Text

License Renewal Commitments, Reactor Vessel Internals Aging Management Plan
	ML15300A574
Person / Time
Site:	Saint Lucie
Issue date:	09/28/2015
From:	Costanzo C; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2015-229
	Download: ML15300A574 (38)
	v • d • e

0September F=PLo 28, 2015 L-201 5-229 10 CFR 54 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555-0001 St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 License Renewal Commitments Reactor Vessel Internals Apqing Manaaqement Plan

References:

1. NUREG 1779, Safety Evaluation Report Related to License Renewal of St. Lucie Nuclear Plant, Units 1 and 2, September 2003.

2. Safety Evaluation by the Office of Nuclear Reactor Regulation related to Amendment No.

213 to Facility Operating License No. DPR-67, Florida Power and Light Company, St. Lucie Plant Unit No. 1, Docket No. 50-335.

3. Safety Evaluation by the Office of Nuclear Reactor Regulation related to Amendment No.

163 to Facility Operating License No. NPF-16, Florida Power and Light Company, St. Lucie Plant Unit No. 2, Docket No. 50-389.

4. FPL Letter from Joseph Jensen to U.S. Nuclear Regulatory Commission (L-2014-1 92) "St.

Lucie Units I and 2 Docket Nos. 50-335 and 50-389, Reactor Vessel Internals Inspection Program Plans and Inspection Dates," June 25, 2014.

The following License Renewal (LR) commitments have been made regarding the St. Lucie Units 1 and 2 Reactor Vessel Internals (RVI) Inspection Program to be implemented during the period of extended operation (PEO). Each of these commitments and the manner in which it has been addressed is described below.

Commitment No. 4 of NUREG 1779 (Reference 1), the Safety Evaluation Report (SER) for the renewed operating licenses of St. Lucie Units 1 and 2, requires the submission of a report summarizing the aging effects applicable to the Reactor Vessel Internals (RVI),

including a description of the inspection plan prior to the end of the initial period of operation for St. Lucie Unit 1.

FPL's response:

As discussed in Reference 4, the RVI inspection plan for St. Lucie Unit I is scheduled for submittal to the NRC by September 30, 2015 and the RVI inspection plan for St. Lucie Unit 2 would be submitted at a later date. The attached RVI Aging Management Plan summarizes the St. Lucie Units I and 2 RVI Inspection Program and provides the age related degradation effects applicable to the RVI components, the schedule of inspections to be performed and the acceptance criteria.

Florida Power & Light Company~k

6501 S. Ocean Drive, Jensen Beach, FL 34957 ,* I

L-201 5-229 Page 2 of 2

Commitment No. 5 of NUREG 1779 requires that FPL perform a one-time inspection of the reactor vessel internals.

FPL'sresponse:

Reference 4 discussed and reaffirmed FPL's adoption of MRP-22 7-A which requires the implementation of periodic inspections for both St. Lucie Unit I and 2, and supersedes the prior commitment for a one-time inspection. As also discussed in Reference 4, the first inspection of St. Lucie Unit I RVI is currently scheduled for the Spring Outage of 2018. The first inspection of St. Lucie Unit 2 RVI will be scheduled within 3 years after PEO.

Commitment No. 12 of the SER for the Extended Power Uprate License Amendment of St. Lucie Unit 1 (Reference 2) and the fourth in a series of commitments of the SER for the Extended Power Uprate License Amendment of St. Lucie Unit 2 (Reference 3) require that FPL adopt MRP-227-A in place of its previously approved RVI Inspection Program.

FPL's response:

The attached RVI Aging Management Plan summarizes the revised St. Lucie Units 1 and 2 RVI Inspection Program which is based upon MRP-227-A.

Should you have any questions, please contact Mr. Eric Katzman, Licensing Manager, at 772-467-7734.

Very truly yours, Christopher Costanzo Site Vice President St. Lucie Plant Attachments: 1) St. Lucie Units 1 and 2 RVI Aging Management Plan

2) Proposed St. Lucie Units 1 and 2 UFSAR Revisions cc: USNRC Regional Administrator, Region II USNRC Project Manager, St. Lucie Nuclear Plant USNRC Senior Resident Inspector, St. Lucie Nuclear Plant

Attachment I to Letter L-2015-229 St. Lucie Units I and 2 RVI Aging Management Plan

St. Lucie Units I and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-201 5-229

SUMMARY

OF RVI INSPECTION PROGRAM The RVI Inspection Program was developed utilizing the EPRI MRP-227-A, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." Applicability of MRP-227-A is demonstrated by the responses to the Licensee Action Items (LAI) in Section 3. The methodology of MRP-227-A is described below.

1.1 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; and 8) irradiation and thermal enhanced stress relaxation/creep. 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 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 (>10s) cycles. Low cycle fatigue results from relatively high cyclic stress (ayield strength) applied for low number of cycles. The aging effect of fatigue is cracking.

Irradiation Embrittlement (IF)

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.

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St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-201 5-229 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 55 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.

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 RVl 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.

1.2 COMPONENT CATEGORIZATION The RVI components were screened for susceptibility to the eight degradation mechanisms based upon their chemical compositions, neutron fluence exposures, operating temperatures and stress levels. Functionality assessments were then performed on the screened-in components to determine the effects of the applicable degradation mechanism(s) on functionality. Each of the RVI components was then categorized as an Existing Program, Primary, Expansion or No Additional Measurements Component based upon the functionality analysis, component accessibility, operating history, existing evaluations and prior examination results. A description of the component categories follows:

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 Page 2 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 degradation mechanism(s). Details of the required inspections for Primary Components are provided in Table 1, CE 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 2, CE Plants Expansion Components.

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 3, Existing Programs 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.

1.3 INSPECTION OF RVI COMPONENTS Inspections detailed in Table 1, CE Plants Primary Components, and Table 3, CE Plants Existing Program Components, are required to manage aging effects in Primary Components. Additionally, inspections detailed in Table 2, CE Plants Expansion Components, are required should evidence of aging degradation be detected in linked Primary Components.

Inspection Methodologies 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 exams to monitor for general degradation associated with loss of material or preload

EVT-1 exams to monitor for surface breaking linear discontinuities indicative of cracking

UT exams to monitor directly for cracking

ECT to further characterize conditions detected by visual (VT-3, VT-i and EVT-1) exams 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.

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St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 Inspection Frequencies 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 Coverag~e The required inspection coverage for Primary and Expansion Components is specified in Tables 1 and 2, respectively. If the specified coverage 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 4, CE Plants Examination Acceptance and Expansion Criteria. 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, described in WCAP-17096-NP-A, "Reactor Internals Acceptance Criteria Methodology/and Data Requirements". The potential loss of fracture toughness must be considered in any flaw evaluations.

Additionally, plant specific acceptance criteria have been developed for the core shroud gap measurements, should they be required. The allowable gap size to insure continued functionality is based upon design and as-built conditions, fluence, circumferential bounds of the gap (how far around the core shroud can the gap exist),stress, impact on adjacent reactor vessel internals components, impact on core and bypass flow rates, and potential effects on fuel management schemes.

Expansion Components The criteria for expanding the scope of examination from the Primary to the linked Expansion Components are also provided in Table 4. Generally, the inspection of the Expansion Components is required in the RFO following that in which degradation of the linked Primary Component was detected.

It should be noted that the component categorizations and associated inspection requirements described above do not replace or relieve current ASME Section Xl inspection requirements for the RVl components.

2 INSPECTION PROGRAM ATTRIBUTES The attributes of the St. Lucie RVI Inspection Program and compliance with NUREG-1801 (GALL Report),Section XI.M16, "PWR Vessel Internals" are described in this section. The GALL identifies 10 attributes for successful component aging management. The framework for assessing the effectiveness of the projected program is established by the use of the 10 elements of the GALL.

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St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 Plan Approach and supplemental information Attribute 1 Scope of The St. Lucie RVI Inspection Program includes all Units 1 and 2 RVI Program components which were built to the CE NSSS design. Using the guidance provided in MRP-227-A, the St. Lucie RVI Inspection Program was developed to manage the aging of these components during the initial and extended periods of operation. Components considered for inspection under MRP-227-A include core support structures, RVI components that serve an intended license renewal safety function pursuant to criteria in 10CFR54.4(a)(1), and other RVI components whose failure could prevent satisfactory accomplishment of any other functions identified in 10 CFR 54.4(a)(i), (ii), or (iii). The program does not included consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation. 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 by the St. Lucie Reactor Vessel Integrity Program and AMP.

2 Preventive The St. Lucie Chemistry Control Program is credited for limiting the Measures levels of corrosive chemical species (e.g. halogens, sulfur compounds, oxygen) in the RCS to extremely low levels as a preventative measure for corrosion related degradation mechanisms including pitting, crevice corrosion, SCC, PWSCC and IASCC.

3 Parameters The St. Lucie RVl Inspection Program manages the following age-Monitored related degradation effects and mechanisms: 1) cracking induced by SCC, PWSCC, IASCC, or fatigue; 2) loss of material induced by wear; 3) loss of fracture toughness induced by either thermal aging or neutron irradiation embrittlement; 4) changes in dimension due to void swelling and irradiation growth, distortion or defection; and 5) loss of preload caused by thermal and irradiation-enhanced stress relaxation or creep.

For the management of cracking, the St. Lucie RVI Inspection Program monitors for evidence of surface breaking linear discontinuities using visual (EVTI-1) exams, or directly using volumetric (UT) or surface (ECT) exams. For the management of loss of material, the RVI Inspection Program monitors for surface conditions that may be indicative of wear using visual (VT-3) exams. For the management of changes in dimension and loss of preload, the RVI Inspection Program monitors for gross surface conditions using visual (VT-3) exams or direct physical measurements. The RVI Inspection Program does not directly monitor for loss of fracture toughness but relies on visual or volumetric examination techniques to monitor for cracking in components.

Specifically, the St. Lucie RVI Inspection Program implements the parameters monitored/inspected criteria for CE Designed Primary Components in Table 4-2 of MRP-227-A. Additionally, the program implements the parameters monitored/inspected criteria for CE designed Expansion Components in Table 4-5 of MRP-227-A. The parameters

________monitored/inspected for Existing Program Components follow the bases Page 5 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 Plan Approach and supplemental information

_______ Attribute ______________________________

___________________for the ASME Section Xl Program.

4 Detection of Discussion and justification of the inspection methods selected for Aging Effects detection of the aging effects managed by the St. Lucie RVI Inspection Program are provided in MRP-227-A and MRP-228. In all cases, well established methods described above were selected. Additionally, the RVI Inspection Program adopts the recommended guidance in MRP-227-A for defining Expansion criteria that need to be applied to inspections of Primary and Existing Program Components and for expanding the examinations to include additional Expansion Components. As a result, inspections performed on the RVI components are 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-I.

Specifically, the St. Lucie RVI Inspection Program implements the parameters monitored/inspected criteria and bases for inspecting the relevant parameter conditions for CE designed Primary Components in Table 4-2 of MRP-227-A and for CE designed Expansion Components in Table 4-5 of MRP-227-A.

The St. Lucie RVI Inspection Program is supplemented by the addition of core support barrel expandable plugs and patches to the Primary Program Components for Unit 1 only. These components were used to repair the core barrel damage associated with the loss of the thermal shield early in plant life. The aging effects monitored for included cracking due to IASCC, SCC and fatigue. Enhanced visual examinations (EVT-1) will be performed no later than 2 refueling outages from the beginning of the PEO and every 10 years thereafter.

The St. Lucie RVI Inspection Program Primary Component inspections include visual inspection (VT-3) for the presence of distortion of the core shroud due to void swelling, as evidence by separation of the assembly's upper and lower portions. If a gap exists, physical measurements are performed from the core side at the core shroud re-entrant corners. Plant specific acceptance criteria for these measurements have been developed as described below.

5 Monitoring and The methods for monitoring, recording, evaluating, and trending the data Trending that result from the St. Lucie RVI Inspection Program inspections are given in Section 6 of MRP-227-A. The evaluation methods include recommendations for flaw depth sizing and for crack growth determinations as well for performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications. 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 Page 6 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-2015-229 Plan Approach and supplemental information Attribute 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 the examination acceptance criteria Criteria for the Primary and Expansion Components in the St. Lucie RVI Inspection Program. For Existing Program components referenced to ASME Section Xl, the IWB-3500 acceptance criteria apply.

Plant specific acceptance criteria has been developed for the core shroud gap measurements, should they be required. The allowable gap size to insure continued functionality is based upon design and as-built conditions, fluence, circumferential bounds of the gap (how far around the core shroud can the gap exist),stress, impact on adjacent reactor vessel internals components, impact on core and bypass flow rates, and potential effects on fuel management schemes.

7 Corrective Components with identified relevant conditions shall be entered into the Actions St. Lucie Corrective action Program (CAP). The disposition may include a supplementary examination to further characterize the relevant condition, an engineering evaluation to show that the component is capable of continued operation with a known relevant condition until the next planned inspection, or repair/replacement to remediate the relevant condition. Additional inspections of expansion category components may also be required. The disposition will insure that the design basis function of the RVI will continue to be fulfilled for all licensing basis loads and events.

8 Confirmation The PSL quality assurance procedures, review and approval processes, Process and Self and administrative controls are implemented in accordance with the Assessment requirements of 10 CFR Part 50, Appendix B. 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 St. Lucie RVI that are addressed in accordance the 10 CFR Part 50, Appendix B confirmation process and administrative controls.

9 Cdmnitrativ The St. Lucie RVl Inspection Program is implemented by 0-ADM-17.29.

10 Operating Experience FPL actively participates in joint industry programs addressing RVI issues including EPRI and PWROG. In accordance with 0-ADM-17.29, operating experience gained from these groups as well as INPO, WANO and international sites will be incorporated into the St. Lucie RVl Inspection Program in a timely manner.

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St. Lucie Units I and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 Table 1 CE Plants Primary Components ExpasionExamination Item Applicability Effect (ehns) ExasoLink Method/Frequency Examination Coverage (ehns) (Note I) (Note I)

Core Shroud Bolted plant Cracking Core support Baseline volumetric (UT) 100% of accessible bolts (see Note 3).

Assembly (Bolted) designs (IASCC), column bolts, examination between 25 and Heads are accessible from the core side.

Core shroud bolts NA for PSL Fatigue Barrel-shroud 35 EFPY, with subsequent UT accessibility may be affected by Aging bolts examination on a ten-year complexity of head and locking device Management (IE interval, designs.

and ISR)

(Note 2) See Figure 4-24, MRP-227-A.

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

sections examination on a ten-year Applicable for (Note 2) interval. See Figures 4-12 and 4-14, MRP-227-A.

PSL Core Shroud Plant designs Cracking Remaining Enhanced visual (EVT-1) Axial weld seams at the core shroud re-Assembly with core (IASCC) axial welds, examination no later than 2 entrant corners, at the core mid-plane (Welded) shrouds Aging ribs and rings refueling outages from the (+/-three feet in height) as visible from the Shroud plates assembled with Management beginning of the license core side of the shroud.

full-height (IE) renewal period and subsequent shroud plates examination on a ten-year See Figure 4-13, MRP-227-A.

NA for PSL (Note 2) interval.

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St. Lucie Units I and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 Table I CE Plants Primary Components Efet Expansion Examination Item Applicability (ehns) Link Method/Frequency Examination Coverage (ehns) (Note I) (Note 1)

Core Shroud Bolted plant Distortion None Visual (VT-3) examination no Core side surfaces as indicated.

Assembly designs later than 2 refueling outages NAfrPL (Void Swelling), from the beginning of the See Figures 4-25 and 4-26, MRP-227-A.

(Bolted) NAfrPL including: license renewal period.

Assembly Subsequent examinations on a

Abnormal ten-year interval.

interaction with fuel assemblies

Gaps along high fluence shroud plate joints

Vertical displaceme nt of shroud plates near high fluence joint Aging Management (IE) ______ ____________________________

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St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. I to Letter L-2015-229 Table I CE Plants Primary Components Efet Expansion Examination Item Applicability (ehns) Link Method/Frequency Examination Coverage (ehns) (Note I) (Note I)

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

PSL segments Aging See Figures 4-12 and 4-14, MRP-227-A.

Management (IE)

Core Support Barrel All plants Cracking (SCC) Lower core Enhanced visual (EVT-I--) 100% of the accessible surfaces of the Assembly Applicable for support beams. examination no later than 2 upper flange weld.

Upper (core support PSL Core support refueling outages from the barrel) flange weld barrel beginning of the license See Figure 4-15, MRP-227-A.

assembly renewal period. Subsequent upper cylinder examinations on a ten-year Upper core interval.

barrel flange Core Support Barrel All plants Cracking (SCC, Lower Cylinder Enhanced visual (EVT-1) 100% of the accessible surfaces of the Assembly Applicable for IASCC) Axial Welds examination no later than 2 lower cylinder welds. (Note 4)

Lower cylinder girth PSL Aging refueling outages from the welds Management beginning of the license See Figure 4-1 5, MRP-227-A.

(IE) renewal period. Subsequent examinations on a ten-year interval.

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St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. I to Letter L-2015-229 Table 1 CE Plants Primary Components Efet Expansion Examination Item Applicability Link Method/Frequency Examination Coverage (Mechnism) (Note I) (Note I)___________________

Lower Support All plants Cracking Lower Cylinder Visual (VT-3) examination no 100% of the accessible surfaces of the Structure Applicable for (SCC,9 IASCC) Axial Welds later than 2 refueling outages core support column welds. (Note 5)

Cesuprcoun PSL Aging from the beginning of the welds) Management license renewal period. See Figure 4-16 and 4-31, MRP-227-A.

(IE) Subsequent examinations on a ten-year interval.

Core Support Barrel All plants Cracking None Iffatigue life cannot be Examination coverage to be defined by Assembly No inspections (Fatigue) demonstrated by time-limited plant-specific fatigue analysis.

required for aging analysis (TLAA),

Lower flange weld PSL Units 1 enhanced visual (EVT-1) See Figure 4-15 and 4-16, MRP-227-A.

and 2 as TLAA examination, no later than 2 exists, refueling outages from the beginning of the license renewal period. Subsequent examination on a ten-year interval.

Core Support Barrel PSL Unit 1 Cracking None Enhanced visual (EVT-1) Repair region of core support barrel Assembly Only (IASCC, SCC, examination no later than 2 ExadbepusadFatigue) refueling outages from the Expanableplug andbeginning of the license patches renewal period. Subsequent examinations on a ten-year interval Page 11 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. I to Letter L-2015-229 Table I CE Plants Primary Components ExpasionExamination Item Applicability Effect Exaso Link Method/Frequency Examination Coverage (ehns) (Note I) (Note I)

Lower Support All plants with a Cracking None If fatigue life cannot be Examination coverage to be defined by Structure core support (Fatigue) demonstrated by time-limited evaluation to determine the potential Core support plate plate Aging aging analysis (TLAA), location and extent of fatigue cracking.

No inspections Management enhanced visual (EVT-1) required for (lE) examination, no later than 2 See Figure 4-16, MRP-227-A.

PSL Units 1 refueling outages from the and 2 as TLAA beginning of the license exists, renewal period. Subsequent examination on a ten-year interval.

Upper Internals All plants with Cracking None If fatigue life cannot be Examination coverage to be defined by Assembly core shrouds (Fatigue) demonstrated by time-limited plant-specific fatigue analysis.

Fuel alignment plate assembled with aging analysis (TLAA),

full-height enhanced visual (EVT-1) See Figure 4-17, MRP-227-A.

shroud plates examination, no later than 2 NA for PSL refueling outages from the beginning of the license renewal period. Subsequent examination on a ten-year interval.

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St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. I to Letter L-2015-229 Table 1 CE Plants Primary Components Efet Expansion Examination Item Applicability (ehns) Link Method/Frequency Examination Coverage (ehns) (Note I) (Note 1)

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

Instrument guide tubes the CEA shroud missing within the CEA license renewal period.

assembly supports or shroud Subsequent examination on a separation at the assemblies ten-year interval.

Applicable for welded joint See Figure 4-18, MRP-227-A.

PSL between the tubes and supports Plant-specific component integrity assessments may be required if degradation is detected and remedial action is needed.

Lower Support All plants with Cracking None Enhanced visual (EVT-1) Examine beam-to-beam welds, in the axial Structure core shrouds (Fatigue) that examination, no later than 2 elevation from the beam top surface to Deep beams assembled with results in a refueling outages from the four inches below.

full-height detectable beginning of the license shroud plates surface-breaking renewal period. Subsequent See Figure 4-19, MRP-227-A.

NA for PSL indication in the examination on a ten-year welds or beams interval, if adequacy of Aging remaining fatigue life cannot be Management demonstrated.

(IE)

NOTE:

1) Examination acceptance criteria and expansion criteria are in Table 4.

2) Void swelling effects on this component is managed through management of void swelling on the entire core shroud assembly.

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

5) A minimum of 75% of the total population of core support column welds Page 13 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 Table 2 CE Plants Expansion Components Item Applicability Effect Primary Link Examination Method Examination Coverage (Mechanism) (Note 1) (Note 1)

Core Shroud Assembly Bolted plant designs Cracking (IASCC, Core shroud Volumetric (UT) 100% (or as supported by plant-(Bolted) Fatigue) bolts examination, specific justification; Note 2) of NA for PSL barrel-shroud and guide lug insert Barrel-shroud bolts Aging Re-inspection every 10 bolts with neutron fluence exposures Management (IE years following initial > 3 displacements per atom (dpa).

and ISR) inspection.

See Figure 4-23, MRP-227-A.

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

Lower core barrel flange flange weld every 10 years following the inital ispettio.

.See Figure 4-15, MRP-227-A.

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

Applicable for PSL Aging flange weld every 10 years following Upper cylinder (including Management (IE) initial inspection. See Figure 4-15, MRP-227-A.

welds)

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

Applicable for PSL flange weld every 10 years following Upper core barrel flange initial inspection.

See Figure 4-15, MRP-227-A.

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

axial welds core barrel assembly girth Page 14 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 Table 2 CE Plants Expansion Components Item Applicability Effect Primary Link Examination Method Examination Coverage (Mechanism) (Note 1) (Note I) weld examinations. . See Figure 4-15, MRP-227-A.

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

shrouds assembled damaged or flange weld every 10 years following Lower support column with full-height fractured material, initial inspection.

beams shroud plates. SeFgr -6ad43,MP27 Aging A. iue41 nd43,MP27 Applicable for PSL Management (IE) A Core Shroud Assembly Bolted plant designs Cracking (IASCC, Core shroud Ultrasonic (UT) examination. 100% (or as supported by plant-(Bolted) Fatigue) bolts Re-inspection every 10 specific analysis) of core support NA for PSL years following initial column bolts with neutron fluence Core support column Aging inspection, exposures > 3 dpa. (Note 2) bolts Management (IE)

See Figures 4-16 and 4-33, MRP-227-A.

Core Shroud Assembly Plant designs with Cracking (IASCC) Core shroud Enhanced visual (EVT-1) Axial weld seams other than the core (Welded) core shrouds plate-former examination, shroud re-entrant corner welds at the Remaining axial welds assembled in two plate weld Re-inspection every 10 core mid-plane.

vertical sections years following initial Applicable for PSL inspection. See Figure 4-12, MRP-227-A.

Core Shroud Assembly Plant designs with Cracking (IASCC) Shroud plates Enhanced visual (EVT-1) Axial weld seams other than the core (Welded) core shrouds of welded core examination, with initial and shroud re-entrant corner welds at the assembled with full- shroud subsequent examination core mid-plane, plus ribs and rings.

Remaining axial welds, height shroud plates assemblies frequencies dependent on the results of the core Page 15 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. I to Letter L-2015-229 Table 2 CE Plants Expansion Components Item Applicability Effect Primary Link Examination Method Examination Coverage (Mechanism) (Note I) (Note 1)

-Ribs and rings NA for PSL shroud weld examinations.

See Figure 4-13, MRP-227-A.

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

Remaining instrument tubes in the CEA results in missing guide tubes examinations dependent on guide tubes shroud assembly supports or within the CEA the results of the instrument Applicable for PSL separation at the shroud guide tubes examinations. See Figure 4-18, MRP-227-A.

welded joint assemblies between the tubes and supports.'

NOTE:

1) Examination acceptance criteria and expansion criteria are in Table 4.

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).

Page 16 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 Table 3 CE Plants Existing Program Components Item Applicability Effect Reference Examination Method Examination Coverage (Mechanism)

Core Shroud Assembly All plants Loss of material ASME Code Visual (VT-3) examination, First 10-year ISI after 40 years of Guide lugs Applicable for (Wear) Section XI general condition examination for operation, and at each subsequent Guide lug inserts and bolts PSL Aging detection of excessive or inspection interval.

Management asymmetrical wear.

(ISR)

Lower Support Structure All plants with Cracking (SCC, ASME Code Visual (VT-3) examination to Accessible surfaces at specified Fuel alignment pins core shrouds IASCC, Fatigue) Section Xl detect severed fuel alignment frequency.

assembled with pins, missing locking tabs, or full-height shroud excessive wear on the fuel plates Aging alignment pin nose or flange.

NA for PSL Management (IE and ISR)

Lower Support Structure All plants with Loss of material ASME Code Visual (VT-3) examination. Accessible surfaces at specified Fuel alignment pins core shrouds (Wear) Section Xl frequency.

assembled in two Aging vertical sections Management (IE Applicable for and ISR)

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

PSL SPage 17 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. I to Letter L-2015-229 Table 4 CE Plants Examination Acceptance and Expansion Criteria Examination Additional Item Applicability Acceptance Criteria EpninExpansion Criteria Examination (Note 1) Lns)Acceptance Criteria Core Shroud Assembly Bolted plant Volumetric (UT) examination, a. Core support a. Confirmation that >5% of the core shroud a and b. The examination (Bolted) designs column bolts bolts in the four plates at the largest distance acceptance criteria for the Coe hrudblt N fr S The examination acceptance b. Barrel- from the core contain unacceptable indications UT of the core support criteria for the UT of the core shroud bolts shall require UT examination of the lower column bolts and barrel-shrud sallbesupport ols column bolts barrel within the next 3 shroud bolts shall be esalse spr fterefueling cyc~les, established as part of the estalised a pat oftheexamination technical examination technicaluticaon justification. b. Confirmation that >5% of the core support jsiiain column bolts contain unacceptable indications shall require UT examination of the barrel-shroud bolts within the next 3 refueling cycles.

Core Shroud Assembly Plant designs Visual (EVT-1) examination. Remaining Confirmation that a surface-breaking indication The specific relevant (Welded) with core axial welds > 2 inches in length has been detected and condition is a detectable shrouds sized in the core shroud plate-former plate crack-like surface indication.

Coreshrod plte-frmerThe specific relevant Coesrult-omr assembled in weld at the core shroud re-entrant corners (as plaeoel cndtintsicdtetalevisible from the core side of the shroud), within vetical two crack-like surface indication.

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

Page 18 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 Table 4 CE Plants Examination Acceptance and Expansion Criteria Examination Additional Expansion Item Applicability Acceptance Criteria Expansion Criteria Examination (Note 1) i sAcceptance Criteria core Shroud Assembly Plant designs Visual (EVT-1) examination, a. Remaining a. Confirmation that a surface-breaking The specific relevant (Welded) with core axial welds indication > 2 inches in length has been condition is a detectable Shroud plates shrouds The specific relevant b. Ribs and detected and sized in the axial weld seams at crack-like surface indication.

assembled condition is a detectable rings the core shroud re-entrant corners at the core with full-height crack-like surface indication, mid-plane shall require EVT-1 or UT shroud plates examination of all remaining axial welds by the NA for PSL completion of the next refueling outage.

b. If extensive cracking is detected in the remaining axial welds, an EVT-1 examination shall be required of all accessible rib and ring welds by the completion of the next refueling outage.

core Shroud Assembly Bolted plant Visual (VT-3) examination. None N/A N/A (Bolted) designs Assembly NA for PSL The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, and vertical displacement of shroud plates near high fluence joints.

Page 19 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 Table 4 CE Plants Examination Acceptance and Expansion Criteria Examination Additional Item Applicability Acceptance Criteria EpninExpansion Criteria Examination Link(s)

(Note 1) Acceptance Criteria Core Shroud Assembly Plant designs Visual (VT-i) examination. None N/A N/A (Welded) with core Assemly shouds The specific relevant assembed in condition is evidence of two vetical physical separation between sections the upper and lower core Applicable for shroud sections.

PSL Core Support Barrel Assembly All plants Visual (EVT-1) examination. Lower core Confirmation that a surface-breaking indication The specific relevant Applicable to support beams >2 inches in length has been detected and sized condition is a detectable Upper (core support barrel) PSL The specific relevant Upper core in the upper flange weld shall require that an crack-like surface indication.

flange weld condition is a detectable barrel cylinder EVIT-1 examination of the lower core support crack-like surface indication. (including beams, upper core barrel cylinder and upper welds) core barrel flange be performed by the Upper core completion of the next refueling outage.

barrel flange (cast)

Page 20 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-2015-229 Table 4 CE Plants Examination Acceptance and Expansion Criteria Examination Additional Expansion Item Applicability Acceptance Criteria Lns)Expansion Criteria Examination (Note 1) Acceptance Criteria Core Support Barrel All plants Visual (EVT-1) examination. Lower cylinder a. Confirmation that a surface-breaking The specific relevant Assembly Applicable to axial welds indication >2 inches in length has been condition for the expansion Lower cylinder girth welds PSL The specific relevant detected and sized in the lower cylinder girth lower cylinder axial welds is condition is a detectable weld shall require an EVT-1 examination of all a detectable crack-like crack-like surface indication, accessible lower cylinder axial welds by surface indication.

completion of the next refueling outage.

Lower Support Structure All plants Visual (VT-3) examination. None None Core support column welds Applicable to PSL The specific relevant condition is missing or separated welds.

Core Support Barrel Assembly All plants Visual (EVT-1) examination. None N/A N/A Lowe flnge eldApplicable to LwrfagwedPSL The specific relevant condition is a detectable crack-like indication.

core Support Barrel Assembly PSL Unit 1 Only Visual (EVT-1) examination. None N/A Expandable plugs and patches The specific relevant condition is a detectable crack-like surface indication.

Lower Support Structure All plants with Visual (EVT-1) examination. None N/A N/A Core support plate a core support plate The specific relevant Applicable to condition is a detectable PSL crack-like surface indication.

Page 21 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-2015-229 Table 4 CE Plants Examination Acceptance and Expansion Criteria Examination Additional Item Applicability Acceptance Criteria EpninExpansion Criteria Examination (Note 1) Lns)Acceptance Criteria Upper Internals Assembly All plants with Visual (EVT-1) examination. None N/A N/A Fuel alignment plate core shrouds assembled *The specific relevant with full-height condition is a detectable shroud plates crack-like surface indication.

NA for PSL Control Element Assembly All plants with Visual (VT-3) examination. Remaining Confirmed evidence of missing supports or The specific relevant Instrument Guide Tubes instruments instrument separation at the welded joint between the conditions are missing tubs nhete peifi rlevnttubes within tubes and supports shall require the visual (VT- supports and separation at tuesinth Te peifc elvatthe CEA shroud 3) examination to be expanded to the the welded joint between CEA shroud conditions are missing assemblies remaining instrument tubes within the CEA the tubes and the supports.

assembly supports and separation at shroud assemblies by completion of the next Applicable to the welded joint between the refueling outage.

PSL tubes and the supports.

Lower Support Structure All plants with Visual (EVT-1) examination. None N/A N/A Deep beams core shrouds assembled The specific relevant with full-height condition is a detectable shrod pltes crack-like indication.

NA for PSL NOTE

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

Page 22 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-2015-229 3 LICENSEE ACTION ITEMS This section provides the FPL response to the eight Licensee Action Items (LAI) noted in the NRC Safety Evaluation Report (SER) issued by the NRC for the report EPRI-MRP-227-A, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" dated December 2011. Additionally, the three bounding assumptions included in Section 2.4 of MRP-227-A are also addressed in the response to LAI #1.

LAI #1 Applicability of FMECA and Functionality Analysis Assumptions As addressedin Section 3.2.5.1 of this SE, each applicant/licensee is responsible for assessing its plant's design and operatinghistory and demonstrating that the approved version of MRP-227 is applicable to the facility. Each applicant/licensee shall refer, in particular,to the assumptions regardingplant design and operatinghistory made in the EMECA 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/licenseeshall 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.

The response to LAI #1 is based directly upon Westinghouse Letter LTR-RIAM-13-75, Rev. 0, Final Summary Report for St. Lucie Units I and 2 for PWROG PA-MSC-0983 Cafeteria Task Deliyerables.

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

The process used to provide reasonable assurance that St. Lucie Units 1 and 2 are reasonably represented by the generic industry program assumptions (with regard to neutron fluence, temperature, stress values, and materials used in the development of MRP-227-A) is:

1. Identification of typical Combustion Engineering (CE)-designed pressurized water reactor (PWR) reactor vessel internals (RVI) components (Table 4-5 of MRP-191).

2. Identification of St. Lucie Units 1 and 2 PWR components.

3. Comparison of the typical CE-designed PWR RVI components to the St. Lucie Units 1 and 2 RVI components:

a. Confirmation that no additional items were identified by this comparison (primarily supports Applicant/Licensee Action Item 2).

b. Confirmation that the materials from Table 4-5 of MRP-191 are consistent with St. Lucie Units 1 and 2 RVI component materials.

Page 23 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229

c. Confirmation that the design and fabrication of St. Lucie Units 1 and 2 RVI components are the same as, or equivalent to, the typical CE-designed PWR RVI components.

4. Confirmation that the St. Lucie Units 1 and 2 operating history is consistent with the assumptions in MRP-227-A regarding core loading patterns and base load operation.

5. Confirmation that the St. Lucie Units I and 2 RVI materials operated at temperatures within the original design basis parameters.

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

7. Confirmation that any changes to the St. Lucie Units 1 and 2 RVJ components do not impact the application of the MRP-227-A generic aging management strategy.

The St. Lucie Units 1 and 2 RVI components are reasonably represented by the design and operating history assumptions regarding neutron fluence, temperature, materials, and stress values in the MRP-191 generic FMECA and in the MRP-232 functionality analysis based on the following:

1. St. Lucie Units 1 and 2 operating history is consistent with the assumptions in MRP-227-A with regard to neutron fluence and fuel management.

a. FMECA and functionality analysis for MRP-227-A made the following assumption of 30 years of operation with high-leakage core loading patterns followed by 30 years of low-leakage core fuel management strategy. The St.

Lucie Units 1 and 2 fuel management program changed from a high to a low leakage core loading pattern prior to 30 years of operation. Therefore, St. Lucie Units 1 and 2 meet the fluence and fuel management assumptions in MRP-191 and requirements for MRP-227-A application.

b. St. Lucie Units 1 and 2 have operated under base load conditions over the life of the plant Therefore, St. Lucie Units 1 and 2 satisfy the assumptions in Materials Reliability Program (MRP) documents regarding operational parameters affecting fluence.

2. The St. Lucie Units 1 and 2 reactor coolant system operates between Tco*d and Thor.

Tco~d is not less than 532°F and there were no changes to TCO~d due to extended power uprate (EPU). Thot was no higher than 594°F prior to EPU and no higher than 608.2°F after EPU for Unit 1. Thot was no higher than 598°F prior to EPU and no higher than 607.9°F after EPU for Unit 2. The design temperature for the vessel is 650°F. Therefore, St. Lucie Units 1 and 2 operating history is within original design basis parameters and is Page 24 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-201 5-229 consistent with the assumptions used to develop the MRP- 227-A aging management strategy with regard to temperature operational parameters.

3. With the exceptions discussed below, the St. Lucie Units 1 and 2 RVI components and materials are comparable to the typical CE-designed PWR RVI components (MRP-191, Table 4-5).

a. There are two additional components for St. Lucie Unit 1 and one component for Unit 2 that are not included in MRP-1 91. In Unit 1, core support barrel patches and core support barrel expandable plugs were installed following the discovery of damage to the core barrel caused by fatigue of the thermal shield attachment points. CE developed and analyzed the repair method. For Unit 2, there are four specialized control element assembly (CEA) shroud assemblies that are fitted with flow bypass inserts. Other than the core support barrel patches, core support barrel expandable plugs, and flow bypass inserts, the components required for inclusion in the St. Lucie Units 1 and 2 program are consistent with those contained in MRP-1 91.

b. St. Lucie Units 1 and 2 RVI component materials are consistent with, or nearly equivalent to, those materials identified in Table 4-5 of MRP-1 91 for CE-designed plants. Where differences exist, there is no impact on the St. Lucie Units 1 and 2 RVI program or the component is already credited as being managed under an alternate St. Lucie Units I and 2 aging management program.

c. Design and fabrication of St. Lucie Units 1 and 2 RVI components are the same as, or equivalent to, the typical CE-designed PWR RVI components.

4. An 11.85% EPU was performed on St. Lucie Units 1 and 2. Evaluations performed by Westinghouse determined that the associated changes in temperature, fluence and loading on the RVI components did not affect the bounding assumptions or applicability of MRP-227-A. With the exception of the thermal shield removal for Unit 1, the modifications to the St. Lucie Units I and 2 RVI made over the lifetime of the plants are those identified in general industry practice or specifically directed by the original equipment manufacturer (OEM). The Unit 1 thermal shield removal was analyzed to be acceptable. Repairs to the core barrel, as a result of the thermal shield removal, were in accordance with recommendations and guidance of the OEM. Therefore, the design has been maintained over the lifetime of the plant as specified by the OEM and operational parameters with regard to fluence and temperature are compliant with MRP-227-A requirements. With the exception of two components for Unit 1 and one for Unit 2, the components are consistent with those considered in MRP-1 91. The materials for those components are also consistent with MRP-1 91, or where differences exist, there is no impact. The additional three components have no impact on the assumptions summarized above; therefore, the St. Lucie Units 1 and 2 RVI are represented by the Page 25 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1Ito Letter L-2015-229 assumptions in MRP-1 91, MRP-227-A, and MRP-232, confirming the applicability of the generic FMECA.

Conclusion St. Lucie Units 1 and 2 comply with LAI #1 of the Nuclear Regulatory Commission Safety Evaluation on MRP-227, Revision 0. Therefore, the requirement is met for application of MRP-227-A as a strategy for managing age-related material degradation in the RVI components.

LAI #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 addressedin 10 CFR 54.4, each applicant/licenseeis 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-189, Revision 1, and Tables 4-4 and 4-5Sin MRP-191 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 applicantor 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-specificAMP. 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.

The response to LAI #2 is based directly upon Westinghouse Letter LTR-RIAM-13-75, Rev. 0, Final Summary Report for St. Lucie Units 1 and 2 for PWROG PA-MSC-0983 Cafeteria Task Deliverables.

FPL Response:

This Applicant/Licensee Action Item requires comparison of the St. Lucie Units 1 and 2 RVI components that are within the scope of license renewal for St. Lucie Units I and 2 to those components contained in Table 4-5 of MRP-191. MRP-189, Tables 4-1 and 4-2 are not applicable to St. Lucie Units 1 and 2 since those tables are applicable to a B&W-plant design, while St. Lucie Units 1 and 2 are CE-plant design. There are two additional components for St.

Lucie Unit 1 and one component for Unit 2 identified in the plant-specific aging management review (AMR) that are not included in MRP-1 91. In Unit 1, core support barrel patches and core support barrel expandable plugs were installed following the discovery of damage to the core barrel caused by fatigue at the thermal shield attachment points, and for Unit 2, there are four specialized CEA shroud assemblies that are fitted with flow bypass inserts. Other than the core support barrel patches, core support barrel expandable plugs, and flow bypass inserts, all components in the St. Lucie Units I and 2 license renewal program are consistent with those contained in MRP-191.

Page 26 of 32

St. Lucie Units I and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-201 5-229 The in-core instrumentation (ICI) guide tubes for both units have a different material than that specified in MRP-1 91, but the difference has no effect on the recommended MRP aging strategy or is already managed by an alternate St. Lucie Units 1 and 2 program; therefore, no modifications to the program details in MRP-227-A need to be proposed. This supports the requirement that the NRC-AMP shall provide assurance that the effects of aging on the St.

Lucie Units 1 and 2 RVI components within the scope of license renewal, but not included in the generic CE-designed PWR RVI components from Table 4-5 of MRP-1 91, 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 the RVI specified in MRP-227- A are applicable to St. Lucie Units 1 and 2 with no modifications for the St. Lucie components that are consistent with those contained in MRP-1 91. For the three components that are not included in MRP-191, the aging management strategy has been determined on a plant-specific basis. FPL has conservatively categorized the Unit 1 core support barrel patches and core support barrel expandable plugs as Primary components for aging management during the period of extended operation. Plant-specific augmented inspections are required On a periodic basis to manage the associated aging effects on Primary components. St. Lucie Unit 2 has four specialized CEA shroud assemblies that are fitted with flow bypass inserts. MRP-191 categorized all components of the CEA shroud assemblies as Category A. Therefore, FPL categorized the Unit 2 flow bypass inserts consistently, making them No Additional Measures components. No Additional Measures components are either not susceptible to any degradation mechanism, or if susceptible the impact of failure on the functionality of the RVI components is insignificant. No further action is required for managing aging of these RVI components.

Conclusion St. Lucie Units 1 and 2 comply with LAI #2 of the Nuclear Regulatory Commission Safety Evaluation on MRP-227, Revision 0. The assessment performed identified three additional components that are not identified~in MRP-191. The aging management strategy for these additional components has been included in the plant-specific program to ensure aging is managed for components that are not included within the scope of MRP-227-A. Therefore, St.

Lucie Units 1 and 2 meet the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components.

LAI #3 Evaluation of the Adequacy of Plant-Specific Existing Programs As addressedin Section 3.2.5.3 in this SE, applicants/licenseesof CE and Westinghouse are required to perform plant-specific analysis either to justify the acceptabilityof an applicant's/licensee'sexisting 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 programsbeing relied on to manage aging of these components shall be submitted as part of the applicant's/licensee'sAMP application. The CE and Westinghouse components identified for Page 27 of 32

St. Lucie Units I and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-2015-229 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-22 7), and Westinghouse guide tube support pins (split pins) (Section 4.3.3 in MRP-22 7). This is Applicant/Licensee Action Item 3.

FPL Response:

There are no thermal shields or thermal shield positioning pins installed on the core barrels of St. Lucie Units 1 and 2.

The St. Lucie Units 1 and 2 in-core instrumentation flux thimble tubes are considered out-of-scope for license renewal based upon the component screening performed in accordance with the Nuclear License Renewal Rule (10 CFR 54). All in-core instrumentation flux thimble tubes for St. Lucie Unit 1 were replaced during the Cycle 21 outage (Spring 2007) (WO 35010464),

and those for St. Lucie Unit 2 were replaced during the Cycle 19 outage (Spring 2011) (WO 35010467). The replacement thimbles have been designed with sufficient margin to accommodate growth of thimbles' zircalloy sections during the PEG.

Conclusion LAI #3 is not applicable to St. Lucie Units 1 and 2.

LAI #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/licenseesshall confirm that the core support structure upper flange weld was stress relieved during the originalfabrication of the Reactor Pressure Vessel in orderto 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 "Primaiy"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 recommendationsin 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 describedin Sections 3.3.1 and 4.3.1 of this SE. The applicant's/licensee'sresolution 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: LAi #4 pertains to B&W Core Support Structure Upper Flange Stress Relief issue and is not applicable to St. Lucie Units 1 and 2 which are CE NSSS designs.

LA! #5 Application of Physical Measurements as part of l&E Guidelines for B&W. CE, and Westinghouse RVI Components As addressedin Section 3.3.5 in this SE, applicants/licenseesshall identify plant-specific acceptance criteria to be applied when performing the physical measurements requiredby the Page 28 of 32

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-201 5-229 NRC-approved version of MRP-227 for loss of compressibilityfor Westinghouse hold down springs, and for distortion in the gap between the top and bottom core shroud segments in CE units with core barrelshrouds assembled in two vertical sections. The applicant/licenseeshall include its proposedacceptance 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:

The response to LAI #5 is based directly upon Westinghouse Letter LTR-RIAM-13-147, Rev. 0, Transmittal of Final Summary Letter for Acceptance Criteria for Visual Examination of Gaps between Upper and Lower Core Shroud Subassemblies at Calvert Cliffs Units 1 and 2 and St.

Lucie Units 1 and 2.

FPL participated in a PWROG Project Authorization (PA) to justify a gap size for the St. Lucie Units I and 2 core shrouds. Basic assumptions of the PA were that the gap be measurable using the specified VTI-I inspection resolution and that it satisfy functionality requirements. The Units I and 2 core shrouds differ slightly in design - Unit 1 uses a mechanical attachment (via tie rods) between the upper and lower core shroud sections, whereas Unit 2 uses a welded attachment. The postulated gap would include both thermal and void swelling contributions.

The thermal contribution would be present only during power operation. The void swelling contribution would be present under all conditions including plant shutdown, during which the physical examination of the core shroud will be performed.

Core shroud gap acceptance criteria have been developed for St. Lucie Units 1 and 2 that are resolvable using the specified VT-I-- inspection method of MRP-227-A. 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.j[F1]

LAI #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 barrelcylinders (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-formerbolts and their locking devices, and B&W core barrel assembly internalbaffle-to-baffle bolts. The MRP also identified that although the B&W core barrel assembly internal baffle-to-baffle bolts are accessible, the bolts are non-inspectableusing currently available examination techniques. Applicants/licenseesshalljustify the acceptabilityof these components for continued operation through the period of extended operationby 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/licenseesshall provide theirjustification for the continued operabilityof each of the inaccessible components Page 29 of 32

St. Lucie Units I and 2 Reactor Vessel Internals Aging Management Plan Attachment No. I to Letter L-2015-229 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: LAI #6 pertains to B&W Inaccessible Components and is not applicable to St.

Lucie Units I and 2 which are CE NSSS designs.

LAI #7 Plant-Specific Evaluation of CASS Materials As discussed in Section 3.3.7 of this SE, the applicants/licenseesof B&W, CE, and Westinghouse reactors are requiredto develop plant-specific analyses to be appliedfor 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 additionalRVI components that may be fabricated from CASS, martensiticstainless steel or precipitationhardened stainless steel materials. These analyses shall also consider the possible loss of fracture toughness in these components due to thermal and irradiationembrittlement, and may also need to consider limitations on accessibilityfor inspection and the resolution/sensitivityof the inspection techniques. The requirement may not apply to components that were previously evaluated as not requiringaging management during development of MRP-227. That is, the requirement would apply to components fabricated from susceptible materialsfor which an individual licensee has determined aging management is required, for example during their review performed in accordancewith 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/licenseeshall 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.

The response to LAI #7 is based directly upon Westinghouse Letter LTR-RIAM-13-75, Rev. 0, Final Summary Report for St. Lucie Units I and 2 for PWROG PA-MSC-0983 Cafeteria Task Deliverables.

FPL Response:

Applicant/Licensee Action Item 7 from the NRC's final Safety Evaluation on MRP-227, Revision o states that, for assessment of cast austenitic stainless steel (CASS) materials, the licensees or applicant for license renewal may apply the criteria in the NRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Stainless Steel Components" (NRC ADAMS Accession No.ML003717179) as the basis for determining whether the CASS materials are susceptible to the thermal aging mechanism. If the application of the screening criteria for the component material demonstrates that the components are not susceptible to either thermal embrittlement (TE) or irradiation embrittlement (IE), or the synergistic effects of TE and IE combined, then no other evaluation would be necessary.

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St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-2015-229 The St. Lucie Units 1 and 2 RVI CASS components and the assessment of their susceptibility to TE are summarized in Table I and as follows:

The St. Lucie Unit 1 core support columns are low molybdenum and static cast. A certified material test report (CMTR) was located for one two-legged column. Its calculated ferrite content is less than 20%; thus, it is not susceptible to TE. The remaining St. Lucie Unit I core support columns are potentially susceptible to TE. The support columns were previously screened in for the age-related degradation mechanism of TE, along with stress corrosion cracking (SCC) of the weld, irradiation-assisted stress corrosion cracking (IASCC), fatigue, and IE in MRP-191, Table 4-7 and the inspection and evaluation guidelines for this Primary component are in MRP-227-A.

The St. Lucie Unit 2 core support columns are 304 5S; thus, A/LAI 7 is not applicable to the St. Lucie Unit 2 core support columns

The St. Lucie Units 1 and 2 control element assembly (CEA) shroud tubes are low molybdenum and centrifugal cast; thus they are not susceptible to TE. The CEA shroud tubes were also previously screened in for the age-related degradation mechanism of SCC of the weld in MRP-1 91, Table 4-7.

The St. Lucie Unit 2 flow bypass inserts are low molybdenum, static cast, and have ferrite content -<20%; thus they are not susceptible to TE. The flow bypass inserts were not identified in MRP-1 91. FPL has categorized the St. Lucie Unit 2 flow bypass inserts as No Additional Measures Components.

CASS Component Molybdenum Content Casting Method Calculated ]Susceptibility to TE (Wt.%) Ferrite Content St. Lucie Unit 1 Core Support Low, 0.5 max Static -<20%*' One column not Columns susceptible to TE Potentially Remaining columns

>20%(2) potentially susceptible to TE( 2 )

CEA Shroud Tubes Low, 0.5 max Centrifugal All Not susceptible to TE St. Lucie Unit 2 CEA Shroud Tubes Low, 0.5 max Centrifugal All Not susceptible to TE Flow Bypass Inserts Low, 0.5 max Static -<20%*'* Not susceptible to TE Notes:

1. Calculated ferrite content is based on CMTR data, input into Hull's formula per the guidance of NUREG/CR-451 3, Rev. 1.

Where molybdenum is not listed on the CMTR, a value of 0.5 percent is used. Where nitrogen is not listed on the CMTR a value of 0.04 percent is used.

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St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Attachment No. 1 to Letter L-201 5-229

2. Where component-specific CMTR data are not available, the ferrite content is calculated based on permitted variations in ASTM A351, Grade CF8 chemical requirement. Allowable variants of Grade CF8 chemical requirements may result in ferrite content greater than 20%; thus, the ferrite content is identified as potentially exceeding 20%.

The St. Lucie Units 1 and 2 mnartensitic stain less steel RVI components include only a 403 SS Hold-down Ring in each unit. There are no martensitic PH-SS RVI components in St. Lucie Units 1 and 2.

Conclusion The results of this evaluation do not conflict with strategy for aging management of RVI provided in MRP-227-A. It is concluded that continued application of the strategies in MRP-227-A and the St. Lucie Units 1 and 2 RVI Inspection Program will meet the requirements for managing age-related degradation of the St. Lucie Units 1 and 2 CASS and martensitic SS RVI components.

LAI #8 Submittal of Information for Staff Review and Approval As addressedin Section 3.5.1 in this SE, applicants/licenseesshall make a submittal for NRC review and approvalto credit theirimplementation 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:

During the license renewal process, St. Lucie Units 1 and 2 prepared and gained approval for RVI Inspection Program from the NRC, as documented in NUREG 1759. Subsequently, during the EPU LAR review, St. Lucie Units 1 and 2 committed to revise the RVI Inspection Program to align with MRP-227-A.

The St. Lucie RVI Inspection Program is summarized in Sections 1 and 2. It provides the following items: 1) components to be inspected; 2) the degradation mechanisms of concern; 3) the inspection methods; 4) the examination coverage; and 5) the examination acceptance criteria. And the responses to the eight Licensee Action Items of MRP-227-A are provided in Section 3. These sections satisfy the requirements of LAI #8.

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Attachment 2 to Letter L-2015-229 St. Lucie Units 1 and 2 Proposed UFSAR Revisions

St. Lucie Nuclear Plant Attachment No. 2 to FPL Letter L-2015-229 Proposed Revision to Updated Safety Analysis Report (UFSAR)

Unit 1 UFSAR Proposed Revision 18.1.4 REACTOR VESSEL INTERNALS INSPECTION PROGRAM The Reactor Vessel Internals (RVI) Inspection Program manages the aging effects on the RVI during the period of extended operation. The RVI consists of three major structuralassemblies, plus three other sets of major components. The three major assemblies include: 1) upper internals assembly; 2) core support barrel assembly, and 3) lower internals assembly. In addition, the three other sets of major components are the control1 element assembly (CEA) shroud assemblies, core shroud assembly, and in-core instrumentation support system. The RV.I Inspection Program is applicable to passive RVI structural components and specifically excludes welded attachments to the reactor vessel and consumable items such as fuel assemblies, control element assemblies (CEAs) and in core instrumentation (ICI).

Aging effects and the causative degradation mechanisms addressed by the RVI Inspection Program include: 1) cracking due to stress corrosion cracking (SCC), irradiationassisted stress corrosion cracking (IA SCC) or fatigue; 2) reduction in fracture toughness due to irradiationor 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 relaxationor creep.

The RVI Inspection Programis based upon the guidance provided in EPRI MRP-22 7-A, "EPRI Materials Reliability Program, PressurizedWater Reactor Internals Inspection and Evaluation Guidelines." The RVI Inspection Program is a living program that will be revised as necessaryin response to ongoingjoint industry efforts aimed at further understandingthe aging effects of the RV Internals.

FPL has satisfied the following commitments concerning the RVI Inspection Program: 1) Submit an integratedreport for St. Lucie Units 1 and 2 to the NRC priorto the end of the initial operating license term for St. Lucie Unit I that summarizes its understandingof the aging effects applicable to the reactorvessel internals and contains a descriptionof the St. Lucie inspection plan, including methods for detection and sizing of cracks and acceptance criteria, and 2) Adopt MPR-22 7-A in place of the previously approved RVI Inspection Programthat was included in the St. Lucie Units I and 2 License Renewal Applications.

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Unit 2 UFSAR Proposed Revision 18.1.3 REACTOR VESSEL INTERNALS INSPECTION PROGRAM The Reactor Vessel Internals (RVI) Inspection Program manages the aging effects on the RVI during the period of extended operation. The RVI consists of three major structuralassemblies, plus three other sets of major components. The three major assemblies include: 1) upper internals assembly; 2) core support barrel assembly, and 3) lower internals assembly. In addition, the three other sets of major components are the control element assembly (CEA) shroud assemblies, core shroud assembly, and in-core instrumentation support system. The RVI Inspection Program is applicable to passive RVI structural components and specifically excludes welded attachments to the reactor vessel and consumable items such as fuel assemblies, con trol element assemblies (CEAs) and in core instrumentation (ICI).

Aging effects and the causative degradation mechanisms addressed by the RVI Inspection Programinclude: 1) cracking due to stress corrosion cracking (SCC), irradiationassisted stress corrosion cracking (IA SCC) 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.

The RVI Inspection Program is based upon the guidance provided in EPRI MRP-22 7-A, "EPRI Materials Reliability Program,Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The RVI Inspection Programis a living program that will be revised as necessary in response to on goingjoint industry efforts aimed at further understandingthe aging effects of the RV Intemnals.

FPL has satisfied the following commitments concerning the RVI Inspection Program: 1) Submit an integratedreport for St. Lucie Units 1 and 2 to the NRC prior to the end of the initial operating license term for St. Lucie Unit I that summarizes its understandingof the aging effects applicable to the reactorvessel internals and contains a description of the St. Lucie inspection plan, including methods for detection and sizing of cracks and acceptance criteria;and 2) Adopt MPR-22 7-A in place of the previously approved RVI Inspection Programthat was included in the i* St. Lucie Units 1 and 2 License Renewal Applications.

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