L-2017-015, License Renewal Commitments, Reactor Vessel Internals Aging Management Plan Response to Request for Additional Information

License Renewal Commitments, Reactor Vessel Internals Aging Management Plan Response to Request for Additional Information

ML17075A194
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Site:	Saint Lucie
Issue date:	03/07/2017
From:	Deboer D Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
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ML17075A197	List: L-2017-015, License Renewal Commitments, Reactor Vessel Internals Aging Management Plan Response to Request for Additional Information ML17075A195
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L-2017-015
Download: ML17075A194 (37)

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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 March 7, 2017 Reactor Vessel Internals Aging Management Plan Response to Request for Additional Information

References:

L-2017-015 10 CFR 54

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.

Electric Power Research Institute (EPRI) Materials Reliability Program Report 1022863 (MRP-227-A), "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," ADAMS Accession Nos.ML12017A194,ML12017A196,ML12017A197,ML12017A191,ML12017A192, ML12017A195, and ML12017A199.

5.

FPL Letter from Joseph Jensen to U.S. Nuclear Regulatory Commission (L-2014-192) "St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389, Reactor Vessel Internals Inspection Program Plans and Inspection Dates," June 25, 2014.

6.

FPL Letter from Christopher Costanzo to U.S. Nuclear Regulatory Commission (L-2015-229) "St.

Lucie Units 1and2 Docket Nos. 50-335 and 50-389, License Renewal Commitments - Reactor Vessel Internals Aging Management Plan," Dated September 28, 2015.

7.

NRC e-Mail from Perry Buckberg to Ken Frehafer, Request for Additional Information, St. Lucie Plant Units 1 and 2, Reactor Vessel Internals Aging Management Plan, Docket Nos. 50-335 and 50-389, TAC Nos. MF6777 and MF6778. ADAMS Accession No. ML16013A215.

8.

FPL Letter from Christopher Costanzo to U.S. Nuclear Regulatory Commission (L-2016-040) "St.

Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389, License Renewal Commitments - Reactor Vessel Internals Aging Management Plan," Dated February 26, 2016.

Florida Power & Light Company 6501 S. Ocean Drive, Jensen Beach, FL 34957

L-2017-015 10 CFR54 By letter to the U.S. Nuclear Regulatory Commission (NRC) dated September 28, 2015 (Reference 6), Florida Power & Light Company (FPL) submitted its License Renewal Reactor Internals Program :MRP-227-A (Reference 4) at St. Lucie Nuclear Plants Units 1 and 2 for NRC staff review.

The NRC staff reviewed the information provided by FPL in its submittal and requested additional information to complete their review (Reference 7). The responses to RAI-1 through RAI-4 and RAI-7 through RAI-10 were previously submitted to the NRC on February 26, 2016 (Reference 8). Based on the additional work required to address RAI--5 and RAI-6, those responses were deferred to a later time.

FPL responses to the RAI-5 and RAI-6 are provided in Attachment No. 1.

The revised Reactor Vessel Internals Aging Management Plan with changes incorporated from the responses to RAI-1 through RAI-10 is contained in Attachment No. 2.

In RAI-3 (Reference 7), the NRC inquired whether St. Lucie Plant Units 1 and 2 have non-weld or bolting austenitic stainless steel components with 20 percent cold work or greater, and if so, do the affected components have operating stresses greater than 30 kilo pounds-per-square inch? RAI-3 also requested that FPL provide a plant-specific aging management recommendation for stress corrosion cracking (SCC) for those components meeting the specified conditions at St. Lucie.

In FPL's initial response to RAI-3 (Reference 8), it was stated that FPL would be crediting the efforts of the PWROG, which is addressing this issue on a generic basis. FPL is a participant in this PWROG Program.

Report PWROG-15105-NP, which provides the technical basis for stating that no non-fastener cold-work greater than 20 percent is present in the RV internals of domestic B&W, CE and Westinghouse designed PWRs (this includes St. Lucie Units 1 and 2), was submitted to the NRC in June 2016. The results and conclusions of PWROG-15105-NP were discussed during a meeting with the NRC on September 7, 2016.

PWROG-15105-NP, which addresses FPL's response to RAI-3, is provided in Attachment No. 3.

The response to RAI-9 described ongoing work for the CASS lower support column welds functionality analysis (Reference 8). This work is ongoing, and the projected completion date remains unchanged (summer 2017).

Should you have any questions, please contact Mr. Michael Snyder, Licensing Manager, at 772-467-7036.

Very truly yours, d~u~

Daniel DeBoer Site Director St. Lucie Plant Attachments:

1) St. Lucie Units 1and2 Responses to RAI-5 and RAI-6

2) St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan

3) PWROG-15105-NP, "PA-MSC-1288 PWR RV Internals Cold-Work Assessment" cc:

NRC Region II Administrator NRC Project Manager, St. Lucie Nuclear Plant NRC Senior Resident Inspector, St. Lucie Nuclear Plant Page2of2 St. Lucie Units 1 and 2 Responses to RAI-5 and RAI-6

-L-2017-015 10 CFR 54

St. Lucie Units 1 and-2 Responses to NRC RAl-5 and RAI-6 dated to 1/13/2016 FPL Letter L-2017-015 Page 1 of6 FPL Responses to NRC Request for Additional Information, RAI-5 and RAI-6:

RAI-MF6777/MF6778-EVJB-05 In the licensee's response to AILA! 1, the licensee stated that an 11.85% EPU was performed on St. Lucie Plant, and that evaluations performed by Westinghouse determined that the associated changes in temperature, fluence, and loading on the RV! components did not affect the bounding assumptions or applicability of MRP-227-A. For St.

Lucie Plant Unit 1, the response to RAJ CVIB-5 related to the EPU (Ref 8) stated that a detailedfluence analysis of the reactor pressure vessel (from the interior of the core shroud plates through the vessel wall around the mid-plane) was used to determine fluence through the various RV! components, and that the fluence calculation adhered to the requirements of Regulatory Guide 1.190 with regard to method and uncertainty.

For St. Lucie Plant Unit 2, the EPU Licensing Report (Ref 9) also implies that a detailed neutronjluence analysis was performed similar to that for St. Lucie Plant Unit 1. The staff therefore requests that the licensee describe how the fluence analysis of the St. Lucie Plant Unit 2 RV! was pe1formed in support of the EPU, or confirm the methodology used was the same as for St. Lucie Plant Unit 1. This is RAI-5.

FPL Response:

The fluence analysis methodology performed in support of the St. Lucie Plant Unit 2 EPU was the same as that previously described for St. Lucie Unit 1.

RAl-MAF6777/MF6778-EVJB-06:

In the staff's safety evaluation related to the EPU for St. Lucie Plant Unit I.(Ref 2), the stqff concluded that it has reviewed the licensee's evaluation of the effects of the proposed EPU on the susceptibility of RV! to known degradation mechanisms and concludes that the licensee has identified appropriate degradation management programs to address the effects of changes in operating temperature and neutron fluence on the integrity of these components. The staff reached a similar conclusion in its safety evaluation related to the EPU for St. Lucie Plant Unit 2. However, the staff notes that in its evaluation of RV! aging considering EPU, the licensee determined that some components are susceptible to certain aging mechanisms, which were screened out in the development process of MRP-227-A. For example, the EPU Licensing Reports for St. Lucie Plant Unit 1 (Ref JO) and St. Lucie Plant Unit 2 (Ref 9) list the fuel alignment plate, upper guide structure support plate, control element assembly (CEA) shroud tubes, and CEA shroud bolts and locking bars as susceptible to loss of fracture toughness due to irradiation embrittlement (IE), while MRP-191 screened out these components for IE. The EPU licensing reports also identified the CEA flow channel parts as susceptible to IE, which are a plant-specific component. There is no equivalent generic component in MRP-191. Similarly, the EPU Licensing Report for St.

Lucie Plant Unit 1, and St. Lucie Plant Unit 2, list the fuel alignment plate, upper guide structure support plate, CEA shrouds (lower part), and CEA shroud bolts and locking bars as components susceptible to irradiation assisted stress corrosion cracking (JASCC), while MRP-191 screened out these components for IASCC. The staff therefore requests the licensee:

This is RAI-6.

1.

Provide the fluence screening criteria it used for IE and IASCC, if different than the screening criteria of MRP-191.

2.

Confirm whether the components listed above actually exceed the MRP-191 fluence screening criteria.

3. If any of the components listed above exceed the MRP-191 fluence screening criteria, provide the estimated fluencefor those components considering EPU at the end of life.

4. If the components do exceed the screening criteria, explain how MRP-227-A is bounding (provides for appropriate aging management) for St. Lucie Plant Units 1 and 2, considering that the fluences for these components exceed the MRP-191 screening limits.

5.

Finally, if MRP-227-A is not bounding for any specific components, provide a plant-specific aging management recommendation for such components.

St. Lucie Units 1 and 2 Responses to NRC RAI-5 and RAI-6 dated to 1/13/2016 FPL Letter L-2017-015 Page2 of6 FPL Response:

1. The irradiation embrittlement (IE) and irradiation assisted stress corrosion cracking (IASCC) fluence screening criteria used in the St. Lucie Units 1 and 2 EPU Licensing Reports for wrought austenitic stainless steels (SS) are shown in Table 1 below. The fluence screening criteria in MRP-191, Rev. 0 and MRP-227-A is based on the fluence screening criteria in MRP-175; it is shown for comparison below.

Table-1, Fluence Screening Limits Degradation Fluence (E>l.OMeV)

Mechanism (DM)

EPU Licensin!! Reports MRP-175 IE

_:::1E+20 n/cm 2

_:::1E+21 n/cm 2

IASCC

_:::1E+21 n/cm 2

>2E+21 n/cm 2 for stress >89 ksi

>6.7E+21 n/cm2 for stres~ >62 ksi

l.3E+22 n/cm2 for stress ;46 ksi

~2.7E+22 n/cm2 for stress ~30 ksi It is noted that the EPU Licensing Reports used more conservative fluence criteria for IE and IASCC than MRP-175. Also, the EPU Licensing Reports did not consider minimum stress thresholds for IASCC at various fluence levels.

Subsequent to the submittal of the St. Lucie Units 1 and 2 EPU Licensing Reports (Refs.10 and 9), FPL committed to adopt MRP-227-A in place of its previously approved RVI Inspection Program for St. Lucie Units 1 (Ref. 14) and Unit 2 (Ref. 15). Consistent with MRP-227-A and its supporting documents, FPL will utilize the fluence thresholds ofMRP-175 for the development of the St. Lucie Units 1 & 2 RVI Aging Management Programs in place of those cited in the previously submitted EPU Licensing Reports.

2. Subsequent to the receipt of the NRC's Request for Additional Information for the St. Lucie Units 1 & 2 RVI Aging Management Plan (AMP), FPL contracted with Westinghouse to perform a more detailed analysis of the fluence for the St. Lucie Units 1 and 2 RVI components. The estimated end oflife (EOL) (60 years of operation) fluences for the components specified in RAI-MAF6777 /MF6778-EVIB-06 are shown below. The estimated fluence ranges for these components in MRP-191 Rev. 0 and MRP-191 Rev. 1 are included for comparison. The latter revision of MRP-191 was issued in October 2016 to reflect the industry's current state of knowledge for the RVI components. It should be noted that MRP-191 Rev. 1 did not dictate any changes to the aging management methodology ofMRP-227-A. The affected fluence-related degradation mechanism (DM) screening criteria of MRP-175 are also shown for comparison.

i

St. Lucie Units 1 and 2 Responses to NRC RAl-5 and RAl-6 dated to 1113/2016 FPL Letter L-2017-015 Page 3 of6 Table-2, Cumulative Fluence at 60 Years of Operation Group Component Fluence (n/cm2) (E>l.OMev)

St. Lucie Units 1 MRP-191, MRP-191, MRP-175

&2 Rev.O Rev.1 DM Estimated Estimated Screening Range Range Criteria Exceeded Upper Fuel Alignment 2:1.0E+21 to 1E+20 to 1E+21 to 2:1E+21 (IE)

Internals Plate

<2.0E+21 7E+20 1E+22 Assembly UGS Support Plate

<1.3E+20

<1E+20

<1E+20 None CEA CEA Shroud Tubes' 2:1.0E+21 to 1E+20 to 1E+20 to 2:1E+21 (IE)

Shroud

<2.0E+21 7E+20 7E+20 Assembly CEA Shroud 2:1.0E+21 to 1E+20 to 1E+20 to 2:1E+21 (IE)

Bolts/Lock Bars

<2.0E+21 7E+20 7E+20 CEA Flow Channel 2:l.OE+21 to 1E+20 to 1E+20 to 2:1E+21 (IE)

Parts2

<2.0E+21 7E+203 7E+203 Notes:

1. CEA Shroud Tubes are equivalent to CEA Shrouds in MRP-191. Highest fluences are at CEA shroud (shroud tube) bases.

2. CEA Flow Channel Parts are a subcomponent of the CEA Shrouds that extend into the Fuel Alignment Plate (FAP).

3. Not listed specifically in MRP-191 Rev. 0 or Rev. 1. Fluence assumed to be same as shroud bases.

The table above shows that for all components listed in RAI-MAF6777/.MF6778-EVIB-06, the cumulative 60-year fluences exceed the MRP-191 Rev. 0 estimated fluence ranges. Additionally, four components exceed the MRP-175 fluence screening criteria for IE. FPL Response Item 5 ofRAI-MAF6777/.MF6778-EVIB-06 provides the plant-specific aging management recommendation for these components that exceed the MRP-175 fluenct;:

screening criteria and are not bounded by MRP-227-A.

3. Table 2 in FPL Response Item 2 ofRAI-MAF6777/.MF6778-EVIB-06 provides the estimated fluences considering EPU at the EOL for the all the components listed in RAl-MAF6777/MF6778-EVIB-06.

4. The St. Lucie Units 1 & 2 Fuel Alignment Plates (FAPs) will exceed the estimated fluence ranges in MRP-191, Rev. 0, but will not exceed the estimated fluence ranges in MRP-191, Rev. I. The estimated fluences for the St.

Lucie Units 1 & 2 Fuel Alignment Plates (F APs) introduce IE as an age-related degradation mechanism not considered in MRP-191 Rev. 0. However, the elevated fluence estimates for the FAP in MRP-191Rev.1 also exceed the screening value for IE and bound the St. Lucie Units 1 & 2 estimated fluences for these components.

Since MRP-191 Rev. 1 did not dictate any changes to the aging management methodology ofMRP-227-A, the St. Lucie Units 1 & 2 F APs are considered to be bounded by MRP-227-A.

The St. Lucie Units l & 2 UGS Support Plates are assumed to be bounded by MRP-227-A, as they fall below the MRP-17 5 fluence screening criteria for all age-related degradation mechanisms involving fluence.

The St. Lucie Units 1 & 2 CEA Shroud Assembly subcomponents (Bases, Bolts/Lock Bars, and Flow Channel Parts) exceed the estimated fluence ranges ofMRP-191, Rev. 0 and 1, and introduce IE as an additional age-related degradation mechanism. The aging management methodology for the St. Lucie Units 1 & 2 CEA Shroud Assembly subcomponents is discussed in FPL Response Item 5 ofRAI-MAF6777/.MF6778-EVIB-06.

5.

As noted above, the fluence estimates for the St. Lucie Units 1 & 2 CEA Shroud Bases, CEA Shroud Bolts/Lock Bars, and CEA Flow Channel Parts introduce IE as an additional age-related degradation mechanism, not currently included in MRP-191, Revs. 0 and I. Age-related degradation mechanisms recognized in MRP-191, Revs. 0 and 1, for 304 SS CEA Shroud Bases include SCC of the welded regions. Age-related degradation

St. Lucie Units 1 and 2 Responses to NRC RAl-"S and RAI-6 dated to 1/13/2016 FPL Letter L-2017-015 Page 4 of6 mechanisms recognized in MRP-191, Revs. 0 and 1, for the CEA Shroud Bolts/Lock Bars include wear, fatigue and irradiation stress relaxation/creep (ISRIIC). The 304 SS CEA Flow Channel Parts are not listed specifically in MRP-191 Rev. 0 or Rev. 1, and the fluence and degradation mechanisms are assumed to be same as the 304 SS CEA Shroud Bases. The CEA Shroud Bases and CEA Shroud Bolts/Lock Bars were classified as Category A in MRP-191, Revs. 0 and 1, and placed in the No Additional Measures Category ofMRP-227-A. FPL currently inspects the accessible CEA Shroud Bases, Bolts/Lock Bars, and Flow Channel Parts under its ASME Section XI Program and proposes to add them to the Existing Programs Components Table of the St. Lucie Units 1 & 2 RV!

AMP. IE will be considered in any future flaw evaluations.

The recent fluence analyses performed by Westinghouse, revealed several additional components (listed below) that exceed the fluence estimates ofMRP-191, Revs. 0 and 1, at EOL (60 years).

Table-3, Additional Components Cumulative Fluence at 60 Years of Ooeration Group Component Fluence (n/cm2) (E>l.OMev)

St. Lucie MRP-191, Rev.

MRP-191, Rev. 1 MRP-175 Units 1 & 2 0

DM Screening Criteria Exceeded Core Shroud Guide Lug 2:2E+21 to 1E+20 to 7E+20 1E+20 to 7E+20 2:1E+21 (IE)

Assembly

<1.3E+22 2:2E+21 (IASCC)

Guide Lug Bolt1 2:1E+21 to No Value No Value Note 2

<2E+21 2 Provided2 Provided2 Guide Lug Insert 2:1E+21 to 1E+20 to 7E+20 1E+20 to 7E+20 2:1E+21 (IE)

<2E+21 Guide Lug Insert Bolt 2:1E+21 to 1E+20 to 7E+20 1E+20 to 7E+20 2:1E+21 (IE)

<2E+21 Tie Rods and Nuts, 2:1.3E+22 1E+21to1E+22 1E+21to1E+22 2:1.3E+22 (VS) 2:2E+21 (!ASCC)

CEA Shroud CEA Instrument Guide 2:1E+21 to 1E+20 to 7E+20 1E+20 to 7E+20 2:1E+21 (IE)

Assembly Tube

<2E+21 Core Support Upper Cylinder 2:2E+21 to 1E+20 to 7E+20 1E+20 to 7E+20 2:2E+21 (IASCC)

Barrel*

<1.3E+22 2:1E+21 (IE)

Upper Flange

<1.3E+20

<1E+20

<1E+20 None Lower Core Suooort Plate

>1.3E+22 1E+21to1E+22 1E+21to1E+22 None Support Fuel Alignment Pins 2:1.3E+22 1E+21to1E+22 1E+21to1E+22 2:2E+21 (IASCC)

Structure Note:

1.

Unit 1 only, Unit 2 Guide Lugs are welded to Core Shroud.

2.

Grouped with Guide Lug Insert Bolts under LAI 2.

3.

Unit 1 only as Unit 2 Core Shroud sections are welded.

The elevated fluence levels of the St. Lucie Units 1 & 2 Guide Lugs, Guide Lug Bolts, Guide Lug Inserts, and Guide Lug Insert Bolts relative to MRP-191, Revs. 0 and 1, exceed the fluence criteria for IE. (It should be noted that the Guide Lug Bolts were not called out specifically in MRP-191 but were grouped with the Guide Lug Insert Bolts during FPL's completion of Licensee Action Item (LAI) 2.) Age-related degradation mechanisms recognized in MRP-191 Revs. 0 and 1, include SCC for the welded regions of the Guide Lugs; wear, fatigue and ISR/IC for the Guide Lug Insert Bolts; and wear for the Guide Lug Inserts. The Guide Lugs and Guide Lug Inserts were ranked as Category A in MRP-191 Revs. 0 and 1, and placed in the Existing Program Components Category ofMRP-227-A. The Guide Lug Insert Bolts were ranked as Category B in MRP-191 Revs. 0 and 1, and placed in Existing Program Components Category ofMRP-227-A. FPL proposes to add IE as an Effect (Mechanism) for the Guide Lugs, Guide Lug Inserts and Bolts in the Existing Programs Components Table of the St. Lucie Units 1 and 2 RV! AMP. IE will be considered in any fature flaw evaluations.

The elevated fluence levels of the St. Lucie Units 1 & 2 Guide Lugs relative t~ MRP-191, Revs. 0 and 1, also exceed the fluence criteria for IASCC. However, the stress level for the Guide Lugs is estimated to be well below the corresponding 62 ksi load threshold in all but possibly the weld regions. FPL proposes to add IASCC as an

St. Lucie Units 1 and 2 R~sponses to NRC RAl-5 and RAI-6 dated to 1113/2016 FPL Letter L-2017-015 Page 5 of6 Effect (Mechanism) for the Guide Lugs in the Existing Program Components Table of the St. Lucie Units 1 & 2 RVIAMP.

The elevated fluence levels of the St. Lucie Unit 1 Core Shroud Tie Rods and Nuts relative to MRP-191, Revs. 0 and 1, exceed the fluence criteria for void swelling in the regions of these components and have temperatures

608°F. These elevated fluence levels also introduce IASCC as a potential age-related degradation mechanism should the applied stress levels be sufficient. Age-related degradation mechanisms recognized in MRP-191, Revs. 0 and 1, include wear, fatigue, IE and ISR/IC. The Core Shroud Tie Rods and Nuts were ranked as Category B under MRP-191, Revs. 0 and 1, but later placed in the No Additional Measures Category ofMRP-227-A based upon subsequent evaluations. FPL currently inspects the visible portions of the Core Shroud Tie Rods and Nuts under its ASME Section XI Program and proposes to add them to the Existing Programs Components Table in the St. Lucie Units 1 & 2 RV! AMP.

The elevated fluence levels of the St. Lucie Units 1 & 2 CEA Instrument Guide Tubes relative to MRP-191, Revs. 0 and 1, exceed the fluence criteria for IE. Age-related degradation mechanisms recognized in MRP-191, Revs. 0 and 1, include SCC and fatigue. The CEA Instrument Guide Tubes are included in the Primary Components Table ofMRP-227-A and the St. Lucie Units 1 & 2 RVI AMP. FPL proposes to add IE as an Effect (Mechanism) for the CEA Instrument Guide Tubes in the Primary Components Table of the St. Lucie U~its 1 & 2 RV! AMP and will consider IE should any future flaw evaluations be pe1formed The elevated fluence levels of the St. Lucie Units 1 & 2 Core Support Barrel (CSB) Upper Cylinder relative to MRP-191, Revs. 0 and 1, exceed the fluence criteria for IE and IASCC. However, the estimated fluence levels of the CSB upper flange is essentially equivalent the estimated fluence levels in MRP-191, Revs. 0 and 1, and do not introduce any additional age-related degradation mechanisms. Therefore, the additional degradation mechanisms (IE and IASCC) are assumed to be applicable to the lower regions of the CSB Upper Cylinder, not inclusive of the CSB upper flange weld. Age-related degradation mechanisms recognized in MRP-191, Revs. 0 and 1, and MRP-227-A include SCC of the CSB Upper Cylinder Welds. The CSB Upper Flange Weld is included in the Primary Components Table ofMRP-227-A and the St. Lucie Units 1 and 2 RVI AMP.

Additionally, the Upper Cylinder is included in the Expansion Components Table ofMRP-227-A and the St.

Lucie Units 1 and 2 RVI AMP. FPL proposes to add IASCC as an Effect (Mechanism) for the CSB Upper Cylinder in the Expansion Components Table of the St. Lucie Units I & 2 AMP; IE is already included No change to the CSB Upper Flange Weld in the Primary Components Table is required since the additional age-related degradation mechanisms (IE and IASCC) are not applicable.

Age-related degradation mechanisms of the Core Support Plate recognized in MRP-191, Revs. 0 and 1, include SCC, IASCC, wear, fatigue, and IE. The elevated fluence levels of the St. Lucie Units 1 & 2 Core Support Plates relative to MRP-191, Revs. 0 and 1, also exceeds the fluence criteria for Void Swelling. However, the estimated temperature of the Core Support Plate is T-cold (532°F) which is below the 608°F minimum temperature threshold for void swelling, noted in MRP-191, Revs. 0 and 1. Therefore, the St. Lucie Units 1 & 2 Core Support Plates are bounded by the aging management methodology ofMRP-227-A.

Age-related degradation mechanisms of the 304 SS Fuel Alignment Pins recognized in MRP-191, Revs. 0 and 1, include IE. IASCC is not recognized as a potential degradation mechanism in these documents (for plants with core shrouds assembled in two vertical sections), presumably due to low operating stress at the estimated fluence level of 1E+21 n/cm2 to 1E+22 n/cm2* The 304 SS Fuel Alignment Pins were ranked as Category A under MRP-191, Revs. 0 and 1, and placed in the Existing Programs Components Category ofMRP-227-A. FPL recognizes that the elevated fluence of the St. Lucie Units 1 & 2 Fuel Alignment Pins may render them more susceptible to IASCC. To address this increased risk for IASCC, FPL proposes to add IASCC as an Effect (Mechanism) for the Fuel Alignment Pins in the Existing Programs Component Table in the St. Lucie Units 1 and 2 RV! AMP.

The table below summarizes the plant-specific aging management for the components not bounded by MRP-227-A.

St. Lucie Un!ts 1 and 2 Responses to NRC RAI-5 and RAl-6 dated to 1/13/2016 FPL Letter L-2017-015 Page 6 of6 Table-4, Summary Table Group Component Fluence (n/cm2) (E>l.OMev)

Aging Management Approach Chan!!es St. Lucie Exceeded Screening Units 1 & 2 Fluence Threshold Guide Lug 2:2E+21 to 2: l.OE+21 (IE)

Add IE and IASCC as DM in

<1.3E+22 2: 2.0E+2l(IASCC)

Existing Program Components Table Guide Lug 2:1E+21 to 2: 1.0E+21 (IE)

Add IE as DM in Existing Program Core Insert

<2E+21 Components Table Shroud Guide Lug 2:1E+21 to 2: l.OE+21 (IE)

Add IE as DM in Existing Program Assembly Insert Bolt

<2E+21 Components Table 2: l.3E+22 (VS)

Add Tie Rods and Nuts in Existing Tie Rods and

?:l.3E+22 2: 2.0E+21 (IASCC)

Program Components Table with Nuts2 wear, fatigue, IE, ISR4, VS and IASCC as the DMs CEA Shroud 2:1.0E+21 to 2: 1.0E+21 (IE)

Add CEA Shroud Bases in Existing Tubes1

<2.0E+21 Program Components Table with SCC and IE as the DMs CEA Shroud

~l.OE+21 to

~ l.OE+21 (IE)

Add CEA Bolts/Lock Bars in Bolts/Lock Bars

<2.0E+21 Existing Program Components Table CEA with wear, fatigue, ISR4 and IE as the Shroud DMs Assembly CEA Flow 2:1.0E+21 to

~ l.OE+21 (IE)

Add CEA Flow Channel Parts in Channel Parts3

<2.0E+21 Existing Program Components Table with SCC and IE as the DMs CEA

~l.OE+21 to

~ l.OE+21 (IE)

Add IE as DM in Primary Program Instrument

<2.0E+21 Components and Expansion Tube Components Table Core Upper Cylinder

~2E+21 to

~ l.OE+21 (IE)

Support

<l.3E+22 2: 2.0E+21 (IASCC)

Add IASCC as DM in Expansion Barrel Components Table Lower Fuel Alignment

?:l.3E+22

~ 2.0E+21 (IASCC)

Add IASCC as DM in Existing Support Pins Program Components Table Structure Notes:

1. CEA Shroud Tubes are equivalent to CEA Shrouds in MRP-191. Highest fluences are at CEA shroud (shroud tube) bases.

2. Unit 1 only as Unit 2 Core Shroud sections are welded.

3. CEA Flow Channel Parts are a subcomponent of the CEA Shrouds that extend into the FAP.

4. ISR/IC designated as ISR in MRP-227-A Primary, Expansion and Existing Component Tables.

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan L-2017-015 10 CFR54

St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Page 1

of 27 1

SUMMARY

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

1.1 Degradation Mechanisms A totai 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 trans granular or intergranular in nature. The aggressive contaminants most commonly associated with sec of austenitic stainless steels are dissolved chlorides and oxygen. Nickel base alloys such as Alloy 600 and X-750 have exhibited susceptibility to intergranular sec 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 CTASCC)

IASCC is a form of intergranular SCC that results from the combined influence of neutron irradiation and an aggressive environment. A limited number ofIASCC failures ofRVI components, specifically fasteners, constructed of austenitic stainless steels and nickel base alloys have been observed. The aging effect ofIASCC 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 (2'.:yield strength) applied for low number of cycles. The aging effect 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.

Page 1 of27

St. Lucie Units 1and2 to FPL Letter L-2017-015 Reactor Vessel Internals Aging Management Plan Page 2 of 27 Void Swelling CVS)

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 ofRVI 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 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.

Page 2 of27

St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Page 3 of 27 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-1 and EVT-1) exams Requirements for the inspection methodologies and qualification ofNDE systems used to perform those inspections are provided in EPRI MRP-228, Inspection Standard for PWR Internals.

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 Coverage 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.

I 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 XI inspection requirements for the RVI components.

2 INSPECTION PROGRAM ATTRIBUTES The attributes of the St. Lucie RVI Inspection Program and compliance with NUREG-1801 (GALL Report),

Section XI.Ml 6, "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.

Page 3 of27

St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Page 4 of 27 Plan Approach and supplemental information Attribute 1

Scope of Program The St. Lucie RVI Inspection Program includes all Units 1 and 2 RVI 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)(l), 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 levels of Measures 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 RVI Inspection Program manages the following age-related Monitored 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 ofpreload 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 (EVT-1) exams, or directly using volumetric (UT) or surface (ECT) exams. For the management ofloss of material, the RVI Inspection Program monitors for surface conditions that may be indicative of wear using visual (VT-3) exams.

For the management ofchanges in dimension and Joss 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 ofMRP-227-A. Additionally, the program implements the parameters monitored/inspected criteria for CE designed Expansion Components in Table 4-5 ofMRP-227-A. The parameters monitored/inspected for Existing Program Components follow the bases for the ASME Section XI Program.

4 Detection of Discussion and justification of the inspection methods selected for detection of Aging Effects 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 Page 4 of27

St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Page 5 of 27 Plan Approach and supplemental information Attribute 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 ofNRC Branch Position RLSB-1.

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 ofMRP-227-A and for CE designed Expansion Components in Table 4-5 ofMRP-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 comers. 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 that Trending result from the St. Lucie RVI Inspection Program inspections are given in Section 6 ofMRP-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 ifthe 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 ofMRP-227-A provides the examination acceptance criteria for the Criteria Primary and Expansion Components in the St. Lucie RVI Inspection Program.

For Existing Program components referenced to ASME Section XI, the IWB-3500 acceptance criteria apply.

Page 5 of27

St. Lucie Units 1 and 2 to FPL Letter L-2017-015 Reactor Vessel Internals Aging Management Plan Page 6 of 27 Plan Approach and supplemental information Attribute 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 Actions Components with identified relevant conditions shall be entered into the 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, and Process and Self administrative controls are implemented in accordance with the Assessment recommendations ofNEI 03-08 and the requirements of 10 CFR Part 50, Appendix B. The implementation of the guidance in MRP-227-A, in conjunction with NEI 03-08, and other guidance documents, reports or methodologies referenced in this AMP, provide an acceptable level of quality and basis for confirming the quality of inspections, flaw evaluations and corrective actions.

9 Administrative The administrative controls for the St. Lucie RVI Inspection Program, including Controls its implementing procedure and review and approval processes, are implemented in accordance with the recommended industry guidelines and criteria in NEI 03-08, and are under the site 10 CFR 50 Appendix B, Quality Assurance Program.

10 Operating The review and assessment of relevant operating experience for impact on the Experience St. Lucie RVI Inspection Program are governed by NEI 03-08 and Appendix A ofMRP-227-A. The reporting of inspection results and operating experience is treated as a "Needed" category item under NEI 03-08.

Page 6 of27

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Table 1 CE Plants Primary Components Effect Expansion Examination Item Applicability (Mechanism)

Link Method/Frequency (Note 1)

(Note n Core Shroud Assembly Bolted plant Cracking Core support Baseline volumetric (UT)

(Bolted) designs (IASCC), Fatigue column bolts, examination between 25 and 35 Core shroud bolts NAforPSL Agiug Barrel-shroud EFPY, with subsequent Management (IE bolts examination on a ten-year interval.

and!SR)

CNote 2)

Core Shroud Assembly Plant designs Cracking Remaining axial Enhanced visual (EVT-1)

(Welded) with core shrouds (IASCC) welds examination no later than 2 assembled in two Agiug refueling outages from the Core shroud plate-former vertical sections Management (IE) begiuning of the license renewal plate weld Applica hie for period and subsequent examination PSL (Note 2) on a ten-year interval.

Core Shroud Assembly Plant designs Cracking Remaining axial Enhanced visual (EVT-1)

(Welded) with core shrouds (IASCC) welds, ribs and examination no later than 2 assembled with Agiug rings refueling outages from the Shroud plates full-height Management (IE) beginning of the license renewal shroud plates period and subsequent examination NAforPSL (Note 2) on a ten-year interval.

Page 7 of27 to FPL Letter L-2017-015 Pa e 7 of 27 Examination Coverage 100% of accessible bolts (see Note 3).

are accessible from the core side. UT Heads accessibility may be affected by complexity of head and locking device designs.

See Figure 4-24, MRP-227-A Axial and horizontal weld seams at the core shroud re-entrant comers as visible from the core side of the shroud, within six inches of central flange and horizontal stiffeners.

See Figures 4-12 and 4-14, MRP-227-A Axial weld seams at the core shroud re-entrant comers, at the core mid-plane (+/-three feet in height) as visible from the core side of the shroud.

See Figure 4-13, MRP-227-A

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Table 1 CE Plants Primary Components Effect Expansion Examination Item Applicability (Mechanism)

Link Method/Frequency (Note 1)

<Note 1)

Core Shroud Assembly Bolted plant Distortion.

None Visual (VT-3) examination no (Bolted) designs (Void Swelling),

later than 2 refueling outages from Assembly NA forPSL including:

the beginning of the license

. Abnormal renewal period. Subsequent interaction examinations on a ten-year with fuel interval.

assemblies

. Gaps along higb fluence shroud plate joints

. Vertical displacemen t of shroud plates near higb fluence joint Aging Management (IE)

Core Shroud Assembly Plant designs Distortion None Visual (VT-I) examination no later (Welded) with core shrouds (Void Swelling),

than 2 refueling outages from the Assembly assembled in two as evidenced by beginning of the license renewal vertical sections separation period. Subsequent examinations Applicable for between the on a ten-year interval.

PSL upper and lower core shroud segments Aging Management (IE)

Page 8 of27 to FPL Letter L-2017-015 Pa e 8 of 27 Examination Coverage Core side surfaces as indicated.

See Figures 4-25 and 4-26, MRP-227-A If a gap exists, make three to five measurements of gap opening from the core side at the core shroud re-entrant corners.

Then, evaluate the swelling on a plant-specific basis to determine frequency and method for additional examinations.

See Figures 4-12 and 4-14, MRP-227-A

Item Core Support Barrel Assembly Upper (core support barrel) flange weld Core Support Barrel Assembly Lower cylinder girth welds Lower Support Structure Core support column welds Core Support Barrel Assembly Lower flange weld Core Support Barrel Assembly Expandable plugs and patches St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Table 1 CE Plants Primary Components Effect Expansion Examination Applicability (Mechanism)

Link Method/Freqnency

!Note 1)

!Note 1)

All plants Cracking (SCC)

Lower core Enhanced visual (EVT-1)

Applicable for support beams.

examination no later than 2 PSL Core support refueling outages from the barrel assembly beginning of the license renewal upper cylinder period. Subsequent examinations Upper core barrel on a ten-year interval.

flange All plants Cracking (SCC, Lower Cylinder Enhanced visual (EVT-1)

Applicable for IASCC)

Axial Welds examination no later than 2 PSL Aging refueling outages from the Management (IE) beginning of the license renewal period. Subsequent examinations on a ten-year interval.

All plants Cracking (SCC, None Visual (VT-3) examination no later Applicable for IASCC) than 2 refueling outages from the PSL Aging beginning of the license renewal Management (IE) period. Subsequent examinations on a ten-year interval.

All plants Cracking None If fatigue life cannot be No inspections (Fatigue) demonstrated by time-limited required for aging analysis (TLAA), enhanced PSL Units 1 and visual (EVT-1) examination, no 2asTLAA later than 2 refueling outages from exists.

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

PSL Unit 1 Only Cracking None Enhanced visual (EVT-1)

(IASCC, sec, examination no later than 2 Fatigue) refueling outages from the beginning of the license renewal period. Subsequent examinations on a ten-year interval Page 9 of27 to FPL Letter L-2017-015 Pa!!e 9 of 27 Examination Coverage 100% of the accessible surfaces of the upper flange weld.

See ~igure 4-15, MRP-227-A.

100% of the accessible surfaces of the lower cylinder welds. (Note 4)

See Figure 4-15, MRP-227-A 100% of the accessible surfaces of the core support column welds. (Note 5)

See Figure 4-16 and4-31, MRP-227-A Examination coverage to be defined by plant-specific fatigue analysis.

See Figure 4-15 and 4-16, MRP-227-A Repair region of core support barrel

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Table 1 CE Plants Primary Components Effect Expansion Examination Item Applicability (Mechanism)

Link Method/Frequency (Note 1)

(Note 1)

Lower Support Structure All plants with a Cracking None If futigue life cannot be Core support plate core support plate (Fatigue) demonstrated by time-limited No inspections Aging aging analysis (1LAA), enhanced required for Management (IE) visual (EVT-1) examination, no PSL Units 1 and later than 2 refueling outages from 2asTLAA the 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 Assembly core shrouds (Fatigue) demonstrated by time-limited Fuel alignment plate assembled with aging analysis (1LAA), enhanced full-height visual (EVT-1) examination, no shroud plates later than 2 refueling outages from NAforPSL the beginning of the license renewal period. Subsequent examination on a ten-year interval.

Control Element All plants with Cracking (SCC, Remaining Visual (VT-3) examination, no Assembly instrument guide Fatigue) that instrument guide later than 2 refueling outages from Instrument guide tubes tubes in the CEA results in missing tubes within the the beginning of the license shroud assembly supports or CEA shroud renewal period. Subsequent Applicable for separation at the assemblies examination on a ten-year interval.

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

Mana!!ement !IE)

Page 10 of27 to FPL Letter L-2017-015 Pa e 10 of 27 Examination Coverage Examination coverage to be defined by evaluation to determine the potential location and extent of fatigue cracking.

See Figure 4-16, MRP-227-A Examination coverage to be defined by plant-specific fatigue analysis.

See Figure 4-17, MRP-227-A I 00% of tubes in peripheral CEA shroud assemblies (i.e., those adjacent to the perimeter of the fuel alignment plate).

See Figure 4-18, MRP-227-A

St. Lucie Units 1 and 2 Reactor Vessel Internals Aging Management Plan Table 1 CE Plants Primary Components Effect Expansion Examination Item Applicability (Mechanism)

Link Method/Frequency

<Note 1)

(Note 1)

Lower Support Structure All plants with Cracking None Enhanced visual (EVT-1)

Deep beams core shrouds (Fatigue) that examination, no later than 2 assembled with results in a refueling outages from the full-height detectable beginning of the license renewal shroud plates surface-breaking period. Subsequent examination on NA forPSL indication in the a ten-year interval, if adequacy of welds or beams remaining fatigue life cannot be Aging demonstrated.

Management (IE)

NOTE:

I) Examination acceptance criteria an~ 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. to FPL Letter L-2017-015 Pa e 11 of 27 Examination Coverage Examine beam-to-beam welds, in the axial elevation from the beam top surface to four inches below.

See Figure 4-19, MRP-227-A.

3) A minimum of75% of the total population (examined+ unexamined), including coverage consistent with the Expansion criteria in Table4, must be examined for inspection credit.

4) A minimum of75% 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 of75% of the total population of core support column welds Page 11 of27

Item Core Shroud Assembly (Bolted)

Barrel-shroud bolts Core Support Barrel Assembly Lower core barrel flange Core Support Barrel Assembly Upper cylinder (including welds)

Core Support Barrel Assembly Upper core barrel flange Core Support Barrel Assembly Core barre! assembly axial welds Lower Support Structure Lower support columo bearos St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pa e 12 of 27 Table 2 CE Plants Expansion Components Applicability Effect (Mechanism)

Primary Link Examination Method Examination Coverage (Note 1)

(Note 1)

Bolted plant designs Cracking (IASCC, Core shroud Volumetric (Ul) examination.

100% (or as supported by plant-NAforPSL Fatigue) bolts Re-inspection every 10 years specific justification; Note 2) of Aging Management following initial inspection.

barrel-shroud and guide lug insert (IE and JSR) bolts with neutron fluence exposures

> 3 displacements per atom (dpa).

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

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

flange weld every 10 years following the initial inspection.

See Figure 4-15, MRP-227-A All plants Cracking (SCC, Upper (core Enhanced visual (EVT-1) 100% of accessible surfaces of the Applicable for PSL IASCC) support barrel) examination. Re-inspection welds and base metal (Note 2).

Aging Management flange weld every I 0 years following initial (IE) inspection.

See Figure 4-15, MRP-227-A All plants Cracking (SCC)

Upper(core Enhanced visual (EVT-1) 100% of accessible bottom surfuce of Applicable for PSL support barrel) examination. Re-inspection the flange (Note 2).

flange weld every 10 years following initial insoection.

See Fi<mre 4-15, MRP-227-A.

All plants Cracking (SCC)

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

barrel assembly girth weld examinations.

See Fi= 4-15, MRP-227-A All plants except those Cracking (SCC, Upper(core Enhanced visual (EVT-1) 100% of accessible surfuces (Note 2).

with core shrouds fatigue) including support barrel) examination. Re-inspection assembled with damaged or flange weld every 10 years following initial See Figure 4-16 and 4-31, MRP-227-full-height shroud fractured material.

inspection.

A.

plates.

Aging Management Applicable for PSL (IE)

Page 12 of27

Item Core Shroud Assembly (Bolted)

Core support column bolts Core Shroud Assembly (Welded)

Remaining axial welds Core Shroud Assembly (Welded)

Remaining axial welds, Ribs and rings Control Element Assembly Remaining instrument guide tubes NOTE:

St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pa e 13 of 27 Table 2 CE Plants Expansion Components Applicability Effect (Mechanism)

Primary Link Examination Method Examination Coverage (Note 1)

(Note 1)

Bolted plant designs Cracking (JASCC, Core shroud Ultrasonic (UT) examination.

100% (or as supported by plant-NAforPSL Fatigue) bolts Re-inspection every I 0 years specific analysis) of core support Aging Management following initial inspection.

column bolts with neutron fluence (IE) exposures > 3 dpa. (Note 2)

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

Plant designs with core Cracking (IASCC)

Core shroud Enhanced visual (EVT-1)

Axial weld seams other than the core shrouds assembled in plate-former examination.

shroud re-entrant comer welds at the two vertical sections plate weld core mid-plane.

Re-inspection every 10 years Applicable for PSL following initial inspection.

See Figure 4-12, MRP-227-A Plant designs with core Cracking (IASCC)

Shroud plates of Enhanced visual (EVT-1)

Axial weld seams other than the core shrouds assembled with welded core examination, with initial and shroud re-entrant comer welds at the full-heigbt shroud plates shroud subsequent examination core mid-plane, plus ribs and rings.

NAforPSL assemblies frequencies dependent on the results of the core shroud weld See Figure 4-13, MRP-227-A.

examinations.

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

in the CEA shroud in missing supports tubes within the examinations dependent on the assembly or separation at the CEA shroud results of the instrument guide welded joint assemblies tubes examinations.

See Figure 4-18, MRP-227-A Applicable for PSL between the tubes and supports.

Aging Management (IE)

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 of75% of the total population oflike components of the examination is required (including both the accessible and inaccessible portions).

Page 13 of27

Item Core Shroud Assembly Guide lugs Guide lug inserts and bolts r

Core Sbrond Assembly Tie Rods and Nuts Lower Support Structure Fuel aligmnent pins Lower Support Structure Fuel alignment pins St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pa e 14 of 27 Table 3 CE Plants Existing Program Components Applicability Effect Primary Link Examination Method Examination Coverage (Mechanism)

All plants Loss of material ASMECode Visual (VT-3) examination, general First I 0-year !SI after 40 years of (Wear)

Section XI condition examination for detection operation, and at each subsequent Applicable for of excessive or asymmetrical wear inspection interval.

PSL Cracking (IASCC) and Joss of integrity in weld regions Aging of guide lugs.

Management (IE, ISR4)

Applicable for Loss of material ASMECode Visual (VT-3) examination to detect Accessible surfaces at specified PSL (Wear)

Section XI missing tie rods and nuts.

frequency.

Distortion (VS)

Cracking (IASCC, Fatigue)

Aging Management (IE,

!SR)

All plants with core Cracking (SCC, ASMECode Visual (VT-3) examination to detect Accessible surfaces at specified shrouds assembled IASCC, Fatigue) severed fuel aligmnent pins, missing frequency.

with full-height Section XI Jocking tabs, or excessive wear on the shroud plates Aging fuel aligmnent pin nose or flange.

Management (IE NA forPSL and ISR)

All plants with core Loss of material ASMECode Visual (VT-3) examination to detect Accessible surfaces at specified shrouds assembled (Wear)

Section XI excessive wear and severed/missing frequency.

in two vertical fuel alignment pins.

sections Cracking (IASCC)

Applicable for Aging PSL Management (IE and !SR)

Page 14 of27 I

St. Lucie Units 1 and 2 to I

Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pa2e 15

• of 27 Table 3 CE Plants Existing Program Components Core Barrel Assembly All plants Loss of material ASMECode Visual (VT-3) examination.

Area of the upper flange potentially (Wear)

Section XI susceptible to wear.

Upper flange Applicable for PSL Core Element Assembly Applicable for PSL Loss of material ASMECode Visual (VT-3) examination to Accessible surfaces at specified (Wear')

Section XI determine general condition including frequency.

CEA Shroud Bases loss of integrity of CEA Shroud CEA Shroud Bolts/Lock Bars Cracking (SCC2, Bases' welded connections, and fatigue3) missing CEA Shroud Bolt/Lock Bars.

CEA Flow Channel Parts Aging Management (IE and!SR3)

Notes:

I. The potential for IASCC is only applicable to the weld regions of the Guide Lugs.

2. Applicable to welded regions of CEA Shroud Bases and Flow Channel Parts.

3. Applicable to the CEA Shroud Bolts/Lock Bars.

4. Applicable to the Guide Lug Insert Bolts.

Page 15 of27

Item Core Shroud Assembly (Bolted)

Core shroud bolts Core Shroud Assembly (Welded)

Core shroud plate-former plate weld St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pa e 16 of 27 Table 4 CE Plants Examination Acceptance and Expansion Criteria Applicability Examination Acceptance Expansion Expansion Criteria Additional Examination Criteria (Note 1)

Link(s)

Acceptance Criteria Bolted plant Volwnetric (U1) examination. a. Core support a. Confirmation that >5% of the core shroud a and b. Tue examination designs column bolts bolts in the four plates at the largest distance acceptance criteria for the NAforPSL

b. Barrel-from the core contain unacceptable indications UT of the core support The examination acceptance shroud bolts shall require UT examination of the lower column bolts and barrel-criteria for the UT of the core support column bolts barrel within the next 3 shroud bolts shall be shroud bolts shall be refueling cycles.

established as part of the established as part of the examination technical examination technical justification.

justification.

b. Confirmation that >5% of the core support colunm bolts contain unacceptable indications shall require UT examination of the barrel-shroud bolts within the next 3 refueling cycles.

Plant designs Visual (EVT-1) examination.

Remaining Confirmation that a surface-breaking indication Tue specific relevant with core axial welds

> 2 inches in length has been detected and sized condition is a detectable shrouds in the core shroud plate-former plate weld at the crack-like surface assembled in The specific relevant core shroud re-entrant corners (as visible from indication.

two vertical condition is a detectable the core side of the shroud), within 6 inches of sections crack-like surface indication.

the central flange and horizontal stiffeners, shall Applicable for require EVT

• l examination of all remaining PSL axial welds by the completion of the ne>.t refueling outage.

Page 16 of27

Item Core Shroud Assembly (Welded)

Shroud plates Core Shroud Assembly (Bolted)

Assembly Core Shroud Assembly (Welded)

Assembly St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pa e 17 of 27 Table 4 CE Plants Examination Acceptance and Expansion Criteria Applicability Examination Acceptance Expansion Expansion Criteria Additional Examination Criteria (Note 1)

Link(s)

Acceptance Criteria Plant designs Visual (EVT-1) examination.

a. Remaining

a. Confirmation that a surface-breaking The specific relevant with core axial welds indication > 2 inches in length has been detected condition is a detectable shrouds

b. Ribs and and sized in the axial weld seams at the core crack-like surface assembled with The specific relevant shroud re-entrant comers at the core mid-plane indication.

full-height condition is a detectable rings shall require EVT-1 or UT examination of all shroud plates crack-like surface indication.

remaining axial welds by the completion of the NAforPSL 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.

Bolted plant Visual (VT-3) examination.

None NIA NIA designs NAforPSL 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.

Plant designs Visual (VT-I) examination.

None NIA NIA with core shrouds assembled in The specific relevant two vertical condition is evidence of sections physical separation between Applicable for the upper and lower core PSL shroud sections.

Page 17 of27

Item Core Support Barrel Assembly Upper (core support barrel) flange weld St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pa e 18 of 27 Table 4 CE Plants Examination Acceptance and Expansion Criteria Applicability Examination Acceptance Expansion Expansion Criteria Additional Examination Criteria (Note 1)

Link(s)

Acceptance Criteria All plants Visual (EVT-1) examination.

Lower core Confirmation that a swiilce-breaking indication The specific relevant Applicable to support beams

> 2 inches in length has been detected and sized condition is a detectable PSL Upper core in the upper flange weld shall require that an crack-like surface The specific relevant barrel cylinder EVT-1 examination of the lower core support indication.

condition is a detectable (including beams, upper core barrel cylinder and upper crack-like swiilce indication.

welds) core barrel flange be performed by the Upper core completion of the next refueling outage.

barrel flange (cast)

Page 18 of27

Item Core Support Barrel Assembly Lower cylinder girth welds Lower Support Structnre Core support column welds Core Support Barrel Assembly Lower flange weld Core Support Barrel Assembly Expandable plugs and patches Lower Support Structnre Core support plate St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pa e 19 of 27 Table 4 CE Plants Examination Acceptance and Expansion Criteria Applicability Examination Acceptance Expansion Expansion Criteria Additional Examination Criteria (Note 1)

Link(s)

Acceptance Criteria All plants Visual (EVT-1) examination.

Lower cylinder a. Confirmation that a surface-breaking The specific relevant Applicable to axial welds iudication >2 inches iu leogth has been detected condition for the expansion PSL The specific relevant and sized in the lower cylinder girth weld shall lower cyliuder axial welds is require an EVT-1 examination of all accessible a detectable crack-like condition is a detectable lower cylinder axial welds by completion of the surface indication.

crack-like surface iudication.

ne>..'t refueling outage.

All plants Visual (VT-3) examination.

None None Applicable to PSL The specific relevant condition is missing or separated welds.

All plants Visual (EVT-1) examination.

None NIA NIA Applicable to PSL The specific relevant condition is a detectable crack-like iudication.

PSL Unitl Visual (EVT-1) examination.

None NIA Only The specific relevant condition is a detectable crack-like surface indication.

All plants with Visual (EVT-1) examination.

None NIA NIA a core support plate The specific relevant Applicable to condition is a detectable PSL crack-like surface iudication.

Page 19 of27

Item Upper Internals Assembly Fuel alignment plate Control Element Assembly Instrument Guide Tubes Lower Support Structure Deep beams NOIB St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pa e 20 of 27 Table 4 CE Plants Examination Acceptance and Expansion Criteria Applicability Examination Acceptance Expansion Expansion Criteria Additional Examination Criteria (Note 1)

Link(s)

Acceptance Criteria All plants with Visual (EVT-1) examination.

None NIA NIA core shrouds assembled with The specific relevant full-height condition is a detectable shroud plates crack-like surface indication.

NAforPSL All plants with Visual (VT-3) examination.

Remaining Confirmed evidence of missing supports or The specific relevant instruments instrument separation at the welded joint between the tubes conditions are missing tubes in the tubes within and supports shall require the visual (VT-3) supports and separation at CEA shroud The specific relevant the CEA examination to be expanded to the remaining the welded joint between the assembly conditions are missing shroud instrument tubes within the CEA shroud tubes and the supports.

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

PSL tubes and the supports.

All plants with Visual (EVT-1) examination.

None NIA NIA core shrouds assembled with full-height The specific relevant shroud plates condition is a detectable NA forPSL crack-like indication.

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

Page 20 of27

St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Page 21 of 27 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 ofMRP-227-A are also addressed in the response to LAI #1.

LAI #1 Applicability ofFMECA and Functionality Analysis Assumptions As addressed in Section 3.2. 5.1 of this SE, 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 RV/ 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-22 7. This is Applicant/Licensee Action Item 1.

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

FPL Response to LAI #1 and Bounding Assumption ofMRP-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 ofMRP-227-A) is:

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

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

3. Comparison of the typical CE-designed PWRRVI components to the St. Lucie Units 1and2 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 ofMRP-191 are consistent with St. Lucie Units 1 and 2 RVI component materials.

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 1 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 1and2 RVI components do not impact the application of the MRP-227-A generic aging management strategy.

Page 21 of27

St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pa2e 22 of 27 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 :rvJRP-191 generic FMECA and in the :rvJRP-232 functionality analysis based on the following:

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

a. FMECA and functionality analysis for :rvJRP-227-A made the following assumption of30 years of operation with high-leakage core loading patterns followed by 30 years oflow-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 :rvJRP-191 and requirements for :rvJRP-227-A application.

b. St. Lucie Units 1and2 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 (:rvJRP) documents regarding operational parameters affecting fluence.

2. The St. Lucie Units 1and2 reactor coolant system operates between T001d and Thot* Tcold is not less than 532°F and there were no changes to Tcold due to extended power uprate (EPU). T hot 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 consistent with the assumptions used to develop the :rvJRP-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 (:rvJRP-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 :rvJRP-191. 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 :rvJRP-191.

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:rvJRP-191 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 1and2 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:rvJRP-227-A. With the exception of the thermal shield removal for Unit 1, the modifications to the St. Lucie Units 1 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 :rvJRP-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 :rvJRP-191. The materials for Page 22 of27

St. Lucie Units 1and2 -to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Page 23 of 27 those components are also consistent with MRP-191, 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 assumptions in MRP-191, MRP-227-A, and MRP-232, confirming the applicability of the generic FMECA.

==

Conclusion:==

St. Lucie Units 1and2 comply with LAI #1 of the Nuclear Regulatory Commission Safety Evaluation on MRP-227, Revision 0. Therefore, the requirement is met for application ofMRP-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 addressed in I 0 CFR 54.4, each applicant/licensee is responsible for identifj;ingwhich RV! components are within the scope ofLRfor its facility.

Applicants/licensees shall review the information in Tables 4-1 and 4-2 in MRP-189, Revision I, and Tables 4-4 and 4-5 in MRP-191 and identifY whether these tables contain all of the RV! components that are within the scope of LR for their facilities in accordance with JO CFR 54.4. If the tables do not identifY all the RV! 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.

The response to LAI #2 is based directly upon Westinghouse Letter LTR-RIAM-13-75, Rev. O, Final Summary Report for St. Lucie Units 1 and 2 for PWROG P A-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 oflicense renewal for St. Lucie Units 1 and 2 to those components contained in Table 4-5 ofMRP-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 (Al\\1R) that are not included in MRP-191. 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 1 and 2 license renewal program are consistent with those contained in MRP-191.

The in-core instrumentation (ICI) guide tubes for both units have a different material than that specified in MRP-191, 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 oflicense renewal, but not included in the generic CE-designed PWRRVI components from Table 4-5 ofMRP-191, will be managed for the period ofextended 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-191. 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.

Page 23 of27

St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pa2e 24 of 27

==

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 ofMRP-227-A. Therefore, St. Lucie Units 1 and 2 meet the requirement for application ofMRP-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 addressed in Section 3. 2. 5. 3 in this SE, applicants/licensees of CE and Westinghouse are required to pe1form 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:

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 PEO.

==

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/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&Wflange weld shall coriform 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:

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.

LAI #5 Application of Physical Measurements as part ofl&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 Page 24 of27

St. Lucie Units 1 and 2 to Reactor Vessel Internals Aging Management Plan FPL Letter L-2017-015 Pal!e 25 of 27 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 ofFinal 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 1 and 2 core shrouds.

Basic assumptions of the PA were that the gap be measurable using the specified VT-1 inspection resolution and that it satisfy functionality requirements. The Units 1 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-1 inspection method ofMRP-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

1. 5.

Note: A non-proprietary copy of the calculation was submitted to the NRC in response to the RAis in Feb 2016.

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 barrel cylinders (including vertical and circumferential seam welds), B&W former plates, B&W external bafjl.e-to-baffl.e bolts and their locking devices, B&W core barrel-to-former bolts and their locking devices, and B&W core barrel assembly internal bafjl.e-to-bafjle bolts. The MRP also identified that although the B&W core barrel assembly internal baffl.e-to-baffl.e 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-22 7, applicants/licensees shall provide their justification for the continued operability of each of the inaccessible components and, if necessa1y, 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 1and2 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/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 IM! 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 RV! 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 mcry also need to consider limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques. The requirement mcry not apply to components that were previously evaluated as not requiring aging management during development of MRP-22 7. 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 ofMRP-227. This is Applicant/Licensee Action Item 7.

Page 25 of27

St. Lucie Units 1and2 to FPL Letter L-2017-015 Reactor Vessel Internals Aging Management Plan Page 26 of 27 The response to LAI #7 is based directly upon Westinghouse Letter LTR-RIAM-13-75, Rev. 0, Final Summary Report for St. Lucie Units 1 and 2 for PWROG P A-MSC-0983 Cafeteria Task Deliverables.

FPL Response:

FPL Response:

Applicant/Licensee Action Item 7 from the NRC's final Safety Evaluation on MRP-227, Revision 0 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.

The St. Lucie Units 1and2 RVI CASS components and the assessment of their susceptibility to TE are summarized in Table 1 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 1 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 SS; 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 ofSCC of the weld in MRP-191, 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-191. 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 Columns Low, 0.5 max Static

S 20%ll)

One column not susceptible to 1E Potentially >20%l'J Remaining columns I

potentially susceptible to IB<2J CEA Shroud Tubes Low, 0.5 max Centrifugal All Not susceptible to 1E 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

S 20%l'J Not susceptible to TE Notes:

I.

Calculated ferrite content is based on CMTR data, input into Hull's formula per the guidance ofNUREG/CR-4513, Rev. 1. Where molybdenum is not listed on the CMTR, a value of0.5 percent is used. Where nitrogen is not listed on the CMTRa value of0.04 percent is used.

2.

Where component-specific CMTR data are not available, the ferrite content is calculated based on permitted variations in AS1M A35 l, 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 martensitic stainless 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.

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St. Lucie Units 1 and 2 to FPL Letter L-2017-015 Reactor Vessel Internals Aging Management Plan Page 27 of 27

==

Conclusion:==

The results of this evaluation do not conflict with strategy for aging management ofRVI 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 addressed in Section 3.5.l 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 RV/ 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 ofMRP-227-A are provided in Section 3. These sections satisfy the requirements of LAI #8.

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