NL-17-020, License Renewal Application - Revisions to Reactor Vessel Internals Aging Management Program and Inspection Plan

License Renewal Application - Revisions to Reactor Vessel Internals Aging Management Program and Inspection Plan

ML17047A541
Person / Time
Site:	Indian Point
Issue date:	02/06/2017
From:	Vitale A Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-17-020
Download: ML17047A541 (78)

Text

• ~*Entergx Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel (914) 254-6700 Anthony J Vitale Site Vice President NL-17-020 February 6, 2017

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• U.S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738

SUBJECT:

License Renewal Application - Revisions to Reactor Vessel Internals Aging Management Program and Inspection Plan 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 NL-07-039, "Indian Point Energy Center License .Renewal Application" (Apr. 23, 2007) (ML071210507) *

• 2) Entergy Letter.NL-10-063, "Amendment 9 to License Renewal Application.

(LRA) - Reactor Vessel Internals Program" (July 14, 2010) (ML102010102) 3). Entergy Letter NL-1"1-107, "License Renewal Application-:- Completion of

• Commitment# 30 Regarding.the Reactor Vessel Internals Inspection Plan" (Sepf 28, 2011) (ML11280A121)

4) Electric Power Research Institute, MRP-227-A, "Materials 'Reliability Program: Pressurized Water.Reactor Internals Inspection and Evaluatio.n Guidelines" (Dec. 2011) (ML120170453) '

5) *Entergy Letter NL-12~037, "License Renewal Application - Revised Reactor Vessel Internals Program and Inspection Plan. Compliant with MRP-227-A" (Feb. 17, 2012) (ML12060A312)

6) NUREG-1930, Supp. 2, "Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3" (Nov.

2014) (ML15188A383)

7) Entergy Letter NL-16-053, "License Event Report# 2016-004-00, Unanalyzed Condition Due to Degraded Reactor Baffle-Former Bolts" (May 31, 2016) (ML16159A219)

Dear Sir or Madam:

By letter dated April 23, 2007 (Reference 1}, Entergy Nuclear Operations, Inc. (Entergy) submitted an application pursuant to 10 CFR Part 54 and 10 CFR Part 51, to renew the operating licenses for Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3}, for review by the U.S. Nuclear Regulatory Commission (NRC). Entergy provided a description of the Indian Point Energy Center (IPEC) Reactor Vessel Internals (RVI) aging management program (AMP) in Amendment 9 to the License Renewal Application (LRA) (Reference 2).

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NL-17-020 Docket Nos. 50-247 and 50-286 Page 2 of 3 I

Consistent with License Renewal Commitment 30, Entergy submitted its Reactor Vessel Internals (RVI) Inspection Plan on September 28, 2011, two years prior to entering the period of extended operation (PEO) for IP2 (Reference 3). Although only IP2 was within two years of entering the PEO at that time, the RVI Inspection Plan covered both units.

Entergy submitted the RVI Inspection Plan based on the new aging management program (AMP) in NUREG~1801, Revision 2. After the Electric Power Research Institute (EPRI) issued the NRG-approved generic industry aging management guidance for RVls in MRP-227-A (Reference 4), Entergy submitted a revised RVI AMP and Inspection Plan for both IP2 and IP3 based on MRP-227-A on February 17, 2012 (Reference 5). Following Entergy's submission.of additional technical information in response to Staff requests for additional information, the Staff approved Entergy's revised RVI AMP and Inspection Plan, as documented in Safety Evaluation Report Supplement 2 issued in November 2014 (Reference 6).

During the Spring 2016 IP2 refueling outage (2R22), Entergy performed ultrasonic (UT) examinations and/or visual inspections of all 832 baffle-former bolts (bolts) in accordance with the MRP-227-A guidelines. As a result of the inspection findings, Entergy replaced all 227 bolts with actual and assumed indications. It also replaced an additional 51 bolts to reduce the probability of future failures as well as minimize-the probability of clusters of failed bolts, resulting in a total of 278 replaced bolts. See Reference 7.

As a result of the IP2 inspection findings and other industry op~rating experience (OE) indicating a significant number of failed bolts at other similarly-designed PWR plants, Entergy recently revised portions of the Indian Point Energy Center (IPEC) PWR Vessel Internals Program (SEP-PVl-IPEC-001, Rev. 1). The revisions included the addition of new Section 6.2 to incorporate discussion of the recent Unit 2 OE described above, including Entergy's related aecision to arrange for offsite fractographic examination of eight baffle-former bolts removed from the IP2 baffle structure during the Spring 2016 outage. The revisions also reflect changes to Entergy's schedule and plans for conducting future UJ and visual inspections as well as replacement of baffle-former bolts at I P2 and IP3. *

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The purpose of this submittal is to make corresponding revisions to the IPEC RVI AMP and IPEC Reactor Vessel Internals Plan, as submitted to the NRC on February 17, 2012. The revisions include: (1) an updated discussion of IPEC and industry operating experience involving baffle-former bolts in Section B.1.42 (Reactor Vessel Internals Program) of the LRA; (2) the addition of new Section 6.0 (Operation Experience and Additional Considerations), which coinCides with current Section 6.0 of SEP-PVl-IPEC-001, to the IPEC Reactor Vessel Internals Plan; and (3) a related revision to Table 5-2 (Primary Components at IPEiC Units 2 and 3) of the IPEC Reactor Vessel Internals Plan to cross-reference newly-added Section 6.2 (Spring 2016 Operating Experience) and the IPEC-specific baffle-former bolt examination/inspection plans discussed therein. ,

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.

NL-17-020 Docket Nos. 50-247 and 50-286 Page 3 of 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on

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~/CJ:,

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,2017.

Sincerely, Attachments: 1. Indian Point Energy Center Revised Reactor Vessel Internals Program, .

2. Indian Point Energy Center Revised Reactor Vessel Internals Inspection Plan cc: Mr. Daniel H. Dorman, Regional Administrator, NRC Region I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel Mr. William Burton, NRC Senior Project Manager, Division of License Renewal Mr. Douglas Pickett, NRA Senior Project Manager Ms. Bridget Frymire, New York State Department of Public Service Mr. John B. Rhodes, President and CEO NYSERDA NRC Resident Inspector's Office

ATTACHMENT 1 TO NL-17-020 INDIAN POINT ENERGY CENTER REVISED REACTOR VESSEL INTERNALS PROGRAM Additions Underlined Deletions Lined Out ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286

NL-17-020 Attachment 1 Page 1of10 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 (MAP). The MAP inspection and evaluation (l&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 MAP also developed inspection requirements specific to the inspection methods delineated in MRP-227-A, as well as requirements for qualification of the nondestructive examination (NOE) 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. 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 dimensiont!-1 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 resolved in accordance with the NEI 03-08 implementation protocol. The RVI Program will be implemented prior to the period of extended operation. ,

NL-17-020 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 (l&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 (NOE) 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. 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 crac:king, loss of material, reduction of fracture toughness, loss of preload and dimensional changes of vessel internals components in accordance with MRP-227-A inspection require,ments 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, <;>r good practice implementation requirements established in MRP-227-A or MRP-228, will be resolved in acc:ordance with the NEI 03-08 implementation protocol. The RVI Program will be implemented prior to the period of extended operation.

NL-17-020 Attachment 1 Page 3 of 10 B.1.42 Reactor Vessel Internals Program Program Description The Reactor Vessel Internals (RVI) Program is a new plant-specific program. Revision 1 of NUREG-1801 (Reference B.2-1) includes no aging management program description for PWR reactor vessel internals. NUREG-1801, Section Xl.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 (l&E) guidelines for managing the effects of aging on pressurized water reactor vessel internals. These guidelines, as reviewed and accepted by the NRC (Reference B.2-2), are presented in MRP-227-A, "Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The l&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 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 (NOE) systems used to perform those inspections. These inspection procedure requirements are presented in MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals."

Revision 2 of NUREG-1801 (Reference B.2-3) contains a new aging management program (Xl.M16A) addressing PWR RVls. This new aging management program relies on the implementation of EPRI report MRP-227, and applies the guidance in that document. In 2013, the NRC Staff issued Final License Renewal Interim Staff Guidance LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors (Reference B.2-4), which revised the recommendations in NUREG-1801. Revision 2

NL-17-020 Attachment 1 Page 4 of 10 and the NRC Staff's acceptance criteria and review procedures to ensure consistency with MRP-227-A and provide a framework to adequately address age-related degradation and aging management of RVI components during the period of extended operation.

MRP-227-A and MRP-228 provide the basis of the IPEC Roaster Vessel Internals (RVlt Program. Revisions to MRP-227-A and MRP-228 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 eyaluate 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 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, ne~ded, 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-§a). 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 i~spections,

• Criteria for sampling and coverage,

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

• Inspection acceptance criteria,

• Methods for evaluating ~xamination results not meeting the acceptance criteria,

• Provisions for updating the pr9gram based on industry-wide resuHs, 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

NL-17-020 Attachment 1 Page 5 of 10 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-227-A);.

• 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

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in Tables 4-3,~4-6 or 4-9 of MRP-227-A).

The categorization of internals components for Westinghouse PW Rs, 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 hionitoring 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), primary 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.

NL-17-020 Attachment 1 Page 6 of 10

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 forth the parameters monitored by the IPEC RVI Program inspections.

The program will use NOE techniques to detect loss of material through wear, identify changes in dimension due to void swelling and irradiation growth, distortion, or deflection, and locate cracks induced by SCC, PWSCC, IASCC, or fatigue/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 aging 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.

Vis.ual 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 broke11 components and crack-like surface flaws of components and welds. Volumetric (ultrasonic) examinations will be used to locate cracking of bolting.

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 NOE 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 delineated in MRP-227-A, Table 4-3. Indications from EVT-1 or UT inspections may result in additional inspections of expansion group components, as determined by expansion criteria delineated in MRP-227-1\, Table 5-3. The relationships between primary group component inspection findings and additional inspections of expansion group components are as follows described in MRP-227A, Table 4-6.

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

NL-17-020 l,

Attachment 1 Page 7 of 10 I

expansion beyond the primary group. Records of inspection results are maintained allowing for comparison with subsequent inspection results.

• In accordance with MRP-227-A, IPEC will provide a summary report of all inspections and monitoring, items requiring evaluation, and new repairs to the MAP Program Manager. The IPEC-specific 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 from provided in Section 5 of MRP-227-A. Table 5-3 and Sections 5.1 through 5.3 of MRP-227-A provide 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; (iii) requirements for system-level assessment of bolted assemblies with unacqeptable 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, a*nd 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.

Any detected condition that 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 MAP- .

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 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

NL-17-020 Attachment 1 Page 8of10 This attribute is discussed in Section B.0.3.

10. Operating Experience From an overall fleet-wide perspective, Rrelatively few incidents of PWR internals aging degradation have been reportep in operating U.S. commercial PWR plants. However, PWR internals aging degradation has been observed in European PWRs, (and. more recently, in certain U.S. 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 th_e UT excimination techniques used in.

the inspections. - -

Since the Spring of 2016. baffle-former bolt degradation has been detected at some operating

• plants in the United States. As a result of the baffle-former bolt inspection findings.

Westinghouse issued Nuclear Safety Advisory Letter NSAL-16-1 which contains a description I

of the issue. a technical evaluation. and recommended actions for utilities to follow. This NSAL -

recommended that tier 1a plants (Le. Westinghouse; 4-loop. downflow plants with 347 stainless steel bolting) similar to the IPEC Units should perform volumetric examinations of the baffle-former bolts at the next refueling outage. As a result. IP3 moved the inspections from the 2019 outage to the 2017 outage to comply with the NSAL recommendations as well as the Interim Guidance issued by the EPRI MRP Program in MRP 2016-022. Transmittal of NEl-'-03-08 "Needed" Interim Guidance Regarding Baffle Former Bolt inspections for Tier 1 plants as Defined in Westinghouse NSAL 16-01.

In the spring of 2016. during IP2 outage 2R22. ultrasonic (UT) and/or visual inspections of all 832 baffle-former bolts (bolts) were performed in accordance with the NRC approved guidelines in MRP-227-A. Visual-inspection of the baffle plates and bolts identified 31 degraded bolts. The UT inspections identified indications on 182 bolts and also determined that 14 bolt locations were not testable. The locations that were not testable were conservatively assumed to

- possess*bolts that failed to meet the acceptance criteria. As a result of the inspection findings.

all 227 bolts (31+182+14) with actual and assumed indications were replaced .. An additional 51 bolts were replaced to reduce the probability of future failures as well as minimize the probability of clusters of failed bolts. Therefore. during 2R22. a total of 278 bolts (227+51) were replaced.

As a result of the IP2 inspection findings and other industry Operating Experience (OE) indicating a significant number of fafled bolts at other similarly-designed PWR plants. the IPEC PWR Vessel Internals Program. SEP-PVl-IPEC-001 was revised. In view of the 2R22. -

inspection findings. Entergy arranged for the fractographic examination of eight baffle-former bolts removed from the IP2 baffle structure during the Spring 2016 outage at Westinghouse Electric Company's hot cell laboratory in Churchill, PA. The results of those fractographic examinations are documented in Westinghouse Report MCOE-TR-16-18. Revision 0.

"Fractography of Indian Point Unit 2 Baffle Former Bolts" (Nov. 30. 2016). Industry-sponsored metallurgical analysis and materials property testing of additional baffle former bolt specimens -

from IP2 and other PWRs is still in progress.

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

NL-17-020 Attachment 1 Page 9 of 10 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 XI Code inspections. Internals inspections (VT-3) have been conducted at IPEC in accordance with ASME Section XI 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 incidents in both U.S. and foreign plants were considered in the deyelopment 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-17-020 Attachment 1 Page 10 of 10 B.2 REFERENCES B.2 1 NUREG 1800, Standard Review Plan for Review of Lioense RenO'tval Applioations for Nuolear Power Plants, U.S. t>Juolear Regulatory Commission, September 2005.

B.2-12 NUREG-1801, Revision 1, Generic Aging Lessons Learned (GALL) Report, U.S. Nuclear Regulatory Commission, September 2005.

B.2-2 Letter from R. Nelson. U.S. Nuclear Regulatory Commission. to N. Wilmshurst. Electric Power Research Institute,' Final Safety Evolution of EPRI Report. Materials Reliability Program Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor (PWR)

Internals Inspection and Evaluation Guidelines (TAC No. ME0680), June 2011 B.2-3 NUREG-1801. Revision 2 Generic Aging Lessons Learned (GALL) Report, U.S. Nuclear Regulatory Commission. December 201 o

• B-2-4 Final Interim Staff Guidance LR-ISG-2011-04; Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors. 78 Fed. Reg.

33, 120 (June 3, 2013)

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

. ATTACHMENT 2 TO NL-17-020 INDIAN POINT ENERGY CENTER REVISED REACTOR VESSEL INTERNALS INSPECTION PLAN Additions Underlined Deletions Lined Out ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286

NL-17-020 Attachment 2 Page 1 of 63 Indian Point Energy Ceriter Reactor Vessel Internals Inspection Plan 1

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 guidelines 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 subsequent inspections and identified the components and locations for supplemental examination.

This inspection plan includes:

• Identification of items for inspection,

• Specification of the type of exanlination 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.

Page 1

NL-17-020 Attachment 2 Page 2 of 63 Indian Point Energy Center Reactor Vessellntemals Inspection 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-17-020 Attachment 2 Page 3 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan ROD TRAVEL HOUSING CONTROL ROD INSTRUMENTATION DRIVE MECHANISM PORTS THERMAL SLEEVE UPPER SUPPORT PLATE LIFTING LUG INTERNALS / CLOSURE HEAD SUPPORT / ASSEMBLY LEDGE HOLD-DOWN SPRING CORE BARREL CONTROL ROD GUIDE TUBE SUPPORT COLUMN CONTROL ROD DRIVE SHAFT UPPER CORE PLATE OUTLET NOZZLE INLET NOZZLE BAFFLE RADIAL CONTROL ROD SUPPORT CLUSTER (WITHDRAW!

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-17-020 Attachment 2 Page 4 of 63 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 Page4

NL-17-020 Attachment 2 Page 5 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan control power generation. Guide tube cards are located within each control rod guide tube to guide the absorber rods. The control rod guide tubes are also slotted in their lower sections 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 comers by edge bolts. In addition, at IPEC, comer 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.

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NL-17-020 Attachment 2 Page 6 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection 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.

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

Page 6

NL-17-020 Attachment 2 Page 7 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan IPEC does not have CRGT flexures Lower Flange Weld CRGT Split Pins Figure 2-3 Typical Westinghouse control rod guide tube assembly Page 7

NL-17-020 Attachment 2 Page 8 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Flange Weld Axial Weld 0 .

Upper Core Barrel to Lower Core Barrel Circumferential Weld Lower Barrel Axial Weld

.......

Lower 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-17-020 Attachment 2 Page 9 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan BAFFLE PLAlE EDGE BOLT BAFFLE TO F'OR.MER BOLT (I.ONG .I: SHOR'J)

CORNER EDGE BRACKET BAFFLE TO FORMER BOLT Figure 2-5

. Bolt locations in typical Westinghouse baffle-former-barrel structure.

Page 9

NL-17-020 Attachment 2 Page 10 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan 00000 0

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

NL-17-020 Attachment 2 Indian Point Energy Center Page 11 of 63 Reactor Vessel Internals Inspection Plan High Fluence Seams Figure 2-7 High fluence seam locations in Westinghouse baffle-former assembly Page 11

NL-17-020 Attachment 2 Page 12 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Potential Gaps at Baffle-Former Plate Levels Potential Bowing Along High Fluence Seam Figure 2-8 Exaggerated view of void swelling induced distortion in Westinghouse baffle-former assembly.

Page 12

'

NL-17-020 Attachment 2 Page 13 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Figure 2-9 Vertical displacement of Westinghouse baffle plates caused by void swelling.

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NL-17-020 Attachment 2 Page 14 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan

• TOP SUPPORT.:PlATE':

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CORE~ARREL Figure 2-10 Schematic cross-sections of the Westinghouse hold-down springs Page 14

NL-17-020 Attachment 2 Page 15 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan o . . *o* . . ~

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Thermal Shield

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Thermal Shield Flexure Core Support Figure 2-11 Location of Westinghouse thermal shield flexures Page 15

NL-17-020 Attachment 2 Page 16 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Lower Core Support Structure Figure 2-12 Schematic indicating loc;_ation of Westinghouse lower core support structure. Additional details shown in Figure 2-13 LOWER CORE PLATE DIFFUSER PLATE CORE SUPPORT PLATE/FORGING CORE~

SUPPORT COLUMN \ _ BOTIOM 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-17-020 Attachment 2 Page 17 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Figure 2-14 Typical Westinghouse core support column. Core support column bolts fasten the top of the support column to the lower core plate

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Figure 2-15 Examples of Westinghouse bottom mounted instrumentation column designs Page 17

NL-17-020 Attachment 2 Page 18 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Figure 2-16 Typical Westinghouse thermal shield flexure Page 18

NL-17-020 Attachment 2 Page 19 of 63 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 exp~rience 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 .

.3.1 Component Inspection and Evalu.ation 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 tile screening criteria, so that aging degradation significance is minimal. Primary, expansion, and 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.

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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 of 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 methods required for Primary and Expansion components W((re 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 inspectio~ technical justification or inspection personnel training (equirements.

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 MRP7227-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.

The Reactor Vessel Component Inspections conducted as part of the ISi Program for IPEC Units 2 and 3 are listed in Table 5-6. The ISi Program inspections are implemented in accordance ~ith ASME Section XI schedule requirements.

3.5 Examination Systems Equipment, techniques, procedures and personnel used to perform examinations required under this program will be consistent with the requirements of MRP-228. Indications detected during Page 20

NL-17-020 Attachment 2 Page 21 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan these examinations will be characterized and reported in accordance with the requirements 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 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. These items have been

~ddressed in the appropriate sections of this document. 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 1

replacement engineering and walkdowns in 2014 and replace the split pins in 2016.

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 Page 21

NL-17-020 Attachment 2 Page 22 of 63

• Indian Point Energy Center Reactor Vessel Internals Inspection Plan 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 t_he hold down springs under all licensing basis conditions will be developed prior to the first required physical measureme~t. The 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 aq;eptance criterion is not developed for the hold down spring, IPEC ~ll 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 du.ring the period of extended operation will consider the possible loss of fracture toughness in these components due to thermal and i~adiation. embrittlement. The analyses will bt1 consistent with the IPEC licensing basis and the 'neeg ~o maintain the .

• functionality of the lower support column bodies under all licensing basis conditions of

.operation. IPEC will submit this information to the NRC prior to the perl.od of extended operation.

SER Section 4.2.8, Applicant/Licensee Action Item 8 A Reactor Vesse.l Intem~ls AMP description for IPEC was included in Amendment 9 to the

.Licerise Renewal Application (NL-10-063, July 14, 2010). The AMP descriptipn has been revised to.be consistent with MRP-227-A. The revised AMP description has been submitted under letter NL-12-03 7.

This document comprises an Inspection plan which addres~es 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.

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NL-17-020 Attachment 2 Page 23 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan 4

EXAMINATION ACCEPTANCE AND EXPANSION CRITERIA AND IMPLEMENTATION REQUIREMENTS 4.1 Ex,amination 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 gro~p, these general relevant conditions are sufficient.

However, for components where visual (VT-3) is specified in thy 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.

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NL-17-020 Attachment 2 Page 24 of 63 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 examina~ion 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 ~ispositioned. 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 24

NL-17-020 Attachment 2 Page 25 of 63 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 area have 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 4.2.l Consistent with the requirements bf 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, br 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 25

NL-17-020 Attachment 2 Page 26 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan 1 an internal independent review is performed and that concurrence is obtained from the responsible utility executive.

1 4.2.2 Examinations contained in this inspection plan shall be conducted in accordance with MRP-228.

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

• 4.2.4 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.

4.2.5 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 Qf MRP-227-A

• are examined.

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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-Table4-4 Table 5-8 IPEC Response to the NRC Revision 1 to the Final Safety Evaluation of MRP-227 Page 27

NL-17-020 Attachment 2 Page 28 of 63 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-191Table4-4 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 (Tables 3-3, 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 and 4-9)

Core Barrel Flange P~ge 28

NL-17-020 Attachment 2 Page 29 of 63 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 Core Barrel Assembly -

Assembly - Core Upper and Lower Core Core Barr~l Assembly - Shell Barrel Barrel Cylinder Girth Welds (Tables 3-3, 4.:3 and Core Barrel Assembly-Thermal Upper Core Barrel 2 places in 5-3)

Shield

- Lower Core Barrel Core Barrel Assembly -

Upper and Lower Core Lower Internals Barrel Cylinder Axial

• Assembly-Neutron Welds (Tables 3-3, 4-6 and panels/thermal. shield 2 places in 5-3)

Thermal shield 7 Core Barrel Assembly - Lower Lower Internals Core Barrel Assembly -

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

Core Barrel Flange Weld Core Barrel Flange Core Barrel Assembly -

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

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

Core Barrel Outlet Nozzles 9 Lower Internals Assembly - Interfacing

• Alignment ahd Interfacing Clevis Insert Bolt Components - Components - Clevis Insert Interfacing Bolts (Tables 3-3 and 4-9)

Components Clevis Insert Bolts Page 29

NL-17-020 Attachment 2 Page 30 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan 1*

Table 5-1 Indian Point 2 & 3 Compc;>nent Cross Reference Item Letter NL-10-063 Component MRP-191Table4-4 MRP-227-A 10 Lower Internals Assembly - Interfacing No additional measures Clevis Insert Components -

Interfacing Components Clevis Inserts 11 Lower Internals Assembly - Lower Internals No additional measures

• Intermediate Diffuser Plate Assembly - Diffuser Plate Diffuser Plate 12 Lower Internals Assembly - Fuel Lower Internals No 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 (Tables Core Plate and Fuel 3-3, and 2 places in 4-9)

Alignment Pins Lower Core Plate Page 30

NL-17-020 Attachment 2 Page 31of63 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-191Table4-4 MRP-227-A 14 Lower Internals Assembly - Lower Internals Lower Internals Assetnbly Assembly - Lower - Lower Support Casting

• Lower Core Support Castings Support Casting or (Tables 3-3, and 4-6)

Forging

• Column Cap Lower Support

• Lower Core Support Column Casting Bodies Lower Internals Assembly-Lower Support Column Assembly No additional measures Lower Support Column Nuts Lower Support Assembly-Lower Support Column Lower Support Bodies (Cast) (Tables 3-3 Column Bodies 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 No additional measures Lower Core Support Plate Assembly - Lower Colmpn Sleeves Support Column Assembly Lower Support Column Sleeves

• Page 31

NL-17-020 Attachment 2 Page 32 of 63 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 17 Lower Internals Assembly~ Lower Internals No additional measures

,..J-Radial Key Assembly- Radial Support Keys Radial Support Keys 18 Lower Internals Assembly- Lower Internals . I No additional measures Secondary Core Support Assembly- Secondary Core Support (SCS)

Assembly SCS Base Plate 19 RCCA Guide Tube Assembly - Upper Internals No 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 32

NL-17-020 Attachment 2 Page 33 of 63 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 22 RCCA Guide Tube Assembly - Upper Internals No 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 and 4-9)

Upper Core Plate Alignment Pins 24 Head/Vessel Alignment Pin Interfacing No 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 No additional measures Assembly - Mixing

- Support Column Orifice Base Devices

- Support Column Mixer Mixing devices Page 33

NL-17-020 Attachment 2 Page 34 of 63

• 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 No additional measures Assembly- Upper Support Column Assemblies Column Bodies 28 Upper Core Plate, Fuel Alignment Upper Internals No additional measures Pin .Assembly- Upper Core Plate and Fuel Alignment Pins Fuel Alignment Pins 29 Upper Support Plate, Support Upper Internals Assembly (Including Ring) Assembly- Upper Support Plate No additional measures for Assembly the upper support plate 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 No 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 34

NL-17-020

.Attachment 2 Page 35 of 63 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-191Table4-4 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 No additional measures Assembly - Upper Instrumentation Conduit and Support Conduits Page 35

NL-17-020 Attachment 2 Page 36 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3

• Effect Examination Item Applicability Expansion Link Examination. Coverage (Mechanism) Method/Frequency Control Rod Guide Tube IPEC Units 2 and Loss of Material None Visual (VT-3) examination no 20% examination of the Assembly 3 (Wear) later than 2 refueling outages number of CRGT Guide plates (cards)  : from the beginning of the assemblies, with all guide license renewal period. cards within each selected

.. Subsequent examinations are CRGT assembly examined .

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)

Assembly 3 Fatigue) instrumentation examination to determine the CRGT lower flange weld Lower flange welds Aging Management (BMI) column presence of crack-like surface surfaces and adjacent base (IE and TE) bodies, flaws in flange welds no later metal on,the individual

• Lower support than 2 refueling outages from periphery CRGT column bodies the beginning of the license assemblies. A minimum of (cast) renewal period and subsequent 75% of the total identified Upper core plate examination on a ten-year sample population must be Lower support interval. examined.

casting

- See Figure 2-3

/

Page 36

NL-17-020 Attachment 2 Page 37 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability Expansion Link Examination Coverage (Mechanism) Method/Frequency Core Barrel Assembly IPEC Units 2 and Cracking (SCC) Core barrel outlet Periodic enhanced visual (EVT- I 00% of one side of the Upper core barrel flange weld 3 nozzle welds 1) examination, no later than 2 accessible surfaces of the refueling outages from the selected weld and adjacent beginning of the license renewal base metal. A minimum of period and subsequent 75% of the total weld length examination on a ten-year (examined+ unexamined),

interval. including coverage consistent with 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, Upper and lower Periodic enhanced visual (EVT- I 00% of one side of the Upper and lower core barrel 3 IASCC, Fatigue) core barrel cylinder 1) examination, no later than 2 accessible surfaces of the cylinder girth welds axial welds refueling outages from the selected weld and adjacent beginning of the license renewal base metal. A minimum of

' period and subsequent 75% of the total weld length examination on a ten-year (examined+ unexamined),

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

See Figure 2-4 Page 37

NL-17-020 Attachment 2 Page 38 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability Expansion Link Examination Coverage (Mechanism) Method/Frequency Core Barrel Assembly IPEC Units 2 and Cracking (SCC, None Periodic enhanced visual (EVT- I 00% of one side of the Lower core barrel flange weld 3 Fatigue) I) examination, no later than 2 accessible surfaces of the refueling outages from the selected weld and adjacent (AtJPEC this weld is the lower beginning of the license renewal base metal. A minim:um of core barrel to lower support period and subsequent 75% of the total weld length casting weld. IPEC does not examination on a ten-year (examined+ unexamined),

have a lower core barrel flange) interval. including coverage consistent with the Expansion criteria in Table 5-5, must be examined from either the inner or outer diameter for inspection

' credit.

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

Page 38

NL-17-020 Attachment 2 Page 39 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability Expansion Link Examination Coverage (Mechanism) Method/Frequency Bame-Former Assembly IPEC Units 2 and Cracking (IASCC, None Visual (VT-3) examination, with Bolts and locking devices Baffle-edge bolts 3 Fatigue) that results baseline examination between on high tluence seams.

m 20 and 40 EFPY and subsequent I 00% of components

• Lost c;ir broken examinations on a ten-year accessible from core side.

locking devices interval. A minimum of 75% of the

• Failed or missing total population (examined bolts + unexamined), including

• Protrusion of bolt coverage consistent with the

' heads Expansion criteria in Table Aging Management 5-5, must be examined for (IE and ISR) inspection credit.

Void swelling effects on this See Figure 2-5 component is managed through management of

• void swelling on -

the entire baffle-former assembly.

Page 39

NL-17-020 Attachment 2 Page 40 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability Expansion Link Examination Coverage (Mechanism) Method/Frequency Bame-Former Assembly IPEC Units 2 and Cracking (IASCC, Lower support Baseline volumetric (UT) 100% of accessible bolts. A Baffle-former bolts 3 Fatigue) column bolts, examination between 25 and 35 minimum of75% of the Aging management Barrel-former bolts EFPY, with subsequent total population (examined (IE and JSR) examination on a ten-year + unexamined), including Void swelling interval. coverage consistent with the effects on this Expansion criteria in Table component is See additional IPEC sgecific 5-5, must be examined for managed through examination reguirements in inspection credit. Heads management of Section 6.2. accessible from the core void swelling on side. UT accessibility may the entire baffle- be affected by complexity former assembly. of head and locking device designs.

See Figures 2-5 and 2-6.

Page 40

NL-17-020 Attachment 2 Page 41 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination

• Item Applicability Expansion Link Examination Coverage (Mechanism) Method/Frequency Bame-Former Assembly IPEC Units 2 and Distortion (Void None Visual (VT-3) examination to Core side surface as Assembly 3 Swelling), or check for evidence of distortion, indicated.

(Includes: Baffle plates, baffle Cracking (IASCC) with baseline examination edge bolts and indirect effects of that results in between 20 and 40 EFPY and See Figures 2-6, 2-7, 2-8 void swelling in former plates) *Abnormal subsequent examinations 9n a and 2-9.

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 flue*nce baffle joint Alignment and Interfacing IPEC Units 2 and Distortion (Loss of None Direct measurement of spring Measurements should be Components 3 Load) height within three cycles of the taken at several points Internals hold down spring beginning of the license renewal around the circumference of Note: This period. If the first set of the spring, with a mechanism was not measurements is not sufficient to statistically adequate strictly identified in determine life, spring height number of measurements at the original list of measurements must be taken each point to minimize age-related during the next two outages, in uncertainty.

degradation order to extrapolate the expected mechanisms. spring height to 60 years. See Figure 2-10 Page 41

NL-17-020 Attachment 2 Page 42 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-2 Primary Components at IPEC Units 2 and 3 Effect Examination Item Applicability Expansion Link Examination Coverage (Mechanism) Method/Frequency Thermal Shield Assembly IPEC Units 2 and Cracking (Fatigue) None Visual (VT-3) no later than 2 l 00% of thermal shield Thermal shield tlexures 3 or Loss of refueling outages from the tlexures Materials (Wear) beginning of the license renewal

_)

that results in period. Subsequent See Figures 2-11 and 2-16 thermal shield examinations on a ten year tlexures excessive interval.

wear, fracture or complete separation Page 42

. NL-17-020 Attachment 2 Page 43 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Effect Examination Method Item Applicability Primary Link Examination Coverage (Mechanism)

Upper Internals Assembly IPEC Units 2 Cracking Control rod guide Enhanced visual (EVT-1) 100% of accessible and 3 (Fatigue, Wear) tube (CRGT) examination. surfaces. A minimum of Upper core plate lower flange weld 75% coverage of the entire Re-inspection every IO years examination area or

-- following initial inspection.

volume, or a minimum sample size of 75% of the total population of like comp.onents of the examination is required

- (including both the accessible and inaccessible portions).

See Figure 2-1 Lower Internals Assembly IPEC Units 2 Cracking Control rod guide Enhanced visual (EVT-1) 100% of accessible and 3 tube (CRGT) examination. surfaces. A minimum of Lower support casting Aging lower flange weld 75% coverage of the entire Management (TE Re-inspection every IO years examination area or in Casting) following initial inspection.

- 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 2-1 (Core Support)

Page 43

NL-17-020 Attachment 2 Page 44 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Effect Examination Method Item Applicability Primary Link Examination Coverage (Mechanism)

Core Barrel Assembly IPEC Units 2 Cracking (IASCC, Baffle-former Volumetric (UT) examination. 100% of accessible bolts.

and 3 Fatigue) bolts Accessibility may be Barrel-former bolts Re-inspection every 10 years limited by presence of Aging following initial inspection.

thermal shields. A Management (IE, minimum of 75%

Void Swelling coverage of the entire and IS.R) 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 2-5 Page 44

NL-17-020 Attachment 2 Page 45 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 ~nd 3 Effect Examination Method Item Applicability

• Primary Link Examination Coverage (Mechanism)

Lower Support Assembly IPEC Units 2 Cracking (IASCC, Baffle-former Volumetric (UT) examination. 100% of accessible bolts and 3 Fatigue) bolts or as supported by plant-Lower support column bolts Re-inspection every 10 years .

specific justification. A Aging following initial inspection.

minimum of 75%

Management (IE, coverage of the entire and ISR) 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 Figures 2-12 and 2-13 Core Barrel Assembly IPEC Units 2 Cracking (SCC, _Upper core barrel Enhanced visual (EVT-1) 100% of one side of the and 3 Fatigue) flange weld examination. accessible surfaces of the Core barrel outlet nozzi'e welds selected weld and adjacent Aging Re-inspection every I 0 years base metal. A minimum of Management (IE following initial inspection.

75% coverage of the entire of lower sections) 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 Page 45

NL-17-020 Attachment 2 Page 46 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPEC Units 2 and 3 Effect Examination Method Item Applicability Primary Link Examination Coverage (Mechanism)

Core Barrel Assembly IPEC Units 2 Cracking (SCC, Upper and lower Enhanced visual (EVT-1) I 00% of one side of the and 3 IASCC) core barrel examination. accessible surfaces of the Upper and lower core barrel cylinder girth selected weld and adjacent cylinder axial welds Aging Re-inspection every IO years welds base metal. A minimum of Management (IE) following initial inspection.

75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the .

examination is required (including both the accessible and inaccessible portions)

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

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

Page 46

NL-17-020 Attachment 2 Page 47 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-3 Expansion Components at IPE~ Units 2 and 3 Effect Examination Method Item Applicability Primary Link Examination Coverage (Mechanism)

Lower Support Assembly IPEC Units 2 Cracking Control rod guide Enhanced visual (EVT-1) 100% of accessible and 3 (IASCC) tube (CRGT) examination. support columns. A Lower support column bodies including the lower flanges minimum of 75%

Re-inspection every 10 years (cast) detection of coverage of the entire following initial inspection.

fractured support examination area or columns volume, or a minimum sampl~ size of 75% of the Aging total population of like Management (IE) components of the examination is required (including both the accessible and inaccessible portions)

See Figure 2-14 Bottom Mounted IPEC Units 2 Cracking Control rod guide Visual (VT-3) examination of 100% of BMI column Instrumentation System and 3 (Fatigue) tube (CRGT) BMI column bodies as bodies for which difficulty including the lower flanges indicated by difficulty of is detected during flux Bottom-mounted detection of *insertion/withdrawal of flux thimble instrumentation (BMI) column completely

• thimbles. insertion/withdrawal.

bodies fractured column Re-inspection every 10 years bodies following initial inspection.

See Figure 2-15 Aging Flux thimble Management (IE) insertion/withdrawal to be monitored at each inspection interval.

Page 47

NL-17-020 Attachment 2 Page 48 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-4 Existing Program Components at IPEC Units 2 and 3 Effect Item Applicability Reference Examination Method Examination Coverage (Mechanism)

Core Barrer Assembly IPEC Units 2 Loss of material ASMECode Visual (VT-3) examination All. accessible surfaces at Core barrel flange and 3 (Wear) Section XI to determine general specified frequency.

condition for excessive wear.

Upper Internals Assembly IPEC Units 2 Cracking (SCC, ASMECode Visual (VT-3) All accessible surfaces at Upper support ring or skirt and 3 Fatigue) Section XI examination. specified frequency.

'

IPEC has a tophat design, therefore there is no support ring or skirt, however the vertical sections of the tophat will be inspected Lower Internals Assembly IPEC Units 2 Cracking (IASCC, ASMECode Visual (VT-3) examination All accessible surfaces at Lower core plate and 3 Fatigue) Section XI of the lower core plates to specified frequency.

Aging Management detect evidence of (IE) distortion and/or loss of bolt integrity.

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

Bottom Mounted IPEC Units 2 Loss of material NUREG-1801 Surface (ET) examination. Eddy current surface Instrumentation System (ll)d 3 (Wear) Rev. 1 examination as defined in Flux thimble tubes plant response to IEB 88-09 Alignment and Interfacing IPEC Units 2 Loss of material ASMECode Visual (VT-3) All accessible surfaces at Components - and 3 (Wear) Section XI examination. specified frequency.

Clevis insert bolts .

Alignment and Interfacing IPEC Units 2 Loss of material ASMECode Visual (VT-3) -All accessible surfaces at Components and 3 (Wear) Section XI examination. specified frequency.

Upper core plate alignment pins -

Page 48

NL-17-020 Attachment 2 Page 49 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-5

• Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Additional Examination Item Applicability Expansion Lirik(s) Expansion Criteria Criteria (Note 1) Acceptance Criteria Control Rod Guide Tube IPEC Units 2 Visual (VT-3) None NIA NIA 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 visu~l (EVT-1) a. Bottdm-mounted a. Confirmation of surface- a. For BMI column bodies, Assembly and 3 examination. instrumentation (BMI) breaking indicati.ons 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 column bodies (cast), examination of BMI 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

- plate and lower support

b. Confirmation of surface-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 49

NL-17-020 Attachment 2 Page Sb of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Additional Examination Item Applicability Expansion Link(s) Expansion Criteria Criteria (Note 1) Acceptance Criteria Core Barrel Assembly IPEC Units 2 Periodic enhanced visual a. Core barrel outlet a. The confirmed detection and a. The specific relevant and 3 (EVT-1) examination. nozzle welds sizing of a surface-breaking condition for the expansion Upper core barrel flange indication with a length greater core barrel outlet nozzle weld The specific relevant b. Lower support than two inches in the upper core weld examination is a condition is a detectable column bodies (non barrel flange weld shall require detectable crack-like crack-like surface cast) that the EVT-1 examination be surface indication.

indication. expanded to include the core

• IPEC lower support b.N/A column bodies are cast, barrel outlet nozzle welds by the completion of the next refueling outage.

b.N/A 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 50

NL-17-020 Attachment 2 Page 51of63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Additional Examination Item Applicability Expansion Link(s) Expansion Criteria 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.

Bame-Former Assembly IPEC Units 2 Visual (VT-3) None NIA NIA 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 51

NL-17-020 Attachment 2 Page 52 of 63 Indian Point Energy Center

. Reactor Vessel Internals Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Additional Examination Item Applicability Expansion Link(s) Expansion Criteria Criteria (Note 1) Acceptance Criteria Bame-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 baffle plates at the largest distance column bolts and the acceptance criteria for the from the core (presumed to be the barrel-former bolts shall be UT of the baffle-former b. Barrel-former bolts 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.

Page 52

NL-17-020 Attachment 2 Page 53 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Additional Examination Item Applicability Expansion Link(s) Expansion Criteria Criteria (Note 1) Acceptance Criteria Bame-Former Assembly IPEC Units 2 Visual (VT-3) None NIA NIA and 3 examination.

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

Alignment and Interfacing IPEC Units 2 Direct physical None NIA NIA Components and 3 measurement of spring height.

Internals hold down spring 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.

Page 53

NL-17-020 Attachment 2 Page 54 of 63 Indian Point Energy Center Reactor Vesselintemals Inspection Plan Table 5-5 Examination Acceptance and Expansion Criteria at IPEC Units 2 and 3 Examination Acceptance Additional Examination Item Applicability Expansion Link(s) Expansion Criteria Criteria (Note 1) Acceptance Criteria Thermal Shield Assembly IPEC Units 2 Visual (VT-3) None NIA NIA and 3 examination.

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

Notes:

I. The examination acceptance criterion for visual examination is the absence of the specified relevant condition Page 54

NL-17-020 Attachment 2 Page 55 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-6 Reactor Vessel Component ISi Program Inspection Plan for IPEC Units 2 and 3 Code Examination Component Extent of Exam Category Method Lower Internals - Exterior B-N-3 VT-3 Components and areas as accessible Core barrel surface Lower Internals - Exterior B-N-3 VT-3 Components and areas as accessible Thermal Shield Lower Internals - Exterior B-N-3 VT-3 Components and areas as accessible Irradiation specimen tubes and guides Lower Internals - Exterior B-N-3 VT-3 Components and areas as accessible Flexures Lower Internals - Exterior B-N-3 VT-3 Components and areas as accessible Fasteners and locking devices Lower Internals - Exterior B-N-3 VT-3 Components and areas as accessible Outlet nozzles at 22 deg, 158 deg, 202 deg, and 338 deg 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 Page-55

NL-17-020 Attachment 2 Page 56 of 63 Indian Point Energy Center R.eactor Vessel Internals Inspection Plan Table 5-6 Reactor Vessel Component ISi Program Inspection Plan for IPEC Units 2 and 3 Code Examination Component Extent of Exam Category Method Lower Internals - Exterior Bottom B-N-3 VT-3 Components and areas as accessible

-v 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 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 nozzles at 22 deg, 158 deg, 202 deg, and 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 Page 56

NL-17-020 Attachment 2 Page 57 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-6 Reactor Vessel Component ISi Program Inspection Plan for IPEC Units 2 and 3 Code Examination Component Extent of Exam Category Method Core Barrel Assembly B-N-3 VT-3 Components and areas as accessible Core barrel flange 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 57

NL-17-020 Attachment 2 Page 58 of* 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-7 List of IPEC Reactor Vessel Interior Components and Materials Based on MRP-191-Table 4-4 UPPER INTERNALS ASSEMBLY Category from MRP-Sub Assembly Component Material 191Table7-2 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 Control rod guide Guide tube support pins (split pins) AX-750 (IP2 only) c Guide tube support pins (split pins) Stainless steel (IP3 only) A tube assemblies and Housing plates Stainless steel A flow downcomers Inserts Stainless steel A Lock bars Stainless steel A Sheaths Stainless steel c Support pin cover plate Stainless steel A Suooort pin cover plate cap screws Stainless steel A Support pin cover plate locking caps Stainless steel A and tie straps Suooort 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 instrumentatjon 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 UHi flow column bases CASS A Upper plenum UHi flow columns Stainless steel A Page 58

NL-17-020 Attachment 2 Page 59 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-7 List of IPEC Reactor Vessel Interior Components and Materials Based on MRP-191-Table4-4 UPPER INTERNALS ASSEMBLY*

Category from MRP-Sub Assembly Component Material 191Table7-2 Adapters Stainless steel A

. Bolts Column bases Stainless steel CASS A

A Upper support Column bodies Stainless steel A column assemblies Extension tubes Stainless steel A Flanges Stainless steel A Lock keys Stainles.s 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 I LOWER INTERNALS ASSEMBLY Category from MRP-Sub Assembly Component Material 191 Table 7-2 Baffle bolting locking bar Stainless steel A Baffle edge bolts Stainless steel c Baffle and forriier Baffle plates Stainless steel B assembly Baffle former bolts Stainless steel c Barrel former bolts Stainless steel c I 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 outlet nozzles Stainless steel B Core barrel Upper core barrel Stainless steel c Lower core barrel Stainless steel c Diffuser plate Diffuser plate Stainless steel A Page 59

NL-17-020 Attachment 2 Page 60 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-7 List of IPEC Reactor Vessel Interior Components and Materials Based on MRP-191-Table4-4 LOWER INTERNALS ASSEMBLY Category from MRP-Sub Assembly Component Material

• 191 Table 7-2 Flux thimble tube plugs - IPEC does Stainless steel B Flux thimbles (tubes) not use tube plugs 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 LCP fuel alignment pin lock caps Stainless steel A alignment pins 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 Lower support casting CASS A casting or forging Thermal shield bolts Stainless steel A Neutron Thermal shield dowels Stainless steel A panels/thermal shield Thermal shield tlexures 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 SCS energy absorber Stainless steel A support (SCS)

SCS guide posts Stainless steel A assembly 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 Head and vessel allignment pin bolts Stainless steel A '-'

Interfacing Head and vessel alignment pin lock Components 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 60

NL-17-020 Attachment 2

• Page 61 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan Table 5-8 IPEC Response to the NRC Revision 1 to the Final Safety Evaluation of MRP-227 MRP-227-A 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 1 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. l 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 IO-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 IO-year re-Condition 6 Periodicity of the re- examination interval for all Expansion inspection category components once examinati~n 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 No plant-specific action required.

Condition 7 Updating of industry guideline SER Section 4.2.l, 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 Aoolicant/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.

Aoolicant/Licensee Action Item 4 SER Section 4.2.5,

• The IPEC discussion regarding hold down springs is contained in Section 3.6.

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

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

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NL-17-020 Attachment 2 Page 62 of 63 Indian PointEnergy Center Reactor Vessel Internals Inspection Plan 6.0 OPERATING EXPERIENCE AND ADDITIONAL CONSIDERATIONS 6.1 Internal and External Operating Experience Operating experience related to degradation of reactor internal components covered in this program will be reviewed on a periodic basis. This review wilJ: include both domestic and international experience and will be documented in accordance with the Entergy operating

.experience process. Results of reactor internal components inspected in accordance with MRP-227-A will be collected and summarized in accordance with NEI 03-08 guidelines.

6.2 Spring 2016 Operating Experience In the spring of 2016, during IP2 outage 2R22, ultrasonic (UT) and/or visual inspections of all .

832 baffle former bolts (bolts) were performed in accordance with the NRC approved guidelines.

in MRP-227-A. Visual inspection of the baffle plates and bolts identified 31 degraded bolts.

The UT inspections;.identified indications on 182 bolts a~d also determined that 14 bolt locations were not testable. The locations that were not testable were conservatively assume.ct to possess bolts that fai°led to meet the acceptance criteria. As a result of the inspection findings. all 227 bolts (31+182+ 14) w~th actual and assumed indications were replaced. An additional 51 bolts were replaced to reduce the probability of future failures. as well as minimize the probability of clusters of failed bolts. Therefore, during 2R22. a total of 278 bolts (227 +51) were replaced.*

• As a result of the IP2 inspection findings and other industry Operating Experience COE) indicating a significant number of failed bolts at other similarly-designed PWR plants, the IPEC PWR Vessel Internals Program was revised. 'In view of the 2R22 inspection findings, Entergy arranged for the fractographic examination of eight baffle former bolts removed from the IP2 baffle structure during the Spring 2016 outage at Westinghouse Electric Company's hot cell laboratory in Churchill, PA. The results of those fractographic examinations are documented in Westinghouse Report MCOE-TR-16-18, Revisfon 0, "Fractography of Indian Point Unit 2 ~affle Former Bolts" (Nov. 30, 2016). Industry-sponsored metallurgical analysis and materials property testing of addi.tional baffle former bolt specimens from IP2 and other PWRs are still in progress.

Based on as-found conditions and current industry knowledge, including the results of the fractographic examinations of the eight IP2 baffle former bolts discussed in Westinghouse Report MCOE-TR-16-18, IPEC concludes that perfdrming a volumetric examination (i.e., UT) of the required original bolts during each refueling outage, and replacing those bolts found to be degraded until none of the remaining original bolts are required to be credited for the baffle structure to be capable of performing its intended safety function, is a reasonable and acceptable approach. Accordingly. IPEC plans to take the actions specified in paragraphs 1-5 below. These actions are subject to possible revision per the OE program based on the results of ongoing and Page 62

NL-17-020 Attachment 2 Page 63 of 63 Indian Point Energy Center Reactor Vessel Internals Inspection Plan planned future inspection and testing of baffle former bolts from IP2 and other PWR plants. Any findings that result from the following actions will be input to the Corrective Action Program.

1. The IP3 baffle bolt inspections that were previously scheduled to be performed in 3R20 (Spring 2019) will be performed in 3R19 (Spring 2017). Visual and UT inspections on 100% of all accessible baffle former bolts. and a visual inspection of the baffle-edge bolts and baffle former assembly. will be performed in 3R19.

2. Entergy will perform a UT inspection of 100% of the original bolts at IP2 and IP3 during each of the subsequent refueling outages if any of the original bolts are required to remain structurally capable of carrying their design load to ensure structural integrity of the baffle structure during all design conditions.

3. Entergy will also perform a general visual inspection to identify anomalies in the baffle structure at IP2 and IP3 during each _subsequent refueling outage.

4. Entergy will perform a UT inspection of inservice replaced (new) bolts if the general visual inspections performed in accordance with paragraph 3. above identify degraded new bolts.

5. Entergy will replace all bolts with indications that are needed to remain structurally capable of carrying their design load to ensure structural integrity of the baffle structure during all design conditions. Additional "good" or anti-cluster bolts will also be replaced to ensure that sufficient margin is maintained to accommodate the same failure rate until the next inspection as the failure rate identified during the current refueling outage. This margin will ensure. compliance with the intent of the guidelines provided in WCAP-17096, Revision 2, "Reactor Internals Acceptance Criteria Methodology and Data Requirements."

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