Text

License Renewal (Lr) Reactor Vessel Internals (Rvi) Commitment Implementation Report and Inspection Plan
	ML12363A103
Person / Time
Site:	Turkey Point
Issue date:	12/14/2012
From:	Kiley M; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2012-438
	Download: ML12363A103 (39)
	v • d • e

0IPL.

December 14, 2012 L-2012-438 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555-0001 Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 License Renewal (LR) Reactor Vessel Internals (RVI) Commitment Implementation Report and Inspection Plan The original renewed license commitment (Outstanding Commitment No. 6 listed in Attachment 1.5 of Reference 1) required Turkey Point to implement the Reactor Vessel Internals (RVI) Inspection Program prior to the end of the initial operating term (July 19, 2012, and April 10, 2013 for Turkey Point Units 3 and 4, respectively). This commitment was based on the specific Turkey Point RVI Inspection Program that was previously approved by the Nuclear Regulatory Commission (NRC).

On May 11,2011, Florida Power and Light Company (FPL) submitted letter L-2011-176, (Reference 2) to inform the NRC of FPL's intent to adopt and implement the NRC approved Reactor Vessel Internals (RVI) MRP-227 Guidelines for Turkey Point Units 3 and 4.

On June 22, 2011, the NRC approved and issued the Final Safety Evaluation Report (SE) for EPRI Report MRP-227 (Reference 3). Subsequently, on July 21, 2011, NRC issued Regulatory Issue Summary (RIS) 2011-07, "License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management" (Reference 4) to provide guidance to Licensees with renewed licenses for the submittal of Aging Management Programs (AMPs)/Inspection Plans. On December 22, 2011, FPL submitted letter L-2011-531 (Reference 5), to document the deferral of the implementation of the RVI Inspection Program crediting MRP-227-A to no later than December 31, 2012 for Turkey Point Unit 3, and prior to the end of the initial operating term (April 10, 2013) for Turkey Point Unit 4.

The evaluation to revise the previous commitment to implement the Turkey Point RVI Inspection Program in order to credit implementation of MRP-227-A guidelines (Reference 6) was performed under the controls of the 10 CFR 50.59 rule.

The purpose of this letter is to obtain a closeout review of the revised commitment and of the RVI Inspection Plan. FPL hereby submits to NRC, in Attachment 1, a commitment implementation report that describes the 10 key attributes of the RVI AMP and the Turkey Point site specific RVI Inspection Plan, which credits the implementation of MRP-227-A at Turkey Point. Additionally, in Attachment 2, FPL provides confirmation and acceptability of the implementation of MRP-227-A for Turkey Point site by addressing each of the site-specific licensee action items outlined in section 4.2 of the NRC SE (Reference 3).

Florida Power , Light Company 9760 SW. 344 Street Homestead, FL 33035

L-2012-438 Page 2 Should there be any questions, please contact Mr. Robert J. Tomonto, Licensing Manager at 305- 246-7327.

Very truly yours, Michael Kiley Vice President Turkey Point Nuclear Plant SM Attachments Enclosure cc: Regional Administrator, USNRC Region II USNRC Senior Resident Inspector - Turkey Point Plant

L-2012-438 Page 3

References:

1. Memorandum, USNRC to File, "Commitment Lists for Renewed Operating License (ROL)

Plants with No Commitment Appendix Attached to Its ROL Safety Evaluation Reports/NUREGs for Use with IP-71003," Attachment 1.5, "Site Specific List of Commitments for Post-Renewal Inspection at Turkey Point Nuclear Plant," dated March 6, 2007

2. FPL letter L-2011-176 to the USNRC, "License Renewal (LR) Reactor Vessel Internals (RVI)

Inspection Program, Letter of Intent to Adopt Materials Reliability Program (MRP) Report MRP-227, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" and Revision of Commitments," dated May 11, 2011,

3. NRC, Final Safety Evaluation of EPRI Report, Materials Reliability Program Report 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," issued June 22, 2011

4. NRC Regulatory Issue Summary 2011-07, "License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management," issued July 21, 2011

5. FPL letter L-2011-531 to USNRC, "License Renewal (LR) Reactor Vessel Internals (RVI)

Inspection Program Implementation Commitment Revision Notification," dated December 22, 2011.

6. EPRI-Materials Reliability Program, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A),Final Report, Technical Report 1022863, issued December 2011.

L-2012-438 Attachment 1 Florida Power and Light Company Turkey Point Units 3 and 4 License Renewal Reactor Vessel Internals Commitment Implementation Report Inspection Plan and Key Attributes of Aging Management Program

TURKEY POINT UNITS 3 & 4 REACTOR VESSEL INTERNALS (RVI) Attachment 1 COMMITMENT IMPLEMENTATION REPORT Page 1 of2, TABLE OF CONTENTS Page No.

1 Background 2 2 RVI Aging Management Program Implementation 2 3 RVI Aging Management Program Key Attributes 3 4 RVI Inspection Plan 6 5 References 10 6 Tables 10

TURKEY POINT UNITS 3 &4 -vtIhmnt REACTOR VESSEL INTERNALS (RVI) Attachment 1 COMMITMENT IMPLEMENTATION REPORT Page 2 of22

1. BACKGROUND Florida Power & Light Company (FPL) previously developed an Aging Management Program (AMP) for the Turkey Point Units 3 and 4 Reactor Vessel Internals (RVI) that was included in the Turkey Point License Renewal Application (LRA) (Ref. 5.1) and subsequently approved by the NRC in NUREG 1759 (Ref. 5.2). Included in the RVI AMP was a commitment to participate in ongoing, joint industry efforts aimed at further understanding the aging effects of the RVI and to revise the Turkey Point Units 3 and 4 RVI AMP as needed. These joint industry efforts culminated in the issuance of EPRI MRP-227 Rev. 0, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" in December 2008 (Ref.

5. 3). This MRP report was submitted to the NRC for review and approval through the Nuclear Energy Institute in January 2009.

On June 22, 2011, the NRC issued Revision 0 of the Safety Evaluation Report (SER) for MRP-227 (Ref.

5.4), endorsing the guidance provided that eight plant specific actions and seven topical report condition items were implemented. On July 21, 2011, the NRC issued Regulatory Issue Summary 2011-07, "License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management" (Ref. 5.5), to inform the industry about the availability of MRP-227 as an acceptable approach for aging management of reactor vessel internals for license renewal. On December 16, 2011, the NRC issued Revision 1 of the Safety Evaluation Report (Ref. 5.6) to incorporate technical changes required to ensure the final approved version of MRP-227 (i.e. MRP-227-A) included all NRC required changes.

MRP-227-A (Ref. 5.7) was issued in December 2011.

On December 22, 2011, FPL submitted letter L-2011-531 (Ref. 5.8) to the NRC. The purpose of this letter was to notify the NRC of a revised LR commitment to defer implementation of the RVI Inspection Program for Turkey Point Units 3, and to document previous FPL/NRC discussion addressing related submittals for Units 3 and 4. The letter summarized that FPL would perform the commitment revision to adopt MRP-227-A under the controls of the 10 CFR 50.59 rule (Ref. 5.9); deferred the due date for implementation of the Unit 3 RVI AMP from July 19, 2012 (end of the initial operating license term) to no later than December 31, 2012; committed to submit to the NRC the RVI Inspection Plan no later than December 31, 2012 and to implement Unit 4 RVI Inspection Program by April 10, 2013 (end of the initial operating term).

2. RVI AGING MANAGEMENT PROGRAM IMPLEMENTATION The Turkey Point RVI consists of two basic assemblies: (1) an upper internals assembly that is removed during each refueling cycle to obtain access to the reactor core; and (2) a lower internals assembly that can be removed, if desired, following a complete core unload.

The RVI Aging Management Program (AMP) applies to both Turkey Point Units 3 and 4 and addresses passive RVI structural components contained within the upper and lower internals assemblies. The RVI AMP specifically excludes welded attachments to the reactor vessel and consumable items such as fuel assemblies and the rod control cluster assemblies (RCCAs). The functions of the included passive RVI components are to provide: 1) core support; 2) flow distribution; 3) guidance and support of instrumentation and RCCAs; 4) shielding to the vessel from neutron irradiation and gamma heating.

The revised RVI AMP was developed crediting implementation of the EPRI MRP-227-A, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines."

The 10 key attributes (program elements as defined in NUREG-1801, Revision 2, AMP XI.M16A) as apply to the Turkey Point Units 3 and 4 RVI AMP are discussed in Section 3 of this Attachment. The corresponding UFSAR description of the revised program is provided in Enclosure 1 of this Attachment.

TURKEY POINT UNITS 3 &4 REACTOR VESSEL INTERNALS (RVI)

COMMITMENT IMPLEMENTATION REPORT

3. RVI AGING MANAGEMENT PROGRAM KEY ATTRIBUTES The 10 key attributes of the Turkey Point Units 3 and 4 that are used to describe the RVI AMP are discussed below:

Plan Attribute Approach and Supplemental Information Scope of The scope of the RVI AMP includes all RVI components at Turkey Point Units 3 & 4, which are built to a Westinghouse NSSS design. The scope of the Program program applies the methodology and guidance in MRP-227-A, which provides augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related internals in commercial operating U.S.

PWR nuclear power plants designed by B&W, CE, and Westinghouse. The scope of components considered for inspection under MRP-227-A guidance includes core support structures (typically denoted as Examination Category B-N-3 by the ASME Code, Section Xl), those RVI components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4(a)(1), and other RVI components whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a)(1)(i), (ii),

or (iii). The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation, because these components are not typically within the scope of the components that are required to be subject to an aging management review (AMR), as defined by the criteria set in 10 CFR 54.21(a)(1). The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are considered to be ASME Code Class 1 appurtenances to the reactor vessel and are adequately managed in accordance with the Turkey Point Units 3 & 4 ASME Section Xl Program.

2 Preventive The guidance in MRP-227-A relies on PWR water chemistry control to prevent Measures or mitigate aging effects that can be induced by corrosive aging mechanisms (e.g., loss of material induced by general, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of its forms [SCC, PWSCC, or IASCC]).

The Turkey Point Units 3 & 4 reactor coolant water chemistry is monitored and maintained in accordance with the Turkey Point Chemistry Control Program.

3 Parameters The RVI AMP manages the following age-related degradation effects and Monitored mechanisms that are applicable in general to the RVI components at the facility: (a) cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness induced by either thermal aging or neutron irradiation embrittlement; (d) changes in dimension due to void swelling and irradiation growth, distortion, or deflection; and (e) loss of preload caused by thermal and irradiation-enhanced stress relaxation or creep.

For the management of cracking, the program monitors for evidence of surface breaking linear discontinuities if a visual inspection technique is used as the non-destruction examination (NDE) method, or for relevant flaw presentation signals if a volumetric UT method is used as the NDE method.

For the management of loss of material, the program monitors for gross or abnormal surface conditions that may be indicative of loss of material occurring in the components. For the management of loss of preload, the program monitors for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections. The program does not directly monitor for loss of fracture toughness that is induced by thermal aging or neutron irradiation embrittlement, or by void swelling and

TURKEY POINT UNITS 3 & 4 Attachment81 REACTOR VESSEL INTERNALS (RVI) Afcmn COMMITMENT IMPLEMENTATION REPORTPae4o2 Plan Attribute Approach and Supplemental Information irradiation growth; instead, the impact of loss of fracture toughness on component integrity is indirectly managed by using visual or volumetric examination techniques to monitor for cracking in the components and by applying applicable reduced fracture toughness properties in the flaw evaluations if cracking is detected in the components and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation under the MRP-227-A guidance or ASME Code,Section XI requirements. The program uses physical measurements to monitor for any dimensional changes due to void swelling, irradiation growth, distortion, or deflection.

4 Detection of The detection of aging effects is covered in two places: (a) the guidance in Aging Effects Section 4 of MRP-227-A provides an introductory discussion and justification of the examination methods selected for detecting the aging effects of interest; and (b) standards for examination methods, procedures, and personnel are provided in a companion document, MRP-228. In all cases, well-established methods were selected. These methods include volumetric UT examination methods for detecting flaws in bolting, physical measurements for detecting changes in dimension, and various visual (VT-3, VT-1, and EVT-1) examinations for detecting effects ranging from general conditions to detection and sizing of surface-breaking discontinuities. Surface examinations may also be used as an alternative to visual examinations for detection and sizing of surface-breaking discontinuities.

Cracking caused by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1 examination (for internals other than bolting) or by volumetric UT examination (bolting). The VT-3 visual methods may be applied for the detection of cracking only when the flaw tolerance of the component or affected assembly, as evaluated for reduced fracture toughness properties, is known and has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions. In addition, VT-3 examinations are used to monitor/inspect for loss of material induced by wear and for general aging conditions, such as gross distortion caused by void swelling and irradiation growth or by gross effects of loss of preload caused by thermal and irradiation-enhanced stress relaxation and creep.

In addition, the program adopts the recommended guidance in MRP-227-A for defining the Expansion criteria that need to be applied to inspections of Primary Components and Existing Requirement Components and for expanding the examinations to include additional Expansion Components. As a result, inspections performed on the RVI components are performed consistent with the inspection frequency and sampling bases for Primary Components, Existing Requirement Components, and Expansion Components in MRP-227-A, which have been demonstrated to be in conformance with the inspection criteria, sampling basis criteria, and sample Expansion criteria in Section A.1.2.3.4 of NRC Branch Position RLSB-1.

Specifically, the program implements the parameters monitored/inspected criteria and bases for inspecting the relevant parameter conditions for Westinghouse designed Primary Components in Table 4-3 of MRP-227-A and for Westinghouse designed Expansion Components in Table 4-6 of MRP-227-A.

In addition, in some cases (as defined in MRP-227-A), physical measurements are used as supplemental techniques to manage for the gross effects of wear, loss of preload due to stress relaxation, or for chanqes in dimension due to

TURKEY POINT UNITS 3 &4 REACTOR VESSEL INTERNALS (RVI)

COMMITMENT IMPLEMENTATION REPORT Plan Attribute Approach and Supplemental Information void swelling, deflection or distortion. The physical measurements methods applied in accordance with this program include measurement of hold down spring height.

5 Monitoring The methods for monitoring, recording, evaluating, and trending the data that and Trending result from the program's inspections are given in Section 6 of MRP-227-A and its subsections. 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. The examinations and re-examinations required by the MRP-227-A guidance, together with the requirements specified in MRP-228 for inspection methodologies, inspection procedures, and inspection personnel, provide timely detection, reporting, and corrective actions with respect to the effects of the age-related degradation mechanisms within the scope of the program. The extent of the examinations, beginning with the sample of susceptible PWR internals component locations identified as Primary Component locations, with the potential for inclusion of Expansion Component locations if the effects are greater than anticipated, plus the continuation of the Existing Programs activities, such as the ASME Code,Section XI, Examination Category B-N-3 examinations for core support structures, provides a high degree of confidence in the total program.

6 Acceptance Section 5 of MRP-227-A provides specific examination acceptance criteria for Criteria the Primary and Expansion Component examinations. For components addressed by examinations referenced to ASME Code,Section XI, the IWB-3500 acceptance criteria apply. For other components covered by Existing Programs, the examination acceptance criteria are described within the Existing Program reference document.

The guidance in MRP-227-A contains three types of examination acceptance criteria:

For visual examination (and surface examination as an alternative to visual examination), the examination acceptance criterion is the absence of any of the specific, descriptive relevant conditions; in addition, there are requirements to record and disposition surface breaking indications that are detected and sized for length by VT-I/EVT-1 examinations;

For volumetric examination, the examination acceptance criterion is the capability for reliable detection of indications in bolting, as demonstrated in the examination Technical Justification; in addition, there are requirements for system-level assessment of bolted or pinned assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits; and

The acceptance criterion for physical measurements performed on the height limits of the Westinghouse-designed hold-down springs are described in the response to Licensee Action Item (LAI) #5 described in Attachment 2 of this document.

7 Corrective Corrective actions following the detection of unacceptable conditions are Actions fundamentally provided for in the Turkey Point Corrective Action Program. Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which

TURKEY POINT UNITS 3 & 4 REACTOR VESSEL INTERNALS (RVI)

COMMITMENT IMPLEMENTATION REPORT Plan Attribute Approach and Supplemental Information may require repair, replacement, or analytical evaluation for continued service until the next inspection. The disposition will ensure that design basis functions of the reactor internals components will continue to be fulfilled for all licensing basis loads and events. Examples of methodologies that can be used to analytically disposition unacceptable conditions are found in the ASME Code, Section Xl or in Section 6 of MRP-227-A. Section 6 of MRP-227-A describes the options that are available for disposition of detected conditions that exceed the examination acceptance criteria of Section 5 of the report. These include engineering evaluation methods conducted in accordance with WCAP-17096, as well as supplementary examinations to further characterize the detected condition, or the alternative of component repair and replacement procedures.

The latter are subject to the requirements of the ASME Code, Section Xl. The implementation of the guidance in MRP-227-A, plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

8 Confirmation Turkey Point quality assurance procedures, review and approval processes, Process and and administrative controls are implemented in accordance with the Self requirements of 10 CFR Part 50, Appendix B, or their equivalent, as Assessment applicable. It is expected that the implementation of the guidance in MRP-227-A will provide an acceptable level of quality for inspection, flaw evaluation, and other elements of aging management of the PWR internals that are addressed in accordance with the 10 CFR Part 50, Appendix B, or their equivalent (as applicable), confirmation process, and administrative controls.

9 Administrative The administrative controls for License Renewal RVI AMP including the Controls implementing procedures, review and approval processes, are under existing Turkey Point site 10 CFR 50 Appendix B Quality Assurance Programs. The RVI AMP is established with a sufficient level of documentation and administrative controls to ensure effective long-term implementation. The implementing procedure for the Turkey Point Units 3 and 4 RVI AMP is 0-ADM-563, Reactor Vessel Internals Aging Management Program.

10 Operating Relatively few incidents of PWR internals aging degradation have been Experience reported in operating U.S. commercial PWR plants. A summary of observations to date is provided in Appendix A of MRP-227-A. Operating experience gained through Industry groups such as the EPRI MRP, the PWROG, INPO, WANO and International Sites shall be evaluated and incorporated into this program as needed in a timeframe consistent with the significance. Operation experience (OE) reports are continuously reviewed by Turkey Point personnel to ensure relevant OE is reviewed for impact on aging effects and/or aging management programs.

4. RVI INSPECTION PLAN The Turkey Point Units 3 and 4 RVI Inspection Plan is based on the implementation of MRP-227-A. It contains a description summary of the degradation mechanisms of concern; the categorization of components; the components to be inspected; the inspection methodology; the examination/inspection coverage; and the examination acceptance criteria.

TURKEY POINT UNITS 3 & 4 L.-dV I 'dou, REACTOR VESSEL INTERNALS (RVI) Attachment 1 COMMITMENT IMPLEMENTATION REPORT Page 7 of22 DEGRADATION MECHANISMS A total of eight age related degradation mechanisms are considered applicable to the RVI: 1) stress corrosion cracking (SCC); 2) irradiation assisted stress corrosion cracking (IASCC); 3) fatigue; 4) irradiation embrittlement (IE); 5) thermal embrittlement (TE); 6) wear; 7) void swelling (VS); and 8) irradiation enhanced stress relaxation/creep (ISR/IC). A brief description of these degradation mechanisms and the associated aging effects follows:

Stress Corrosion Cracking (SCC)

SCC is a localized, non-ductile failure caused by a combination of stress, susceptible material, and an aggressive environment. The fracture path of SCC can be either transgranular or intergranular in nature. The aggressive contaminants most commonly associated with SCC of austenitic stainless steels are dissolved chlorides and oxygen. Nickel base alloys such as Alloy 600 and X-750 have exhibited susceptibility to intergranular SCC in primary water without the presence of aggressive contaminants, commonly referred to as primary water stress corrosion cracking (PWSCC). SCC of Stainless Steel (SS) in primary water is also considered feasible at high stress levels. The aging effect of SCC is cracking.

Irradiation Assisted SCC (IASCC)

IASCC is a form of intergranular SCC that results from the combined influence of neutron irradiation and an aggressive environment. A limited number of IASCC failures of RVI components, specifically fasteners, constructed of austenitic stainless steels and nickel base alloys have been observed. The aging effect of IASCC is cracking.

Fatigue Fatigue is defined as the structural deterioration that can occur as a result of the periodic application of stress by mechanical, thermal, or combined effects. High cycle fatigue results from relatively low cyclic stress (<yield strength) applied for many (>105) cycles. Low cycle fatigue results from relatively high cyclic stress (>yield strength) applied for low number of cycles. The aging affect of fatigue is cracking.

Irradiation Embrittlement (IE)

IE refers to a gradual and progressive change in mechanical properties of a material resulting from exposure to high levels of neutron irradiation. These changes include an increase in yield and tensile strengths, and a corresponding decrease in ductility and toughness. The aging effect of IE is loss of fracture toughness.

Thermal Embrittlement (TE)

Thermal embrittlement refers to the same gradual and progressive change in mechanical properties of a material as IE except it results from exposure to elevated temperatures rather than neutron irradiation.

For the RVI components, TE is only a concern for SS castings and welds with duplex microstructures containing both ferrite and austenite. The aging effect of TE is loss of fracture toughness.

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

Void Swelling (VS)

Void swelling is the gradual increase in volume of a component caused by the formation of microscopic cavities. These cavities result from the nucleation and growth of vacancies created by exposure to high levels of neutron irradiation. During the initial licensing periods of domestic PWRs, field experience has not revealed any evidence of VS in RVI components; however it is postulated as a possibility during periods of extended operation based upon accelerated laboratory testing. The aging effect of VS is dimensional change.

TURKEY POINT UNITS 3 &4 .... ...

REACTOR VESSEL INTERNALS (RVI) Attachment 1 COMMITMENT IMPLEMENTATION REPORT Page 8 of22 Irradiation and Thermally Enhanced Stress Relaxation/Creep (SR/C)

Stress relaxation involves the short term unloading of preloaded components upon exposure to elevated temperatures or high levels of neutron irradiation. Creep is a longer term process in which plastic deformation occurs within a loaded component. The temperatures of RVI are typically not high enough to support creep; however it can develop upon exposure to high levels of neutron irradiation over an extended period. The aging effect of stress relaxation and creep is loss of preload.

COMPONENT CATEGORIZATION The RVI components that required an aging management review were identified during the Turkey Point Units 3 &4 license renewal process (Ref 5.10) in accordance with the Nuclear Plant License Renewal Rule, 10 CFR 54 (Ref. 5.11). These components were then categorized as Existing Program, Primary, Expansion or No Additional Measurements, based upon the guidance provided in MRP-227-A.

A complete listing of the Turkey Point Units 3 and 4 RVI components that required an aging management review, their material of construction, intended function and categorization is provided in Table 1, RVI Component Details and Categorization. A description of the component categories follows:

Existing Program Components Existing Program Components are susceptible to at least one of the eight degradation mechanisms, for which existing plant programs are capable of managing the associated aging effect(s). Details of the required inspections for Existing Program Components are provided in Table 2, Westinghouse Plants Existing Program Components.

Primary Components Primary Components are highly susceptible to at least one of the eight degradation mechanisms, for which augmented inspections are required on a periodic basis to manage the associated aging effect(s). Primary Components are considered lead indicators for the onset of the applicable degradation mechanism(s). Details of the required inspections for Primary Components are provided in Table 3, Westinghouse Plants Primary Components.

Expansion Components Expansion Components are highly or moderately susceptible to at least one of the eight degradation mechanisms, but exhibit a high degree of tolerance to the associated aging effect(s). Augmented inspections are required once a specified level of degradation is detected in a linked Primary Component. Details of the required inspections for Expansion Components are provided in Table 4, Westinghouse Plants Expansion Components.

No Additional Measures Components No Additional Measures Components are either not susceptible to any of the eight degradation mechanisms, or if susceptible the impact of failure on the functionality of the RVI components is insignificant. No further action is required for managing the aging of these RVI components.

INSPECTION OF RVI COMPONENTS Based on MRP-227-A implementation at Turkey Point Units 3 and 4, the RVI Inspections detailed in Table 2, Westinghouse Plants Existing Program Components, and Table 3, Westinghouse Plants Primary Components are required to manage the aging effects in Primary and Existing Program Components. Additionally, the inspections detailed in Table 4, Westinghouse Plants Expansion Components, are required should evidence of aging degradation be detected in linked Primary Components. The expansion criteria are found in Table 5, Westinghouse Plants Examination Acceptance and Expansion Criteria.

TURKEY POINT UNITS 3 &4 1-- --.

REACTOR VESSEL INTERNALS (RVI) Attachment 1 COMMITMENT IMPLEMENTATION REPORT Page 9 of22 Inspection Methods Proven inspection methodologies are utilized to detect evidence of the relevant aging mechanism(s) for the Existing Programs, Primary and Expansion Components.

These include the following:

" Direct physical measurements to monitor for loss of material or preload

" VT-3 (visual) exams to monitor for general degradation associated with loss of material or preload

EVT-1 (enhanced visual) exams to monitor for surface breaking linear discontinuities indicative of cracking

UT (ultrasonic) exams to monitor directly for cracking 0 ECT (eddy current testing) to further characterize conditions detected by VT-3 and EVT-1 examination.

The requirements for the inspection methodologies and qualification of NDE systems used to perform those inspections are provided in EPRI MRP-228, Inspection Standard for PWR Internals (Ref. 5.12).

Inspection Frequencies The required inspection frequencies for Existing, Primary and Expansion Components are specified in Tables 2, 3 and 4, respectively. Specified inspection frequencies are considered adequate to manage aging effects; however more frequent inspections may be warranted based upon an internal and external OE.

Inspection Coverage The required inspection coverage for Primary and Expansion Components are specified in Tables 3 and 4, respectively. The required inspection coverage for the Existing Program Components is as specified in the applicable program document (e.g. ASME Section Xl). If the specified coverage for any of these components cannot be obtained, the condition shall be addressed in the Corrective Action Program (CAP).

Acceptance Criteria The acceptance criteria for Primary and Expansion Components are provided in Table 5, Westinghouse Plants Examination Acceptance and Expansion Criteria. These criteria are based upon the requirements of ASME Section XI. All detected relevant conditions must be addressed in the CAP prior to plant start-up. Possible disposition options include: 1) supplemental exams to further characterize a detected condition; 2) engineering evaluation for continued service until the next inspection; 3) repair; or

4) replacement.

Engineering evaluations for continued service shall be conducted in accordance with NRC approved methodologies. WCAP-1 7096, "Reactor Internals Acceptance Criteria Methodology and Data Requirements" (Ref. 5.13), is currently under NRC review for this purpose. The potential loss of fracture toughness must be considered in any flaw evaluations.

Expansion Criteria The expansion criteria for expanding the scope of examination from the Primary to the linked Expansion Components, including the timing of inspections, are provided in Table 5, Westinghouse Plants Examination Acceptance and Expansion Criteria.

It should be noted that the categorizations and associated inspection requirements described above do not replace or relieve any of the current ASME Section XI inspection requirements for the RVI components.

TURKEY POINT UNITS 3 &4 REACTOR VESSEL INTERNALS Attachment 1 COMMITMENT IMPLEMENTATION REPORT Page 10 of22

5. REFERENCES

1. Application for Renewed Operating Licenses, Turkey Point Units 3 and 4, submitted to NRC September 8, 2000.

2. NUREG 1759, Safety Evaluation Report Related to the License Renewal of Turkey Point Nuclear Plant, Units 3 and 4, Docket Nos. 50-250 and 50-251.

3. EPRI Materials Reliability Program, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Rev. 0), Technical Report 1016596.

4. Final Safety Evaluation of EPRI Report, Materials Reliability Program 1016596 (MRP-227),

Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines" (TAC NO. ME0680).

5. NRC Regulatory Issue Summary 2011-07, License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management, July 21, 2011.

6. Revision 1 to the Final Safety Evaluation of EPRI Report, Materials Reliability Program 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines" (TAC NO. ME0680).

7. EPRI Materials Reliability Program, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), Technical Report 1022863.

8. FPL Letter L-2011-531, Michael Kiley to USNRC, License Renewal (LR) Reactor Vessel Internals (RVI) Inspection Program Implementation Commitment Revision Notice, December 22, 2011.

9. 10 CFR 50.59, Changes, tests and experiments.

10. PTN-ENG-LRAM-99-0075, "License Renewal Aging Management Review - Reactor Vessel Internals."

11. Title 10 Code of Federal Regulation Part 54 - Requirements for Renewal of Operating Licenses for Nuclear Power Plants

12. EPRI Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228), 10166609.

13. WCAP-17096-NP, Reactor Internals Acceptance Criteria Methodology and Data Requirements.

14. NUREG 1801, Rev. 2, Generic Aging Lessons Learned (GALL) Report, December 2010.

6. TABLES TABLE 1 - RVI Component Details and Categorization TABLE 2 - Westinghouse Plants Existing Programs Components TABLE 3 - Westinghouse Plants Primary Components TABLE 4 - Westinghouse Plants Expansion Components TABLE 5 - Westinghouse Plants Examination Acceptance and Expansion Criteria.

L-2012-438, Turkey Point Units 3 & 4 Attachment 1 TABLE 1 RVI Component Details and Categorization Page 11 of 22 COMPONENT MATERIAL INTENDED FUNCTION CATEGORY Lower Internals Assembly Lower Core Plate 304 SS Core Support, Flow Distribution Existing Program Fuel Alignment Pins 304 SS Core Support No Additional Measures Lower Support Forging 304 SS Core Support, Flow Distribution Expansion Lower Support Columns CF8 cast SS Core Support Expansion Lower Support Column Bolting 316 SS Core Support Expansion Radial Keys 304 SS Core Support No Additional Measures Diffuser Plate 304 SS Flow Distribution No Additional Measures Secondary Core Support 304 SS Core Support, Flow Distribution No Additional Measures Bottom Mounted Instrument 304 SS Guide and Support Instrumentation Expansion Column Bodies Bottom Mounted Instrumentation CF8 cast SS/304 SS Guide and Support Instrumentation No Additional Measures Bases and Extension Bars Head Cooling Spray Nozzles 304 SS Flow Distribution No Additional Measures Interfacing Components Clevis Insert Alloy 600 Core Support No Additional Measures Upper Core Plate Alignment Pins 304 SS w/Stellite hardfacing Guide and Support RCCAs Existing Program Internals Hold-Down Spring 304 SS Core Support Primary HeadNessel Alignment Pins 304 SS Core Support No Additional Measures Clevis Insert Bolting X-750 Core Support Existing Program Flux Thimble Tubes 316 SS Guide Instrumentation Existing Program Lower Internals Assembly/Core Barrel Subassembly Core Barrel Flange 304 SS Core Support, Flow Distribution Existing Program Primary Core Barrel Outlet Nozzles 304 SS Core Support, Flow Distribution Expansion Lower Core Barrel 304 SS Core Support, Flow Distribution Primary Upper Core Barrel 304 SS Core Support, Flow Distribution Primary Thermal Shield 304 SS Shield Vessel No Additional Measures Thermal Shield Flexures 316 SS Shield Vessel Primary

L-2012-438, Turkey Point Units 3 & 4 Attachment 1 TABLE 1 RVI Component Details and Categorization Page 12 of 22 COMPONENT MATERIAL INTENDED FUNCTION CATEGORY Lower Internals Assembly/Baffle-Former Subassembly Baffle and Former Assembly 304 SS Core Support, Flow Distribution Primary Baffle-Former Bolts 347 SS Core Support, Flow Distribution Primary Baffle Edge Bolts 316 SS Core Support, Flow Distribution Primary Barrel-Former Bolts 316 SS Core Support, Flow Distribution Expansion Upper Internals Assembly Upper Support Plate 304 SS Guide and Support RCCAs No Additional Measures Upper Support Ring 304 SS Guide and Support RCCAs Existing Program Upper Core Plate 304 SS Core Support, Flow Distribution Expansion Upper Support Column Bases CF8 cast SS Guide and Support RCCAs No Additional Measures Upper Support Columns 304 SS Guide and Support RCCAs No Additional Measures Upper Support Column Bolting 316 SS Core Support No Additional Measures Upper Instrumentation Column 304 SS/Unit 3 Guide and Support Thermocouples No Additional Measures 316 SS/Unit 4 Guide Tube Assemblies (GTA)

GTA Lower Flanges 304 SS Guide and Support RCCAs Primary Guide Cards 304 SS Guide and Support RCCAs Primary GTA C-Tubes 304 SS Guide and Support RCCAs No Additional Measures GTA Sheaths 304 SS Guide and Support RCCAs No Additional Measures GTA Support Pins 316 SS Guide and Support RCCAs Existing Program (Note 1)

GTA Bolting 316 SS Guide and Support RCCAs No Additional Measures Note 1: There is no formal Program Document for the GTA Support Pin Replacement Program; however appropriate actions will be taken upon receiving further recommendations from Westinghouse. Details about this program are provided in the response to LAI #3 in Attachment 2.

Turkey Point Units 3 & 4 L-2012-438, TABLE 2 Attachment 1 Westinghouse Plants Existing Programs Components Page 13 of 22 EFFECT REFERENCE GENERIC REQUIREMENT EXAMINATION METHOD AND (MECHANISM) DOCUMENT DESCRIPTION FREQUENCY Core Barrel Assembly Turkey Point 3 Loss of material ASME Code Visual (VT-3) examination to All accessible surfaces; one time Core barrel flange Turkey Point 4 (wear) Section Xl determine general condition for per interval excessive wear Upper Internals Assembly Turkey Point 3 Cracking (SCC, ASME Code Visual (VT-3) examination All accessible surfaces; one time Upper support ring or skirt Turkey Point 4 Fatigue) Section X1 per interval Lower Internals Assembly Turkey Point 3 Cracking ASME Code Visual (VT-3) examination of the lower All accessible surfaces; one time Lower core plate Turkey Point 4 (IASCC, fatigue), Section Xl core plates to detect evidence of per interval distortion and/or loss of bolt integrity Lower Internals Assembly Turkey Point 3 Loss of material ASME Code Visual (VT-3) examination All accessible surfaces; one time Lower core plate Turkey Point 4 (wear) Section Xl per interval Bottom Mounted Turkey Point 3 Loss of material BMI-FTT-IP; Surface (ET) examination ET surface examination of full Instrumentation System Turkey Point 4 (wear) NUREG-1801, length tubes at frequency specified Flux thimble tubes Rev. 1 in BMI-FTT-IP. Tube selection and frequency based upon engineering evaluation of previous examination results.

Alignment and Interfacing Turkey Point 3 Loss of material ASME Code Visual (VT-3) examination All accessible surfaces; one time Components Turkey Point 4 (wear) (Note 1) Section Xl per interval Clevis insert bolt Alignment and Interfacing Turkey Point 3 Loss of material ASME Code Visual (VT-3) examination All accessible surfaces; one time Components Turkey Point 4 (wear) Section Xl per interval Upper core plate alignment pins (wear)

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

Turkey Point Units 3 & 4 TABLE 3 Westinghouse Plants Primary Components EFFECT EXPANSION EXAMINATION EXAMINATION COMPONENT APPLICABILITY (MECHANISM) (NOTE 1) METHOD COVERAGE Control Rod Guide Tube Turkey Point 3 Loss of material None Visual (VT-3) examination no later 20% examination of the Assembly (CRGT) Turkey Point 4 (wear) than 2 refueling outages from the number of CRGT assemblies, Guide plates (cards) beginning of the license renewal with all guide cards within period, and no earlier than two each selected CRGT refueling outages prior to the start of assembly examined.

the license renewal period.

Subsequent examinations are required on a ten-year interval.

Control Rod Guide Tube Turkey Point 3 Cracking (SCC, Bottom-mounted Enhanced visual (EVT-1) 100% of outer (accessible)

Assembly Turkey Point 4 fatigue) instrumentation examination to determine the CRGT lower flange weld Lower flange Welds Management (BMI) column presence of crack-like surface flaws surfaces and adjacent base (IE and TE) bodies, Lower in flange welds no later than 2 metal on the adjacent base support column refueling outages from the beginning metal on the individual bodies (cast), of the license renewal period and periphery CRGT assemblies.

Upper core plate, subsequent examination on a ten- (Note 2)

Lower support year interval.

casting/forging Core Barrel Assembly Turkey Point 3 Cracking (SCC) Lower support Periodic enhanced visual (EVT-1) 100% of one side of the Upper Core Barrel Flange Turkey Point 4 column bodies examination, no later than 2 refueling accessible surfaces of the Weld (non cast) outages from the beginning of the selected weld and adjacent Core barrel outlet license renewal period and base metal (Note 4) nozzle welds subsequent examination on a ten-year interval.

Core Barrel Assembly Turkey Point 3 Cracking (SCC, Upper and lower Periodic enhanced visual (EVT-1) 100% of one side of the Upper and lower core barrel Turkey Point 4 IASCC, Fatigue) cylinder axial examination, no later than 2 refueling accessible surfaces of the cylinder girth welds welds outages from the beginning of the selected weld and adjacent license renewal period and base metal (Note 4) subsequent examination on a ten-year interval.

Core Barrel Assembly Turkey Point 3 Cracking (SCC, None Periodic enhanced visual (EVT-1) 100% of one side of the Lower core barrel flange Turkey Point 4 Fatigue) examination, no later than 2 refueling accessible surfaces of the weld (Note 5) outages from the beginning of the selected weld and adjacent license renewal period and base metal (Note 4) subsequent examination on a ten-year interval.

Turkey Point Units 3 & 4 TABLE 3 Westinghouse Plants Primary Components EFFECT EXPANSION EXAMINATION EXAMINATION COMPONENT APPLICABILITY (MECHANISM) (NOTE 1) METHOD COVERAGE Baffle-Former Assembly Turkey Point 3 Cracking (IASCC, None Visual (VT-3) examination, with Bolts and locking devices on Baffle-edge bolts Turkey Point 4 fatigue) that results baseline examination between 20 high fluence seams. 100% of in and 40 EFPY and subsequent components accessible from

" Lost or broken examinations on a ten-year interval, core side. (Note 3) locking devices

Failed or missing bolts

Protrusion of bolt heads Aging Management (IE and ISR) (Note 6)

Baffle-Former Assembly Turkey Point 3 Cracking (IASCC, Lower support Baseline volumetric (UT) 100% of accessible bolts Baffle-former bolts Turkey Point 4 fatigue) Aging column bolts, examination between 25 and 35 (Note 3). Heads accessible Management (IE Barrel-former EFPY, with subsequent examination from the core side. UT and ISR) (Note 6) bolts on a ten-year interval, accessibility may be affected by complexity of head and locking device designs.

Baffle-Former Assembly Turkey Point 3 Distortion (void None Visual (VT-3) examination to check Core side surface as Assembly (Includes: Baffle Turkey Point 4 swelling), or for evidence of distortion, with indicated cracking (IASCC) baseline examination between 20

plates, indirect baffle effectsedge bolts and of void that results in and a d 40 4 EFPY F Yand a d subsequent s b e u n swingdic eformerts plates Abnormal examinations on a ten-year interval.

swelling in former plates) interaction with fuel assemblies

Gaps along high fluence baffle joint

Vertical displacement of baffle plates near high fluence joint

Broken or damaged edge bolt locking systems along high fluence baffle joint

Turkey Point Units 3 & 4 L-2012-438, TABLE 3 Attachment 1 Westinghouse Plants Primary Components Page 16 of 22 EFFECT EXPANSION EXAMINATION EXAMINATION COMPONENT APPLICABILITY (MECHANISM) (NOTE 1) METHOD COVERAGE Alignment and Interfacing Turkey Point 3 Distortion (loss of None Direct measurement of spring height Measurements should be Components Turkey Point 4 load) within three cycles of the beginning taken at several points springNote: This of the license renewal period. Ifthe around the circumference of mechanism was not first set of measurements is not the spring, with a statistically strictly identified in sufficient to determine life, spring adequate number of the original list of height measurements must be taken measurements at each point age-related during the next two outages, in order to eliminate uncertainty.

degradation to extrapolate the expected spring mechanisms. height to 60 years.

Thermal Shield Assembly Turkey Point 3 Cracking (fatigue) None Visual (VT-3) examination no later 100% of thermal shield Thermal Shield Flexures Turkey Point 4 Loss of material than 2 refueling outages from the flexures.

(wear) that results in beginning of the license renewal thermal shield period. Subsequent examinations flexures excessive on a ten-year interval.

wear, fracture, or complete separation.

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

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

3. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5, must be examined for inspection credit.

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

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

6. Void swelling effects on this component is managed through management of void swelling on the entire baffle-former assembly.

Turkey Point Units 3 & 4 L-2012-438, TABLE 4 Attachment 1 Westinghouse Plants Expansion Components Page 17 of 22 Effect Component Applicability (Mechanism) Primary Link Examination Method Examination Coverage Upper Internals Assembly Turkey Point 3 Cracking (Fatigue, CRGT lower Enhanced visual (EVT-1) 100% of accessible Upper core plate Turkey Point 4 Wear) flange weld examination, surfaces (Note 2)

Aging Management Reexamination every 10 years (IE) 3 following initial inspection.

Lower Internals Assembly Turkey Point 3 Cracking CRGT lower Enhanced visual (EVT-1) 100% of accessible Lower support forging or Turkey Point 4 Aging Management flange weld examination, surfaces (Note 2) castings (both Units (TE in casting) Reexamination every 10 years have forgings) following initial inspection.

Core Barrel Assembly Turkey Point 3 Cracking (IASCC, Baffle-former Volumetric (UT) examination. 100% of accessible bolts.

Barrel-former bolts Turkey Point 4 fatigue) bolts Re-inspection every 10 years Accessibility may be Aging Management following initial inspection, limited by presence of (IE, Void Swelling thermal shields or neutron and ISR) pads. (Note 2)

Lower Support Assembly Turkey Point 3 Cracking (IASCC, Baffle-former Volumetric (UT) examination. 100% of accessible bolts or Lower support column bolts Turkey Point 4 fatigue) bolts Re-inspection every 10 years as supported by plant Aging Management following initial inspection, specific justification.

(IE and ISR) (Note 2)

Core Barrel Assembly Turkey Point 3 Cracking (SCC, Upper core Enhanced visual (EVT-1) 100% of one side of the Core barrel outlet nozzle welds Turkey Point 4 fatigue) barrel flange examination, accessible surfaces of the Aging Management weld Re-inspection every 10 years selected weld and (IE of lower following initial inspection. adjacent base metal.

sections) (Note 2)

Core Barrel Assembly Turkey Point 3 Cracking (SCC, Upper and Enhanced visual (EVT-1) 100% of one side of the Upper and lower core barrel Turkey Point 4 IASCC) lower core examination, accessible surfaces of the cylinder axial welds Aging Management barrel cylinder Re-inspection every 10 years selected weld and (IE) girth welds following initial inspection. adjacent base metal.

(Note 2)

Lower Support Assembly NA to Turkey Cracking (IASCC) Upper core Enhanced visual (EVT-1) 100% of accessible Lower support column bodies Point 3 Aging Management barrel flange examination, surfaces (Note 2)

(non cast) NA to Turkey (IE) weld Re-inspection every 10 years Point 4 following initial inspection.

Turkey Point Units 3 &4 L-2012-438, TABLE 4 Attachment 1 Westinghouse Plants Expansion Components Page 18 of 22 Effect Component Applicability (Mechanism) Primary Link Examination Method Examination Coverage Lower Support Assembly Turkey Point 3 Cracking (IASCC) Control rod Visual (EVT-1) examination. 100% of accessible support Lower support column bodies Turkey Point 4 including the guide tube Reexamination every 10 years columns (Note 2)

(cast) detection of (CRGT) lower following initial inspection.

flanges fractured support columns Aging Management (IE)

Bottom Mounted Turkey Point 3 Cracking (fatigue) Control rod Visual (VT-3) examination of BMI 100% of BMI column Instrumentation System Turkey Point 4 including the guide tube column bodies as indicated by bodies for which difficulty Bottom-mounted detection of (CRGT) lower difficulty of insertion/withdrawal of is detected during flux instrumentation (BMI) column completely flanges flux thimbles. thimble bodies fractured column Reexamination every 10 years insertion/withdrawal.

bodies. following initial inspection.

Aging Management Flux thimble insertion/withdrawal to (IE) be monitored at each inspection interval.

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

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

3. IE must be considered in a flaw evaluation for the upper core plate. Included due to fluence change resulting from EPU implemented in 2012 (Unit 3) and 2013 (Unit 4)

Turkey Point Units 3 & 4L20 24 8 TABLE 5 Attachment8, Westinghouse Plants Examination Acceptance and Expansion Criteria Atcmn Page 19 of 22 Additional Examination Acceptance Expansion Examination item Applicability Criteria (Note 1) Link(s) Expansion Critena Acceptance Criteria Control Rod Guide Turkey Point 3 Visual (VT-3) examination. None N/A N/A Tube Assembly Turkey Point 4 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 Turkey Point 3 Enhanced visual (EVT-1) examination, a. Bottom- a. Confirmation of surface-breaking a. For BMI column Tube Assembly Turkey Point 4 mounted indications in two or more CRGT bodies, the specific Lower flange welds instrumentation lower flange welds, combined relevant condition The specific relevant condition is a (BMI) column with flux thimble for the VT-3 detectable crack-like surface indication, bodies insertion/withdrawal difficulty, examination is shall require visual (VT-3) completely fractured examination of BMI column column bodies.

b. Lower support bodies by the completion of the column bodies next refueling outage.

(cast), and b. For cast lower upper core support column plate and lower b. Confirmation of surface-breaking bodies, upper core support forging indications in two or more CRGT plate and lower or casting lower flange welds shall require support EVT-1 examination of cast lower forgings/castings, support column bodies, upper the specific relevant core plate and lower support condition is a forgings/castings within three detectable crack-fuel cycles following the initial like surface observation, indication.

Turkey Point Units 3 &4 L-2012-438, TABLE 5 L-2012-438, Westinghouse Plants Examination Acceptance and Expansion Criteria Attachment 1 Page 20 of 22 Additional Item Applicability Examination Acceptance Expansion Criteria Examination Criteria (Note 1) Link(s) Expansion Acceptance Criteria Core Barrel Turkey Point 3 Periodic enhanced visual (EVT-1) a. Core barrel a. The confirmed detection and a and b. The specific Assembly Turkey Point 4 examination, outlet nozzle sizing of a surface-breaking relevant condition Upper core barrel welds indication with a length greater for the expansion flange weld than two inches in the upper core barrel outlet The specific relevant condition is a core barrel flange weld shall nozzle weld and detectable crack-like surface indication. b. Lower support require that the EVT-1 lower support column bodies examination be expanded to column body (non cast) include the core barrel outlet examination is a nozzle welds by the completion detectable crack-of the next refueling outage. like surface

b. If extensive cracking in the indication.

remaining core barrel outlet nozzle welds is detected, EVT-1 examination shall be expanded to include the upper six inches of the accessible surfaces of the non-cast lower support column bodies within three fuel cycles following the initial observation.

Core Barrel Turkey Point 3 Periodic enhanced visual (EVT-1) None None None Assembly Turkey Point 4 examination.

Lower core barrel The specific relevant condition is a flange weld (Note 2) detectable crack-like surface indication.

Core Barrel Turkey Point 3 Periodic enhanced visual (EVT-1) Upper core barrel The confirmed detection and sizing of The specific relevant Assembly Turkey Point 4 examination, cylinder axial welds a surface-breaking indication with a condition for the Upper core barrel The specific relevant condition is a length greater than two inches in the expansion upper core cylinder girth welds detectable crack-like surface indication, upper core barrel cylinder girth welds barrel cylinder axial shall require that the EVT-1 weld examination is a examination be expanded to include detectable crack-like the upper core barrel cylinder axial surface indication.

welds by the completion of the next refueling outage.

Turkey Point Units 3 &4 L-2012-438, TABLE 5 Attachment 1 Westinghouse Plants Examination Acceptance and Expansion Criteria Page 21 of 22 Additional Item Applicability Examination Acceptance Expansion Examination Criteria (Note 1) Link(s) Expansion Criteria Acceptance Criteria Core Barrel Turkey Point 3 Periodic enhanced visual (EVT-1) Lower core barrel The confirmed detection and sizing of The specific relevant Assembly Turkey Point 4 examination, cylinder axial welds a surface-breaking indication with a condition for the Lower core barrel The specific relevant condition is a length greater than two inches in the expansion lower core cylinder girth welds detectable crack-like surface indications lower core barrel cylinder girth welds barrel cylinder axial shall require that the EVT-1 weld examination is a examination be expanded to include detectable crack-like the lower core barrel cylinder axial surface indication.

welds by the completion of the next refueling outage.

Baffle-Former Turkey Point 3 Visual (VT-3) examination. None N/A N/A Assembly Turkey Point 4 Baffle-edge bolts The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.

Baffle-Former Turkey Point 3 Volumetric (UT) examination, a. Lower support a. Confirmation that more than 5% a and b. The Assembly Turkey Point 4 column bolts of the baffle-former bolts actually examination Baffle-former bolts examined on the four baffle acceptance criteria for The examination acceptance criteria for the plates at the largest distance the UT of the lower UT of the baffle-former bolts shall be b. Barrel-former from the core (presumed to be support column bolts established as part of the examination bolts the lowest dose locations) and the barrel-former technical justification. contain unacceptable indications bolts shall be shall require UT examination of established as part of the lower support column bolts the examination within the next three fuel cycles, technical justification.

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.

Turkey Point Units 3 &4 L-2012-438, TABLE 5 L-2012-438, Westinghouse Plants Examination Acceptance and Expansion Criteria Attachment 1 Page 22 of 22 Additional Item Applicability Examination Acceptance Expansion Examination Criteria (Note 1) Link(s)Expansion Criteria Acceptance Criteria Baffle-Former Turkey Point 3 Visual (VT-3) examination. None N/A N/A Assembly Turkey Point 4 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 Turkey Point 3 Direct physical measurement of spring None N/A N/A Interfacing Turkey Point 4 height.

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

Thermal Shield Turkey Point 3 Visual (VT-3) examination. None N/A N/A Assembly Turkey Point 4 Thermal shield The specific relevant conditions for thermal flexures shield flexures are excessive wear, fracture, or complete separation.

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

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

L-2012-438 Attachment I Enclosure 1 Florida Power and Light Company Turkey Point Units 3 and 4 License Renewal Reactor Vessel Internals Commitment Implementation Report UFSAR Update Mark-UP Insert

16.1.6 REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM Florida Power & Light Company (FPL) previously developed a Reactor vessel Internals (RVI) Aging Management Program (AMP) for Turkey Point units 3 and 4 that was included in its License Renewal Application (LRA) and subsequently approved by the NRC in NUREG 1759. Included in the RVI AMP was a commitment to participate in ongoing, joint industry efforts aimed at further understanding the aging effects of the RVI and to revise the Turkey Point Units 3 and 4 RVI AMP as needed. These joint industry efforts culminated in the issuance of EPRI MRP-227-A.

On December 22, 2011, Florida Power and Light Company submitted letter L-2011-531 to notify the Nuclear Regulatory commission of its intent to revise the original RVI AMP to align with MRP-227-A. Furthermore, for Turkey Point unit 3, Florida Power and Light Company committed to submit to the NRC the revised Reactor vessel Internals Inspection Plan, based upon MPR-227-A, no later than December 31, 2012.

For Turkey Point unit 4, Florida Power and Light Company committed to submit to the NRC the revised Reactor Vessel Internals Inspection Plan, based upon MRP-227-A, prior to the end of the initial operating licensing term (April 10, 2013).

The Turkey Point Reactor vessel Internals (RVI) Aging Management Program (AMP),

based upon MRP-227-A, manages the effects of aging on the RVI during the extended periods of operation. The RVI AMP is applicable to passive RVI structural components contained within the upper and lower internals assemblies. The RVI AMP specifically excludes welded attachments to the reactor vessel and consumable items such as fuel assemblies and RCCAs.

Specific aging effects/degradation mechanisms managed by the RVI AMP include: 1) cracking due to SCC, IASCC or fatigue; 2) reduction in fracture toughness due to irradiation or thermal embrittlement; 3) loss of material due to wear; 4) dimensional change due to void swelling; 5) loss of mechanical closure integrity (or preload) due to irradiation and thermal enhanced stress relaxation or creep.

Methods employed to manage these aging effects for the above components include periodic visual, volumetric and surface inspections of primary or lead components; similar inspections of expansion components when degradation is detected in primary components; strict control of corrosive chemical species in the RCS as a preventative measure for corrosion related degradation mechanisms; and periodic replacement of components when required. The RVI relies on current ASME Section XI inspection requirements for the management of aging in certain components but in no way replaces or relieves current ASME Section XI inspection requirements. The RVI AMP is a living program that will be revised as necessary in response to ongoing joint industry efforts aimed at further understanding the aging effects of the RVI.

L-2012-438 Attachment 2 Florida Power and Light Company Turkey Point Units 3 and 4 License Renewal Reactor Vessel Internals Commitment Implementation Report Plant Site Specific Confirmation and Applicability of MRP-227-A

Turkey Point Units 3 and 4 L-2012-438, RVI Commitment Implementation Report Attachment 2 Confirmation and Acceptability of Implementing MRP-227-A at Turkey Point Page 1 of 10 The NRC Staff in the NRC SE for MRP-227, Revision 0, identified some issues and concerns that were not adequately resolved and are related to specific licensee action items related to the use of MRP-227. These plant-specific action items (LAIs) address topics related to the implementation of MRP-227 that could not be effectively addressed on a generic basis in MRP-227, Revision 0 and are identified in Section 4.2 of the NRC SE.

FPL provides confirmation and acceptability of the implementation of MRP-227-A for Turkey Point site by addressing each of the applicable site-specific licensee actions items outlined in section 4.2 of the NRC SE.

1. Applicability of Failure Modes Effects and Criticality Analysis (FMECA) and Functionality Analysis Assumptions As addressed in Section 3.2.5.1 of the NRC SE for MRP-227, each applicant/licensee is responsible for assessing its plant's design and operating history and demonstrating that the approved version of MRP-227 is applicable to the facility. Each applicant/licensee shall refer, in particular, to the assumptions regarding plant design and operating history made in the FMECA and functionality analyses for reactors of their design (i.e., Westinghouse, CE, or B&W) which support MRP-227 and describe the process used for determining plant-specific differences in the design of their RVI components or plant operating conditions, which result in different component inspection categories. The applicant/licensee shall submit this evaluation for NRC review and approval as part of its application to implement the approved version of MRP-227. This is Applicant/Licensee Action Item 1.

FPL Response to LAI #1 and Bounding Assumption of MRP-227-A:

MRP-227-A, Section 2.4, includes three bounding assumptions concerning its applicability to individual licensees: 1) Operation of 30 years or less with high-leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel strategy for the remaining 30 years of operation; 2) Base load operation, i.e., typically operates at fixed power levels and does not vary power on a calendar or load demand schedule; and 3) No design changes beyond those identified in general industry guidance or recommended by the original vendors. These bounding assumptions are directly related to LAI #1 and, as such, are addressed in the response to this LAI.

The response to LAI #1 is based directly upon Westinghouse calculation document, CN-RIDA-12-54, "Turkey Point Units 3 and 4 Reactor Internals MRP-227-A Applicant/Licensee Action Items 1 and 2."

The process used to verify that Turkey Point Units 3 and 4 are reasonably represented by the generic industry program assumptions with regard to neutron fluence, temperature, materials, and stress values used in the development of MRP-227-A is as follows:

1. Identification of typical Westinghouse pressurized water reactor (PWR) internal components (MRP-1 91, Table 4-4).

2. Identification of Turkey Point Units 3 and 4 PWR internal components.

3. Comparison of the typical Westinghouse PWR internal components to the Turkey Point Units 3 and 4 PWR internal components. Confirm that:

a.) No additional items were identified by this comparison (primarily supports A/LAI 2).

b.) The materials identified for Turkey Point Units 3 and 4 are consistent with those materials identified in (MRP-191, Table 4-4).

c.) Turkey Point Units 3 and 4 internals are the same as, or equivalent to, the typical Westinghouse PWR internals regarding design and fabrication.

4. Confirmation that the Turkey Point Units 3 and 4 operating history is consistent with the assumptions in MRP-227-A regarding core loading patterns.

Turkey Point Units 3 and 4 L-2012-438, RVI Commitment Implementation Report Attachment 2 Confirmation and Acceptability of Implementing MRP-227-A at Turkey Point Page 2 of 10

5. Confirmation that Turkey Point Units 3 and 4 operated at base load:

6. Confirm that the Turkey Point Units 3 and 4 RVI materials operated at temperatures within the original design basis parameters.

7. Determination of stress values based on design basis documents.

8. Confirmation that any changes to the Turkey Point Units 3 and 4 RVI components do not impact the application of the MRP-227-A generic aging management strategy.

Turkey Point Units 3 and 4 reactor internal components are represented by the design and operating history assumptions regarding neutron fluence, temperature, materials, and stress values in the generic FMECA of MRP-191 and the functionality analyses of MRP-232 based on the following:

1. Turkey Point Units 3 and 4 operating history is consistent with the assumptions in MRP-227-A with regard to neutron fluence.

a. The FMECA and functionality analyses for MRP-227-A were based on the assumption of 30 years of operation with high-leakage core loading patterns followed by 30 years of low-leakage core fuel management strategy. Turkey Point Units 3 and 4 fuel management program changed from a high leakage to low-leakage core loading pattern prior to 30 years of operation. Unit 3 switched from high-leakage to low-leakage loading pattern at 24 years of operation. Unit 4 switched from high-leakage to low-leakage loading pattern at 23 years of operation. Turkey Point Units 3 and 4 continues to use a low-leakage core design for all subsequent fuel cycles.

Therefore, Turkey Point Units 3 and 4 meets the fluence and fuel management assumptions in MRP-191 and requirements for MRP-227-A application.

b. Turkey Point Units 3 and 4 have operated under base load conditions over the life of the plant.

Therefore, Turkey Point Units 3 and 4 satisfies the assumptions in MRP documents regarding operational parameters affecting fluence.

2. At the highest Tavg, the Turkey Point Units 3 and 4 RVI operate between Thot and TcoId, which are not less than approximately 549.2 0 F (546.6 0 F prior to extended power uprate (EPU) for Tcold or higher than 616.8 0F (607.8°F prior to EPU) for Thot. The no-load temperature for the vessel is 547'F.

Therefore, Turkey Point Units 3 and 4 historical operation is within the original design basis parameters and is consistent with the assumptions used to develop the MRP-227-A aging management strategy with regard to temperature operational parameters.

3. Turkey Point Units 3 and 4 internal components and materials are comparable to the typical Westinghouse PWR internal components (MRP-191, Table 4-4).

a. No additional components were identified for Turkey Point Units 3 and 4 by this comparison.

b. Most of the Turkey Point Units 3 and 4 RVI component materials are consistent or nearly equivalent with those materials identified in MRP-1 91, Table 4-4, for Westinghouse-designed plants, except for one component, Upper Instrumentation Columns and Supports, that was fabricated from CF8 cast austenitic stainless steel (CASS) material rather than the Type 304 stainless steel (SS) called out in MRP-1 91. This item is further addressed in the response to LAI #7 below.

c. The design and fabrication of Turkey Point Units 3 and 4 RVI components are the same as, or equivalent to, the typical Westinghouse-designed PWR RVI components.

4. The EPU to 2644 MWt has been implemented for Turkey Point Unit 3 and is planned for Turkey Point Unit 4. An evaluation of the EPU impact on the temperature, fluence and loading of the2 RVI revealed that the estimated fluence for the upper core plate was elevated above lx1 1 n/cm (E>1.0 MeV) whereas MRP-191 assumed a fluence range of 7x10 20 to 1x10 21 n/cm 2 (E>1.0 MeV).

As a result, the upper core plate would now exceed the threshold values for IASCC and IE in MRP-

Turkey Point Units 3 and 4 L-2012-438 RVI Commitment Implementation Report Attachment 2 Confirmation and Acceptability of Implementing MRP-227-A at Turkey Point Page 3 of 10 175. The upper portions of the BMI support columns have also been identified as exceeding the MRP-175 fluence thresholds for IASCC and IE under EPU conditions. The impact of these changes on the applicability of MRP-227-A is discussed below.

As noted in MRP-191, both fluence and stress thresholds must be exceeded for IASCC to occur.

All fluences exceeding the screening value are found at distances less than 58 inches from the upper core plate center point. A review of the information calculated for stress shows that the stress exceeds the screening value only at the outside perimeter of the upper core plate, or approximately 64 inches from the upper core plate center location. The stress calculation considered all plant operational levels (A to D).

Since the fluence and stress that exceed the screening values do not occur simultaneously at the same locations and are at a relatively distant proximity, it is concluded that the potential for IASCC to manifest as a mechanism is not credible for the upper core plate. Reduction of fracture toughness due to IE is only relevant if a flaw is discovered and fracture mechanics analysis is required. Since IASCC is not considered credible, there is no reasonable expectation of flaws in the upper core plate. Furthermore, the upper core plate is included as an Expansion Component in MRP-227-A, linked to the detection of surface breaking indications in the Control Rod Guide Tube Assembly Lower Flange Welds. It is therefore concluded that the effects of IE and IASCC as a result of the marginally higher fluence value following the EPU are adequately managed by the implementation of the MRP-227-A inspection strategy for management of material degradation. IE has been added as a degradation mechanism for the upper core plate in Table 4, Westinghouse Plants Expansion Components, and will be considered in any flaw evaluation for the upper core plate.

The BMI columns screened in for both IASCC and IE in MRP-191 as the assumed fluence values exceeded the thresholds of MRP-175. Therefore, MRP-227-A effectively addresses both degradation mechanisms for the Turkey Point Units 3 and 4 BMI Columns under EPU conditions.

Modifications to the Turkey Point Units 3 and 4 RVI made over the lifetime of the plant are those identified in general industry guidance or specifically directed by Westinghouse, and no modifications have been made since May 2007, other than the split pin replacements in 2007 and 2008 for Unit 3 and Unit 4, respectively. Therefore, the Turkey Point Units 3 and 4 components have remained within the original structural design configuration and stress values are represented by the assumptions in MRP-227-A, MRP-191, and MRP-232.

Conclusion Turkey Point Units 3 and 4 comply with LAI #1 of the NRC SE on MRP-227, Revision 0, and therefore meets the requirement for application of MRP-227-A as a strategy for managing age- related material degradation in reactor internal components. Additionally, Turkey Point Units 3 and 4 satisfy the three bounding assumptions of MRP-227-A, outlined in Section 2.4. This satisfies the requirements of LAI #1.

2. PWR Vessel Internal Components Within the Scope of License Renewal As discussed in Section 3.2.5.2 of this SE, consistent with the requirements addressed in 10 CFR 54.4, each applicant/licensee is responsible for identifying which RVI components are within the scope of LR for its facility. Applicants/licensees shall review the information in Tables 4-1 and 4-2 in MRP-1 89, Revision 1, and Tables 4-4 and 4-5 in MRP-1 91 and identify whether these tables contain all of the RVI components that are within the scope of LR for their facilities in accordance with 10 CFR 54.4. If the tables do not identify all the RVI components that are within the scope of LR for its facility, the applicant or licensee shall identify the missing component(s) and propose any necessary modifications to the program defined in MRP-227, as modified by this SE, when submitting its plant-specific AMP. The AMP shall provide assurance that the effects of aging on the missing component(s) will be managed for the period of extended operation. This issue is Applicant/Licensee Action Item 2.

Turkey Point Units 3 and 4 L-2012-438 RVI Commitment Implementation Report Attachment 2 Confirmation and Acceptability of Implementing MRP-227-A at Turkey Point Page 4 of 10 FPL Response to LAI# 2:

The response to LAI #2 is based directly upon Westinghouse calculation CN-RIDA-12-54 "Turkey Point Units 3 and 4 Reactor Internals MRP-227-A Applicant/Licensee Action Items 1 and 2."

This action item requires comparison of the RVI components that are within the scope of license renewal for Turkey Point Units 3 and 4 to those components contained in MRP-1191, Table 4-4. All components required to be included in the Turkey Point Units 3 and 4 RVI AMP are consistent with those contained in MRP-1 91.

The Turkey Point RVI components, listed below, have different materials than specified in MRP-191, Table 4-4, but these have no effect on the recommended MRP aging management strategy or are already managed by an alternate Turkey Point aging management program (Thimble Tube Inspection Program) and no modifications to the program contained in MRP-227-A need to be proposed.

Component MRP-191 Turkey Point Units Disposition Material 314 Material Upper Core Plate Fuel 316 SS 304 SS Both are wrought austenitic SS alloys Alignment Pin with the same screening criteria for all degradation mechanisms. No change to aging management strategy is required.

Upper Instrumentation 304 SS CF8 (Unit 4 only) Component did not screen in for any Conduit & Supports degradation mechanisms under MRP-191. See response to LAI #7 for further discussion of CASS susceptibility to IE and TE.

Upper Support Plate 316 SS 304 SS Both are wrought austenitic SS alloys Assembly Lock Key with the same screening criteria for all degradation mechanisms. No change to aging management strategy is required.

Flux Thimble Tube Plug 304 SS 308 SS Both are wrought austenitic SS alloys with the same screening criteria for all degradation mechanisms. No change to aging management strategy is required.

Lower Core Plate Fuel 316 SS 304 SS Both are wrought austenitic SS alloys Alignment Pins with the same screening criteria for all degradation mechanisms. No change to aging management strategy is required.

Lower Support Column 304 SS 316 SS Both are wrought austenitic SS alloys Bolts with the same screening criteria for all degradation mechanisms. No change to aging management strategy is required.

Thermal Shield Dowel 316 SS 304 SS Both are wrought austenitic SS alloys with the same screening criteria for all degradation mechanisms. No change to aging management strategy is required.

Thermal Shield Flexures 304 SS 316 SS Both are wrought austenitic SS alloys with the same screening criteria for all degradation mechanisms. No change to aging management strategy is required.

Radial Support Key Bolts 304 SS 316 SS Both are wrought austenitic SS alloys with the same screening criteria for all degradation mechanisms. No change to aging management strategy is required.

Turkey Point Units 3 and 4 L-201A2-438 RVI Commitment Implementation Report Attachment 2 Confirmation and Acceptability of Implementing MRP-227-A at Turkey Point Page 5 of 10 One of these components, the Upper Instrumentation Columns and Supports, was found to be constructed of CASS rather than wrought 304 SS and is further addressed in the response to LAI #7.

The above information addresses the requirement that the Turkey Point Units 3 and 4 RVI AMP provide reasonable assurance that the effects of aging on missing RVI components will be managed for the period of extended operation. The generic scoping and screening of the RVI as summarized in MRP-191 and MRP-232 to support the inspection sampling approach for aging management of RVI specified in MRP-227-A is applicable to Turkey Point Units 3 and 4 with no modifications.

Conclusion Turkey Point Units 3 and 4 comply with LAI# 2 of the NRC SE on MRP-227, Revision 0, and therefore meet the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internal components. This satisfies the requirements of LAI #2.

3. Evaluation of the Adequacy of Plant-Specific Existing Programs As addressed in Section 3.2.5.3 in this SE, applicants/licensees of CE and Westinghouse are required to perform plant-specific analysis either to justify the acceptability of an applicant's/licensee's existing programs, or to identify changes to the programs that should be implemented to manage the aging of these components for the period of extended operation. The results of this plant-specific analyses and a description of the plant-specific programs being relied on to manage aging of these components shall be submitted as part of the applicant's/licensee's AMP application. The CE and Westinghouse components identified for this type of plant-specific evaluation include: CE thermal shield positioning pins and CE in-core instrumentation thimble tubes (Section 4.3.2 in MRP-227), and Westinghouse guide tube support pins (split pins) (Section 4.3.3 in MRP-227). This is Applicant/Licensee Action Item 3.

FPL Response to LAI# 3:

Existing plant programs credited for adequately managing specific aging effects of selected RVI components in the Turkey Point Units 3 and 4 RVI AMP include:

" Chemistry Control Program

ASME Section Xl In-Service Inspection Program

Control Rod Guide Tube (CRGT) Support Pin (Split) Replacement

Flux Thimble Tube Inspection Program Chemistry Control Program The Turkey Point Units 3 and 4 Chemistry Control Program is credited for controlling the levels of corrosive contaminants in the Primary Water System, thereby preventing or mitigating cracking of RVI components by SCC and IASCC. The Chemistry Control Program is not listed in the Attachment 2, RVI Component Details, Turkey Point Units 3 & 4 since it does not include any inspections of RVI components.

ASME Section Xl In-Service Inspection Program ASME Section Xl, IWB-2500, Examination Category B-N-3 provides inspection requirements for removable RVI components categorized as core support structures in the ASME Code. Visual inspections (VT-3) of the applicable components' accessible surfaces are required one time per interval. Relevant conditions requiring correction are described in ASME Section Xl, IWB-3520. The Turkey Point Units 3 and 4 ASME Section Xl Program is implemented under QI 11-1 0-PTN-4. The licensing basis, program basis documents and implementing procedures of this program were reviewed by the NRC (Region 2) Inspectors during the License Renewal Inspections performed during June 2012.

Turkey Point Units 3 and 4 L-2012-438 RVI Commitment Implementation Report Attachment 2 Confirmation and Acceptability of Implementing MRP-227-A at Turkey Point Page 6 of 10 CRGT Support Pins (Split Pins) Replacement The original Control Rod Guide Tube (CRGT) support pins were fabricated from X-750 alloy with a heat treatment (Rev. A) that was later determined to have rendered them susceptible to stress corrosion cracking based upon external operating experience (OE). Replacement support pins, also manufactured of X-750 but with a modified heat treatment (Rev. B), were installed in Units 3 and 4 in 1985 and 1986, respectively, per recommendations from Westinghouse. The second generation X-750 (Rev. B) support pins were later found to also be susceptible to SCC, again based upon external OE. Based on the recommendations from Westinghouse replacement support pins constructed of cold worked (CW) 316 SS were installed in Units 3 and 4 in 2007 and 2008, respectively, Prior to installation, the replacement CW 316 SS were evaluated for resistance to the eight degradation mechanisms of concern for the RVI : 1) stress corrosion cracking (SCC); 2) irradiation assisted stress corrosion cracking (IASCC); 3) fatigue; 4) irradiation embrittlement (IE); 5) thermal embrittlement (TE); 6) wear; 7) void swelling; and 8) irradiation and thermal enhanced stress relaxation/creep. None of the eight degradation mechanisms were found to be a concern for the CW 316 SS material over a 40 year design life. Therefore, there is no need for augmented inspections of the CW 316 SS CRGT support pins under the RVI AMP.

There is no formal Aging Management Program Document for the CRGT Support Pin Replacement Program, however appropriate actions will be taken upon receiving further recommendations from Westinghouse. This program is included in the Aging Management Program (AMP) as an 'Existing Program' for RV Internals.

Flux Thimble Tubes.

The Thimble Tube Inspection Program manages the aging effect of material loss due to fretting wear. This program consists of an eddy current test (ECT) inspection of thimble tubes.

Wear on the BMI flux thimble tubes is detected by the eddy current testing (ECT), a volumetric technique that reads changes in the "volume" of the thimble wall. This technique is well known in the industry, and has been used to detect imperfections in other components tubing such as steam generators, heat exchangers etc. The ECT technique has been already used (at least two times) to detect wall thinning on the thimble tubes with satisfactory results. The ECT is performed in accordance with approved plant procedure by qualified personnel. This program has been effective in identifying thimble tube wall thinning.

Eddy current testing of thimble tubes was initiated in response to the NRC Bulletin 88-09, "Thimble Tube Thinning in Westinghouse Reactors". The original commitment under the Bulletin was to perform a one-time inspection of thimble tube N-05 in Unit 3. Since degradation was identified in thimble tube N-05 and other examined locations that could not be demonstrated to last throughout the period of extended operation, a periodic inspection program was developed in order to properly manage aging. The program is the Bottom Mounted Instrumentation Flux Thimble Tubing Inspection Program for Turkey Point Units 3 & 4 (BMI-FTT-IP). With this inspection program, the structural integrity of the RPV thimble tubes regarding fretting wear concerns is ensured for the extended operating period of 60 years. The licensing basis, program basis documents, implementing procedures of this program were reviewed by the NRC (Region 2)

Inspectors during the Phase 2 License Renewal Post-Approval Site Inspections performed during June 2012.

Conclusion Turkey Point Units 3 and 4 Existing Programs adequately manage the effects of aging for applicable components during the period of extended operation. No changes to the current programs are required.

This satisfies the requirements of LAI #3.

Turkey Point Units 3 and 4 L-201A2-438 RVI Commitment Implementation Report Attachment 2 Confirmation and Acceptability of Implementing MRP-227-A at Turkey Point Page 7 of 10

4. B&W Core Support Structure Upper Flange Stress Relief As discussed in Section 3.2.5.4 of this SE, the B&W applicants/licensees shall confirm that the core support structure upper flange weld was stress relieved during the original fabrication of the Reactor Pressure Vessel in order to confirm the applicability of MRP-227, as approved by the NRC, to their facility. If the upper flange weld has not been stress relieved, then this component shall be inspected as a "Primary" inspection category component. If necessary, the examination methods and frequency for non-stress relieved B&W core support structure upper flange welds shall be consistent with the recommendations in MRP-227, as approved by the NRC, for the Westinghouse and CE upper core support barrel welds. The examination coverage for this B&W flange weld shall conform to the staff's imposed criteria as described in Sections 3.3.1 and 4.3.1 of this SE. The applicant's/licensee's resolution of this plant-specific action item shall be submitted to the NRC for review and approval. This is applicant /Licensee Action Item 4.

FPL Response to LAI# 4:

This item pertains to B&W Core Support Structure Upper Flange Stress Relief issue. Turkey Point Units 3 and 4 are Westinghouse NSSS plants, therefore this LAI is not applicable.

5. Application of Physical Measurements as part of M&E Guidelines for B&W. CE. and Westinghouse RVI Components As addressed in Section 3.3.5 in this SE, applicants/licensees shall identify plant-specific acceptance criteria to be applied when performing the physical measurements required by the NRC-approved version of MRP-227 for loss of compressibility for Westinghouse hold down springs, and for distortion in the gap between the top and bottom core shroud segments in CE units with core barrel shrouds assembled in two vertical sections. The applicant/licensee shall include its proposed acceptance criteria and an explanation of how the proposed acceptance criteria are consistent with the plants' licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version of MRP-227.

This is Applicant/Licensee Action Item 5.

FPL Response to LA# 5:

The Turkey Point Units 3 and 4 Hold Down Spring (HDS) acceptance criteria are based on the measured height of the springs as a function of time relative to the required hold-down force. The criteria is based upon a conservative assumption that relaxation occurs linearly over time. The approach used to develop the HDS height acceptance criteria included consideration of the actual HDS height at plant start-up and the required HDS heights at the end of 60 years, under applicable plant loading conditions. A linear interpolation at the time of the HDS height measurement determines the required minimum HDS height.

For HDS height measurements less than the required minimum HDS heights, re-evaluation and successive measurements or a replacement HDS is required.

Time dependent details for the HDS height measurements are summarized in Westinghouse document

[LTR-RIDA-12-170]. Plant-specific details are proprietary and not typically released publicly. If the NRC requests additional details, the calculation can be made available for review. This satisfies the requirements of LAI #5.

6. Evaluation of Inaccessible B&W Components As addressed in Section 3.3.6 in this SE, MRP-227 does not propose to inspect the following inaccessible components: the B&W core barrel cylinders (including vertical and circumferential seam welds), B&W former plates, B&W external baffle-to-baffle bolts and their locking devices, B&W core barrel-to-former bolts and their locking devices, and B&W core barrel assembly internal baffle-to-baffle bolts. The MRP also

L-201 2-438 Turkey Point Units 3 and 4 Lttac2 ent8 RVI Commitment Implementation Report Attachment 2 Confirmation and Acceptability of Implementing MRP-227-A at Turkey Point Page 8 of 10 identified that although the B&W core barrel assembly internal baffle-to-baffle bolts are accessible, the bolts are non-inspectable using currently available examination techniques. Applicants/licensees shall justify the acceptability of these components for continued operation through the period of extended operation by performing an evaluation, or by proposing a scheduled replacement of the components. As part of their application to implement the approved version of MRP-227, applicants/licensees shall provide their justification for the continued operability of each of the inaccessible components and, if necessary, provide their plan for the replacement of the components for NRC review and approval. This is Applicant/Licensee Action Item 6.

FPL Response to LAI# 6:

This item pertains to B&W Core Support Structure Upper Flange Stress Relief issue. Turkey Point Units 3 and 4 are Westinghouse NSSS plants; therefore this LAI is not applicable.

7. Plant-Specific Evaluation of CASS Materials As discussed in Section 3.3.7 of this SE, the applicants/licensees of B&W, CE, and Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate that B&W IMI guide tube assembly spiders and CRGT spacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their functionality during the period of extended operation or for additional RVI components that may be fabricated from CASS, martensitic stainless steel or precipitation hardened stainless steel materials. These analyses shall also consider the possible loss of fracture toughness in these components due to thermal and irradiation embrittlement, and may also need to consider limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques.

The requirement may not apply to components that were previously evaluated as not requiring aging management during development of MRP-227. That is, the requirement would apply to components fabricated from susceptible materials for which an individual licensee has determined aging management is required, for example during their review performed in accordance with Applicant/Licensee Action Item 2.

The plant-specific analysis shall be consistent with the plant's licensing basis and the need to maintain the functionality of the components being evaluated under all licensing basis conditions of operation. The applicant/licensee shall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-227. This is Applicant/Licensee Action Item 7.

FPL Response to LAI# 7:

As noted in the response to LAI #2, a comparison of the Turkey Point Units 3 and 4 RVI components within the scope of LR to those included in MRP-191 did reveal one additional Unit 4 component, the Upper Instrumentation Columns and Supports, that was manufactured from CASS. MRP-191 had only identified wrought 304 SS as a possible materials of construction for this component. Therefore, the Upper Instrumentation Columns and Supports component as well as other previously identified CASS components are addressed in this response to LAI #7.

Four Turkey Point RVI components within the scope of LR are constructed of CASS:

1) Lower Support Columns;

2) Upper Support Column Bases;

3) BMI Cruciforms; and

4) Upper Instrumentation Columns and Supports (Unit 4 only).

The first three of these components screened in for thermal embrittlement (TE) and irradiation embrittlement (IE) during the initial screening process for MRP-227-A, as documented in Table 5-1 of MRP-191. This screening was based only upon the temperature (500'F) and fluence (6.7 x 1020 n/cm 2, E>1.0 MeV) thresholds for TE and IE, respectively, without consideration of the components' delta ferrite contents.

L-201 2-438 Turkey Point Units 3 and 4 Lttac2 ent8 RVI Commitment Implementation Report Attachment 2 Confirmation and Acceptability of Implementing MRP-227-A at Turkey Point Page 9 of 10 MRP-1 91 did not consider CASS as a possible material of construction for the fourth item, but it is being included in the response to this LAI. The evaluation of the four CASS is described below.

Two of these items, Upper Support Column Bases and BMI Cruciforms, were placed in the No Additional Measures Group by the FMECA, categorization and rankings described in MRP-191 and MRP-227-A. The third item, Upper Instrumentation Columns and Supports, was not evaluated as a CASS constructed component during the MRP-191 process. Upon discovery of the CASS construction, a Turkey Point plant specific FMECA was performed by Westinghouse. This FMECA also categorized the CASS constructed Upper Instrumentation Columns and Support as a No Additional Measures component. As such, no further action is required for the Upper Support Column Bases, BMI Cruciforms, and Upper Instrumentation Columns and Supports to insure functionality during the period of extended operation.

The Lower Support Columns were placed in the Expansion Group by the FMECA, categorization and rankings described in MRP-191 and MRP-227-A. As such, supplemental EVT-1 inspections are required for the Lower Support Columns if cracking is detected in the linked Primary Component, the Guide Tube Assembly Lower Flange Welds. Further evaluation of the Lower Support Columns for TE and IE utilizing the guidance provided in NRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Stainless Steel Components", is described below.

In an effort to better determine the susceptibility of the Lower Support Columns to TE, FPL has performed a search of the original manufacturing records. Certified Material Test Reports (CMTR) were located for all of the Unit 3 and Unit 4 lower support columns, sixty eight (68) total for each Unit. The ferrite content for the Lower Support Columns was then calculated using Hull's equivalent factors (Ref. NUREG/CR4513). The calculated ferrite for these components ranged from 4.29% to 14.83%, well below the 20% threshold for TE susceptibility as described in NRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Stainless Steel Components".

To address IE, FPL has developed acceptance criteria for the Lower Support Columns in accordance with WCAP-17096 which is currently under review by the NRC (Ref. WCAP-17449-P). This analysis determined the minimum functional requirements and number of Lower Support Columns required to maintain structural and functional stability under normal and faulted conditions, considering the effects of TE and IE. The analysis results showed that redundancy did exist for the Lower Support Columns. Specifically, the dimensional stability of the Lower Core Plate is maintained if up to two columns have failed, and the columns are more than 16 inches apart.

Finally, the manufacturing records for the Lower Support Columns also indicated that the components were inspected by liquid penetrant prior to delivery. This inspection would have minimized the chances of a preexisting fabrication defect contributing to an in-service failure resulting from a reduction in fracture toughness due to TE and/or IE.

Conclusion The results of the manufacturing record search and the acceptance criteria development for the Turkey Point Units 3 and 4 Lower Support Columns demonstrate that the current inspection requirements of MRP-227-A will adequately manage the aging effects associated with TE and IE, and will insure functionality during the period of extended operation. This evaluation was performed in accordance with the guidance provided in NRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Stainless Steel Components". Therefore, the requirements of LAI #7 are satisfied.

Turkey Point Units 3 and 4 RVI Commitment Implementation Report Attachment 2 Confirmation and Acceptability of Implementing MRP-227-A at Turkey Point Page 10 of 10

8. Submittal of Information for Staff Review and Approval As addressed in Section 3.5.1 in this SE, applicants/licensees shall make a submittal for NRC review and approval to credit their implementation of MRP-227, as amended by this SE, as an AMP for the RVI components at their facility. This submittal shall include the information identified in Section 3.5.1 of this SE.

This is Applicant/Licensee Action Item 8.

FPL Response to LAI# 8:

During the license renewal process, Turkey Point Units 3 and 4 prepared and gained approval for RVI Aging Management Programs (AMPs) from the NRC, as documented in NUREG 1759. Subsequently, Turkey Point Units 3 and 4 committed to revise these Aging Management Programs to align with MRP-227-A (Ref. FPL Letter L-2011-531 to NRC). Revisions to the License Renewal basis document for the Turkey Point Units 3 and 4 RVI AMPs have been completed under the 50.59 Process.

The Turkey Point Unit 3 current license bases (CLB) include the following commitments: 1) for Turkey Point Unit 3 implement its revised RVI AMP by December 31, 2012; and 2) submit its RVI Inspection Plan to the NRC. Similarly, the Turkey Point Unit 4 CLB includes the following commitments: 1) implement its revised RVI AMP prior to the end of the initial operating licensing term (April 10, 2013); and 2) submit its RVI Inspection Plan to the NRC prior to the end of the initial operating licensing term (April 10, 2013).

The implementing procedure for the Turkey Point Units 3 and 4 RVI AMP is 0-ADM-563, Reactor Vessel Internals Aging Management Program. Revision 0 of this procedure was based upon MRP-227, Rev. 0 and was in place prior to December 31, 2011 to satisfy an NEI 03-08 Materials Initiative requirement. Revision 1 of 0-ADM-563 has been issued to incorporate the changes associated with the implementation of MRP-227-A for Turkey Point Units 3 and 4.

The 10 key attributes of the revised RVI AMP are provided in Attachment 1 of this letter. The Turkey Point Units 3 and 4 RVI Inspection Plan, has been prepared based upon the LR AMP basis document, (PTN-ENG-LRAM-00-0041, Rev. 4) which credits implementation of MRP-227-A. As discussed in Attachment 1 of this letter, the RVI Inspection Plan includes the following items: the degradation mechanisms of concern; the components to be inspected; the inspection methods and frequency; the examination coverage; and the examination acceptance criteria.

The RVI commitment implementation report provided in Attachments 1 and 2 of this letter provides the required submittals to NRC as requested in LAI #8.

L-2012-438, License Renewal (Lr) Reactor Vessel Internals (Rvi) Commitment Implementation Report and Inspection Plan

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