 19

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan existing examinations verify that the chemical control program has been effective at controlling stress corrosion cracking and loss of material due to corrosion for Category A components.

Additional components were placed in the No Additional Measures group as a result of Failure Modes, Effects and Criticality Analysis (FMECA) and the functionality assessment. No further action is required for managing the effects of aging onf the No Additional Measures components.

However, any core support structures subject to ASME Section XI Examination Category B-N-3 requirements continue to be subject to those ASME Code requirements throughout the period of extended operation.

The inspection&+aqtid methods for Primary and Expansion components were selected from visual, surface and volumetric examination methods 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 methods include: Visual examinations (VT-3, VT-1, EVT-1),

surface examinations, volumetric examinations (specifically UT) and physical measurements.

MRP-227-A provides detailed justification for the components selected for inspection and the specific examination methods selected for each. The MRP-228 report, PWR Internals Inspection Standards, provides detailed examination standards and any inspection technical justification or inspection personnel training requirements.

3.2 Inspection and Evaluation Requirements for .Primary Components The inspection requirements for Primary Components at IPEC Units 2 and 3 from MRP-227-A are provided in Table 5-2.

3.3 Inspection and Evaluation Requirements for Expansion Components The inspection requirements for Expansion Components at IPEC Units 2 and 3 from MRP-227-A are provided in Table 5-3.

3.4 Inspections of Existing Program Components The list of Existing Program Components at IPEC Units 2 and 3 from MRP-227-A are provided in Table 5-4. This includes components in the Section XI ISI Program categories B-N-2 and B-N-3 for IPEC Units 2 and 3 designated a B N 2 and B N 3 lo;atio The Reactor Vessel Component Inspections Plaif conducted as part of the ISI plrogram for IPEC Units 2 and 3 is;provded are listed in Table 5-6. The. .p.nent. are in.p;. ted as pa.-e f*fhe41S Prga.... The ISI Program inspections are implemented in accordance with ASME Section XI schedule requirements.

Page 20

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian PointEnergy Center Reactor Vessel Internals Inspection Plan 3.5 Examination Systems Equipment, techniques, procedures and personnel used to perform examinations required under this program will be consistent with the r requifeifaeits provisions of MRP-228. Indications detected during these examinations will be characterized and reported in accordance with the e ite provisions of MRP-228.

3.6 Information Supplied in Response to the NRC Safety Evaluation of MRP-227-A As part of the NRC Revision 1 to the Final Safety Evaluation of MRP-227, a number of action items and conditions were specified by the staff. Table 5-8 summarizes doeumnen4s the IPEC response to the NRC Final Safety Evaluation of MRP-227. Topical Report Conditions from the NRC Final Safety Evaluation of MRP-227 have been addressed in MRP-227-A. Whefeve possible,4 These items have been addressed in the appropriate sections of this document. Al4 NRC action items and conditi.ns not addressed els*where in this doumen Applicant/Licensee Action Items from the NRC Final Safety Evaluation of MRP-227 are discussed in this section.

SER Section 4.2.1. Applicant/Licensee Action Item 1 IPEC has assessed its plant design and operating history and has determined that MRP-227-A is applicable to the facility. The assumptions regarding plant design and operating history made in MRP-191 are appropriate for IPEC and there are no differences in component inspection categories at IPEC. IPEC Unit 2 (IP2) had the first 8 years of operation with a high leakage core loading pattern. IPEC Unit 3 (IP3) had the first 10 years 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, IPEC is bounded by the assumptions in MRP-191. IPEC has always operated as a base-load plant which operates at fixed power levels and does not vary power on a calendar or load demand schedule.

SER Section 4.2.2, Applicant/Licensee Action Item 2 IPEC reviewed the information in Table 4-4 of MRP-191 and determined that this table contains all of the RVI components that are within the scope of license renewal. This is shown in Table 5-7.

SER Section 4.2.3. Applicant/Licensee Action Item 3 At IP2, the original X750 guide tube support pins (split pins) were replaced in 1995 (after 21 years in service) with an improved X750 Revision B material made from more selective material with more continuous carbide coverage grain boundaries and tighter quality controls, to provide greater resistance to stress corrosion cracking. IP2 plans to begin preliminary split pin replacement engineering and walkdowns in 2014 and replace the split pins in 2016.

Page 21

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan At IP3 the original X750 guide tube support pins (split pins) were replaced in 2009 (after 33 years in service) with cold worked 316 stainless steel. The cold worked 316 stainless steel is a significant improvement over the X750. Based on this operating experience, the IP3 split pins are expected to last through the end of the period of extended operation. At IPEG thc effects of aging on these

. ompn.ents will be managed in the period of e eperation based on industry exprec an.d plantt specifie evaluations-.

SER Section 4.2.4, Applicant/Licensee Action Item 4 This action item does not apply to Westinghouse designed units.

SER Section 4.2.5, Applicant/Licensee Action Item 5 The IPEC plant specific acceptance criteria for hold down springs and an explanation of how the proposed acceptance criteria are consistent with the IPEC licensing basis and the need to maintain the functionality of the hold down spring&under all licensing basis conditions will be developed prior to the first required physical measurement. Ina...r.dan.e withi SER Sectionl.

4.125, IPEC will submit this infor-mation to the NRC as part of the submfittal to apply the approved ver.sien of MR 227,r "TThe acceptance criteria will ensure the remaining compressible height of the spring shall provide hold down forces within the IPEC design tolerance. If a plant specific acceptance criterion is not developed for the hold down spring, JPEC will replace the

,spring in lieu of performing the first required physical measurement.

SER Section 4.2.6, Applicant/Licensee Action Item 6 This action item does not apply to Westinghouse designed units.

SER Section 4.2.7, Applicant/Licensee Action Item 7 The IPEC plant specific analyses to demonstrate the lower support column bodies will maintain their functionality during the period of extended operation will consider the possible loss of fracture toughness in these components due to thermal and irradiation embrittlement. The analyses will be consistent with the IPEC licensing basis and the need to maintain the functionality of the lower support column bodies under all licensing basis conditions of operation. In a.eor.dan.e with SER Sc.tion 4..2.7, IPEC will submit this information to the NRC prior to the period of extended operation. as .part of the submittal to apply the appr.ved ver.si of MRP 7-SER Section 4.2.8. Applicant/Licensee Action Item 8 A Reactor Vessel Internals AMP description for IPEC was included in Amendment 9 to the License Renewal Application (NL-10-063, July 14, 2010). The AMP description has been revised to be consistent with MRP-227-A. The revised AMP description has been submitted under letter NL-12-037.

Page 22

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan This document ifi*ekdes comprises an inspection plan which addresses the identified plant-specific action items contained in the NRC Revision 1 to the Final Safety Evaluation for MRP-227. IPEC is not requesting any deviations from the guidance provided in MRP-227-A-as approved by the 1,, NR, G.T Page 23

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan 4

EXAMINATION ACCEPTANCE AND EXPANSION CRITERIA AND IMPLEMENTATION REQUIREMENTS 4.1 Examination Acceptance Criteria 4.1.1 Visual (VT-3) Examination Visual (VT-3) examination is 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, provides a set of relevant conditions for the visual (VT-3) examination of removable core support structures in Section IWB. These are:

1. structural distortion or displacement of parts to the extent that component function may be impaired;

.2. loose, missing, cracked, or fractured parts, bolting, or fasteners;

.3. corrosion or erosion that reduces the nominal section thickness by more than 5%;

4. wear of mating surfaces that may lead to loss of function; and

5. structural degradation of interior attachments such that the original cross-sectional area is reduced more than 5%.

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 5-5 for the benefit of the examiners. One or more of these specific relevant condition descriptions may be applicable to the Primary and Expansion components listed in Tables 5-2 and 5-3.

The examination acceptance criteria for components requiring visual (VT-3) examination is thus the absence of any of the relevant condition(s) specified in Table 5-5.

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.

Page 24

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan 4.1.2 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." The acceptance criterion for any visual (VT-1) examinations is the absence of any relevant conditions defined by the ASME Code, as supplemented by more specific plant inservice inspection requirements.

4.1.3 Enhanced Visual (EVT-1) Examination Enhanced visual (EVT-1) examination has the same requirements as the ASME Code Section XI visual (VT-1) examination, with additional requirements given in the Inspection Standard, MRP-228. These enhancements are intended to improve the detection and characterization of discontinuities taking into account the remote visual aspect of reactor internals examinations. As a result, EVT- 1 examinations are capable of detecting small surface breaking cracks and sizing surface crack length when used in conjunction with sizing aids (e.g. landmarks, ruler, and tape measure). EVT-1 examination is the appropriate NDE method for detection of cracking in plates or their welded joints. Thus the relevant condition applied for EVT- 1 examination is the same as for cracking in Section XI which is crack-like surface breaking indications.

Therefore, until such time as engineering studies provide a basis by which a quantitative amount of degradation can be shown acceptable for the specific component, any observed relevant condition must be dispositioned. In the interim, the examination acceptance criterion is the absence of any detectable surface breaking indication.

4.1.4 Surface Examination Surface ET (eddy current testing) examination is 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 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 in 4.1.3 (and accompanying entries in Table 5-5) are therefore applied when this method is used as an alternative or supplement to visual examination.

4.1.5 Volumetric Examination The intent of volumetric examinations specified for bolts and pins is to detect planar defects. No flaw sizing measurements are recorded or assumed in the acceptance or rejection of individual Page 25

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan bolts or pins. Individual bolts or pins are accepted based on the absence of relevant indications established as part of the examination technical justification. When a relevant indication is detected in the cross-sectional area of the bolt or pin, it is assumed to be non-functional and the indication is recorded. A bolt or pin that passes the criterion of the examination is considered functional.

Because of this pass/fail acceptance of individual bolts or pins, the examination acceptance criterion for volumetric (UT) examination of bolts and pins is based on a reliable detection of indications as established by the individual technical justification for the proposed examination.

This is in keeping with current industry practice. For example, planar flaws on the order of 30%

of the cross-sectional areahave been determined reliably detectable in previous bolt NDE technical justifications for baffle-former bolting.

Bolted and pinned assemblies are evaluated for acceptance based on a plant specific evaluation.

4.2 Physical Measurements Examination Acceptance Criteria Continued functionality can be confirmed by physical measurements where, for example, loss of material caused by wear, loss of pre-load of clamping force caused by various degradation mechanisms, or distortion/deflection caused by void swelling may occur. For Westinghouse designs, tolerances are available on a design or plant-specific basis. Specific acceptance criteria will be developed as required, and thus are not provided generically in this plan.

4.3 Expansion Criteria The criterion for expanding the scope of examination from the Primary components to their linked Expansion components is contained in Table 5-5 for IPEC.

4.4 Implementation Requirements 4.4.1 Consistent with the requirements of NEI 03-08, if the guidance contained in Tables 5-2, 5-3, 5-4, and 5-5 cannot, need not, or will not be implemented as written, a technical justification must be prepared that clearly states what requirement cannot, need not, or will not be met and why; what alternative action is being taken to satisfy the objective or intent of the guidance: and why the alternative action is acceptable. Since the Expansion components are also "needed" requirements, the technical justification for not fully implementing a Primary component examination or not implementing it in a manner consistent with its intent, would be expected to include disposition of the associated Expansion components.

When submittal of a deviation from work products or elements is required, the justification shall be reviewed and approved in accordance with the applicable plant procedures with the additional responsibility for deviation from a "Needed" element that Page 26

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan an internal independent review is performed and that concurrence is obtained from the responsible utility executive.

4.4.2 Examinations contained in this inspection plan shall be conducted in accordance with MRP-228.

4.4.3 Examination results that do not meet the examination acceptance criteria shall be recorded and entered in the IPEC corrective action program and dispositioned.

4.4.4 If an engineering evaluation is used to disposition an examination result that does not meet the examination acceptance criteria, this engineering evaluation shall be conducted in accordance with a NRC-approved evaluation methodology.

4.4.5 A summary report of all inspections and monitoring, items requiring evaluation, and new repairs shall be provided to the MRP Program Manager within 120 days of the completion of an outage during which PWR internals within the scope of MRP-227-A are examined.

Page 27

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan 5

TABLES Table 5-1 Indian Point 2 & 3 Component Cross Reference Table 5-2 Primary Components at IPEC Units 2 and 3 Table 5-3 Expansion Components at IPEC Units 2 and 3 Table 5-4 Existing Program Components at IPEC Units 2 and 3 Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Table 5-6 Reactor Vessel Component ISI Program Inspection Plan for IPEC Units 2 and 3 Table 5-7 List of IPEC Reactor Vessel Interior Components and Materials Based on MRP-191 - Table 4-4 Table 5-8 IPEC Response to the NRC Revision 1 to the Final Safety Evaluation of MRP-227 Page 28

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227-A 1 Core Baffle/Former Assembly - Lower Internals Baffle-Former Assembly -

Bolts Assembly - Baffle and Baffle-Edge Bolts (Tables Former Assembly 3-3, 4-3 and 5-3)

Baffle-Edge Bolts Baffle-Former Assembly -

Baffle-Former Bolts Baffle-Former Bolts (Tables 3-3, 4-3 and 5-3) 2 Core Baffle/Former Assembly - Lower Internals Baffle-Former Assembly -

Plates Assembly - Baffle and Assembly (Tables3,4-3 Former Assembly and 5-3)

Baffle Plates Former Plates 3 Core Barrel Assembly - Bolts and Lower Internals Core Barrel Assembly -

Screws Assembly - Baffle and Barrel-Former Bolts Former Assembly (Tables 3-3 and 4-6)

Barrel-Former Bolts 4 Core Barrel Assembly - Axial Lower Internals Thermal Shield Assembly Flexure Plates (Thermal Shield Assembly - Neutron - Thermal Shield Flexures Flexures) Panels/Thermal Shield (Tables 3-3_ 4-3 and 5-3)

Thermal Shield Flexures 5 Core Barrel Assembly - Flange Lower Internals Core Barrel Assembly -

Assembly - Core Core Barrel Flange (Tables Barrel 3-3, 4-6 and 4-9)

Core Barrel Flange Page 29

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227-A 6 Core Barrel Assembly - Ring Lower Internals None-Core Barrel Assembly - Core Assembly - Upper and Core Barrel Assembly - Shell Barrel Lower Core Barrel Cylinder Girth Welds Core Barrel Assembly - Thermal Upper Core Barrel (Tables 3-3, 4-3 and 2 Shield places in 5-3)

Lower Core Barrel Core Barrel Assembly -

Lower Internals Upper and Lower Core Assembly - Neutron Barrel Cylinder Axial panels/thermal shield Welds (Tables 3-3, 4-6 and 2 places in 5-3)

Thermal shield 7 Core Barrel Assembly - Lower None Core Barrel Assembly -

Core Barrel Flange Weld Lower Core Barrel Flange Lower Internals Weld (Tables 3-3, 4-63 and Core Barrel Assembly - Upper Assembly - Core 5-3)

Core Barrel Flange Weld Barrel Core Barrel Assembly -

Core Barrel Flange Upper Core Barrel Flange Weld (Tables 3-3, 4-3 and 5-3) 8 Core Barrel Assembly - Outlet Lower Internals Core Barrel Assembly -

Nozzles Assembly - Core Core Barrel Outlet Nozzles Barrel Welds (Tables 3-3, and 4-6)

Core Barrel Outlet Nozzles 9 Lower Internals Assembly - Interfacing Alignment and Interfacing Clevis Insert Bolt Components - Components - Clevis Insert Interfacing Bolts (Tables 3-3, and 4-9)

Components Clevis Insert Bolts Page 30

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 &3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227-A 10 Lower Internals Assembly - Interfacing Non*e additional measures Clevis Insert Components -

Interfacing Components Clevis Inserts 11 Lower Internals Assembly - Lower Internals None additional measures Intermediate Diffuser Plate Assembly - Diffuser Plate Diffuser Plate 12 Lower Internals Assembly - Fuel Lower Internals None additional measures Alignment Pin Assembly - Lower Core Plate and Fuel Alignment Pins Fuel Alignment Pins 13 Lower Internals Assembly - Lower Internals Lower Internals Assembly Lower Core Plate Assembly - Lower - Lower Core Plate (

Core Plate and Fuel Tables 3-3, and 2 places in Alignment Pins 4-9,--2-paees)

Lower Core Plate Page 31

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 &3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227-A 14 Lower Internals Assembly - Lower Internals None Lower Internals Assembly - Lower Assembly - Lower Support "Lower Core Support Castings Support Casting or Casting (Tables 3-3, and 4-Forging 6)

"Column Cap Lower Support "Lower Core Support Column Casting Bodies None Lower Internals None-additional measures Assembly - Lower Support Column Assembly Lower Support Column Nuts Lower Support Lower Support Assembly -

Column Bodies Lower Support Column Bodies (Cast) (Tables 3-3, and 4-6) 15 Lower Internals Assembly - Lower Internals Lower Support Assembly -

Lower Core Support Plate Assembly - Lower Lower Support Column Column Bolt Support Column Bolts (Tables 3-3, and 4-6)

Assembly Lower Support Column Bolts 16 Lower Internals Assembly - Lower Internals None-additional measures Lower Core Support Plate Assembly - Lower Column Sleeves Support Column Assembly Lower Support Column Sleeves Page 32

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227-A 17 Lower Internals Assembly - Lower Internals None-additional measures Radial Key Assembly - Radial Support Keys Radial Support Keys 18 Lower Internals Assembly - Lower Internals None-additional measures Secondary Core Support Assembly - Secondary Core Support (SCS)

Assembly SCS Base Plate 19 RCCA Guide Tube Assembly - Upper Internals None-additional measures Bolt Assembly - Control Rod Guide Tube Assemblies and Flow Downcomers Bolts 20 RCCA Guide Tube Assembly - Upper Internals Control Rod Guide Tube Guide Tube (including Lower Assembly - Control Assembly - Lower Flange Flange Welds) Rod Guide Tube Welds (Tables 3-3_,_4-3 and Assemblies and Flow 5-3)

Downcomers Flanges - lower 21 RCCA Guide Tube Assembly - Upper Internals Control Rod Guide Tube Guide Plates Assembly - Control Assembly - Guide Plates Rod Guide Tube (Cards) (Tables 3-3. 4-3 Assemblies and Flow and 5-3)

Downcomers Guide Plates/Cards Page 33

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227-A 22 RCCA Guide Tube Assembly - Upper Internals Nofie-additional measures Support Pin Assembly - Control Rod Guide Tube Assemblies and Flow Downcomers Guide Tube Support Pins 23 Core Plate Alignment Pin Interfacing Alignment and Interfacing Components - Components - Upper Core Interfacing Plate Alignment Pins Components (Tables 3-3,4-9)

Upper Core Plate Alignment Pins 24 Head/Vessel Alignment Pin Interfacing None-additional measures Components -

Interfacing Components Head and Vessel Alignment Pins 25 Hold-down Spring Interfacing Alignment and Interfacing Components - Components - Internals Interfacing Hold Down Spring (Tables Components 3-3, 4-3 and 5-3)

Internals Hold Down Spring 26 Mixing Devices Upper Internals None-additional measures Assembly - Mixing

- Support Column Orifice Base Devices

- Support Column Mixer Mixing devices Page 34

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 &3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227-A 27 Support Column Upper Internals Nonie-additional measures Assembly - Upper Support Column Assemblies Column Bodies 28 Upper Core Plate, Fuel Alignment Upper Internals None-additional measures Pin Assembly - Upper Core Plate and Fuel Alignment Pins Fuel Alignment Pins 29 Upper Support Plate, Support Upper Internals No additional measures for Assembly (Including Ring) Assembly - Upper the upper support plate Support Plate Assembly Upper Support Plate Upper Internals Assembly

- Upper Support Ring or Upper Support Ring or Skirt (Tables 3-3 and 4-9)

Skirt 30 Upper Support Column Bolt Upper Internals None-additional measures Assembly - Upper Support Column Assemblies Bolts 31 Bottom Mounted Instrumentation Lower Internals Bottom Mounted Column Assembly - Bottom- Instrumentation System -

Mounted Bottom Mounted Instrumentation (BMI) Instrumentation (BMI)

Column Assemblies Column Bodies (Tables 3-3 and 4-6)

BMI Column Bodies Page 35

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-1 Indian Point 2 & 3 Component Cross Reference Item Letter NL-10-063 Component MRP-191 Table 4-4 MRP-227-A 32 Flux Thimble Guide Tube Lower Internals Bottom Mounted Assembly - Flux Instrumentation System -

Thimbles (Tubes) Flux Thimble Tubes (Tables 3-3 and 4-9)

Flux Thimbles (Tubes) 33 Thermocouple Conduit Upper Internals Nonte-additional measures Assembly - Upper Instrumentation Conduit and Support Conduits Page 36

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability (Mechanism) Expansion Link Method/Frequency Examination Coverage Control Rod Guide Tube IPEC Units 2 and Loss of Material None Visual (VT-3) examination no 20% examination of the number of Assembly 3 (Wear) later than 2 refueling outages CRGT assemblies, with all guide Guide plates (cards) from the beginning of the cards within each selected CRGT license renewal period, assembly examined.

Subsequent examinations are required on a ten-year interval. See Figure 2-2 Control Rod Guide Tube IPEC Units 2 and Cracking (SCC, Bottom-mounted Enhanced visual (EVT-1) 100% of outer (accessible) CRGT Assembly 3 Fatigue) instrumentation examination to determine the lower flange weld surfaces and Lower flange welds Aging Management (BMI) column presence of crack-like surface adjacent base metal on the (lE and TE) bodies, flaws in flange welds no later individual periphery CRGT Lower support than 2 refueling outages from assemblies. A minimum of 75% of column bodies the beginning of the license the total identified sample (cast) renewal period and subsequent population must be examined.

Upper core plate examination on a ten-year Lower support interval. See Figure 2-3 casting Core Barrel Assembly IPEC Units 2 and Cracking (SCC) NenieCore barrel Periodic eEnhanced visual 100% of one side of the accessible Upper core barrel flange 3 outlet nozzle welds (EVT- 1) examination, no later surfaces of the selected weld and weld than 2 refueling outages from adjacent base metal. A minimum of the beginning of the license 75% of the total weld length renewal period and subsequent (examined + unexamined),

examination on a ten-year including coverage consistent with interval, the Expansion criteria in Table 5-5, must be examined from either the inner or outer diameter for inspection credit.

ISee Figure 2-4 Page 37

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 IndianPoint Energy Center Reactor Vessel InternalsInspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability (Mechanism) Expansion Link Method/Frequency Examination Coverage Core Barrel Assembly IPEC Units 2 and Cracking (SCC, Upper and lower Periodic enhanced visual (EVT- 100% of one side of the accessible Upper and lower core barrel 3 IASCC, Fatigue) core barrel cylinder 1) examination, no later than 2 surfaces of the selected weld and cylinder girth welds axial welds refueling outages from the adjacent base metal. A minimum of beginning of the license renewal 75% of the total weld length period and subsequent (examined + unexamined),

examination on a ten-year including coverage consistent with interval. the Expansion criteria in Table 5-5, must be examined from either the inner or outer diameter for inspection credit.

See Figure 2-4 Core Barrel Assembly IPEC Units 2 and Cracking (SCC, None Periodic enhanced visual (EVT- 100% of one side of the accessible Lower core barrel flange 3 Fatigue) 1) examination, no later than 2 surfaces of the selected weld and weld refueling outages from the adjacent base metal. A minimum of beginning of the license renewal 75% of the total weld length (At IPEC this weld is the period and subsequent (examined + unexamined),

lower core barrel to lower examination on a ten-year including coverage consistent with support casting weld. IPEC interval, the Expansion criteria in Table 5-5, does not have a lower core must be examined from either the barrel flange) inner or outer diameter for inspection credit.

See Figure 2-4 (Core Barrel to ISupport Plate Weld)

Page 38

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability (Mechanism) Expansion Link Method/Frequency Examination Coverage Baffle-Former Assembly IPEC Units 2 and Cracking (IASCC, None Visual (VT-3) examination, with Bolts and locking devices on high Baffle-edge bolts 3 Fatigue) that results baseline examination between fluence seams. 100% of components in 20 and 40 EFPY and subsequent accessible from core side. A

" Lost or broken examinations on a ten-year minimum of 75% of the total locking devices interval, population (examined +

" Failed or missing unexamined), including coverage bolts consistent with the Expansion

" Protrusion of bolt criteria in Table 5-5, must be heads examined for inspection credit. a Aging Management components total (arccessible (IE and ISR) inace.ile inspe.tion area or Void swelling volume will be examined or, when effects on this add-ressig a set of like cemponents component is (e.g., bolting), that the inspection managed through exmn a mnu . ampl.e si.e It management of 75 percen.t of the total p.pulation fa void swelling on like omponents. For the in*Spe.tion the entire baffle- of a set of like component.s, it is former assembly. understood that essen.tially 100% a the volumeiarea of each ac-essible like component will be examined.

________________________________

_____________ ____________ ____________ISee___ FiSeerigur22-Page 39

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability (Mechanism) Expansion Link Method/Frequency Examination Coverage Baffle-Former Assembly IPEC Units 2 and Cracking (IASCC, Lower support Baseline volumetric (UT) 100% of accessible bolts. or--as Baffle-former bolts 3 Fatigue) column bolts, examination between 25 and 35 supported by plant specific Aging management Barrel-former bolts EFPY, with subsequent j;ust eatiefi+*

.. A minimum of 75% of (IE and ISR) examination on a ten-year the total population (examined +

Void swelling interval. after- 0 years to unexamined), including coverage effects on this e ."fir..

stability of bolting consistent with the Expansion component is pattern criteria in Table 5-5, must be managed through examined for inspection credit.

management of Heads accessible from the core side.

void swelling on UT accessibility may be affected by the entire baffle- complexity of head and locking former assembly, device designs. 5% of-a eemponents total (aecessible inacc~essible) inspection eao volume will be eaie. , When.

addrfessing a set of like components (e.g., bolting), that the inspection xa in aiminmum sample size of 75 percent of the total population ot like components. For-the inspectiona of a set of like components, it is understood that essentially 100% oe the volumearea of eacdh acess like component will be examined.

_______________________~~~~~~~~~~~~~~

_________________ _______ ______________See Figures 2-5 and 2-6.

Page 40

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian PointEnergy Center Reactor Vessel InternalsInspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability (Mechanism) Expansion Link Method/Frequency Examination Coverage Baffle-Former Assembly IPEC Units 2 and Distortion (Void None Visual (VT-3) examination to Core side surface as indicated.

Assembly 3 Swelling), or check for evidence of distortion, (Includes: Baffle plates, Cracking (IASCC) with baseline examination See Figures 2-6, 2-7, 2-8 and 2-9.

baffle edge bolts and that results in between 20 and 40 EFPY and indirect effects of void

• Abnormal subsequent examinations on a swelling in former plates) interaction with ten-year interval.

fuel assemblies

- Gaps along high fluence baffle joint

- Vertical displacement of baffle plates near high fluence joint

. Broken or damaged edge bolt locking systems along high fluence baffle joint Alignment and Interfacing IPEC Units 2 and Distortion (Loss of None Direct measurement of spring Measurements should be taken at Components 3 Load) height within three cycles of the several points around the Internals hold down spring beginning of the license renewal circumference of the spring, with a Note: This period. If the first set of statistically adequate number of mechanism was not measurements is not sufficient to measurements at each point to strictly identified in determine life, spring height minimize uncertainty. Replaee...ent the original list of measurements must be taken of 301 springs by 103 springs ic age-related during the next two outages, in r.equired when the spring stiffness is degradation order to extrapolate the expected determined to relax beyond de"ig-mechanisms. spring height to 60 years. teleFanee7 ISee Figure 2-10 Page 41

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability (Mechanism) Expansion Link Method/Frequency Examination Coverage Thermal Shield Assembly IPEC Units 2 and Cracking (Fatigue) None Visual (VT-3) no later than 2 100% of thermal shield flexures Thermal shield flexures 3 or Loss of refueling outages from the Materials (Wear) beginning of the license renewal See Figures 2-11 and 2-16 that results in period. Subsequent thermal shield examinations on a ten year flexures excessive interval.

wear, fracture or complete separation Cor a-FPAr-rel Assembly -=-C-Un1its 2 and Gr-aekifig (1ACC None Enhanced visual (E'T-1) 100% of on p id p. f the cesil Upper-and l0oFe 60ore bane -3 Nleutpe exami-nation, fie later-than 2 surfaes of the selected weld and welds Effbr-ialeffent) refueling outages from the adjacent base metal.

beginning of the license renewal period and subsequent See Pi~gue 24-Coc .Bar.el Assembly IPEC Units 2 and Cractkng (LtSCC, None Enhanced visual (EVT 1) 100% of o*n sidt' of the accessibl Lower-core baffel flange -3Nefften examination, no later- than -2 surfaces of the selected weld and weld Embr-ittement) refuieling ouitages fromn the adjacent base metal.

beginning of the license r-enewal (At WEC this weld is the period and subsequent See Figure 2 4 (Core Baffel to lower core barrel to lower examination on a ten yer SU.pport Plate Weld) suppoet castin-g weld. W-EC does not have a lower-core baffel-4nge) ___

Page 42

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Effect Examination Method Item Applicability (Mechanism) Primary Link Examination Coverage Upper Internals IPEC Units 2 Cracking Control rod guide Enhanced visual (EVT-1) 100% of accessible surfaces. A Assembly and 3 (Fatigue, Wear) tube (CRGT) examination, minimum of 75% coverage of the

!J lower flange weld entire examination area or volume, Upper core lvlate Re-insp~ection every 10 yearsramniu or ape sizeieoof 75%

5 following initial inspection, of athe minimum sample total population of like components of the examination is required (including both the accessible and inaccessible portions).

See Figure 2-1 Lower Internals IPEC Units 2 Cracking Control rod guide Enhanced visual (EVT-1) 100% of accessible surfaces. A Assembly and 3 tube (CRGT) examination, minimum of 75% coverage of the Agiag lower flane weld_ entire examination area or volume, Lower support casting Management (TE letiwd Re-inspection every 10 yearsof 75%

in Casting) following initial inspection, of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

See Figure 2-1 (Core Support)

Page 43

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian PointEnergy Center Reactor Vessel InternalsInspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Effect EaiainMto Item Applicability (Mechanism) Primary Link Examination Method Examination Coverage Core Barrel Assembly IPEC Units 2 Cracking (IASCC, Baffle-former Volumetric (UT) examination. 100% of accessible bolts.

Barrel-former bolts and 3 Fatigue) bolts Re-inspection every 10 years Accessibility may be limited by AginIg following initial inspection.with presence of thermal shields. A Management (1E, initial .... a dependent

......... o minimum of 75% coverage of the

.............. i.tiale.amiation depen.e n entire examination area or volume, Void Swelling results of baffle for.mer and....................

Re bol a or a minimum sample size of 75%

an year

..... ina .s. Re deg.raation.at of the total population of like 1 ear.in the.r...once. . ea-....... components of the examination is identified in the primay component, required (including both the accessible and inaccessible portions). The inspection shall examin a iimumn sample size of 75% of the total population of b-ots. Accessibility may be limited by pr-esence of thermal shields 75% of a compenent's total (accessible . inaccessible) inspection area or volume will be examiaed or-, wheit-n adrsiga set of like components (e.g., boltig)-,

that the inspection examnine a minimm sapie size of 75 percent of the total population of like components. For the inspection of a set of like components, itt s understood that essentially 100% of the volume/area of each accessible like ccmponent will be examained.

See Figure 2-5 Page 44

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Effect EaiainMto Item Applicability EMechanism) Primary Link Examination Method Examination Coverage Lower Support Assembly IPEC Units 2 Cracking and (IASCC, Baffle-former 3Fatiue)oltssupported Volumetric (UT) examination. 100% of accessible bolts or as by plant-specific Lower support column and 3 Fatigue) bolts Re-inspection every 10 years justificationu A minimum of 75%

bolts Agin Manaemet (E, following

• ,ol ... initial

• ,,no inspection. with

• ... * .... justificaton coverage of theA mnimum tof entire examination Management (.E......................... area or volume, or a minimum and ISR) results of baffle former bolt examinations. PRe examinations at sample size of 75% of the total dxairadation

. e population of like components of 10 year-intervals

. .... o naedatinn is the examination is required identified i; the primary component. (including both the accessible and inaccessible portions):-e inspection shall examine a minimm saple size of 75 per.ent of the total populatio bolts. 75% of a eomponent'stoa (accessible 1 inaccessible) inspection ar-ea ofvolum!eill be examined or-, when addressing a set of like compntents (e.g., boting)-,

that the inspection examdine a minimum- saple size of 75 percsent of the total populationo like componients. PFo the kipeetion of a set of like ccmpenents, it is underfstood that essentially 100% of the volume/ar-ea of eachae ssibl like componient will be examined.

See Figures 2-12 and 2-13 Page 45

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel inieknals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Item Applicability Effect Primary Link Examination Method Examination Coverage Core Barrel Assembly JPEC Units 2 Cracking (SCC, Upper core barrel Enhanced visual (EVT-1) 100% of one side of the accessible Core barrel fange, and 3 Fatigue) flange weld examination, surfaces of the selected weld and Aadjacent base metal. A minimum Core barrel outlet nozzle Anagen( Re-inspection every 10 years of 75% coverage of the entire welds Management (IE following initial inspection.

• examination area or volume, or a of lower sections) initial examination frequency minimum sample size of 75% of depnetS Uhe exa on.. lati.. the total population of like reu foamrin up cor barrl flng---- components of the examination is Rne examinations at 10 ear intervals required (including both the on.e degradation is identified in he. accessible and inaccessible primary component, portions). 75% of a compenent's total (accessible , i-naccessible) inspection area or vclume ;iAl be examined or, w;hen addressing a set of like compnenifts (e.g., bolting-)-,

that the inspecation examine -a minimum sample size of 75 percent of the total population ot like compoents. For the inspection of a set of like eacmponents, it is under-stood tha essentially 100% of the volume/area of each accessible like component will be examined.

See Figure 2-4 Page 46

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian PointEnergy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Item Applicability Effect Primary Link Examination Method Examination Coverage Core Barrel Assembly IPEC Units 2 Cracking (SCC, Upper and lower Enhanced visual (EVT-1) 100% of one side of the accessible and 3 IASCCQ core barrel examination, surfaces of the selected weld and Upper and lower coregirth adjacent base metal. A minimum barrel cylinder axial welds Agin ld of 75% coverage of the entire Management (IE) welds following initial inspection, 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)

See Figure 2-4 Lower Support Assembly IPEC lower support column Lower support column bodies are cast.

They are (non cast) captured in the next Item of this table.

Page 47

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian PointEnergy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Effect Examination Method Item Applicability (Mechanism) Primary Link Examination Coverage Lower Support Assembly LPEC Units 2 Cracking Control rod guide Enhanced -visual (EVT- 1) 100% of accessible support and 3 (IASCC) tube (CRGT) examination, columns. A minimum of 75%

Lower support column including the lower flanges coverage of the entire examination detection of fRein initial inspection. Re- area or volume, or a minimum (cast) fractured supportoowg at 10 ya interval sample size of 75% of the total columns .on.edegradatio is identified in the population of like components of

.... . .the examination is required MAgng prmEary component. (including both the accessible and Management (fIE) inaccessible portions) 75% ef a component's total (accessible-inaccessible) inspection area-or volume will be examined or, when addressing a set of like components (e.g., belting), that the inspection examine a minimum sample size of 75 percent oft total population of like components. For-the inispection oa untder-stood thatesnily10 of the volume/area of each acesbelike coempoenift willb e*amiaed.

See Figure 2-14 Page 48

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Item Applicability EMect Primary Link Examination Method Examination Coverage Bottom Mounted IPEC Units 2 Cracking Control rod guide Visual (VT-3) examination of BMI 100% of BMI column bodies for Instrumentation System and 3 (Fatigue) tube (CRGT) column bodies as indicated by which difficulty is detected during Bottom-mounted including the lower flanges difficulty of insertion/withdrawal of flux thimble insertion/withdrawal.

Bottom-nted detection of flux thimbles.

instrumentation (BMI) completely Re-inspection every 10 years column bodies _racturdcolumSeeFiure_2-1 fractured column following initial inspection. See Figure 2-15 bodies.

Flux thimble insertion/withdrawal to Abe monitored at each inspection Management (IE) interval. Re examinations at 10 year interfvals o-nce deg~adatian is identified in the primary component.

Upper-

. nterna. WEC.-,its -, GCaekinag (SCC-, G,,t,-l fdgi Enhanced visual (,VT, 1) 100% of asccessible upper core Assembly and 3 Fatigue) tube R-T' examination, with initial plate. 75% of a comp onent's total plwer-langew examination fequency dependent on (acessible 1inaclessible) the examination results f4r CPRGT inspection area or-volume will be lower flange weld. Re e*aminations examined or, when addressing a set at 10 year-intervals onee degradation of like components (e.g., boltin) is identified i-he pr*ifmry that the inspection examine a GMEonipennt. min-imfum- sample size of 75 percent of the total populationo like components. For the intspecstion of a set of like components, it is unesodthat essentially 100% of the volume/area of each accessible like compenent will be examined. See re44 Page 49

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Effect EaiainMto Item Applicability Mechanism) Primary Link Examination Method Examination Coverage LOW.. Support Assembly

.. -UniC--4 EG..L,.. Confal fed guide Enhanic

... d visual

.. (EVT 1) 100% of accessible lower support OWe- suppor casting and43.atigue) tube RGT. examination, with .. initi casting. 75% of a compenent's l.wef flange-.e.. examination frequency dependent 6n total (accessible inaccessible) the examnination results for C=RGT inspection area or-volume will b lower flange Weld. Re examinations examined or, when addressing a set at 10 year intervals once degadation of Wce components (e.g., boting) is idenftified in the primnary that the inspectien examnine a

.. MPe.i.. minimum sample size of 75 percent of the total populationo like components. For the inspection Of a Set Of like components, it is understood that essentially 100% of the volumelarea of each accessible like compcnent will be examined.

See Figure 2 1 (Core Support)

Page 50

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-4 Existing Program Components at IPEC Units 2 and 3 Item Applicability Effect Primsm) Lin! Examination Method Examination Coverage (Mechanism) Reference __________

Core Barrel Assembly IPEC Units 2 Loss of material ASME Code Visual (VT-3) examination All accessible surfaces at Core barrel flange and 3 (Wear) Section XI to determine general ASME Setion specified condition for excessive frequency.

wear.

Upper Internals Assembly N/A--IPEC N/A-Cracking N/A ASME Code N/k- Visual (VT-3) N/A-All accessible surfaces Upper support ring or skirt (-This Units 2 and 3 (SCC, Fatigue) Section XI examination at specified frequency item is N/A beca.use IPEC has a tophat design, therefore there is no support ring or skirt, however the vertical sections of the tophat will be inspected --7 Lower Internals Assembly IPEC Units 2 Cracking (IASCC, ASME Code Visual (VT-3) examination All accessible surfaces at Lower core plate and 3 Fatigue) Section XI of the lower core plates to ASNME Section XI specified Aging Management detect evidence of frequency.

(IE) distortion and/or loss of bolt integrity.

Lower Internals Assembly IPEC Units 2 Loss of material ASME Code Visual (VT-3) All accessible surfaces at Lower core plate and 3 (Wear) Section XI examination. ASNME Sectein XI specified frequency.

Bottom Mounted IPEC Units 2 Loss of material N/A NUREG- Surface (ET) examination. N/A Eddy current surface Instrumentation System and 3 (Wear) 1801 Rev. 1 examination as defined in Flux thimble tubes plant response to IEB 88-09 Alignment and Interfacing IPEC Units 2 Loss of material ASME Code Visual (VT-3) All accessible surfaces at Components and 3 (Wear) Section XI examination. AS..E Seetion I specified Clevis insert bolts frequency.

Alignment and Interfacing IPEC Units 2 Loss of material ASME Code Visual (VT-3) All accessible surfaces at Components and 3 (Wear) Section XI examination. ASM, Sectien XI specified Upper core plate alignment pins frequency.

Page 51

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 IndianPoint Energy Center Reactor Vessel InternalsInspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Expansion Link(s) Expansion Criteria Additional Examination Item Applicability Criteria (Note 1) Acceptance Criteria Control Rod Guide Tube IPEC Units 2 Visual (VT-3) None N/A N/A Assembly and 3 examination.

Guide plates (cards) The specific relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.

Control Rod Guide Tube IPEC Units 2 Enhanced visual (EVT- 1) a. Bottom-mounted a. Confirmation of surface- a. For BMI column bodies, Assembly and 3 examination, instrumentation (BMI) breaking indications in two or the specific relevant column bodies more CRGT lower flange welds, condition for the VT-3 Lower flange welds The specific relevant combined with flux thimble examination is completely condition is a detectable insertion/withdrawal difficulty, fractured column bodies.

crack-like surface shall require visual (VT-3) indication. b. Lower support shallneqire visual(V-column bodies (cast) examination of BtI column upper core plate and bodies by the completion of the b. For cast lower support lower support casting next refueling outage. column bodies, upper core

b. Confirmation of surface- plate and lower support breaking indications in two or casting, the specific more CRGT lower flange welds relevant condition is a shall require EVT- 1 examination detectable crack-like of cast lower support column surface indication.

bodies, upper core plate and lower support casting, within three fuel cycles following the initial observation.

Page 52

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Intekhals Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Expansion Link(s) Expansion Criteria Additional Examination Item Applicability Criteria (Note 1) Acceptance Criteria Core Barrel Assembly

• PEC Units 2 Periodic Eenhanced visual None a. The confirmed detection and a. The specific relevant and 3 (EVT- 1) examination, sizing of a surface-breaking condition for the expansion Upper core barrel flange a. Core barrel outlet indication with a length greater core barrel outlet nozzle weld The specific relevant nozzle welds than two inches in the upper core weld examination is a condition is a detectable barrel flange weld shall require detectable crack-like Upper and lo..er. re b.e. crack-like surface that the EVT- I examination be surface indication.

indication, column bodies (non expanded to include the core

.....

LO"'Of r flagcast) barrel outlet nozzle welds by the b. N/A

.e.

....eld (At uJpEG this;,A i** IPEC lower support completion of the next refueling thpe lowver cere baF-el to column bodies are cast outage.

lower suppoet .asting Weld. b. N/A JPEG does noet have a lower Gore barre! flange)

Gef baFfel flange CorFe barrel outlet nozzle&

Core Barrel Assembly IPEC Units 2 Periodic enhanced visual None None None and 3 (EVT- 1) examination.

Lower core barrel flange weld (At IPEC this weld is The specific relevant the lower core barrel to condition is a detectable lower support casting weld. crack-like surface IPEC does not have a lower indication.

core barrel flange)

Page 53

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian PointEnergy Center Reactor Vessel Interhals InspectionPlan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Expansion Link(s) Expansion Criteria Additional Examination Item Applicability Criteria (Note 1) Acceptance Criteria Core Barrel Assembly IPEC Units 2 Periodic enhanced visual Upper core barrel The confirmed detection and The specific relevant and 3 (EVT- 1) examination. cylinder axial welds sizing of a surface-breaking condition for the expansion Upper core barrel cylinder indication with a length greater upper core barrel cylinder girth welds The specific relevant than two inches in the upper core axial weld examination is a condition is a detectable barrel cylinder girth welds shall detectable crack-like crack-like surface require that the EVT-1 surface indication.

indication. examination be expanded to include the upper core barrel cylinder axial welds by the completion of the next refueling outage.

Core Barrel Assembly IPEC Units 2 Periodic enhanced visual Lower core barrel The confirmed detection and The specific relevant and 3 (EVT-1) examination, cylinder axial welds sizing of a surface-breaking condition for the expansion Lower core barrel cylinder indication with a length greater lower core barrel cylinder girth welds The specific relevant than two inches in the lower core axial weld examination is a condition is a detectable barrel cylinder girth welds shall detectable crack-like crack-like surface require that the EVT- 1 surface indication.

indication. examination be expanded to include the lower core barrel cylinder axial welds by the completion of the next refueling outage.

Baffle-Former Assembly JPEC Units 2 Visual (VT-3) None N/A N/A and 3 examination.

Baffle-edge bolts The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.

Page 54

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Initerhals Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 I A Examination Acceptance E. Additional Examination Item Applicability Criteria (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Baffle-Former Assembly IPEC Units 2 Volumetric (UT) a. Lower support a. Confirmation that more than a and b. The examination and 3 examination, column bolts 5% of the baffle-former bolts acceptance criteria for the Baffle-former bolts actually examined on the four UT of the lower support The examination acceptance criteria for the baffle plates at the largest distance column bolts and the UT of the baffle-former b. Barrel-former bolts from the core (presumed to be the barrel-former bolts shall be lowest dose locations) contain established as part of the bolts shall be established as unacceptable indications shall examination technical part of the examination require UT examination of the justification.

technical justification. lower support column bolts within the next three fuel cycles.

b. Confirmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require UT examination of the barrel-former bolts.

Baffle-Former Assembly IPEC Units 2 Visual (VT-3) None N/A N/A Assembly and 3 examination.

The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, vertical displacement of shroud plates near high fluence joints, and broken or damaged edge bolt locking systems along high fluence baffle plate joints.

Page 55

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Additional Examination Item Applicability Criteria (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Alignment and Interfacing IPEC Units 2 Direct physical None N/A N/A Components and 3 measurement of spring spring height.

Internals hold down The examination acceptance criterion for this measurement is that the remaining compressible height of the spring shall provide hold-down forces within the plant-specific design tolerance.

Thermal Shield Assembly IPEC Units 2 Visual (VT-3) None N/A N/A and 3 examination.

Thermal shield flexures The specific relevant conditions for thermal shield flexures are excessive wear, fracture, or complete separation.

Uppcr intcr-nals Assembl Fkihaneed visual (EVT 1) *4ote NýA NtA WEGUnis

-4. areofeplate Page 56

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 I A Examination Acceptance Additional Examination ItemCriteria (Note 1) Expansion Link(s) Expansion Criteria Acceptance Criteria Low-er Support Assembly EC-E Units 2 Enhanced visual (EVT 1) N/one -/-A N/A Lower- support castin., and-4 exa atiea.

The specific r-elevant cendition is a detectable crack like surface

_ _ _ ~~indieation. __I_

Notes:

1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition Page 57

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian PointEnergy Center Reactor Vessel InternalsInspection Plan Table 5-6 Reactor Vessel Component ISI Program inspection Plan for IPEC Units 2 and 3 Component Extent of Exam Rcaetor Vessel Inter-ior Components and areas as accessible Radial Support Keys Rcaetor Vessel lntcr-ior Components and areas as accessible Bottom Head Instarmenmtatin Nozzles ReaetorItssenls terior Components and areas as accessible Outlet and ILret Nozzle mating sulfaces and Reatr nerVessel interior Components and areas as accessible Upper internal to vessel mating surface wit keys and agcess slats RcaetOr Vessel Interior Components and areas as accessible vessel4 flange suffaee Lower Internals - Exterior Components and areas as accessible Core barrel surface Lower Internals - Exterior Components and areas as accessible Thermal Shield Lower Internals - Exterior Components and areas as accessible Irradiation specimen tubes and guides Lower Internals - Exterior Components and areas as accessible Flexures Lower Internals - Exterior Components and areas as accessible Fasteners and locking devices Page 58

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-6 Reactor Vessel Component ISI Program Inspection Plan for IPEC Units 2 and 3 Code Examination Component Category Method Extent of Exam Lower Internals - Exterior B-N-3 VT-3 Components and areas as accessible Outlet nozzle at 22 deg, 158 deg, 202 deg, and 338 deg Lowr Iternmals Exterior- R-N 3 ITT Cemnpanentc and areas as acerssible Outlet nozzle at+15 deg Lewer Intcrn-eai Extr'erior N VT-3 Components and areas as accessible Outdet nozzle at 22 e Lower Internals Exterior B-N-3 VT3 Components and areas as accessible Outlet nozzle at 33 ate Lower Internals - Exterior Bottom B-N-3 VT-3 Components and areas as accessible Lower core support plate Lower Internals - Exterior Bottom B-N-3 VT-3 Components and areas as accessible Flow distribution plate Lower Internals - Exterior Bottom B-N-3 VT-3 Components and areas as accessible Lower support casting Lower Internals - Exterior Bottom B-N-3 VT-3 Components and areas as accessible Core support column Lower Internals - Exterior Bottom B-N-3 VT-3 Components and areas as accessible Secondary core support Lower Internals - Exterior Bottom B-N-3 VT-3 Components and areas as accessible Instrumentation guides Page 59

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-6 Reactor Vessel Component ISI Program Inspection Plan for IPEC Units 2 and 3 Code Examination Component Categor Method Extent of Exam Lower Internals - Exterior Bottom B-N-3 VT-3 Components and areas as accessible Radial support keys Lower Internals - Interior Bottom B-N-3 VT-3 Components and areas as accessible Outlet nozzle at 22 deg, 158 deg, 202 deg, and 338 deg Lower Internats Intcrior Bottom B-N3 V-T-3 Components and areas as aecessible Outlet nozzle at 159 deg Lowers intcr-nAls -- crn Bottom B-N 3vq Components and areas as aecessible Outilet nozzle at 202 deg________

Lower Iaternau laftcetio Bettem B--N--34 Components and areas as accessible Outlet nozzle at 338 deg Lower Internals - Interior Bottom B-N-3 VT-3 Components and areas as accessible Core barrel alignment pin Lower Internals - Interior Bottom B-N-3 VT-3 Components and areas as accessible Lower core plate Lower Internals - Interior Bottom B-N-3 VT-3 Components and areas as accessible Fuel alignment pins Upper Internals Assembly B-N-3 VT-3 Components and areas as accessible Vertical sections of tophat Core Barrel Assembly B-N-3 VT-3 Components and areas as accessible Core barrel flange Page 60

NL- 12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan Table 5-6 Reactor Vessel Component ISI Program Inspection Plan for IPEC Units 2 and 3 Code Examination Component Categor Method Extent of Exam Alignment and Interfacing Components B-N-3 VT-3 Components and areas as accessible Clevis insert bolts Alignment and Interfacing Components B-N-3 VT-3 Components and areas as accessible Upper core plate alignment pins Page 61

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan UPPER INTERNALS ASSEMBLY Sub Assembly Component Material Category from MRP-SubAssmby CmpnenMteral191 Table 7-2 Anti-rotation studs and nuts Stainless steel A Bolts Stainless steel A C-tubes Stainless steel C Enclosure pins Stainless steel A Upper guide tube enclosures Stainless steel. A Flanges intermediate Stainless steel A Flanges lower Stainless steel A Flexureless inserts Stainless steel A Guide plates/cards Stainless steel C Guide tube support pins (split pins) A X-750 (IP2 only) C Control rod guide Guide tube support pins (split pins) Stainless steel (IP3 only) A tube assemblies and flow downcomers Housing plates Stainless steel A Inserts Stainless steel A Lock bars Stainless steel A Sheaths Stainless steel C Support pin cover plate Stainless steel A Support pin cover plate cap screws Stainless steel A Support pin cover plate locking caps Stainless steel A and tie straps Support pin nuts Alloy X-750 (IP2 only) A Support pin nuts Stainless steel (IP3 only) A Water flow slot ligaments Stainless steel A Mixing Devices Mixing devices CASS A Upper core plate and Fuel alignment pins Stainless steel A fuel alignment pins Upper core plate Stainless steel A Bolting Stainless steel A Brackets,clamps,terminal blocks, and conduit straps Stainless steel A Upper Conduit seal assembly-body, Stainless steel A instrumentation tubesheets conduit and supports Conduit seal assembly-tubes Stainless steel A Conduits Stainless steel A Flange base Stainless steel A Locking caps Stainless steel A Support tubes Stainless steel A flow column bases CASS A UpperUHI UHI flow columns Stainless steel A Page 62

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan UPPER INTERNALS ASSEMBLY Sub Assembly Component Material Category from MRP-191 Table 7-2 Adapters Stainless steel A Bolts Stainless steel A Column bases CASS A Upper support Column bodies Stainless steel A column assemblies Extension tubes Stainless steel A Flanges Stainless steel A Lock keys Stainless steel A Nuts Stainless steel A Bolts Stainless steel A Deep beam ribs Stainless steel A Deep beam stiffeners Stainless steel A Flange Stainless steel A Upper support plate Inverted top hat flange Stainless steel A assembly Inverted top hat upper support plate Stainless steel A Lock keys Stainless steel A Ribs Stainless steel A Upper suport plate Stainless steel A Upper support ring or skirt Stainless steel B LOWER INTERNALS ASSEMBLY Sub Assembly Component Material Category from MRP-191 Table 7-2 Baffle bolting locking bar Stainless steel A Baffle edge bolts Stainless steel C Baffle and former Baffle plates Stainless steel B assembly Baffle former bolts Stainless steel C Barrel former bolts Stainless steel C Former plates Stainless steel B BMI column bodies Stainless steel B BMI column bolts Stainless steel A Bottom mounted BMI column collars Stainless steel B instrumentation BMI column cruciforms CASS B (BMI) column BMI column extension bars Stainless steel A assemblies BMI column extension tubes Stainless steel B BMI column lock caps Stainless steel A BMI column nuts Stainless steel A Core barrel flange Stainless steel B Core barrel Core barrel outlet nozzles Stainless steel B Upper core barrel Stainless steel C Lower core barrel Stainless steel C Diffuser plate Diffuser plate Stainless steel A Page 63

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel InternalsInspection Plan LOWER INTERNALS ASSEMBLY Sub Assembly Component Material Category from MRP-191 Table 7-2 Flux thimble tube plugs - IPEC does not use tube plugs, instead Stainless steel B capped (P2 has 9 tubes apped,* I...

Flux thimbles (tubes) has 0 tubc~pe d)___________

Flux thimbles (tubes) Stainless steel C Irradiation specimen guide Stainless steel A Irradiation specimen Irradiation specimen guide bolts Stainless steel A guides Irradiation specimen lock caps Stainless steel A Specimen plugs Stainless steel A Fuel alignment pins Stainless steel A Lower core plate LCP fuel alignment pin bolts Stainless steel A (LCP) and fuel alignment pins LCP fuel alignment pin lock caps Stainless steel A Lower core plate Stainless steel C Lower support column bodies CASS B Lower support Lower support column bolts Stainless steel B column assemblies Lower support column nuts Stainless steel A Lower support column sleeves Stainless steel A Lower support casting or forging Lower support casting CASS A Thermal shield bolts Stainless steel A Neutron Thermal shield dowels Stainless steel A panels/thermal shield Thermal shield flexures Stainless steel B Thermal shield Stainless steel A Radial support key bolts Stainless steel A Radial support keys Radial support key lock keys Stainless steel A Radial support keys Stainless steel A SCS base plate Stainless steel A SCS bolts Stainless steel A Secondary core support (SCS) SCS energy absorber Stainless steel A assembly SCS guide posts Stainless steel A SCS housing Stainless steel A SCS lock keys Stainless steel A Clevis insert bolts A X-750 B Clevis insert lock keys Stainless steel A Clevis inserts Alloy 600 A Interfacing Head and vessel allignment pin bolts Stainless steel A Components Head and vessel alignment pin lock Stainless steel A caps Head and vessel allignment pins Stainless steel A Internals hold down spring 304 Stainless steel B Upper core plate alignment pins Stainless steel B Page 64

NL-12-037 Docket Nos. 50-247 and 50-286 Attachment 2 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-8 IPEC Response to the NRC Revision I to the Final Safety Evaluation of MRP-227 MRP-227 SER Item IPEC Response SER Section 4.1.1, Topical Report In accordance with SER Section 4.1.1, the upper core plate and the lower Condition I Moving components to support casting have been added to the IPEC "Expansion" inspection category "Expansion" category from "No and are contained in Table 5-3. The components are linked to the "Primary" additional measures" category. component CRGT lower flange weld. The examination method is consistent with the examinations performed on the CRGT lower flange weld.

SER Section 4.1.2, Topical Report In accordance with SER Section 4.1.2, the upper and lower core barrel cylinder Condition 2 Inspection of girth welds and lower core barrel to lower support casting weld have been components subject to irradiation- added to the IPEC "Primary" inspection category and are contained in Table 5-assisted stress corrosion cracking 2. The examination method is consistent with the MRP recommendations for these components, the examination coverage conforms to the criteria described in Section 3.3.1 of the NRC SE, and the re-examination frequency is on a 10-year interval consistent with other "Primary" inspection category components.

The inspection shall be expanded to axial welds (expansion component) in the event that degradation is observed in the girth welds.

SER Section 4.1.3, Topical Report No action required. This item does not apply to components in Westinghouse Condition 3 Inspection of high designed reactors.

consequence components subject to multiple degradation mechanisms SER Section 4.1.4, Topical Report In accordance with SER Section 4.1.4, IPEC will meet the minimum inspection Condition 4 Minimum examination coverage specified in the SER. The appropriate wording has been added to coverage criteria for "expansion" Table 5-3 examination coverage.

inspection category components SER Section 4.1.5, Topical Report In accordance with SER Section 4.1.5, the examination frequency for baffle-Condition 5 Examination former bolts specifies a 10-year inspection frequency following the baseline frequencies for baffle-former bolts inspection in Table 5-2.

SER Section 4.1.6, Topical Report In accordance with SER Section 4.1.6, Table 5-3 requires a 10-year re-Condition 6 Periodicity of the re- examination interval for all Expansion inspection category components once examination of "expansion" degradation is identified in the associated Primary inspection category inspection category components component and examination of the expansion category component commences.

SER Section 4.1.7, Topical Report This.. ndition applies to update of the industry guidelines. No plant-specific Condition 7 Updating of industry action required.

guideline SER Section 4.2.1, The evaluation of design and operating history demonstrating that MRP-227-A Applicant/Licensee Action Item 1 is applicable to IPEC is contained in Section 3.6.

SER Section 4.2.2, The IPEC review of components within the scope of license renewal against the Applicant/Licensee Action Item 2 information contained in MRP-191 Table 4-4 is discussed in Section 3.6.

SER Section 4.2.3, The IPEC discussion regarding guide tube support pins (split pins) is contained Applicant/Licensee Action Item 3 in Section 3.6.

SER Section 4.2.4, No action required. This item does not apply to Westinghouse designed units.

Applicant/Licensee Action Item 4 SER Section 4.2.5, The IPEC discussion regarding hold down springs is contained in Section 3.6.

Applicant/Licensee Action Item 5 SER Section 4.2.6, No action required. This item does not apply to Westinghouse designed units.

Applicant/Licensee Action Item 6 SER Section 4.2.7, The IPEC discussion regarding lower support column bodies is contained in Applicant/Licensee Action Item 7 Section 3.6.

SER Section 4.2.8, The submittal of information for staff review and approval is discussed in Applicant/Licensee Action Item 8 Section 3.6.

Page 65