L-2018-193, Set 6 Responses to Request for Additional Information (RAI) on Subsequent License Renewal Application Safety Review

Set 6 Responses to Request for Additional Information (RAI) on Subsequent License Renewal Application Safety Review

ML18311A299
Person / Time
Site:	Turkey Point
Issue date:	11/02/2018
From:	Maher W Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAW-18-4826, EPID L-2018-RNW-0002, L-2018-193
Download: ML18311A299 (152)

Text

ATTACHMENTS 12P, 13P AND 25P CONTAIN INFORMATION REQUESTED TO BE WITHHELD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 November 2, 2018 U.S. Nuclear Regulatory CommIBsion Attn: Document Control Desk Washington, D.C. 2055~~-0001 Re: Florida Power & Light Company Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-193 10 CFR 54.17 Turkey Point Units 3 and 4 Subsequent License Renewal Application Safety Review Requests for Additional Information (RAI) Set 6 Responses

References:

1. FPL Letter L-2018-004 to NRC dated January 30, 2018, Turkey Point Units 3 and 4 Subsequent License Renewal Application (ADAMS Accession No. ML18037A812)

2. FPL Letter L-2018-082 to NRC dated April 10, 2018, Turkey Point Units 3 and 4 Subsequent License Renewal Application - Revision 1 (ADAMS Accession No. ML18113A134)

3. NRC RAI E-Mail to FPL dated October 4, 2018, Requests for Additional Information for the Safety Review of the Turkey Point Subsequent License Renewal Application - Set 6 (EPID No. L-2018-RNW-0002) (ADAMS Accession Nos. ML18269A227 and ML18269A228)

Florida Power & Light Company (FPL) submitted a subsequent license renewal application (SLRA) for Turkey Point Units 3 and 4 to the NRC on January 30, 2018 (Reference 1) and SLRA Revision 1 on April 10, 2018 (Reference 2).

The purpose of this letter is to provide, as attachments to this letter, public and certain non-public (proprietary) responses to the safety review RAls issued by the NRC on October 4, 2018 (Reference 3). The RAI responses and corresponding attachments and associated information enclosures are indexed on pages 2 and 3 of this letter. The attachments identify revisions amending the SLRA (if applicable).

Attachments 12P, 13P and 25P have been placed after Attachment 30 of this submittal and contain proprietary information (enclosed within brackets and/or marked 'Withhold from Public Disclosure Under 10 CFR 2.390') that FPL requests.be withheld from public disclosure under 10 CFR 2.390(a)(4). The withholding request applications for this proprietary information are enclosed with, or referenced within, Attachments 12, 12P, 13, 13P, 25 and 25P.

Florida Power & Light Company 700 Universe Boulevard, Juno Beach, FL 33408

ATTACHMENTS 12P, 13P AND 25P CONTAIN INFORMATION REQUESTED TO BE WITHHELD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-193 Page 2 of 3 If you have any questions, or need additional information, please contact me at 561-691-2294.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on November 2, 2018.

Sincerely,

~~~

William Maher Senior Licensing Director Florida Power & Light Company WDM/RFO Attachments: 33 RAI Responses (refer to Letter Attachment Index)

Enclosures:

4 RAI Response Enclosures (refer to Letter Enclosures Index)

LETTER ATTACHMENT INDEX Attachment NRC RAI Attachment NRC RAI 1

4.3.4-1 16 8.2.3.35-3 2

4.3.4-2 17 8.2.3.35-4 3

8.2.3.10-1 18 3.5.1.100-1 4

2.5-1 19 3.5.1.66-1 5

8.2.3.-1-1 20 8.2.3.13-1 6

8.2.3.1-2 21 3.3.1.199-1 7

3.5.2-9-1 22 8.2.3.9-1 8

8.2.3.30-1 23 4.7.6-1 9

8.2.3.30-2 24 4.7.6-2 10 3.5.1.9-1 25,25P 8.2.3.7-1 11 3.5.1.9-2 26 8.2.3.7-2 12,12P 4.7.5-9 27 8.2.3.7-3 13,13P 4.7.5-10 28 8.2.3.7-4 14 8.2.3.35-1 29 8.2.3.7-5 15 8.2.3.35-2 30 B.2.3.7-7

ATTACHMENTS 12P, 13P AND 25P CONTAIN INFORMATION REQUESTED TO BE WITHHELD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-193 Page 3 of 3 LETTER ENCLOSURES INDEX Attachment Enclosure Attachment Enclosure 12, 12P 1

25,25P cc: w/o Attachments 12P, 13P and 25P Senior Resident Inspector, USNRC, Turkey Point Nuclear Regional Administrator, USNRC, Region II Project Manager, USNRC, Turkey Point Nuclear Plant Project Manager, USNRC, SLRA Plant Project Manager, USNRC, SLRA Environmental Ms. Cindy Becker, Florida Department of Health

,.~..

1

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.3.4-1 L-2018-193 Attachment 1 Page 1 of 3 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

1. Reactor Pressure Vessel Underclad Cracking, TLAA 4.3.4 Regulatory Basis:

Pursuant to 10 CFR 54.21 (c), the SLRA shall include an evaluation of time-limited aging analyses (TLAAs). The applicant shall demonstrate that (i) the analyses remain valid for the period of extended operation; (ii) the analyses have been projected to the end of the period of extended operation; or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation. In accordance with 10 CFR 54.21 (c)(1 )(ii), the applicant has proposed to disposition the SLRA Section 4.3.4 TLAA for RPV underclad cracking in accordance with 10 CFR 54.21 (c)(1 )(ii) to demonstrate that the analyses have been projected to the end of the subsequent period of extended operation (SPEO).

Background:

To support its 10 CFR 54.21 (c)(1 )(ii) disposition of the RPV underclad cracking TLAA, the SLRA included PWR Owners Group (PWROG) Report PWROG-17031-NP, Revision O in of the SLRA. The PWROG-17031-NP report provides a generic methodology for analysis of underclad cracks in Westinghouse RPVs, applicable to 80-years of plant operation. PWROG17031-NP is not generically approved by the NRC staff for use SLR applications. Therefore, the staff is reviewing the PWROG-17031-NP report, as included in the SLRA, to determine whether this supports the applicant's TLAA disposition of 10 CFR 54.21 (c)(1 )(ii).

RAI 4.3.4-1 Issue:

PWROG-17031-NP shows that the Code-allowable flaw sizes for normal/upset/test and emergency/faulted loading conditions per IWB-3600 remain the same for 80 year applications as those in the 2002 version of this methodology, WCAP-15338-A, which is NRG-approved for 60 years. This is based on consideration of the same governing transient intensities for 3-Loop plants, as well as the continued use of time-invariant upper shelf fracture toughness (Kie) of 200 ksi°'1in for all transient analyses.

In order for an assumed Kie fracture toughness of 200 ksi°'1in to remain valid for the SPEO, the Level B, C, and D transient temperatures shall exceed the limiting adjusting RT NOT values for the analyzed flaw depths by at least 104.25 °F; this is based on the Kie curve provided in the ASME Code,Section XI, Appendix A.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.3.4-1 L-2018-193 Attachment 1 Page 2 of 3 Request:

Considering the projected state of RPV beltline neutron embrittlement through the end of the SPEO, as analyzed in SLRA Section 4.2, and the limiting temperatures for the governing transients evaluated in PWROG-17031-NP, please justify the continued use of 200 ksi'1in as the RPV material fracture toughness for determining the allowable flaw sizes in Section 5.5 (normal, upset, and test) and Section 5.7 (emergency and faulted conditions) of PWROG-17031-NP.

FPL Response:

PWROG-17031-NP and WCAP-15338-A calculate fracture toughness (Kie) per ASME Section XI, Appendix A, A-4200. Since there is no prescribed upper limit in the ASME code, 200 ksi'1in was conservatively used as a maximum (or "upper shelf') value.

All limiting transients for normal, upset and test conditions have high fluid temperatures, and the calculated Kie exceeds 200 ksi'1in even if the 1 OCFR50.61 PTS screening value of 270°F is used. It is noted that per the Turkey Point SLRA, Turkey Point Units 3 and 4 do not exceed this screening criteria. Therefore, K,e was limited to 200 ksi'1in to maintain conservatism and be in line with industry practices. For transients of emergency and faulted conditions, if T-RTNoT > 104.25 °F, 200 ksi'1in is used; otherwise, the Kie equation per A-4200 is used.

While fluence increases from 60 to 80 years, the level of neutron embrittlement does not significantly increase for limiting materials. As shown in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," Figure 1, the embrittlement curve becomes "saturated" over time, resulting in a significantly smaller increase in embrittlement per increase in fluence. Plants with relatively high copper content in their limiting vessel materials (su~h as Turkey Points Units 3 and 4) are in the saturation zone by 60 years, if not sooner. Therefore, there is only a small increase in limiting RT NOT values from 60 years to 80 years.

As a further conservatism, underclad cracks are assumed to be surface flaws which results in a conservative Kl. The surface flaw assumption results in a higher calculated fatigue crack growth rate as it considers a water environment. The maximum flaw depth due to fatigue crack growth for 80 years is 0.4267 inches as shown in PWROG 17031-NP, Section 5.4. This represents a significant margin compared to the Normal/Upset/Test allowable flaw depth of 0.67 inches and Emergency/Faulted allowable flaw depth of 1.25 inches. This also allows for a small increase in RT NOT to accommodate any additional neutron embrittlement beyond 60 years of operation.

References:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.3.4-1 L-2018-193 Attachment 1 Page 3 of 3 Associated SLRA Revisions:

None Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.3.4-2 L-2018-193 Attachment 2 Page 1 of 1 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 4.3.4-2 Issue:

The staff noted that analysis results for the "Large Steamline Break" transient in Section 5.7 of PWROG-17031-NP and in Appendix A-5 ofWCAP-15338-A (2002) show a slightly more limiting allowable flaw depth for the continuous circumferential flaw (2.21 inches) compared to the allowable flaw depth for the continuous axial flaw (2.50 in.). Considering its reliance on the assumed KIC fracture toughness of 200 ksi'"'1in for the upper shelf temperature regime, this analysis result is inconsistent with RPV beltline shell tensile stress due to internal pressure.

Request:

Considering the RPV shell axial stress versus RPV shell hoop stress due to internal pressure and a fixed Kie value of 200 ksi'"'1in, please explain how the IWB-3600 analysis of Large Streamline Break transient can result in the more limiting allowable flaw depth (2.21 in.) for the continuous circumferential flaw compared to the 2.50 in. allowable flaw depth for the continuous axial flaw. If this is a typographical error, please correct it in both the WCAP-15338-A and PWROG-17031-NP reports.

FPL Response:

The pressure hoop stress for axial flaws is higher than the pressure axial stress for circumferential flaws. The large steam line break transient results in a continuous circumferential flaw size of 2.64 inches.

This is a typographical error that is in both WCAP-15338-A and PWROG-17031-NP Rev.

0. An errata letter has been issued for WCAP-15338-A to document the typographical correction. Report PWROG-17031-NP Revision O has been revised to a Revision 1. The table containing this typographical error has been removed from Revision 1 of PWROG-17031-NP, and no further action is required for this report.

References:

None Associated SLRA Revisions:

None Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRG RAI No. B.2.3.10-1 L-2018-193 Attachment 3 Page 1 of 8 NRC RAI Letter Nos. ML18269A227 and. ML18269A228 Dated October 04, 2018

2. Steam Generators, GALL AMP XI.M19 Regulatory Basis:

Section 54.21 (a)(3) of 10 CFR requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. As described in the SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report.

In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

RAI 8.2.3.10-1

Background:

SRP-SLR Section 3.1.2.2.11, "Cracking Due to Primary Water Stress Corrosion Cracking," recommends actions to manage aging of divider plate assemblies depending on the material of the divider plate assemblies and whether industry analyses (EPRI 3002002850, "Steam Generator Management Program: Investigation of Crack Initiation and Propagation in the Steam Generator Channel Head Assembly," October 2014) are bounding for the applicant's unit(s).

Since the divider plate assemblies in the applicant's SGs are fabricated of Alloy 600 material and the applicant's evaluation of EPRI TR-3002002850 is not yet complete, the following recommendations from SRP-SLR are potentially applicable at Turkey Point:

For units with divider plate assemblies fabricated of Alloy 600 or Alloy 600 type weld materials, if the analyses performed by the industry (EPRI [Electric Power Research Institute] 3002002850) are applicable and bounding for the unit, a plant-specific AMP is not necessary.

For units with divider plate assemblies fabricated of Alloy 600 or Alloy 600 type weld materials, if the industry analyses (EPRI 3002002850) are not bounding for the applicant's unit, a plant-specific AMP is necessary or a rationale is necessary for why such a program is not needed. A plant-specific AMP (one beyond the primary water chemistry and the steam generator programs) may include a one-time inspection that is capable of detecting cracking to verify the effectiveness of

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 8.2.3.10-1 L-2018-193 Attachment 3 Page 2 of 8 Issue:

the water chemistry and steam generator programs and the absence of PWSCC in the divider plate assemblies.

SLRA Section 3.1.2.2.11, "Cracking due to Primary Water Stress Corrosion Cracking,"

states, in part:

Turkey Point has an Alloy 600 divider plate and the EPRI analysis is applicable.

FPL is evaluating the industry analysis (EPRI TR-3002002850) as part of the existing Steam Generators AMP for the current PEO to determine whether it is bounding for Turkey Point. This evaluation is scheduled for completion by the end of 2018.

Until the applicant has completes its evaluation to determine whether the industry analyses are bounding, the NRC staff cannot complete its review.

Request:

Please provide the results of FPL's evaluation regarding EPRI 3002002850 and whether the relevant industry analyses are bounding for the units.

FPL Response:

The industry analysis (EPRI TR-3002002850) is conservatively assumed not to be bounding for PTN. Thus, the PTN steam generator divider plate assemblies will be inspected, through the PTN One-Time Inspection AMP, using a volumetric inspection to verify the effectiveness of the Water Chemistry and Steam Generator AMPs and the absence of PWSCC in the divider plate assemblies. The PTN SLRA is revised to include this inspection.

References:

None Associated SLRA Revisions:

SLRA Section 3.1.2.2.11, Table 3.1-1, Table 3.1.2-5, Appendix A Table 17-3, Appendix 8 Section 8.2.3.10 and 8.2.3.20 are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise SLRA Section 3.1.2.2.11.1 as follows:

Turkey Point has an Alloy 600 divider plate.FPL is evaluating tihe industry analysis (EPRI TR-3002002850) as part of the existing Steam Generators AMP for the current PEG to determine v,hether it is conservatively assumed not to be bounding for Turkey Point. This evaluation is scheduled for completion by the end of 2018.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.10-1 L-2018-193 Attachment 3 Page 3 of 8 If the analysis is determined to be bounding, the Steam Generators /\\MP will be revised to address primary 1.vater stress corrosion cracking in the divider plate for the PEG, and carried forward through the SPEO. A plant specific AMP is no necessary.

If the analysis is determined to not be bounding, a One Time Inspection AMP 'Nill be implemented for SLR toThe One-Time Inspection AMP is used to verify the effectiveness of the Water Chemistry and Steam Generators AMPs. The volumetric examinations will be performed by qualified personnel and the techniques used will be capable of detectieng primary water stress corrosion cracking in the divider plate assemblies and associated welds.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.10-1 L-2018-193 Attachment 3 Page 4 of 8 Revise SLRA Table 3.1-1 as follows:

Table 3.1-1: Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Further Item Aging Effect/

Aging Management Evaluation Number Component Mechanism Program (AMP)/TLAA Recommended Discussion 3.1-1, 025 Steel (with nickel alloy Cracking due to AMP XI.M2, "Water Yes (SRP-SLR The Water Chemistry and cladding) or nickel alloy primary water sec Chemistry," and AMP Sections Steam Generators AMPs steam generator primary XI.M19, "Steam 3.1.2.2.11.1 and will be used to manage side components: divider Generators." In addition, a 3.1.2.2.11.2) primary water sec in the plate and tube-to-tube plant-specific program is to divider plate exposed to sheet welds exposed to be evaluated.

reactor coolant. The reactor coolant One-Time lns~ection AMP is used to verif~

the effectiveness of the Water Chemistrl and Steam Generators AMPsA potential enl=lanseF!'lent to tl=le gteaFfl GeneFatOFS AMP is ieentifiee to aEIEIFess in accordance with EPRI Report 3002002850

• regarding sec in the divider plate.

Further evaluation is documented in Section 3.1.2.2.11.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.10-1 L-2018-193 Attachment 3 Page 5 of 8 Revise SLRA Table 3.1.2-5 as follows:

Table 3.1.2-5: Steam Generators - Summary of Aging Management Evaluation Aging Effect Aging Component Intended Requiring Management Type Function Material Environment Management Program Divider Plate Direct flow Nickel alloy Reactor Cracking Water coolant Chemistry Steam Generators One-Time Ins~ection NUREG-2191 Item Table 1 Item Notes IV.D1.RP-3.1-1, 025 E

367

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.10-1 L-2018-193 Attachment 3 Page 6 of 8 Revise SLRA Appendix A Table 17-3 as follows:

Table 17-3 List of SLR Commitments and Implementation Schedule (Continued)

Aging Management Program or NUREG-2191 No.

Activity (Section)

Section Commitment Implementation Schedule 14 Steam Generators (17.2.2.10)

XI.M19 Continue the existing PTN Steam No later than 6 months prior Generators AMP, including enhancement to the SPEO, i.e.:

to:

PTN3: 1/19/2032 a) Incorporate the latest EPRI steam PTN4: 10/10/2032 generator guidelines per NEI 97-06; If U1e EliviEleF 13late asseFflelies aFe Aet 99!:IAEleEI ee iAEl!:1StFy aAalyses ~PRI JQQ2QQ2!HiQ, ssl:leE11:1le a eAetiFfle iAs13estieA te eeAfiFFfl tl:le effestiveAess ef tl:le WateF Cl:leFflistry aAEI SteaFfl GeAemteFs.A.MPs.

b)

Perform a one-time inspection as Rart of the One-Time lnsRection AMP using qualified techniques capable of detecting primary water stress corrosion cracking in the divider plate assemblies and associated welds.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.10-1 L-2018-193 Attachment 3 Page 7 of 8 Revise SLRA Appendix B Section B.2.3.10 as follows:

Inspections of the divider plate may be are required to be conducted prior to entering fef the SPEO. Nickel-alloy divider plates could experience PWSCC as described in the SRP-SLR (Reference B.3.10). The analysis performed by the industry (EPRI TR 3002002850 (Reference B.3.104)) is applicable as PTN has an aAlloy_-600 divider plate.

The industry analyses are currently being evaluated to determine whether it is conservatively assumed to not be bounding for PTN.:. and 'Nill be completed prior to the SPEO. If the evaluation is not bounding, Thus, PTN will perform a one-time inspection of the divider plates to confirm the effectiveness of the actions currently in place to manage SCC (Water Chemistry AMP and the visual inspections performed for the existing Steam Generator AMP).

• Enhancements The PTN Steam Generators AMP will be enhanced as follows for alignment with NUREG-2191. The changes and enhancements are to be implemented no later than six months prior to entering the SPEO.

Element Affected Enhancement

3. Parameters Monitored or A Rew iRs13estieR may ee FeEfYiFeEl te ee im13lemeRteEl feF SbR AR Inspected evalYatieR ef ePRI ::rn JQQ2QQ2!H:iQ ~RefeFeRse EU.rn4) is sst:ieElYleEl te ee sem13leteEl ey tt:ie eREl ef 2Q~ g as 13aFt ef tt:ie e*iStiR§ steam !:JeReFateF AMP feF tt:ie SYFFeRt PeG te EleteFmiRe if tt:ie P+t>I EliviEleF 13late is eeYREleEl ey tt:ie aRalysis. If tt:ie aRalysis is Ret eeYREliR!:J, a eRe time iRs13estieR mYst ee sst:ieElYleEl Re lateF tt:iaR Si* meRtRS J:lFi9F te eRteFiR§ tt:ie SPeG, aREl mYst ee 13eFfeFmeEl J:lFi9F te eRteFiR§ tt:ie SPeG.

Update AMP procedures to include adding reference lists, which include EPRI documents, and including additional means for monitoring loose parts.

Revise SLRA Appendix B Section B.2.3.20 as follows:

The PTN steam generator transition cone was cut in the middle to replace the bottom part of the steam generator. The resulting new circumferential weld is a field weld, as opposed to the upper and lower transition cone welds which were performed in a controlled

. manufacturing facility. To identify the effectiveness of the water chemistry program, the PTN One-Time Inspection AMP will perform an inspection on the new transition cone weld of the steam generators. This inspection will be a volumetric inspection consistent with the techniques currently in place for the original transition cone welds.

The PTN steam generator divider plate will be inspected prior to entering the SPEO.

Nickel-alloy divider plates could experience PWSCC as described in the SRP-SLR (Reference 8.3.10). The analysis performed by the industry (EPRI TR 3002002850 (Reference 8.3.104)) is applicable as PTN has an Alloy 600 divider plate. The industry analyses are conservatively assumed to not be bounding for PTN. Thus,

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.10-1 L-2018-193 Attachment 3 Page 8 of 8 PTN will perform a one-time inspection of the divider plates to confirm the effectiveness of the actions currently in place to manage sec (Water Chemistry AMP and the visual inspections performed for the existing Steam Generator AMP).

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 2.5-1 L-2018-193 Attachment 4 Page 1 of 3 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

3. Scoping and Screening Results: Electrical and Instrumentation and Controls, SLRA2.5 Regulatory Basis:

The scoping criteria in 10 CFR 54.4(a)(3) require, in part, an applicant to consider "all systems, structures, and components relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for station blackout (SBO) (10 CFR 50.63)."

As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.4(a)(3) by ensuring that its methodology includes SSCs relied upon during the "coping duration"-and "recovery" phase of an SBO event. In addition, because 10 CFR 50.63(c)(1 )(ii) and its associated guidance in RG 1.155 include procedures to recover from an SBO that include offsite and onsite power, the offsite power system that is used to connect the plant to the offsite power source should also be included within the scope of the rule.

SRP-SLR, Section 2.5.2.1.1, "Components Within the Scope of SBO (10 CFR 50.63),"

states in part, that both the offsite and onsite power systems are relied upon to meet the requirements of the SBO Rule and include equipment that is required to cope with an SBO (e.g., alternate ac power sources), and the plant system portion of the offsite power system that is used to connect the plant to the offsite power source meeting the requirements under 10 CFR 54.4(a)(3).

RAI 2.5-1

Background:

SLRA section 2.1.3.4.5 states that Turkey Point's design satisfies the SBO Rule by providing for a unit cross-tie at the 4.16 kV level. It further states that resolution of the SBO issue for the Turkey Point nuclear units is by use of an alternate safety-related, Class 1 E, seismic Class/Category I, power source with the ability to align the source to the SBO unit within 10 minutes of confirmation of a station blackout condition. However, the highlighted electrical drawing the applicant supplied does not include the cross-tie and the alternate safety-related power source.

Issue:

1. It is not clear to the NRC staff whether the cross-tie and alternate AC power source are within the scope of license renewal.

2. The applicant's SLRA and electrical drawing identifying the SBO recovery path do not state the specific type of structures, cabling, and connections (e.g., bus duct, cable bus, metal enclosed bus, iso-phase bus, etc.) used in the SBO recovery path

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 2.5-1 L-2018-193 Attachment 4 Page 2 of 3 from alternate AC source via the unit 3 and 4 cross-tie.

Request:

1. Clarify whether the cross-tie and alternate AC power source (part of the SBO recovery path), are safety-related and are considered in-scope for license renewal.

2. Provide information regarding methods, and type of components and structures used for the unit 3 and 4 cross-tie connection.

FPL Response:

1. The SBO drawings included in the SLRA, depict the restoration power path for offsite power to the 4.16 kV auxiliary buses via the startup transformers following a SBO event. Cr.oss-tie breakers 3AD07 and 4AD07, and the emergency diesel generator (EOG) alternate AC power sources are not part of the restoration power path for offsite power.

Cross-tie breakers 3AD07 and 4AD07 are part of the PTN safety related 4.16 kV electrical system (System 005). Auxiliary 4.16 kV switchgear 30 and 40 (i.e., the swing switchgear) are also part of PTN System 005. The alternate power sources consist of the Unit 3 and Unit 4 EDGs 3A, 38, 4A and 48. These components provide onsite power supplies relied upon to meet safety related requirements, and the requirements of the SBO Rule to cope with a SBO. All of these components are safety related and are within the scope of SLR as shown in SLRA Table 2.2-3.

2. The 4.16 kV system has the capability via the cross-tie and the swing switchgear to connect any EOG with either the "A" or "B" switchgear of the opposite unit. The design provides the capability to perform this function from within the Control Room. The control circuitry includes permissives arid interlocks to prevent inadvertent closure of the station blackout tie breakers during conduct of emergency operating procedures. When the required permissives are satisfied, the station blackout tie breakers can be manually closed by operation of administratively controlled, key-lock control switches (one for each unit) provided in the Control Room. Closure of the breaker associated with the blackout unit will then enable the opposite blackout cross-tie breaker (the breaker associated with the non-blackout unit) to be closed. The status of station blackout permissives and breaker position indicating lights are provided in the Control Room.

Both the 3D and 40 auxiliary switchgear are located in the same Class I Unit 4 EOG Building but in separate rooms. The distance between the rooms is a relatively short span. The 4.16 kV inter-tie between the switchgear consist of 3-1/C 750 kcmil cables run in Class I conduit. There are no cable-bus, bus duct, metal-enclosed bus, iso-phase bus, manholes, etc. utilized in the cross-tie circuit

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 2.5-1 L-2018-193 Attachment 4 Page 3 of 3 configuration. The Unit 4 EOG building and the Control Building are Cla~s I structures and are within the scope of SLR as shown in SLRA Table 2.2-2.

References:

1. SBO Drawings: 561 O-E-1, Sheets 1 and 2 (Ref. 2.5.3.6 of the SLRA)

Associated SLRA Revisions:

None Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.1-1 L-2018-193 Attachment 5 Page 1 of 3 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

4. ASME Section XI lnservice Inspection, Subsections IWB, IWC, AND IWD, GALL AMP XI.M1 Regulatory Basis:

Title 10 of the Code of Federal Regulations (10 CFR) 54.21 (a)(3) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis (CLB) for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to the managing the effects of aging during the subsequent period of extended operation (SPEO) on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the subsequent renewed license will continue to be conducted in accordance with the CLB.

As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL SLR Report. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

Pursuant to 10 CFR 54.4(a)(1 ), safety-related systems, structures, and components which are those relied upon to remain functional during and following design-basis events (as defined in 10 CFR 50.49 (b)(1 )) to ensure the following functions -- (i) The integrity of the reactor coolant pressure boundary; (ii) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (iii) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in § 50.34(a)(1 ), § 50.67(b)(2), or§ 100.11 of this chapter, as applicable.

Pursuant to 10 CFR 54.4(a)(2), all nonsafety-related systems, structures, and components whose failure could prevent satisfactory accomplishment of any of the functions identified in paragraphs (a)(1 )(i), (ii), or (iii) of 54.4.

Background:

Sections 3.1.2.2.10.1 and 3.1.2.2.10.2 of the Turkey Point subsequent license renewal application (SLRA) provide discussions on the reactor pressure vessel head penetration control rod drive mechanisms (CROM) thermal sleeve wear. Specifically, 3.1.2.2.10.2 states the thermal sleeves "do not perform a subsequent license renewal intended function."

Notes for Table 2.3.1-2 of the Turkey Point SLRA, in part, states, "the thermal sleeves are considered to support the pressure boundary component intended function."

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.1-1 L-2018-193 Attachment 5 Page 2 of 3 In a teleconference public meeting dated September 10, 2018, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18262A078)), the applicant asserted that:

The stainless steel (SS) thermal sleeves of the CROM were not scoped into the Turkey Point SLRA because they were determined to be not part of the reactor coolant pressure boundary and not safety-related components, and their failure could not affect satisfactory accomplishment of any of the functions identified under 10 CFR 54.4(a)(1 );

Reference to SS thermal sleeve in both Table 2.3.1-2 and notes for Table 2.3.1-2 was intended for the pressurizer's thermal sleeves and not for the thermal sleeves of the CRDMs.

The enhancements discussed in Section 3.1.2.2.10 include the aging management of loss of material due to wear at the centering tab location and at the bottom of the nickel alloy CROM nozzle tubes which were scoped into Turkey Point SLRA, but not at the SS thermal sleeves of CRDMs which were not scoped into Turkey Point SLRA.

Issue:

Given the recent operating experience documented in Westinghouse letters, L TR-NRC-18-34 "Notification of the Potential Existence of Defects Pursuant to 10 CFR Part 21" dated May 23, 2018 (ML18143B678), L TR-NRC-18-53 "NSAL-18-1 Revision 0, "Thermal Sleeve Flange Wear Leads to Stuck Control Rod"" dated July 17, 2018 (ML18198A275),

and NRC Information Notice 2018-10 "Thermal Sleeve Flange Wear Leads to Stuck Control Rod at Foreign Nuclear Plant" dated August 29, 2018 (ML18214A710), the staff requests the following additional information.

RAI 8.2.3.1-1 Provide the basis for not scoping the SS thermal sleeves of CRDMs into Turkey Point SLRA, and justify why the control rod functionality could not be impacted if the thermal sleeve flange failed due to wear.

FPL Response:

Based on the recent operating experience discussed above regarding the CROM thermal sleeves, FPL has determined that the CROM thermal sleeves should be included in the scope of SLR for PTN. The CROM thermal sleeves have been determined to meet the scoping criteria pursuant to 10 CFR 54.4(a)(2) as failure of the thermal sleeve flange due to wear could adversely impact control rod functionality.

Revisions associated with SLRA Section 2.3.1, which includes the mechanical scoping and screening results for the reactor coolant system, are provided below. Additional

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.1-1 L-2018-193 Attachment 5 Page 3 of 3 SLRA revisions associated with the CROM thermal sleeves are provided in the response to RAI B.2.3.1-2.

References:

None Associated SLRA Revisions:

The SLRA is amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise SLRA Table 2.3.1-3 as follows:

Table 2.3.1-3 Reactor Vessel Components Subject to Aging Management Review Component Type Component Intended Function(s)

CROM thermal sleeves Guide rod control cluster assemblies (RCCAs}

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.1-2 L-2018-193 Attachment 6 Page 1 of 7 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 8.2.3.1-2 Discuss the aging management of loss of material due to wear at the SS thermal sleeve flange location, and at the bottom of the CROM tube location.

FPL Response:

The aging effect of loss of material due to wear of the CROM thermal sleeves and at the bottom of the CROM tube will be managed by the ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD aging management program (AMP). The necessary revisions to the PTN SLRA are provided below.

References:

1. Westinghouse letter L TR-NRC-18-34 "Notification of the Potential Existence of Defects Pursuant to 10 CFR Part 21" dated May 23, 2018 (ML181438678)

2. Westinghouse letter LTR-NRC-18-53 "NSAL-18-1 Revision 0, "Thermal Sleeve Flange Wear Leads to Stuck Control Rod"" dated July 17, *2018 (ML18198A275)

3. NRC Information Notice 2018-10 "Thermal Sleeve Flange Wear Leads to Stuck Control Rod at Foreign Nuclear Plant" dated August 29, 2018 (ML18214A710)

Associated SLRA Revisions:

SLRA Sections 3.1.2.2.10.2, B.2.3.1 and B.3 and Tables 3.1-1, 3.1.2-3 and Appendix A Table 17-3 are amended as indicated by the following text deletion (strikethrough text) and text addition (red underlined font) revisions.

Revise SLRA Section 3.1.2.2.10.2 as follows:

This item is not applicable to Turkey Point as the CROM thermal sleeves do not perform a subsequent license rene'Nal intended function. The wear interaction which impacts the CROM head penetration nozzles, which do perform a subsequent license renmval intended function, is described above in Section 3.1.2.2.10, item 1.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.1-2 L-2018-193 Attachment 6 Page 2 of 7 Based on recent CROM thermal sleeve operating experience documented in Westinghouse letters, L TR-NRC-18-34 "Notification of the Potential Existence of Defects Pursuant to 1 O CFR Part 21" dated May 23, 2018 (ML181438678), L TR-NRC-18-53 "NSAL-18-1 Revision O, "Thermal Sleeve Flange Wear Leads to Stuck Control Rod"" dated July 17, 2018 (ML18198A275), and NRC Information Notice 2018-10 "Thermal Sleeve Flange Wear Leads to Stuck Control Rod at Foreign Nuclear Plant" dated August 29, 2018 (ML18214A710), loss of material due to wear for the CRDM thermal sleeves is an aging effect requiring management during the PTN SPEC. This aging effect will be managed during the SPEC by the ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD AMP. The following enhancements are added to the existing program to adequately monitor the condition of the CRDM thermal sleeves:

Continue to monitor the industry operating experience regarding wear of CRDM thermal sleeves.

Perform CRDM thermal sleeve inspections as specified in Westinghouse NSAL-18-1, Rev. 0. This includes:

o Performance of visual inspections of reactor vessel (RV) upper internals on the top of the upper guide tube (UGT) for wear marks during every refueling outage starting after 2025. Examinations include looking for shiny surfaces on the top edge of the upper guide tube enclosure.

o Performance of visual inspections and measurements of thermal sleeves conjunction with the RV head volumetric examinations starting after 2025. Examinations include: 1 l a visual inspection of the bottom of the thermal sleeve guide funnels to look for any shiny surfaces on the bottom surface of the guide funnel that would indicate that the thermal sleeve guide funnels have dropped to a point where they are in contact with the top of the guide tube, and 2) a visual inspection of thermal sleeve guide funnel elevations to identify whether any sleeves are noticeably lower than others.,

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.1-2 L-2018-193 Attachment 6 Page 3 of 7 Revise SLRA Table 3.1-1 as follows:

Table 3.1-1: Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Aging Effect/

Aging Management Further Component Program Evaluation Discussion Number Mechanism (AMP)/TLAA Recommended

~Jot applicable.

+Re +b!Fkey PoiAt GRQM tl:leFFflal sleeves do Rot perioFFfl a sblbseqbleAt liceAse FeAewal iAteAded fb!ActioA.

Stainless steel, nickel alloy control rod drive Yes (SRP-SLR Consistent with NUREG-2191. The 3.1-1, 117 penetration nozzle Loss of material Plant-specific aging Section ASME Section XI lnservice thermal sleeves exposed due to wear management program 3.1.2.2.10.2) lnsgection 1 Subsections IWB1 IWC 1 to reactor coolant and IWD AMP will be used to manage loss of material due to wear in the control rod drive mechanism thermal sleeves ex~osed to reactor coolant.

Further evaluation is documented in Section 3.1.2.2.10.2.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.1-2 L-2018-193 Attachment 6 Page 4 of 7 Revise SLRA Table 3.1.2-3 as follows:

Table 3.1.2-3: Reactor Vessels -

Summary of Aging Management Evaluation Intended Aging Effect Component Type Material Environment Requiring Function Management Aging Management NUREG-Table 1 Notes Program 2191 Item Item CRDM thermal Guide rod Stainless Reactor Loss of material ASME Section XI IV.A2.R-414 3.1-11 117 E.

sleeves control steel coolant lnservice cluster lnspection1 assemblies Subsections IWB1 (RCCAs)

IWC1andlWD Revise SLRA Appendix A Table 17-3 "List of SLR Commitments and Implementation Schedule" as follows:

Aging Management NUREG-Implementation No.

Program or Activity 2191 Commitment (Section)

Section Schedule 5

ASME Section XI XI.M1 Continue the existing PTN ASME Section XI lnservice Inspection, Subsections IWB, No later than 6 months lnservice Inspection, IWC, and IWD AMP, including enhancement§. ta-for the following components:

prior to the SPEO, i.e.:

Subsections IWB,

1. CRDM Head Penetrations PTN3: 1/19/2032 IWC, and IWD a) Develop a wear depth measurement process for the CROM head penetrations.

(17.2.2.1)

PTN4: 10/10/2032 b) Incorporate inspections using the demonstrated process at accessible locations to measure depth of wear on the CROM housing penetration wall associated with contact.

c) Develop a procedure to estimate the wall thickness of the accessible CROM housing penetration wear in the area of interest at the end of the next reactor vessel head inspection interval and compare that projected wall thickness to the thickness used in the design basis analyses to demonstrate validity of the analyses.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.1-2 L-2018-193 Attachment 6 Page 5 of 7 Aging Management NUREG-No.

Program or Activity 2191 Commitment (Section)

Section d) Evaluate industry experience related to CROM housing penetration wear due to thermal sleeve centering pads and initiatives to measure CROM housing penetration wear and resulting nozzle wall thickness.

2.

CROM Thermal Sleeves a) Continue to monitor the industQ! O(!erating ex(!erience regarding wear of CROM thermal sleeves.

b) Perform visual ins(!ections of reactor vessel {RV} U(!(!er internals on the to(! of the U(!(!er guide tube {UGT} for wear marks during eve!Ji! refueling outage starting after 2025. Examinations include looking for shin~

surfaces on the to(! edge of the U(!(!er guide tube enclosure.

c) Perform visual ins(!ections and measurements of thermal sleeves in conjunction with the RV head volumetric examinations starting after 2025. Examinations includej 1} a visual ins(!ection of the bottom of the thermal sleeve guide funnels to look for an~ shin~ surfaces on the bottom surface of the guide funnel that would indicate that the thermal sleeve guide funnels have drO(!(!ed to a (!Oint where the~ are in contact with the to(! of the guide tube1 and 2} a visual ins(!ection of thermal sleeve guide funnel elevations to identifv whether an~ sleeves are noticeabl~ lower than others.

Implementation Schedule

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.1-2 L-2018-193 Attachment 6 Page 6 of 7 Add a new paragraph at the end of the "Program Description" section and revise the "Enhancement" section of SLRA Section B.2.3.1 as follows:

The program is also enhanced to manage the aging effect of loss of material due to wear in the CROM thermal sleeves. The enhancements include continuing to monitor the industry operating experience regarding wear of CROM thermal sleeves and perform CROM thermal sleeve inspections as specified in Westinghouse NSAL-18-1, Rev. 0. (Reference 8.3.151 ).

Element Affected Enhancement Develop a wear-depth measurement process for the CROM head penetrations to include:7 a)

Incorporate inspections using the demonstrated process at accessible locations to measure depth of wear on the control rod drive mechanism (CROM) housing penetration wall associated with contact.

b)

Develop a procedure to estimate the wall thickness of the accessible CROM housing penetration wear in the area of interest at the end of the next reactor vessel head inspection interval and compare that projected wall thickness to the thickness used in the design basis analyses to demonstrate validity of the analyses.

Develog insgection grocedure for the CROM

4. Detection of Aging Effects thermal sleeves to include:

a) Performance of visual insgections of reactor vessel {RV} ugger internals on the tog of the ugger guide tube {UGT}

for wear marks during eve!)! refueling outage starting after 2025.

Examinations include looking for shin)£ surfaces on the tog edge of the ugger guide tube enclosure.

b) Performance of visual insgections and measurements of thermal sleeves in conjunction with the RV head volumetric examinations starting after 2025. Examinations include; 1} a visual insgection of the bottom of the thermal sleeve guide funnels to look for an)£ shin)£ surfaces on the bottom surface of the guide funnel that would indicate

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.1-2 L-2018-193 Attachment 6 Page 7 of 7 Element Affected Enhancement that the thermal sleeve guide funnels have droeeed to a eoint where thel£ are in contact with the* toe of the guide tube1 and 2) a visual inseection of thermal sleeve guide funnel elevations to identi~ whether anl£ sleeves are noticeabll£ lower than others.

Add a new reference to SLRA Section 8.3 as follows:

B.3.151 Westinghouse letter L TR-NRC-18-53 "NSAL-18-1 Revision O, "Thermal Sleeve Flange Wear Leads to Stuck Control Rod"" dated July 17, 2018 (ML18198A275)

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.2-9-1 L-2018-193 Attachment 7 Page 1 of 3 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

5. Non-GALL AMR, Structural Components Regulatory Basis:

10 CFR 54.21 (a)(3) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under§ 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis (CLB). As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

RAI 3.5.2-9-1

Background:

SLRA Table 3.5.2-9 (SLRA page 3.5-149) states that the Unit 4 carbon steel diesel oil storage tank (DOST) liner embedded in concrete has no aging effect requiring management for the external surface exposed to the concrete.

GALL-SLR Item VII.J.AP-282 addresses carbon steel piping embedded in concrete and lists no aging effect and no AMP but does recommend further evaluation. The associated SRP-SLR Table 3.3-1 ID 112 also recommends further evaluation. The further evaluation section (SRP-SLR 3.3.2.2.9) notes that no aging effect is applicable for steel embedded in concrete if: (a) the concrete was constructed in accordance with ACI 318 or 349 (low

\\/\\(ater-to-cement ratio, low permeability, adequate air entrainment); (b) plant-specific operating experience indicates no degradation of the concrete that could lead to penetration of water to the metal surface; and (c) the piping is not potentially exposed to groundwater.

Issue:

Although the GALL-SLR lists no aging effect for steel embedded in concrete, this is only applicable if the above conditions are met. SLRA Section 3.3.2.2.9 states that a review of operating experience indicates there are occurrences of concrete degradation that could lead to the penetration of water to the metallic surfaces of the tank.

Request:

Based on the operating experience described in Section 3.3.2.2.9 of the SLRA, explain why aging management is unnecessary for the external metallic surface of the Unit 4

I Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251

. FPL Response to NRC RAI No. 3.5.2-9-1 L-2018-193 Attachment 7 Page 2 of 3 DOST tank embedded in concrete. If sealants are relied upon to keep water off the external metallic surfaces of the tank, explain how the sealants are age-managed.

FPL Response:

The Unit 4 DOSTs are constructed of steel lined concrete with leak chases installed behind the liner walls and floors to remove any water that may accumulate, such as rainwater on the roof of the diesel generator building, which serves as the top of the tank.

All of the piping penetrations through the liner and concrete walls are routed to the Unit 4 diesel oil transfer pump rooms which are indoors. As a result, the wall penetrations are not exposed to water. Sealants are not relied on to keep water off of the external surface of the tanks.

The rooftop penetrations were identified by FPL in Attachment 15 of the referenced letter.

These rooftop penetrations include manway enclosures, tank sample connections, and flame arrestors. These penetrations are exposed to outdoor precipitation. According to the RAI B.3.17-2 response, these penetrations do rely on penetration seals (Weatherproofing) and the seals are managed for "Loss of sealing" using the Structures Monitoring Aging Management Program (AMP) as stated in the PTN SLRA, Revision 1, Table 3.5.2-9.

No aging effect is applicable for Unit 4 DOST concrete embedded steel due to the all of the following:

a) The concrete was constructed in accordance with ACI 318-63.

b) Even though concrete degradation has been identified at PTN, degradation of the Unit 4 DOST concrete would not lead to water penetration that could damage the liner, since the installed leak chases would divert the water away. The aging effects for the concrete are managed by the Structures Monitoring Aging Management Program (AMP) which will ensure prope_r function of the leak chase.

c) The through-wall piping and respective piping penetrations are not potentially exposed to groundwater. Although the roof penetrations are exposed to outdoor precipitation, they are sealed, and their respective seals are managed for "Loss of sealing" per SLRA Table 3.5.2-9.

The "Associated SLRA Revisions" section shows how SLRA Section 3.3.2.2.9 is revised to clarify why the concrete exposed liner does not require aging management.

References:

FPL Letter L-2018-175 to NRC dated October 17, 2018, Turkey Point Units 3 and 4 Subsequent License Renewal Application, Safety Review Requests for Additional In-formation (RAI) Set 5 Responses.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.2-9-1 L-2018-193 Attachment 7 Page 3 of 3 Associated SLRA Revisions:

SLRA Section 3.3.2.2.9 paragraph 4 is amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise SLRA Section 3.3.2.2.9 paragraph 4 as follows:

Consistent with the recommendation of GALL-SLR, the Buried and Underground Piping and Tanks AMP is used to manage loss of material in steel piping exposed to concrete.

This AMP provides for the management of aging effects through periodic visual inspection.

Any visual evidence of loss of material will be evaluated for acceptability. Deficiencies will be documented in accordance with the 10 CFR Part 50, Appendix B Corrective Action Program. The Buried and Underground Piping and Tanks AMP is described in Appendix B.

The Unit 4 Diesel Oil Storage Tanks (DOSTs) have leak chases installed which divert potential water intrusion and eliminate the need to manage aging effects of the carbon steel liner surfaces facing the concrete tank.

r Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.30-1 L-2018-193 Attachment 8 Page 1 of 8 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

6. ASME Section XI, Subsection IWE, GALL AMP XI.S1 Regulatory Basis:

Section 54.21 (a)(3) of 10 CFR requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis for the period of extended operation. As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report and when evaluation of the matter in the GALL-SLR Report applies to the plant.

RAI B.2.3.30-1

Background:

The "detection of aging effects" program element of GALL-SLR AMP XI.S1 states, in part:

The requirements of ASME Code Section XI, Subsection IWE and 10 CFR 50.55a are further supplemented to require a one-time volumetric examination of metal shell or liner surfaces that are inaccessible from one side, only if triggered by plant-specific OE [operating experience]. The trigger for this supplemental examination is plant-specific occurrence or recurrence of measurable metal shell or liner corrosion (base metal material loss exceeding 10 percent of nominal plate thickness) initiated on the inaccessible side or areas, identified since the date of issuance of the first renewed license. This supplemental volumetric examination consists of a sample of one-foot square locations that include both randomly-selected and focused areas most likely to experience degradation based on OE and/or other relevant considerations such as environment. Any identified degradation is addressed in accordance with the applicable provisions of the AMP.

The sample size, locations, and any needed scope expansion (based on findings) for this one-time set of volumetric examinations should be determined on a plant-specific basis to demonstrate statistically with 95 percent confidence that 95 percent of the accessible portion of the containment liner is not experiencing corrosion degradation with greater than 10 percent loss of nominal thickness.

Guidance provided in EPRI TR-107514 may be used for sampling considerations.

(emphasis added)

SLRA Section B.2.3.30 states that the Turkey Point ASME Section XI, Subsection IWE AMP, with enhancements, will be consistent with the 10 elements of NUREG-2191 AMP XI.S1. Further, in SLRA Section B.2.3.30, the enhancement to the "detection of aging effects" program element states:

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 8.2.3.30-1 L-2018-193 Attachment 8 Page 2 of 8 Issue:

If site-specific OE identified after the approval of the SLRA triggers the requirement to implement a one-time supplemental volumetric examination, then perform this inspection by sampling randomly-selected, as well as focused, liner locations susceptible to corrosion that are inaccessible from one side. The trigger for this one-time examination is site-specific occurrence or recurrence of liner corrosion that is determined to originate from the inaccessible (concrete) side. Any such instance would be identified through code inspections performed since June 6, 2002. (emphasis added).

The staff is unable to determine that the "detection of aging effects" program element, with the stated enhancement, will be consistent with that in GALL-SLR AMP XI.S1 because of the following issues identified with regard to the enhancement.

1. Contrary to the GALL-SLR specification that the one-time volumetric examination would be triggered by plant-specific OE identified since the date of issuance of the first renewed license (i.e., June 6, 2002 for Turkey Point U3 and U4), the enhancement states the trigger to be "site-specific OE identified after the approval of the SLRA."

2. The trigger specified in the GALL-SLR is the site-specific occurrence or recurrence of the stated plant-specific OE without regard to the method by which (how) it is identified. Contrary to this, the SLRA enhancement states that the triggering OE would be specific to that identified through code inspections.

3. The enhancement does not include the sampling specifications in the GALL-SLR program element that the sample size, locations and any needed scope expansion for this one-time volumetric examination shall demonstrate statistically with 95 percent confidence that 95 percent of the accessible portion of the containment liner is not experiencing corrosion degradation with greater than 10 percent loss of nominal thickness.

4. Based on information provided in the SLRA and on the electronic portal, the staff is unable to positively determine whether or not there has been operating experience of containment liner corrosion initiated on the inaccessible (concrete) side of Turkey Point Unit 3 or Unit 4 identified since the June 6, 2002, issuance of first renewed license.

Request:

1) Provide a revised enhancement to the "detection of aging effects" program element in SLRA Section B.2.3.30, that addresses the issues identified in 1 through 3 above and would make the Turkey Point AMP program element consistent with that in,GALL-SLR AMP XI.S1, or explain why a revised enhancement is

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.30-1 L-2018-193 Attachment 8 Page 3 of 8 unnecessary.

2) State if there has been operating experience of containment liner corrosion initiated on the inaccessible (concrete) side identified at Turkey Point Unit 3 or Unit 4 since the June 6, 2002, issuance of the first renewed license.

3) If the response to Request 2 is yes, then (i) describe the operating experience and how it was addressed in the corrective action program; and (ii) explain how the conduct of the "triggered" supplemental volumetric examination, including schedule, is sufficiently captured in the revised enhancement in response to Request 1.

FPL Response:

1) SLRA Section 17.2.3.30, Table 17-3, item 34, and Section B.2.3.30 are revised as described below to -

provide further clarification of when site specific OE will trigger volumetric examinations, clarify that degradation may be detected by maintenance or testing activities (in addition to code inspections), and to include sample size, location and scope expansion considerations from the AMP basis document.

2) The only containment liner corrosion partially attributed to the inaccessible (concrete) side is the small hole found in the floor of the Unit 4 reactor cavity sump liner plate in 2006 that is described in SLRA Sections B.2.3.4 (pg B-7) and B.2.3.30 (pg B-235).

This operating experience is discussed further in item 3) below.

Otherwise, there has been no operating experience of liner corrosion that initiated on the inaccessible side since issuance of the PTN renewed licenses. Site operating experience since 2011 is expressly described in SLRA Section B.2.3.30 (pg B-233 to B-234). Moisture barrier and toe plate (between the moisture barrier and liner) degradation was identified and corrected for Unit 3 in 2015, with no degradation of the liner itself. Some minor sealant degradation and toe plate degradation where identified for Unit 4 in 2016, with only surface discoloration of the liner due to outage-related activities in a congested area. These indications were also corrected. A 2010 instance of liner degradation in the lower region of the Unit 3 reactor pit was addressed and repaired through augmented visual and ultrasonic examinations. The degradation initiated from the accessible side due to boric acid. There was no evidence of corrosion on the concrete side. The degraded liner section was replaced and a pro'per coating applied. A similar coating was applied to the lower region of the Unit 4 reactor pit.

3) As described in SLRA Section B.2.3.5 (pg B-72), a small hole in the floor of the Unit 4

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRG RAI No. B.2.3.30-1 L-2018-193 Attachment 8 Page 4 of 8 reactor cavity sump liner plate was found and corrected in November 2006. The corrosion was attributed to water trapped behind the liner plate when high pressure water was used to cut a hole in the Containment building to facilitate reactor vessel closure head (RVCH) replacement. Bulges in the liner plate provided a path for retained water to collect beneath the reactor sump.

A walkdown revealed water trickling out of the hole below a steel plate used to support one of the sump pumps, when it was displaced, and evaluation considered the shim material used for the plate. The hole was plugged and welded, the area was left with steel shims, and the steel support plate returned. The repair was leak tested successfully. In addition, periodic inspections of sump areas were added to the ASME Section XI, Subsection IWE program so that future degradation can be identified before the condition adversely impacts structural steel components or coatings.

Subsequent inspections found the area acceptable. Furthermore, numerous UT measurements were taken in the sump pit area around the time of this repair and no additional sample expansion warranted.

Therefore, this apparent localized corrosion that may have originated on the inaccessible (concrete) side, as a result of trapped water associated with the RVCH temporary modification in 2005, does not affect the ASME Section XI, Subsection IWE AMP for SLR beyond the operating experience discussion in the AMP basis document that is summarized in SLRA Section B.2.3.30 and identifying cavity sump pit as a likely area for focused inspection.

SLRA Sections 17.2.2.30, B.2.3.4, and B.2.3.30, as well as Table 17-3, item 34 are revised as described below. The response to Set 5 RAI 3.5.2.1.2-1 submitted via FPL Letter L-2018-175 includes unrelated revisions to SLRA Section 17.2.3.30; Table 17-3, Item 34, and B.2.3.30.

References:

None Associated SLRA Revisions:

SLRA Sections 17.2.2.30, B.2.3.4, B.2.3.30 and Table 17-3 (Item 34) are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise Section 17.2.2.30, paragraph 3 on page A-35 as follows:

If triggered by site-specific OE, this AMP also includes a one-time supplemental volumetric examination by sampling both randomly selected and focused liner locations (such as a reactor cavity sump pit) susceptible to corrosion that are

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.30-1 L-2018-193 Attachment 8 Page 5 of 8 inaccessible from one side. This sampling is conducted to demonstrate, with 95% confidence, that 95% of the accessible portion of the liner is not experiencing greater than 10% loss of wall thickness.

Revise Section B.2.3.4, 1st numbered paragraph on page B-72 as follows:

1. In November 2006, a small hole was found in the floor of the Unit 4 reactor cavity sump liner plate. The corrosion was attributed to water trapped behind the liner plate when high pressure water was used to cut a hole in the Containment building to facilitate reactor vessel closure head (RVCH} replacement. Bulges in the liner plate provided a path for retained water to collect beneath the reactor-sump. The hole was plugged and welded and the area was left with stainless steel shims GA-for a stainless steel support plate. The repair was leak tested successfully.

Periodic inspections of sump areas were added to the ASME Section XI Subsection IWE AMP (Section B.2.3.30) program so that future degradation can be identified before the condition adversely impacts structural steel components or coatings. It appeared to be attributed to Boric Acid.

Revise Section B.2.3.30, 3rd paragraph on page B-230 as follows:

If site-specific OE identified after the approval of the SLRA triggers the requirement to implement a one-time supplemental volumetric examination, then this inspection is performed by sampling randomly-selected, as well as focused (such as cavity sump

.Pill, liner locations susceptible to corrosion that are inaccessible from one side. The trigger for this one-time examination is site-specific occurrence or recurrence of liner corrosion that is determined to originate from the inaccessible (concrete) side. Any such instance would be identified through code inspections or other maintenance/testing activities performed since June 6, 2002.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.30-1 L-2018-193 Attachment 8 Page 6 of 8 Revise Enhancements for Section B.2.3.30 on page B-232 as follows:

Element Affected Enhancement

14. Detection of Aging If site-specific OE identified after the approval of Effects the SLRA triggers the requirement to implement a one-time supplemental volumetric examination, then perform this inspection by sampling randomly-selected, as well as focused (such as cavity sump pit), liner locations susceptible to corrosion that are inaccessible from one side.

This sampling is conducted to demonstrate1 with 95% confidence1 that 95% of the accessible portion of the liner is not ex~riencing greater than 10% wall loss. The trigger for this one-time examination is site-specific occurrence Gf recurrence of liner corrosion that is determined to originate from the inaccessible (concrete) side.

Any such instance would be identified through code inspections or other maintenance/testing activities performed since June 6, 2002.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 8.2.3.30-1 L-2018-193 Attachment 8 Page 7 of 8 Revise 1st paragraph of industry operating experience for Section 8.2.3.30 on page 8-232 as follows:

NRC IN 2010-12 was issued to inform addressees of the then-recent issues involving the corrosion of the steel reactor containing building liner. The NRC expected recipients to review the information for applicability of their facilities and to consider actions, as appropriate, to avoid similar problems. In response, PTN issued an AR which evaluated that the containment liner inspection programs in effect at PTN are effective in detecting and addressing any found degradation of the containment liner due to corrosion, and ensure that the structural integrity and design function of the component are maintained.

Additionally, the planned ASME Section XI Subsection IWE inspection of Unit3 the liner in 2010 effectively located and corrected liner plate corrosion prior to a loss of function.

Further discussion is located in Section iii below.

Revise site-specific operating experience for Section 8.2.3.30 on 3rd full paragraph on page 8-233 as follows:

There has been no evidence of corrosion degradation on the concrete side of the liner plate, apart from the pin hole identified in the Unit 4 cavity sump area in 2006, which was partially attributed to water trapped under the liner plate when high pressure water was used to cut a hole in the Containment building to facilitate RVCH replacement as described in SLRA Section B.2.3.4. The following review of site-specific OE provides examples of how PTN is managing aging effects associated with the PTN ASME Section XI, Subsection IWE.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.30-1 L-2018-193 Attachment 8 Page 8 of 8 Revise the commitment for the ASME Section XI Subsection IWE AMP in Table 17-3, item 34, on page A-103 as follows:

Aging Management Program or Activity NUREG-2191 Implementation No.

(Section)

Section Commitment Schedule 34 ASME Section XI, XI.S1 Continue the existing PTN ASME Section XI, Complete any Subsection IWE AMP Subsection IWE AMP, including enhancement to:

applicable pre-SPEO (17.2.2.30) a) Include preventive actions, consistent with one-time inspections industry guidance, to provide reasonable no later than 6 months assurance that bolting integrity is maintained for or the last RFO prior structural bolting, and if high strength bolting is to SPEO.

used, the appropriate guidance in Section 2 of Corresponding dates Research Council for Structural Connections are as follows:

publication "Specification for Structural Joints PTN3: 1/19/2032 Using High-Strength Bolts" is to be considered.

PTN4: 10/10/2032 b) Implement a one-time inspection of metal liner surfaces that samples randomly selected as well as focused (such as cavitv sumQ Qit} locations susceptible to loss of thickness due to corrosion from the concrete side if triggered by site-specific OE identified through code inspections or other maintenance/testing activities performed since June 6, 2002. This samQling is conducted to demonstrate1 with 95% confidence1 that 95% of the accessible QOrtion of the liner is not exQeriencing greater than 10% wall loss.

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.30-2 L-2018-193 Attachment 9 Page 1 of 4 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 8.2.3.30-2

Background:

The "operating experience" program element of GALL-SLR AMP XI.S1 includes industry operating experience described in NRC Information Notice (IN) 2014-07 concerning

• degradation of inaccessible areas of containment liner due to moisture intrusion into leak-chase channel systems through degraded interfaces at the containment floor level from lack of inspection of these interface components that serve a moisture barrier function.

The staff's review of Drawing 561 O-C-164, Revision 4, Containment Structure Floor Liner Plate Plan, indicates the existence of an air chase system along the circumference as well as at inner locations of the containment floor for Turkey Point Unit 3 and Unit 4. The typical air test connection shown in Details 2, 3 and 9 and "Typical Air Test Connection" details on the drawing appear to indicate that these connections provide pathways, at the containment floor-level interface, for potential intrusion of moisture into inaccessible areas of the liner plate.

Issue:

Based on review of the Program Basis Document for the IWE AMP and the Second IWE Inspection Interval Program Plan, it is not clear if barriers (e.g., pipe cap, pipe plug, etc.)

associated with the air chase system test connections at containment floor-level interfaces, intended to prevent moisture intrusion, are being inspected under IWE program as discussed in NRC IN 2014-07.

Request:

Discuss whether or not the air chase test connection components at the containment floor-level interfaces, that serve a function to prevent moisture intrusion into inaccessible areas of the liner, are examined in the IWE Program as discussed in NRC IN 2014-07. If not, justify the adequacy of the Turkey Point ASME Section XI, Subsection IWE to manage liner degradation in inaccessible areas related to operating experience described in NRC IN 2014-07.

FPL Response:

There are currently no air chase test connections formally being tested or inspected as part of the containment IWE inspections for Unit 3 and Unit 4.

The evaluation of subject NRC Information Notice (IN) 2014-07, through the Turkey Point Corrective Action Program, was based on consideration of previous inspection reports and pictures from the lnservice Inspection (ISi) group. Actions were assigned for ISi personnel to perform walkdowns to search for possible interfaces of the air chase system

Turkey Point Units 3 and 4 Pocket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.30-2 L-2018-193 Attachment 9 Page 2 of 4 with the liner plate and generate new action requests for any accessible interface that is not covered by the current IWE inspection program.

The Unit 4 Containment 14' elevation walkdown in PT4-28 (2014) did not identify any accessible interfaces of the air chase system with the liner plate. However, the Unit 3 Containment 14' elevation walkdown in PT3-28 (2015) identified multiple (test) connections whose conditions were observed to be satisfactory.

As such, though not formally included in the implementing procedures for the ASME Section XI, Subsection IWE AMP, accessible air chase system interfaces with the containment liner plate are inspected. Each air chase channel is seal welded to the containment liner, for leak tight integrity. The test connections for the air chase system include standard threaded pipe caps and a concrete cover with lean grout after tests were completed and approved.

For completeness, the IWE inspection plan will be enhanced to include inspection of the accessible air chase test connections at the containment floor-level interfaces, that serve a function to prevent moisture intrusion into inaccessible areas of the liner. SLRA Section B.2.3.30, as well as Table 17-3, item 34, are revised as described below. The response to Set 5 RAI 3.5.2.1.2-1 in the referenced FPL letter includes unrelated revisions to SLRA Table 17-3, Item 34, and Section B.2.3.30.

References:

FPL Letter L-2018-175 to NRC dated October 17, 2018, Turkey Point Units 3 and 4 Subsequent License Renewal Application, Safety Review Requests for Additional Information (RAI) Set 5 Responses (ML18292A642)

Associated SLRA Revisions:

SLRA Section B.2.3.30 and Table 17-3 (Item 34) are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise enhancement and operating experience discussions in Section B.2.3.30 on page B-232 as follows:

Element Affected Enhancement

10. Operating Ui;!date insE!ection E!rocedure/E!lan to formal!~

Experience include accessible air chase s~stem test connections at the containment floor-level interfaces.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.30-2 L-2018-193 Attachment 9 Page 3 of 4 Operating Experience Industry Operating Experience NRC IN 2010-12 was issued to inform addressees of the then-recent issues involving the corrosion of the steel reactor containing building liner. The NRC expected recipients to review the information for applicability of their facilities and to consider actions, as appropriate, to avoid similar problems. In response, PTN issued an AR which evaluated that the containment liner inspection programs in effect at PTN are effective in detecting and addressing any found degradation of the containment liner due to corrosion, and ensure that the structural integrity and design function of the component are maintained.

Additionally, the planned ASME Section XI Subsection IWE inspection in 2010 effectively located and corrected liner plate corrosion. Further discussion is located in Section iii belovv.

NRC IN 2014-07 was issued to inform addressees of identified issues concerning degradation of floor weld leak channel systems of steel containment shell and concrete containment metallic liner that could affect leak-tightness and aging management of containment structures. This IN provides examples of operating experience at some plants of water accumulation and corrosion degradation in the leak-chase channel system that has the potential to affect the leak-tight integrity of the containment shell or liner plate. In each of the examples, the licensee had no provisions in its ISi plan to inspect any portion of the leak-chase channel system for evidence of moisture intrusion and degradation of the containment metallic shell or liner within it. The moisture intrusion and associated degradation found within leak chase channels, if left uncorrected, could have resulted in more significant corrosion degradation of the containment shell or liner and associated seam welds.

Turkey Point does have an air chase system inside the Unit 3 and Unit 4 containment structures, similar to the leak chase system discussed in IN 2014-07.

Accessible interfaces of the air chase system with the containment liner plate were walked down during a recent outage (PT3/4-28) and test connection (grouted pipe cap) condition determined to be satisfactory or indeterminate (inaccessible). The inspection procedure/plan will be updated to formally include the accessible air chase system test connections in future IWE inspections.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 8.2.3.30-2 L-2018-193 Attachment 9 Page 4 of 4 Revise the commitments for the ASME Section XI, Subsection IWE AMP in Table 17-3, item 34, on page A-103 as follows:

Aging Management NUREG-Program or Activity 2191 Implementation No.

(Section)

Section Commitment Schedule 34 ASME Section XI, XI.S1 Continue the existing PTN ASME Section XI, Subsection Complete any Subsection IWE AMP IWE AMP, including enhancement to:

applicable pre-SPEO (17.2.2.30) a) Include preventive actions, consistent with industry one-time inspections guidance, to provide reasonable assurance that no later than 6 months bolting integrity is maintained for structural bolting, or the last RFO prior to and if high strength bolting is used, the appropriate SPEO.

guidance in Section 2 of Research Council for Corresponding dates Structural Connections publication "Specification for are as follows:

Structural Joints Using High-Strength Bolts" is to be PTN3: 1/19/2032 considered.

PTN4: 10/10/2032 b) (See response to RAI 8.2.3.30-1, Attachment 8 to this letter).

c) (See response to RAI 3.5.2.1.2-1, Attachment 14 to Reference 1).

d} U~date ins~ection ~rocedure/~lan to formal!~

include accessible air chase s~stem test connections at the containment floor-level interfaces.

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-1 L-2018-193 Attachment 1 0 Page 1 of 9 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

7. Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analyses, TLAA 4.6 Regulatory Basis:

Section 54.21 (c)(1) of 10 CFR requires the applicant to evaluate time limited aging analyses (TLAA). Section 54.21 (a)(3) of 10 CFR requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report and when evaluation of the matter in the GALL-SLR Report applies to the plant.

RAI 3.5.1.9-1

Background:

Section 4.6 of the SRP-SLR states that containment metal liner plates, metal containments and penetrations (including personnel airlocks, equipment hatches, sleeves, dissimilar metal welds, and bellows), may be designed in accordance with requirements of Section Ill of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). The SRP-SLR also states that if a plant's code of record requires a fatigue parameter evaluation (fatigue analysis or fatigue waiver), then this analysis may be a time limited aging analyses (TLAA) and must be evaluated in accordance with 10 CFR 54.21 (c)(1) for the subsequent period of extended operation.

SRP-SLR Table 3.5-1, item 009, associated with the further evaluation section 3.5.2.2.1.5, recommends that metal liner, metal plates, and penetrations (including personnel airlocks, equipment hatches, penetration sleeves, bellows, vent lines, etc.) be managed for cumulative fatigue damage due to cyclic loading using the TLAA disposition from SRP-SLR Section 4.6, if a current licensing basis (CLB) fatigue analysis exists.

Otherwise, if a CLB fatigue analysis does not exist for these components, SRP-SLR Table 3.5-1, item 027, recommends these components to be managed for cracking due to cyclic loading using the GALL-SLR Report AMP XI.S1, "ASME Section XI, Subsection IWE," and AMP XI.S4, "10 CFR Part 50, Appendix J" aging management programs (AMPs).

Turkey Point subsequent license renewal application (SLRA) Table 3.5-1, item 3.5-1, 009, states that the containment liner plate fatigue analysis is addressed in SLRA Section 4.6 and that the further evaluation is documented in SLRA Section 3.5.2.2.1.5. The SLRA further evaluation for "Cumulative Fatigue Damage" states that liner and connections to penetration sleeves and hatches for the containment structures is addressed in SLRA Section 4.6. SLRA Section 4.6, "Containment Liner Plate, Metal Containments, and

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-1 L-2018-193 Attachment 10 Page 2 of 9 Penetrations Fatigue," addresses a TLAA for containment liner plate and piping penetrations. Additionally, SLRA Table 3.5-1, item 3.5-1, 027, states that this item is not applicable.

Also, SLRA Section B.2.3.30 "ASME Section XI, Subsection IWE" AMP states: "PTN

[Turkey Point] has no pressure-retaining components subject to cyclic loading without CLB fatigue analysis.... " Further, there are no enhancements proposed in the SLRA AMP to perform recommended supplemental surface examination or other applicable technique capable of detecting fine cracking; and no Appendix J leak rate tests are credited.

Issue:

Based on the information provided in the SLRA, it is not clear if a fatigue analysis or fatigue waiver exists for containment penetrations other than piping penetrations (e.g.

personnel airlocks, equipment hatch and/or personnel hatch, electrical penetrations, etc.),

or how the aging effect of cracking due to cyclic loading will be adequately managed, in accordance with 10 CFR 54.21 (a)(3), for these components.

The staff notes that the SLRA does not clearly state if a CLB fatigue analysis exists for the components described above, or how these components were designed for cyclic loading. If a CLB fatigue analysis or fatigue waiver exists for these components, it is not clear how these analyses were dispositioned in SLRA Section 4.6, or why Table 3.5-1, item 3.5-1, 009, its associated Table 2 items, and Section 3.5.2.2.1.5 of the SLRA does not address these components to demonstrate that the aging effect of cumulative fatigue damage due to cyclic loading will be adequately managed during the subsequent period of extended operation. Likewise, if a CLB fatigue analysis or fatigue waiver does not exist for these components, it is not clear how cracking due to cyclic loading will be adequately managed for these components during the subsequent period of extended operation since SLRA Table 3.5-1, item 3.5-1, 027, states that this item is not applicable.

Request:

1. Clarify if a fatigue analysis or fatigue waiver exists for containment penetrations other than piping penetrations (i.e. personnel airlocks, equipment hatch, personnel hatch, electrical penetrations, etc.)

2. If a fatigue analysis or fatigue waiver exists, address with supporting justification the disposition under 10 CFR 54.21 (c)(1) of each containment penetration fatigue analysis or fatigue waiver, and describe the following for each analyzed component:

a. the name of the transients considered in each analysis,

b. the design cycle limits of each transient,

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-1 L-2018-193 Attachment 10 Page 3 of 9

c. the projected cycles to 80-years of operation for each transient, and

d. the review of the calculated cumulative usage factor (CUF), if applicable.

Otherwise, pursuant to 10 CFR 54.21 (a)(3), if fatigue analysis or fatigue waiver does not exist, clarify how containment penetrations other than piping penetrations will be adequately managed for cracking due to cyclic loading during the subsequent period of extended operation (i.e. SRP-SLR Table 3.5-1, item 027, with GALL-SLR Report recommendation for supplemental surface examinations using AMP XI.S1, "ASME Section XI, Subsection IWE" or identifying and crediting appropriately justified Appendix J leak rate tests).

FPL Response:

Responses to the above numbered requests are as follows:

1. Turkey Point UFSAR Appendix 58, Section 8.2.1 provides a description of the fatigue analysis that was performed for the containment liner plate and penetrations.

However, based on a review of available documentation, FPL has been unable to locate the original fatigue analysis, or confirm if a fatigue waiver exists for the Turkey Point non-piping containment penetrations.

2. Considering the response to item 1 above, the PTN SLRA is revised to indicate cracking due to cyclic loading of non-piping containment penetrations (hatches, electrical penetrations, etc.) will be managed by supplemental surface examinations using the ASME Section XI, Subsection IWE AMP (XI.S1) and the 10 CFR 50, Appendix J AMP (XI.S4).

References:

1. FPL Letter L-2018-175 to NRC dated October 17, 2018, Turkey Point Units 3 and 4 Subsequent License Renewal Application, Safety Review Requests for Additional Information (RAI) Set 5 Responses (ML18292A642)

Associated SLRA Revisions:

SLRA Section 3.5.2.2.1.5, Table 3.5-1, Table 3.5.2-1, Table 3.5.2-15 and Appendix A, Table 17-3, and Section 8.2.3.30 are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions. Note these changes include the changes associated with the response to RAI 3.5.1.9-2 (Attachment 11 ).

SLRA Section 3.5.2.2.1.5 is revised as follows:

Cumulative fatigue damage for the Turkey Point liner plate and connections to piping penetration§. and sleeves and hatches for the containment structures is addressed in the Containment Liner Plate and Penetrations Fatigue Analysis TLAA in Section 4.6.

Cumulative fatigue damage for non-piping penetrations (hatches, electrical

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-1 L-2018-193 Attachment 10 Page 4 of 9 penetrations, etc.), dissimilar metal welds, and the fuel transfer tube expansion joints will be managed by supplemental surface examinations using the ASME Section XI, Subsection IWE AMP (XI.S1) and the 10 CFR Part 50, Appendix J AMP.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-1 L-2018-193 Attachment 10 Page 5 of 9 SLRA Table 3.5-1 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for the Containment, Structures, and Component Supports Further Item Aging Aging Management Evaluation Number Component Effect/Mechanism Program (AMP)/TLAA Recommended Discussion 3.5-1, 009 Metal liner, metal plate, Cumulative fatigue TLAA, SRP-SLR Section Yes (SRP-SLR TLAA applies to liner plate personnel airlock, damage due to cyclic 4.6, "Containment Liner Section and piping penetrations onll£.

equipment hatch, control loading (Only if CLB Plate and Penetration 3.5.2.2.1.5)

Further evaluation for other rod drive (CRD) hatch, fatigue analysis exists)

Fatigue Analysis" components is documented in penetration sleeves; Section 3.5.2.2.1.5.

penetration bellows, steel elements: torus; vent line; vent header; vent line bellows; downcomers, suppression pool shell; unbraced downcomers, steel elements: vent header; downcomers 3.5-1, 027 Metal liner, metal plate, Cracking due to cyclic AMP XI.S1, "ASME Section No Net a1313lisaele. GbB fati§1:1e airlock, equipment hatch, loading (CLB fatigue XI, Section IWE," and AMP aRalysis is ElessFieeEI iR SestieR CRD hatch; penetration analysis does not XI.S4, "10 CFR Part 50, 4.-e,.

sleeves; penetration exist)

Appendix J" Consistent with NUREG-2191 bellows, steel elements:

for non-piping penetrations torus; vent line; vent (hatches 1 electrical header; vent line bellows; penetrations1 etc.}1 dissimilar downcomers, suppression metal welds and fuel transfer pool shell tube exoansion ioints.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-1 L-2018-193 Attachment 10 Page 6 of 9 SLRA Table 3.5.2-1 is revised as follows:

Table 3.5.2-1: Containment Structure and Internal Structural Components -

Summary of Aging Management Evaluation Aging Effect c

Component Intended Requiring Aging Management NUREG-Table 1 Type Function Material Environment Management Program 2191 Item Item Liner plate, Pressure Carbon steel Air-indoor Cumulative TLAA - Containment II.A.3.C-13 3.5-1, 009

.ruID.!!.9.

boundary uncontrolled fatigue damage Liner Plate, Metal 12enetrations Fire Containments, and barrier Penetrations Fatigue Liner 12late1 Pressure Carbon Air-indoor Cracking due to ASME Section Xl 1 II.A3.CP-37 3.5-1. 027 non-(!i(!ing bounda!Y steel uncontrolled Cl£Clic loading Subsection IWE 12enetrations Fire 10 CFR Part 501

{hatches1 barrier AppendixJ electrical 12enetrations1 etc.}.

dissimilar metal welds Notes A

A

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-1 L-2018-193 Attachment 10 Page 7 of 9 SLRA Table 3.5.2-15 is revised as follows:

Table 3.5.2-15: Spent Fuel Storage and Handling -Summary of Aging Management Evaluation Aging Effect Component Intended Requiring Aging Management Type Function Material Environment Management Program Fuel transfer Pressure Stainless Air-indoor Cracking due to ASME Section Xl1 tube boundaD£ steel uncontrolled c~clic loading Subsection IWE

{including 10 CFR Part 501 penetration AppendixJ sleeves and expansion joints)

NUREG-Table 1 2191 Item Item Notes II.A3.CP-37 3.5-11 027 A

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-1 L-2018-193 Attachment 10 Page 8 of 9 SLRA Table 17-3 is revised as follows:

Aging Management Program NUREG-2191 No.

or Activity (Section)

Section 34 ASME Section XI, Subsection XI.S1 IWE (17.2.2.30)

Commitment Implementation Schedule Continue the existing PTN ASME Section XI, Subsection Complete any applicable IWE AMP, including enhancement to:

pre-SPEO one-time a)

Include preventive actions, consistent with industry inspections no later than 6 guidance, to provide reasonable assurance that months or the last RFO prior bolting integrity is maintained for structural bolting, and to SPEO. Corresponding if high strength bolting is used, the appropriate dates are as follows:

guidance in Section 2 of Research Council for PTN3: 1/19/2032 Structural Connections publication "Specification for PTN4: 10/10/2032 Structural Joints Using High-Strength Bolts" is to be considered.

b) (See response to RAI B.2.3.30-1, Attachment 8 to this letter).

c)

(See response to RAI 3.5.2.1.2-1, Attachment 14 to Reference 1).

d)

(See response to RAI B.2.3.30-2, Attachment 9 to this letter).

e) Perform SUE!E!lemental surface examinations to detect cracking due to C)lclic loading of non-E!iE!ing E!enetrations {hatches 1 electrical E!enetrations1 etc.) 1 dissimilar metal welds 1 and fuel transfer tube exoansion ioints.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-1 L-2018-193 Attachment 1 O Page 9 of 9 SLRA Section B.2.3.30, paragraph 6 is revised as follows:

PTN has no pressure-retaining components associated with the liner and piping penetrations subject to cyclic loading without CLB fatigue analysis (Turkey Point UFSAR Appendix 58, Section B.2.1 ). Pressure retaining components associated with the containment liner, including attachments and piping penetrations, are addressed by a fatigue evaluation. Cracking due to cyclic loading of non-piping penetrations (hatches, electrical penetrations, etc.}, dissimilar metal welds, and the fuel transfer tube expansion joints will be managed by supplemental surface examinations using the ASME Section XI, Subsection IWE AMP (XI.S1} and the 10 CFR 50, Appendix J AMP (XI.S4}.

Element Affected Enhancement

4.

Detection of Aging Effects (See response to RAI B.2.3.30-1, Attachment 8 to this letter).

The AMP will be enhanced to eerform sueelemental surface examinations to detect cracking due to c~clic loading of non-eieing eenetrations {hatches1 electrical eenetrations1 etc.}1 dissimilar metal welds 1 and fuel transfer tube exeansion joints.

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-2 L-2018-193 Attachment 11 Page 1 of 3 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 3.5.1.9-2

Background:

Section 4.6 of the SRP-SLR states that dissimilar metal welds are used to connect the piping penetrations to the bellows or stainless steel (SS) plates to provide a leak-tight penetration, and high energy piping penetrations and the fuel transfer tubes in some plants are equipped with SS bellow assemblies. The SRP-SLR also states that these components may be designed in accordance with the requirements of Section Ill of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), and if a plant's code of record requires a fatigue parameter evaluation (fatigue analysis or fatigue waiver), then this analysis may be a time limited aging analyses (TLAA) and must be evaluated in accordance with 10 CFR 54.21 (c)(1) for the subsequent period of extended operation.

GALL-SLR Report item II.A3.C-13, associated with SRP-SLR Table 3.5-1, item 009, addresses, in part, the cumulative fatigue damage due to fatigue for penetrations with dissimilar metal welds and penetration bellows when a current licensing basis (CLB) fatigue analysis exists. Likewise, SLR-GALL Report item II.A3.CP-37, associated with SRP-SLR Table 3.5-1, item 027, addresses, in part, cracking due to cyclic loading for penetrations with dissimilar metal welds and penetration bellows when a CLB fatigue analysis does not exist.

Subsequent license renewal application (SLRA) Section 3.5.2.2.1.6 states that stainless steel piping from high-temperature piping systems that penetrates the containment uses dissimilar metal welds between the flued head of the steel penetration assembly and the outside of the pipe. SLRA Table 3.5-1, item 3.5-1, 009, and the associated further evaluation in SLRA Section 3.5.2.2.1.5, addresses only the disposition of carbon steel penetration sleeves and containment liner plate. SLRA Table 3.5-1, item 3.5-1, 027, states that this item is not applicable.

During the in-office audit, the staff reviewed drawing no. 561 O-C-204, "Containment Structure Reactor Fuel Transfer Tube," and noted that the fuel transfer tube has two "expansion joints" as part of its design.

Issue:

Based on the information provided in the SLRA, it is not clear if (1) a fatigue analysis or fatigue waiver analysis exists for penetrations with dissimilar metal welds and for penetration bellows which may require an evaluation in accordance with 10 CFR 54.21 (c)(1 ), and (2) how the aging effect of cracking due to cyclic loading will be adequately managed, in accordance with 10 CFR 54.21 (a)(3), for these components during the subsequent period of extended operation.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-2 L-2018-193 Attachment 11 Page 2 of 3 The staff noted that the SLRA does not clearly state if a CLB fatigue analysis exists for the penetrations with dissimilar metal welds and the "expansion joints" described in the background section (above), or how these components where evaluated for cumulative fatigue damage due to fatigue. If a CLB fatigue analysis or fatigue waiver exists for these components, it is not clear how these analyses were dispositioned in SLRA Section 4.6, or why Table 3.5-1, item 3.5-1, 009, its associated Table 2 items, and Section 3.5.2.2.1.5 of the SLRA do not address these components to demonstrate that the associated aging effect will be adequately managed during the subsequent period of extended operation.

Likewise, if a CLB fatigue analysis or fatigue waiver does not exist for these components, it is not clear how cracking due to cyclic loading will be adequately managed for these components during the subsequent period of extended operation since SLRA Table 3.5-1, item 27, states that this item is not applicable.

Also, SLRA Section B.2.3.30 "ASME Section XI, Subsection IWE" AMP states: "PTN

[Turkey Point] has no pressure-retaining components subject to cyclic loading without CLB fatigue analysis.... " Further, there are no enhancements proposed in the SLRA AMP to perform recommended supplemental surface examination or other applicable technique capable of detecting fine cracking; and no Appendix J leak rate tests are credited.

Request:

1. Clarify if a fatigue analysis or fatigue waiver analysis exists for dissimilar the piping penetrations with dissimilar metal welds (including the welds) described in SLRA Section 3.5.2.2.1.6.

(

2. Clarify if a fatigue analysis or fatigue waiver exists for the expansion joints illustrated in drawing 561 O-C-204, "Containment Structure Reactor Fuel Transfer Tube,"

3. If a fatigue analysis or fatigue waiver exists for any on the components discussed above, address with supporting justification the disposition under 10 CFR 54.21 (c)(1) of each the fatigue analysis or fatigue waiver, and describe the following for each analyzed component:

the name of the transients considered in each analysis, the design cycle limits of each transient, the projected cycles to 80-years of operation for each transient, and the review of the calculated cumulative usage factor (CUF), if applicable.

Otherwise, pursuant to 10 CFR 54.21 (a)(3), if fatigue analysis or fatigue waiver does not exists, clarify how these components will be adequately managed for cracking due to cyclic loading during the subsequent period of extended operation

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.9-2 L-2018-193 Attachment 11 Page 3 of 3 (i.e. SLRA Table 3.5-1, item 3.5-1, 027, with GALL-SLR Report recommendation for supplemental surface examinations using AMP XI.S1, "ASME Section XI, Subsection IWE" or identifying and crediting appropriately justified Appendix J leak rate tests).

FPL Response:

Responses to the above numbered requests are as follows:

1. Turkey Point UFSAR Appendix 5B, Section B.2.1 provides a description of the fatigue analysis that was performed for the containment liner plate and penetrations.

However, based on a review of available documentation, FPL has been unable to locate the original fatigue analysis, or confirm if a fatigue waiver exists for dissimilar metal welds associated with piping penetrations.

2. Turkey Point UFSAR Appendix 5B, Section B.2.1 provides a description of the fatigue analysis that was performed for the containment liner plate and penetrations.

However, based on a review of available documentation, FPL has been unable to locate the original fatigue analysis, or confirm if a fatigue waiver exists for the fuel transfer tube expansion joints.

3. Considering the responses to items 1 and 2 above, the SLRA is revised to indicate cracking due to cyclic loading of dissimilar metal welds associated with piping penetrations and the fuel transfer tube expansion joints will be managed by supplemental surface examinations using the ASME Section XI, Subsection IWE AMP (XI.S1) and the 10 CFR 50, Appendix J AMP (XI.S4).

References:

None Associated SLRA Revisions:

(Refer to) Attachment 10 Associated SLRA Revisions Associated

Enclosures:

None

WITHHELD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.5-9 L-2018-193 Attachment 12 Page 3 of 3 Table 4.7.5-9.1 - Limiting Fracture Toughness Values Derived using NUREG/CR-4513, Rev. 1 Considering Turkey Point Pump Casing CF8 Material Chemistry J1c (in-lb/in2)

T mat (dimensionless)

Jmax(in-lb/in2)

[

]a,c,e

[

] a,c,e

[

] a,c,e Table 4.7.5-9.2 - Fracture Toughness Values Reported in WCAP-13045 for the Model 93 Pump Casing CF8 Material Chemistry and used in WCAP-15355 for Turkey Point J1c (in-lb/in2)

T mat (dimensionless)

Jmax(in-lb/in2)

[

]a,c,e

[

] a,c,e

[

] a,c,e Table 4.7.5-9.3-Limiting Fracture Toughness Values Derived using NUREG/CR-4513, Rev. 2 Considering Turkey Point Pump Casing CF8 Material Chemistry J1c (in-lb/in2)

T mat (dimensionless)

Jmax(in-lb/in2)

[

]a,c,e

[

] a,c,e

[

] a,c,e

References:

None Associated SLRA Revisions:

None Associated

Enclosures:

Westinghouse Letter CAW-18-4826 dated October 24, 2018, Application for Withholding Proprietary Information from Public Disclosure

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.5-9 L-2018-193 Attachment 12 Page 1 of 3 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

8. Reactor Coolant Pump Integrity Analysis, GALL TLAA 4.7 Regulatory Basis:

Section 54.21 (c) of 10 CFR requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis (CLB) for the subsequent period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to the managing the effects of aging during the subsequent period of extended operation (SPEO) on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the subsequent renewed license will continue to be conducted in accordance with the CLB. As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters

• described below.

Background:

The regulation in 10 CFR 54.21 (c)(1 )(ii) states that, for a specific time limited aging analyses (TLAA) that is dispositioned in accordance with this regulation, the applicant must demonstrate that the analysis has been projected to the end of the SPEO.

Subsequent license renewal application (SLRA) Section 4.7.5, "Code Case N-481 Reactor Coolant Pump Integrity Analysis," identifies the examination reactor coolant pump (RCP) casing in the current licensing basis as a TLAA item.

In 2000, the applicant submitted for NRC review and approval the 60-year license renewal application. As part of that application, the applicant performed a reactor coolcrnt pump (RCP) integrity analysis for Turkey Point Units 3 and 4 as documented in Westinghouse topical reports, WCAP-13045 and WCAP-15355. To demonstrate continued compliance during SPEO, the Pressurized Water Reactor Owner's Group (PWROG) re-evaluated WCAP-13045 associated with the application of Code Case N-481 to the RCP casing during the SPEO as documented in PWROG-17033, Revision 0.

The applicant submitted the topical report PWROG-17033, Revision Oas part of the SLRA.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.5-9 L-2018-193 Attachment 12 Page 2 of 3 RAI 4.7.5-9 Issue:

Section 2.2 of PWROG-17033 discusses fracture toughness calculation based on NUREG/CR4513, Revision 2. The NRC staff notes that Aging Management Program XI.M12, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel, in Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) report, NUREG-2191, Volume 2, discusses fracture toughness values based on the prediction method in NUREG/CR-4513, Revision 1. The GALLSLR report does not reference Revision 2 of NUREG/CR-4513.

Request:

Discuss whether the saturated fracture toughness value used in the crack stability analysis of pump casing at Turkey Point would still be limiting and bounding if the method of predicting fracture toughness in accordance with NUREG/CR-4513, Revision 1 was used. That is, regardless whether the method in revision 1 or revision 2 of NUREG/CR-4513 was used, the fracture toughness value used in the crack stability analysis at Turkey Point would still be limiting.

FPL Response:

Section 4.2 of WCAP-15355 provides the saturated fracture toughness values (see Table 4.7.5-9.1 below for convenience) that were derived using the methodology of NUREG/CR-4513, Revision 1, considering the limiting Turkey Point CF8 material chemistry. The NUREG/CR-4513, Revision 1 values in Table 4.7.5-9.1 can be used for comparison to the generic fracture toughness values used in the stability criteria per Table 5-2 of WCAP-15355 (provided in Table 4.7.5-9.2 below for convenience) for the Turkey Point pump casings. The fracture toughness values in Table 4.7.5-9.2 were reported in WCAP-13045 for the Model 93 pump casing considering CF8 material. Finally, Table 4.7.5-9.3 provides the fracture toughness values reported in the response to RAI 4.7.5-3(d) for the Model 93 pump casing using the methodology of NUREG/CR-4513, Revision 2, also considering the limiting Turkey Point CF8 material chemistry.

As shown by comparing Tables 4.7.5-9.1, 4.7.5-9.2, and 4.7.5-9.3 below, the saturated fracture toughness values derived using NUREG/CR-4513, Revision 1 (Table 4.7.5-9.1) and Revision 2 (Table 4.7.5-9.3) are higher than the generic CF8 material fracture toughness values reported in WCAP-13045 (Table 4.7.5-9.2) and used for Turkey Point.

Therefore, the fracture toughness values used in the crack stability analyses of the pump casing for Turkey Point, as reported in WCAP-15355, are limiting and more bounding than the fracture toughness values determined by either Revision 1 or Revision 2 of NUREG/CR-4513 for the Turkey Point CF8 material chemistry.

WITHHELD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.5-10 L-2018-193 Attachment 13 Page 1 of 2 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 4. 7.5-10 Issue:

The applicant stated that the J1c, Jmax, and T mat values in Table 1 of PWROG-17033-P are bounding and were used to demonstrate the crack stability of pump casing. However, the J1c, Jmax and T mat values used in Tables 11-2 and 11-3 of WCAP-13045 to demonstrate the crack stability of flaw 5-93 were higher than the J1c, Jmax, and T mat values listed in Table 1 of PWROG17033, respectively. If the lower J1c, Jmax and T mat values in Table 1 of PWROG-17033 were used to analyze flaw 5-93, crack stability may not be demonstrated for flaw 5-93. It appears that various fracture toughness value criteria are needed to qualify various flaws to demonstrate crack stability, not a single set of fracture toughness value as specified in Table 1 of PWROG17033-P. Table 5-1 of WCAP-13045 does provide 4 sets of fracture toughness values as end of service life criteria. Therefore, it seems that the fracture toughness values in Table 5-1 ofWCAP-13045 should be compared to the fracture toughness values predicted based on the method in NUREG/CR-4513, Revision 1..

Request:

Discuss whether the 4 sets of fracture toughness values in Table 5-1 of WCAP-13045 satisfy the fracture toughness values as predicted using the method in NUREG/CR-4513 Revision 1. If not please justify.

FPL Response:

The limiting values from WCAP-13045 [

]a,c,e, as reported in Table 1 of PWROG-17033 for the nozzle outer quarter flaw location, were used to generically demonstrate that the methodology in WCAP-13045 will produce more limiting fracture toughness values than those derived using the methodology in NUREG/CR-4513, Revision 2 for the same flaw location with the same chemistry. The aim of PWROG-17033 was to demonstrate that one of the most limiting CF8M material heats for the pump casing has reached the saturated condition based on the latest NRC approved NUREG/CR-4513 Revision 2, and therefore, the stability criteria for the fracture toughness values in WCAP-13045 are limiting.

For the Turkey Point subsequent license renewal application, the fracture toughness values that apply to Turkey Point were derived in WCAP-13045 and are shown in Table 4.7.5-9.2 in RAI response 4.7.5-9. The other three sets of values in Table 5-1 ofWCAP-13045 are for pump casings with CF8M material chemistry and therefore do not apply to Turkey Point. The fracture toughness values that apply to Turkey Point were qualified for crack stability in WCAP-15355. As discussed in the Westinghouse response to RAI 4. 7.5-

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.5-9 L-2018-193 Attachment 12 Enclosure Page 1 of 8 Enclosure Westinghouse Letter CAW-18-4826 dated October 24, 2018 Application for Withholding Proprietary Information from Public Disclosure Westinghouse Affidavit CAW-18-4826 Proprietary Information Notice and Copyright Notice L TR-SDA-18-101-P, Rev. 0, "Responses to the U.S. NRC Request for Additional Information on Pump Casings (PWROG-17033) Relative to the Turkey Point Subsequent License Renewal Application" (Proprietary)

Turkey Point Units 3 and 4 Docket.Nos. 50-250 arid 50-251 L-2018-193 Attachment 12 Enclosure Page 2 of 8 Westinghouse Non-Proprietary Class 3

@ Westinghouse U.S. Nuclear Regulatory Commission Document Control Desk 11555 Rockville Pike Rockville, MD 20852 Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Direct tel: (412) 374-3382 Direct fax: (724) 940-8542 e-mail: russpa@westinghouse.com CAW-18-4826 October 24, 2018 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

LTR-SDA-18-101-P, Rev. 0, "Responses to the U.S. NRC Request for Additional Information on Pump Casings (PWROG-17033) Relative to the Turkey Point Subsequent License Renewal Application" (Proprietary)

The Application for Withholding Proprietary Information from Public Disclosure is submitted by Westinghouse Electric Company LLC ("Westinghouse"), pursuant to the provisions of paragraph (b )(I) of Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations. It contains commercial strategic infonnation proprietary to Westinghouse and customarily held in confidence.

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CA W-18-4826 signed by the owner of the proprietary information, Westinghouse. The Affidavit, which accompanies this letter, sets forth the basis on which the infonnation may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b )( 4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Florida Power and Light Company.

• Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-18-4826, and should be addressed to Camille Zozula,. Manager. Facilities and Infrastructure Licensing, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 2 Suite 256, Cranberry Township, Pennsylvania 16066.

Paul A. Russ, Director Licensing and Regulatory Affairs

Enclosures:

1.

Affidavit CA W-18-4826

2.

Proprietary fuformation Notice and Copyright Notice

3.

L TR-SDA-18-101-P, Rev. 0, "Responses to the U.S. NRC Request for Additional Information on Pump Casings (PWROG-17033) Relative to the Turlcey Point Subsequent License Renewal Application" (Proprietary)

© 2018 Westinghouse Electric Company LLC. All Rights Reserved.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 COMMONWEALTH OF PENNSYLVANIA:

COUNTY OF BUTLER:

AFFIDAVIT ss L-2018-193 Attachment 12 Enclosure Page*3 of 8 CAW-18-4826 I, Paul A. Russ, am authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC ("Westinghouse") and declare that the avennents of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

Executed on: lO /;; f./ f J 75 I

Paul A. Russ, Director Licensing and Regulatory Affairs

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 3

L-2018-193 Attachment 12 Enclosure Page 4 of 8 CA W-18-4826 (1)

I aµi Director, Licensing and Regulatory Affairs, Westinghouse Electric Company LLC.

("Westinghouse"), and as such, I have been specifically delegated the function of reviewing the ptoprietary. information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2)

I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit (3)

I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4)

Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from_public disclosure should be withheld.

(i)

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii)

The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for detennining the types of information customarily held in confidence by it and, in that connection, utilizes a system to detennine when and whether to hold certain types of infonnati<:m in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a)

The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 4

L-2018-193 Attachment 12 Enclosure Page 5 of 8 CAW-18-4826 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b)

It consists of supporting data, including test data, relative to a process ( or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

( c)

Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

( d)

It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

( e)

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f)

It contains patentable ideas, for which patent protection may be desirable.

(iii)

There are sound policy reasons behind the Westinghouse system which include the following:

(a)

.The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b)

It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c)

Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 5

L-2018-193 Attachment 12 Enclosure Page 6 of 8 CAW-18-4826

( d)

Each component of proprietary information pertinent to a pa.i:ctcular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

( e) **

Unrestricted dlsclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those coW1tries.

(f)

The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iv)

The information is being transmitted to the Commission in confidence and. under the provisions of 10 CPR Section 2.390, is to be received in confidence by the Commission.

(v)

The information sought to be protected is not available in public sources or available*

information has not been previously employed in the same original manner or inethod to the best of our knowledge and belief.

(vi)

The proprietary information sought to be withheld in this submittal is that which is appropriately marked in LTR-SDA-18-101-P, Rev. 0, "Responses to the U.S. NRC Request for Additional Information on Pump Casings (PWROG-17033) Relative to the Turkey Point Subsequent License Renewal Application" (Proprietary), for submittal to the Commission, being transmitted by Florida Power & Light Company letter. The proprietary information as submitted by Westinghouse is that associated with Westinghouse's request for NRC approval of LTR-SDA-18-101-P, and may be used only for that purpose.

(a)

This information is part of that which will enable Westinghouse to provide a technical justification for acceptability of RCP components for Turkey Point Units 3 and 4 in support of their subsequent license renewal program.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 6

L-2018-193Attachment 12 Enclosure Page 7 of 8 CAW-18-4826 (b)

Further, this information has substantial commercial value as follows:

(i)

Westinghouse plans to sell the use of similar information to its customers for the purpose of supporting other subsequent license renewal programs.

(ii)

Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications.

(iii)

The information requested to be withheld reveals the distinguishing aspects of a methodology which was _developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial hann to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The developmen~ of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 PROPRIETARY INFORMATION NOTICE L-2018-193 Attachment 12 Enclosure Page 8 of 8 Transmitted herewith are proprietary and non-proprietary versions of a document. furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(l).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as thejssuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instance~ and the proprietary notice if the original was identified as proprietary.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRG RAI No. 4.7.5-10 L-2018-193 Attachment 13 Page 2 of 2 9, the saturated fracture toughness values derived using NUREG/CR-4513, Revision 1 _

and Revision 2, as reported in WCAP-15355 and RAI 4.7.5-3(d), respectively, and shown in Tables 4.7.5-9.1 and 4.7.5-9.3, are greater than the Turkey Point fracture toughness values shown in Table 4.7.5-9.2. Therefore, the fracture toughness values used in the crack stability analyses of the pump casing for Turkey Point, as reported in WCAP-15355, are limiting and more bounding than the fracture toughness values determined by either Revision 1 or Revision 2 of NUREG/CR-4513.

References:

None Associated SLRA Revisions:

None Associated

Enclosures:

(Refer to) Attachment 12 Enclosure

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-1 L-2018-193 Attachment 14 Page 1 of 6 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

9. Structures Monitoring Program, GALL AMP XI.S6 Regulatory Basis:

Section 54.21 (a)(3) of 10 CFR requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended, function(s) will be maintained consistent with the current licensing basis for the period of extended operation. As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report and when evaluation of the matter in the GALL-SLR Report applies to the plant.

RAI 8.2.3.35-1

Background:

The "detection of aging effects" program element of GALL-SLR Report AMP XI.S6, "Structures Monitoring," recommends inspectors to be qualified consistent with industry guidelines and standards and guidelines for implementing the requirements of 10 CFR 50.65. The GALL-SLR Report states that qualifications of inspection and evaluation personnel specified in.AC! 349.3R are acceptable for inspection of concrete structures.

The subsequent license renewal application (SLRA), Section B.2.3.35, "Structures Monitoring," states that inspections are performed and evaluated by qualified personnel.

The SLRA also states that the aging management program (AMP), with exception and enhancements, will be consistent with the 10 elements of NUREG-2191,Section XI.S6, "Structures Monitoring."

During the audit, the staff reviewed procedure O-ADM-561, "Structures Monitoring Program," and report no. FPLCORP020-REPT-107, "Aging Management Program Basis Document - Structures Monitoring," and noted that inspectors and lead reviewers are qualified in accordance with the Engineering Training Program (ETP) or ACI 349.3R, and by having proficiency in detecting structural deficiencies. The staff also noted that the Fleet ETP (i.e. TR-AA-110) and Guideline ACAD 98-004, "Guidelines for Training and Qualification of Engineering Personnel," does not specify qualification requirements for inspectors or lead reviewers performing inspections or evaluations under the Structures Monitoring Program.

Issue:

The Structures Monitoring Program allows an inspector or a lead reviewer to be qualified by only meeting the "Engineering Training Program" requirements, which does not provide specific personnel qualification requirements that is consistent with industry guidelines and standards (e.g. ACI 349.3R) for concrete. Therefore, it is not clear how the Turkey Point Structures Monitoring Program will ensure consistency with the GALL-SLR Report recommendation for inspector qualifications, or what criteria is used by Turkey Point to determine that an inspector or a lead reviewer is qualified. Additionally,

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-1 L-2018-193 Attachment 14 Page 2 of 6 it is not clear how "by having proficiency in detecting structural deficiencies," as stated in the program procedure, supports the qualification program or what criteria is used to establish and maintain proficiency.

Request:

Clarify how Turkey Point Structures Monitoring Program will be consistent with the GALL-SLR Report recommendation for inspectors to be qualified consistent with industry guidelines and standards (e.g. ACI 349.3R for concrete); otherwise, provide adequate justification if an exception is taken to the "detection of aging effects" program element as recommended by the GALL-SLR Report.

FPL Response:

The Turkey Point Structures Monitoring AMP inspector qualification requirements are consistent with the NUREG-2191,Section XI.S6 recommendations in that inspections are performed by personnel qualified to monitor structures and components for applicable aging effects, as described in SLRA Section B.2.3.35. Inspector qualifications are consistent with industry guidelines and standards and the guidelines for implementing the requirements of 10 CFR 50.65.

The individuals who view and examine the structures in the field has the same qualifications as a reviewer. Both the inspector and the reviewer are currently qualified in accordance with the FPL Engineering Training Program (ETP) or ACI 349.3R and have proficiency in detecting structural deficiencies. The judgment of the reviewer determines the acceptability of the results of the inspection. If appropriate, numerical criteria based on structural measurements are established in lieu of, or in combination with, this judgment. A reviewer is either a licensed Professional Engineer experienced in this area or is working under the direction of a licensed Professional Engineer experienced in this area. In addition, inspector proficiency is ascertained by a lead reviewer. A lead reviewer has expertise in the design and inspection of steel, concrete, and masonry structures and is a licensed Professional Engineer in this area. As described in SLRA Section 17.2.2.35, Quantitative acceptance criteria for concrete inspections are based on ACI 349.3R. Inspections and evaluations are performed using criteria derived from industry codes and standards contained in the plant CLB including but not limited to ACI 349.3R, ACI 318, SEI/ASCE 11 and the AISC specifications.

Therefore, the inspectors for the Structures Monitoring AMP are qualified to monitor structures and components consistent with industry guidelines and standards. To make this more clear, SLRA Table 17-3 (Item 39) and Section B.2.3.35 are revised as described below.

References:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-1 L-2018-193 Attachment 14 Page 3 of 6 Associated SLRA Revisions:

SLRA Table 17-3 (Item 39) and Section B.2.3.35 are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-1 L-2018-193 Attachment 14 Page 4 of 6 Revise the commitments for the Structures Monitoring AMP in Table 17-3, item 39, on page A-107 as follows:

Aging Management Program or Activity NUREG-2191 Implementation No.

(Section)

Section Commitment Schedule e) Revise inspection procedures to reference SEI/ASCE 11 and the American Institute of Steel Construction Material, and to clarify that insQector gualification will be Qer ACI 349.3R.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-1 L-2018-193 Attachment 14 Page 5 of 6 Revise the 3rct paragraph of Section B.2.3.35 on page B-255 as follows:

The PTN Structures Monitoring AMP consists primarily of periodic visual inspections by personnel qualified to monitor structures and components for applicable aging effects from degradation mechanisms, such as those described in the ACI 349.3R-02 (Reference B.3.118), "Evaluation of Existing Nuclear Safety-Related Concrete Structures," ACI 201.1 R-08 (Reference B.3.116), "Guide for Conducting a Visual Inspection of Concrete Service," and Structural Engineering Institute/American Society of Civil Engineers Standard (SEI/ASCE) 11-99 (Reference B.3.121 ), "Guideline for Structural Condition Assessment of Existing Buildings." Identified aging effects are evaluated by qualified personnel using criteria derived from industry codes and standards contained in the plant CLB including but not limited to ACI 349.3R, ACI 318 (Reference B.3.117), SEI/ASCE 11, and the American Institute of Steel Construction (AISC) specifications, as applicable. For the SPEC, inspection and evaluation personnel will be qualified per ACI 349.3R.

Revise the pertinent enhancement in SLRA Section B.2.3.35 on page B-257 as follows:

Enhancements The PTN Structures Monitoring AMP will be enhanced as follows, for alignment with NUREG-2191. The changes and enhancements are to be implemented no later than six months prior to entering the SPEO.

Element Affected Enhancement

4. Detection of Aging Effects Update the governing AMP procedure to clarifv that inspector qualification, including proficienc~ criteria and proficienc~ maintenance, be per ACI 349.3R.

Update the governing AMP procedure with a site-specific enhancement that may include evaluations, destructive testing, and/or focused inspections of representative accessible (leading indicator) or below-grade, inaccessible concrete structural elements exposed to aggressive groundwater/soil. The respective evaluation/inspection/testing interval is not to exceed 5 years.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-1 L-2018-193 Attachment 14 Page 6 of 6 Element Affected Associated

Enclosures:

None Enhancement '

Update the governing AMP procedure with guidance on monitoring for indications of cracking and expansion due to reaction with aggregates in concrete structures.

Update the governing AMP procedure to clarify that tactile inspection may be needed for detection of elastomer hardening.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 8.2.3.35-2 L-2018-193 Attachment 15 Page 1 of 4 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 8.2.3.35-2

Background:

The "parameters monitored or inspected" and "detection of aging effects" program elements of GALL-SLR Report AMP XI.S6, "Structures Monitoring," recommends monitoring and trending leakage volumes and chemistry for signs of concrete or steel reinforcement degradation if through-wall leakage or groundwater infiltration is identified. The GALL-SLR Report also recommends, in part, assessing the indication thru engineering evaluation, more frequent inspections, or destructive testing of affected concrete to validate existing concrete properties. Additionally, it recommends to include analysis of the leakage pH, along with mineral, chloride, sulfate and iron content in the water when leakage volumes allow such analyses.

The subsequent license renewal application (SLRA), Section B.2.3.35, "Structures Monitoring," states that structures are monitored to confirm the absence of water in-leakage or signs of concrete leaching, chemical attack or steel reinforcement degradation. The SLRA also states that the aging management program (AMP), with exception and enhancements, will be consistent with the 10 elements of NUREG-2191,Section XI.S6, "Structures Monitoring."

Issue:

During the audit, the staff reviewed procedure O-ADM-561, "Structures Monitoring Program," and Report No. FPLCORP020-REPT-107, "Aging Management Program Basis Document-Structures Monitoring," and was not able to verify consistency with the "parameters monitored or inspected" and "detection of aging effects" program elements of the GALL-SLR Report because the AMP (1) does not provide requirements to monitor and trend leakage volumes and chemistry for signs of concrete or steel reinforcement degradation when through-concrete leakage is identified, and (2) does not clearly identify how indications of groundwater infiltration or through-concrete leakage will be assessed for aging effects.

The staff notes that the program currently monitors structures elements to confirm the absence of water in-leakage. However, no AMP enhancement was provided in the SLRA to include the monitoring, trending and assessment of aging effects if through-concrete leakage is identified, to be consistent with the GALL-SLR Report recommendations.

Request:

Clarify how Turkey Point Structures Monitoring Program will be consistent with the "parameters monitored or inspected," and "detection of aging effects" program elements from the GALL-SLR Report, with respect to through-concrete leakage. Otherwise, provide adequate justification if an exception is taken to the GALL-SLR Report recommendations.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-2 L-2018-193 Attachment 15 Page 2 of 4 FPL Response:

Structures are monitored to confirm the absence of water in-leakage. Structures are acceptable without further evaluation if the absence of ground water in-leakage is confirmed for concrete surfaces. Observed concrete surface conditions with evidence of degradation or that are found to be detrimental to the structural or functional integrity are considered unacceptable and in need of further technical evaluation. This further technical evaluation is performed through the corrective action program, if needed. As such, the Turkey Point Structures Monitoring AMP is consistent with the 'parameters monitored or inspected' and 'detection of aging effects' elements of NUREG-2191, XI.S6, as described in SLRA Section B.2.3.35.

As described in the exception in Section B.2.3.35, the groundwater/soil at PTN is aggressive (chlorides> 500 ppm), and periodic sampling and testing is not necessary.

As such, inclusion of information supporting the 'further technical evaluation' in the AMP governing procedure is warranted for the SPEO. This includes monitoring leakage volumes and chemistry if through-wall leakage or groundwater infiltration is identified, as well as analysis of that leakage for pH, mineral, chloride, sulfate or iron content if possible. Should groundwater infiltration be identified, engineering evaluation, more frequent inspections, or destructive testing, to validate existing properties of the concrete, and analysis of the leakage may be necessary. To that end, and for closer consistency with NUREG-2191, pertinent SLRA sections are revised.

References:

None Associated SLRA Revisions:

SLRA Table 17-3, Item 39, and Section B.2.3.35 are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-2 L-2018-193 Attachment 15 Page 3 of 4 Revise the commitments for the Structures Monitoring AMP in Table 17-3, item 39, on page A-107 as follows:

Aging Management Program or Activity NUREG-2191 No.

(Section)

Section Commitment Implementation Schedule g) Revise inspection procedures to include guidance on monitoring for indications of cracking and expansion due to reaction with aggregates in concrete structures.

h} URdate insRection Rrocedure{s} to include monitoring volumes and chemistrL1 more freguent insRections or destructive testing 1 and anal~zing concrete ~H 1 along with ~H and content of the leakage water when ~ossible 1 IF through-wall groundwater in-leakage is identified.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 8.2.3.35-2 L-2018-193 Attachment 15 Page 4 of 4 Revise the pertinent enhancement in SLRA Section B.2.3.35 on page B-257 as follows:

The PTN Structures Monitoring AMP will be enhanced as follows, for alignment with NUREG-2191. The changes and enhancements are to be implemented no later than six months prior to entering the SPEO.

Element Affected Enhancement

4. Detection of Aging Effects Update the governing AMP procedure with a site-specific enhancement that may include evaluations, destructive testing, and/or focused inspections of representative accessible (leading indicator) or below-grade, inaccessible concrete structural elements exposed to aggressive groundwater/soil. The respective evaluation/inspection/testing interval is not to exceed 5 years.

Update the governing AMP procedure with guidance on monitoring for indications of cracking and expansion due to reaction with aggregates in concrete structures.

Update the governing AMP procedure to clarify that tactile inspection may be needed for detection of elastomer hardening.

U~date the governing AMP ~rocedure to clarifv that engineering evaluation 1 more freguent ins~ections 1 or destructive testing of affected concrete

{to validate ~ro~erties and determine

~H} are reguired 1 along with anal)lsis of the leakage water when ~ossible 1 IF groundwater leakage is identified.

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-3 L-2018-193 Attachment 16 Page 1 of 10 NRC RAI Letter Nos; ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 8.2.3.35-3

Background:

The "detection of aging effects" program element of GALL-SLR Report AMP XI.S6, "Structures Monitoring," recommends that a plant-specific aging management program (AMP) accounting for the extent of the degradation experienced should be implemented to manage the concrete aging during the subsequent period of extended operation if the plant has an aggressive groundwater/soil environment. The GALL-SLR Report provides examples of what actions may be included as part of the plant-specific AMP. The SRP-SLR Appendix A provides the staff positions and guidance for a plant-specific AMP.

The subsequent license renewal application (SLRA), Section B.2.3.35, "Structures Monitoring," states that groundwater/soil at Turkey Point is aggressive (chlorides> 500 ppm), and that the AMP, with exception and enhancements, will be consistent with the 10 elements of NUREG-2191,Section XI.S6, "Structures Monitoring." The SLRA provides an enhancement to address aggressive groundwater/soil that may include evaluations, destructive testing, and/or focused inspections of representative accessible (leading indicator) or below-grade, inaccessible concrete structural elements exposed to aggressive groundwater/soil. The SLRA enhancement also states that the respective evaluation, inspection and testing interval is not to exceed 5 years.

During the on-site audit the staff noted several plant-specific operating experience items related to corrosion degradation in accessible areas of concrete structures exposed to air-outdoor environment. These degradations were attributed to the significant chloride level present at the site, which is the same aging effect mechanism expected from an aggressive groundwater/soil environment.

Issue:

The staff was not able to verify consistency with the "detection of aging effects" program element of the GALL-SLR Report since the enhancement provided in the SLRA restates the general examples provided in the GALL-SLR Report for a plant-specific AMP, and does not provide an adequate plant-specific AMP description or enhancements to the different program elements in accordance with SRP-SLR Appendix A, Section A.1, to ensure that structures and components exposed to an aggressive groundwater/soil environment will be adequately managed as required by 10 CFR 54.21 (a)(3). Staff review of the SLRA AMP program elements did not identify how the applicant plans to address the aging effects of structures and components exposed to an aggressive groundwater/soil environment using the focused inspections, evaluations, and/or destructive testing suggested by GALL-SLR Report, and/or using other acceptable method(s). Also, it is not clear how the plant-specific operating experience associated with corrosion from accessible areas of the structures were considered in the

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-3 L-2018-193 Attachment 16 Page 2 of 10 implementation of the plant-specific AMP to ensure that inaccessible areas exposed to aggressive groundwater/soil environment are adequately managed. The staff notes that the aging effect mechanism present in accessible areas of concrete structures exposed to an air-outdoor environment is the same as in the inaccessible areas of the structures exposed to an aggressive ground/soil environment (i.e. significant chloride levels).

Request:

Provide the Turkey Point plant-specific AMP description or enhancements for each of the program elements in the Structures Monitoring Program (as applicable) to demonstrate that structures and components exposed to an aggressive groundwater/soil environment will be adequately managed for the subsequent period of extended operation. The proposed program or enhancements should account for any plant-specific OE with aggressive groundwater, and the on-going corrosion degradation observed in accessible areas of the structures due to the presence of chloride.

FPL Response:

As stated in SLRA Section B.2.3.35, from comparison with the chloride level for seawater, the groundwater/soil at PTN is considered as aggressive (chlorides> 500 ppm). Since the chloride levels for seawater are much greater than 500 ppm, there is reasonable certainty that any groundwater/soil chemistry tests will consistently result in chloride level readings that are greater than 500 ppm which indicates an aggressive groundwater/soil classification, and periodic sampling and testing is not necessary. Therefore, PTN is required to account for the extent of degradation experienced due the aggressive groundwater/soil aging effects. The PTN Structures Monitoring AMP contains a site-specific enhancement to manage the concrete aging during the SPEO rather than implementing a site~specific AMP. For clarity, the existing enhancement to the detection of aging effects element will be replaced with a site-specific enhancement to the pertinent elements (scope of program, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria). The site-specific enhancement includes the following:

1. A baseline inspection of inaccessible concrete will be conducted prior to the SPEO.

a) The baseline inspection locations will consider site-specific OE. OE considered will include known degradation due to chlorides in ambient air and the potential for further degradation due to the aggressive groundwater.

b) The baseline inspection will include excavation, visual inspection, pH analysis, and a chloride concentration test of inaccessible concrete at a location close to the coastline/intake and a location in the main plant area for comparison.

2. A baseline evaluation will be performed prior to the SPEO.

a) The baseline evaluation will consider the baseline inspection results to

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-3 L-2018-193 Attachment 16 Page 3 of 10 determine the additional actions (if any) that are warranted. Additional actions may include: enhanced inspection techniques and/or frequency, destructive testing, and focused inspections of representative accessible concrete (leading indicator) or below grade, inaccessible concrete structural elements exposed to aggressive groundwater/soil.

b) The baseline inspection and evaluation results will set the subsequent inspection requirements and inspection intervals (not to exceed 5 years) for the SPEO.

3. Periodic inspections at a frequency determined in the baseline evaluation (not to exceed 5 years) will be performed, either focused or opportunistic when locations are excavated for other reasons.

4. Periodic evaluation updates will be performed (not to exceed 5 years).

a) Updates will be based on OE, periodic inspections, and b) will consider the opportunistic or focused inspection results during the interval. The periodic evaluation results will update subsequent inspection requirements and inspection intervals (not to exceed 5 years) for the SPEO as required.

Accessible areas of in-scope concrete structures are inspected through the Structures Monitoring AMP for aging affects related to aggressive chemical attack such as loss of material (spalling, scaling), cracking, and other irregularities (increase in porosity and permeability. Issues related to accessible areas of concrete are entered into the corrective action program. Pertinent SLRA sections are revised to reflect the Structures Monitoring AMP site-specific enhancement.

References:

None Associated SLRA Revisions:

SLRA Section 17.2.2.35, Table 17-3 Item 39, and Section B.2.3.35 Structures Monitoring AMP are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise the Appendix A Section 17.2.2.35 on page A-37 as follows:

The PTN Structures Monitoring AMP is an existing condition monitoring program that consists primarily of periodic visual inspections of plant SCs for evidence of deterioration or degradation, such as described in the American Concrete Institute (ACI) Standards 349.3R, ACI 201.1 R, and Structural Engineering Institute/American Society of Civil Engineers Standard (SEI/ASCE) 11. Quantitative acceptance criteria for concrete inspections are based on ACI 349.3R. Inspections and evaluations are performed using

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-3 L-2018-193 Attachment 16 Page 4 of 10 criteria derived from industry codes and standards contained in the plant CLB including but not limited to ACI 349.3R, ACI 318, SEI/ASCE 11, and the American Institute of Steel Construction (AISC) specifications. The AMP includes preventive actions to ensure structural bolting integrity. Results from periodic inspections are trended. Due the presence of aggressive groundwater chemistry (Chlorides> 500 parts per million (ppm)),

the AMP includes site specific evaluations, destructive testing, if warranted, and/or focused inspections of representative accessible (leading indicator) or below grade, inaccessible concrete structural elements exposed to aggressive ground'l1mter/soil, on an inter1al not to exceed five years. the AMP includes a site-specific enhancement to conduct a baseline visual inspection, pH analysis, a chloride concentration test, and evaluation to address the degradation of concrete due to exposure of aggressive chemical attack. The baseline evaluation will consider site-specific OE and the baseline inspection results and will determine the additional actions that are warranted. Periodic inspections (either focused or opportunistic) and evaluation updates (not to exceed 5 years) will be performed throughout the SPEC to ensure aging of inaccessible concrete is adequately managed.

Revise the Structures Monitoring "Program Description" in Section B.2.3.35 on page B-256 as follows:

A dewatering system is not used or part of the CLB for PTN. Structures are monitored to confirm the absence of water in-leakage or signs of concrete leaching, chemical attack or steel reinforcement degradation. Due to the presence of high chloride levels in the groundwater a site-specific enhancement to manage the concrete aging during SPEO will include evaluations, destructive testing, and/or focused inspections of representative accessible (leading indicator) or below grade, inaccessible concrete structural elements exposed to aggressive groundwater/soil, on an inter1al not to exceed 5 years. include a baseline visual inspection, pH analysis, and a chloride concentration test prior to the SPEC. The inspection will include a location close to the coastline/intake and a location in the main plant area for comparison and consider site-specific OE. The baseline inspection results will be used to conduct a baseline evaluation that will determine the additional actions that are warranted. Additionally, the baseline evaluation results will set the subsequent inspection requirements and inspection intervals (not to exceed 5 years). Periodic inspections (either focused or opportunistic) and evaluation updates (not to exceed 5 years) will be performed throughout the SPEC to ensure aging of inaccessible concrete is adequately managed.

Revise the Exceptions to NUREG-2191 in Section B.2.3.35 on page B-256 and B-257 as follows:

The groundwater/soil at PTN is aggressive (chlorides> 500 ppm). Since the chloride levels for seawater are much greater than 500 ppm, there is reasonable certainty that any

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-3 L-2018-193 Attachment 16 Page 5 of 10 groundwater/soil chemistry tests will consistently result in chloride level readings that are greater than 500 ppm which indicates an aggressive groundwater/soil classification, and periodic sampling and testing is not necessary and of little value. Rather, the PTN Structures Monitoring AMP includes a site-specific enhancement to address aggressive groundwater soil:., that may include evaluations, destructive testing if 1.varranted, and/or focused inspections of representative accessible (leading indicator) or below grade, inaccessible concrete structural elements exposed to aggressive ground1.vater/soil, based on site OE but not to exceed 5 year intervals. The site-specific enhancement includes the following:

1. A baseline inspection of inaccessible concrete will be conducted prior to the SPEC.

a) The baseline inspection locations will consider site-specific OE. OE considered will include known degradation due to chlorides in ambient air and the potential for further degradation due to the aggressive groundwater.

b) The baseline inspection will include excavation, visual inspection, pH analysis, and a chloride concentration test of inaccessible concrete at a location close to the coastline/intake and a location in the main plant area for comparison.

2. A baseline evaluation will be performed prior to the SPEC.

a) The baseline evaluation will consider the baseline inspection results to determine the additional actions (if any) that are warranted. Additional actions may include: enhanced inspection techniques and/or frequency, destructive testing, and focused inspections of representative accessible concrete (leading indicator) or below grade, inaccessible concrete structural elements exposed to aggressive groundwater/soil.

b) The baseline inspection and evaluation results will set the subsequent inspection requirements and inspection intervals (not to exceed 5 years} for the SPEC.

3. Periodic inspections at a frequency determined in the baseline evaluation (not to exceed 5 years) will be performed, either focused or opportunistic when locations are excavated for other reasons.

4. Periodic evaluation updates will be performed (not to exceed 5 years) throughout the SPEC.

a) Updates will be based on OE, periodic inspections, and b) will consider opportunistic or focused inspection results during the

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-3 L-2018-193 Attachment 16 Page 6 of 10 interval. The periodic evaluation results will update subsequent inspection requirements and inspection intervals (not to exceed 5 years) for the SPEO as required.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-3 L-2018-193 Attachment 16 Page 7 of 10 Revise the Enhancements in Section B.2.3.35 on page B-258 as follows:

Element Affected Enhancement

3. Detection of Aging Effects 6J13Elate U1e §evemiR§ AMl=l 13rnseEIHFe witl=1 a site s13esifis eRRaRsemeRt tt:lat may iRslHEle evalHatieRs, ElestFHstive testiR§,

aRElleF fesHseEI iRs13estieRs ef Fe13FeseRtati>.ie assessisle ~leaEliR§ iRElieate11 eF 13elmN §raEle, iAaeeessisle eeRsFete strnetHFal elemeRts e*13eseEI te a§§Fessive §FeHAEl1NateF,lseil. +Re Fes13estive evalHatieAliAs13estieAI testiA§ iRteNal is Ret te e*seeEI 5 yeaFs.

Update the governing AMP procedure with guidance on monitoring for indications of cracking and expansion due to reaction with aggregates in concrete structures.

Update the governing AMP procedure to clarify that tactile inspection may be needed for detection of elastomer hardening.

A new implementing procedure, or new attachment to the AMP governing procedure, for management of concrete exposure to aggressive groundwater/soil will also be developed that addresses:

Element Affected Enhancement Scope Inaccessible concrete/foundations exposed to groundwater/soil in scope.

3.

Parameters Monitored or Monitoring of the condition of Inspected inaccessible concrete, including pH and chloride concentration, of concrete exposed to groundwater/soil.

Detection of Aging Effects Guidance on baseline excavation with visual inspection, pH anal)lsis, and a chloride concentration test of a location near the coastline and a location in the main plant area for comparison prior to the SPEC. Include periodic inspections

{either focused or opportunistic} at a

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-3 L-2018-193 Attachment 16 Page 8 of 10 Element Affected Monitoring and Trending Acceptance Criteria Enhancement freguencl£ determined in the baseline evaluation {not to exceed 5 )£ears}.

Guidance for the evaluation of the baseline inspection results and related OE1 with concrete exposed to ambient air and to groundwater/soil prior to the SPEC to determine subseguent inspection/evaluation reguirements and intervals {not to exceed 5 l£ears}1 with periodic updates based on periodic inspections {either focused or opportunistic} and OE.

Acceptance criteria for inaccessible concrete exposed to groundwater through baseline inspection and evaluation 1 with periodic updates based on periodic inspections {either focused or opportunistic} and OE.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-3 L-2018-193 Attachment 16 Page 9 of 10 Revise the Table 17-3, Item 39 on page A-107 as follows:

Table 17-3 List of SLR Commitments and Implementation Schedule (Continued)

No.

Aging Management Program NUREG-2191 Commitment or Activity (Section)

Section 39 Structures Monitoring XI.S6 f) PeFfeFm e1.iah:1atieRs, ElestF1::1eti1.ie testiR§, aRElleF fes1:1seEI (17.2.2.35) iRs13estieRs ef Fe13FeseRtatii.ie assessil31e (leaEliR§ iRElisateF) eF 13elev.* §FaEle, iRassessil31e seRsFete strnst1:1Fal elements e*13eseEI te a§§Fessii.ie §IFel:lnElwateF.lseil. +Ae Fes13estive eval1:1atieR.lins13estien,l testiR§ iRteFVal is Ret te e*seeEI a yeaFs. Develoe a new imelementing erocedure or attachment to an existing imelementing erocedure to address aging management of inaccessible areas exeosed to groundwater/soil. The document will include guidance to conduct a baseline visual inseection1 QH analisis 1 and a chloride concentration test erior to the SPEO at a location close to the coastline/intake and a location in the main Qlant area for comearison. The baseline inseection results will be used to conduct a baseline evaluation that will determine the additional actions that are warranted. Additionallll1 the baseline evaluation results will set the subseguent inseection reguirements and inseection intervals {not to exceed 5 iears}. Periodic insQections {either focused or OQQOrtunistic} and evaluation uedates {not to exceed 5 llears} will be eerformed throughout the SPEO to ensure aging of inaccessible concrete is adeguatelll managed.

Implementation Schedule

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRG RAI No. B.2.3.35-3 L-2018-193 Attachment 16 Page 10 of 10 Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-4 L-2018-193 Attachment 17 Page 1 of 3 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 8.2.3.35-4

Background:

The "scope of program," "parameters monitored or inspected," and "acceptance criteria" program elements of GALL-SLR Report AMP XI.S6, "Structures Monitoring,"

recommends that sliding surfaces within the scope of Structures Monitoring Program be monitored for indication of significant loss of material due to wear or corrosion, and for accumulation of debris or dirt. The GALL-SLR Report also states that identified conditions in sliding surfaces are acceptable when there are (a) no indications of excessive loss of material due to corrosion or wear and (b) no debris or dirt that could restrict or prevent sliding of the surfaces as required by design.

The "program description" in subsequent license renewal application (SLRA), Section B.2.3.35, "Structures Monitoring," states that the program inspects accessible sliding surfaces for indication of significant loss of material due to wear or corrosion, and for accumulation of debris or dirt. The SLRA does not include an enhancement associated with sliding surfaces, and does not have an aging management review line item associated with sliding surfaces being managed by the Structures Monitoring Program.

Issue:

The staff was unable to verify consistency between O-ADM-561, Revision 5 (Attachment 2), "Turkey Point Plant Structures Monitoring Program," and the GALL-SLR Report AMP XI.S6 program elements associated with sliding surfaces. Also, it is not clear which sliding surface components that are subject to aging management review will be managed by the Structures Monitoring Program, to be consistent with the "Program Description" Section of SLRA, Section B.2.3.35, "Structures Monitoring."

The staff notes that the "Program Description" in SLRA Section B.2.3.35 states that the program manages the effects of aging in accessible sliding surfaces, however, no parameters or acceptance criteria is provided in the Structures Monitoring Program procedure to effectively manage the effects of aging in sliding surfaces. The staff also notes that no enhancement was provided in the SLRA to ensure that the effects of aging for these components will be adequately managed by the Structures Monitoring Program during the subsequent period of extended operations.

Request:

Clarify if there are sliding surfaces components that are within the scope of subsequent license renewal and subject to aging management review that will be managed by the Structures Monitoring Program, as described in the "Program Description" section of SLRA, Section B.2.3.35, "Structures Monitoring."

For those sliding surfaces components managed using the Structures Monitoring Program, clarify how Turkey Point Structures Monitoring Program will be consistent with

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRG RAI No. B.2.3.35-4 L-2018-193 Attachment 17 Page 2 of 3 the "scope of program," "parameters monitored or inspected," and "acceptance criteria" program elements of GALL-SLR Report AMP XI.S6 by addressing the issue described above. Otherwise, provide adequate justification if an exception is taken to the GALL-SLR Report recommendations.

FPL Response:

As described in SLRA Table 3.5-1, item 074, there are no sliding support bearing or sliding support surfaces outside of containment in the Structures and Component Supports group that require aging management. Sliding surfaces located inside containment are managed by the ASME Section XI Subsection IWF program (Ref. SLRA Table 3.5-1, item 075, and Table 3.5.2-1). Since there are no sliding surfaces outside of containment that require aging management at PTN, the "scope of program," "parameters monitored or inspected," and "acceptance criteria" program elements of GALL SLR Report AMP XI.S6 are not applicable to the PTN Structures Monitoring AMP. To make this more clear, pertinent SLRA sections are revised as described below.

References:

None Associated SLRA Revisions:

SLRA Section B.2.3.32 ASME Section XI, Subsection IWF and Section B.2.3.35 Structures Monitoring are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise the Program Description in Section B.2.3.32 on page B-241 as follows:

The ASME Section XI, Subsection IWF AMP is an existing condition monitoring program that consists of periodic visual examination of ASME Code Section XI Class 1, 2, and 3 supports for ASME piping and components for signs of degradation such as corrosion; cracking, deformation; misalignment of supports; missing, detached, or loosened support items; loss of integrity of welds; improper clearances of guides and stops; and improper hot or cold settings of spring supports and constant load supports. Bolting for Class 1, 2, and 3, piping and component supports is also included and inspected for corrosion, loss of integrity of bolted connections due to self-loosening, and material conditions that can affect structural integrity. Associated sliding surfaces are inspected for indication of loss of material due to wear or corrosion, and for accumulation of debris or dirt.,

aoo Associated vibration isolation elements are also inspected for loss of material or mechanical or isolation function due to hardening.

Revise the Program Description in Section B.2.3.35 on page B-255 as follows:

Concrete surfaces are inspected for cracking, scaling, spalling, pitting, erosion, corrosion of reinforcing bars, settlement, deformation, leaching, discoloration, groundwater leakage, rust stains, exposed rebar, rust bleeding, and other surface irregularities. Structural steel

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.35-4 L-2018-193 Attachment 17 Page 3 of 3 is inspected for loss of material (corrosion), deflection, and distortion. Bracing connections associated with masonry walls are also inspected for degradation. The PTN Structures Monitoring AMP inspects accessible sliding surfaces for indication of significant loss of material due to 'Near or corrosion, and for accumulation of debris or dirt. Loose, missing, or damaged anchor bolts are visually inspected. High-strength bolts are visually inspected. Each structure's foundation is monitored for overall settlement and differential settlement. Structures are monitored to confirm the absence of water in-leakage or signs of concrete leaching, chemical attack or steel reinforcement degradation. Elastomers will be inspected for signs of hardening. PTN has no sliding surfaces outside of containment that require aging management.

Revise the Site-Specific Operating Experience in Section B.2.3.35 on page B-259 as follows:

Structures and piping/component support material condition inspections have been performed at PTN since the mid-1980s. The inspection requirements in support of the Maintenance Rule have been in effect since 1996, and have proven effective at maintaining structure and structural component material conditions. Unsatisfactory conditions are detected and resolved through the CAP prior to a loss of intended function.

There have been no instances of degraded sliding surfaces for components in the scope of SLR. While the PTN Structures Monitoring AMP inspects for settlement and differential settlement as described above, there have been no occurrence of settlement at PTN, as described in Sections 3.5.2.2.1.1 and 3.5.2.2.2.1. Elastomer seal degradation has been identified and corrected through visual inspections. The PTN Structures Monitoring AMP will be enhanced to include tactile inspection of elastomeric materials for detection of hardening.

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.100-1 L-2018-193 Attachment 18 Page 1 of 7 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 3.5.1.100-1

Background:

SRP-SLR Table 3.5-1, item 100, recommends that aluminum and stainless steel support members, welds, bolted connections and support anchorage to building structures be managed for loss of material and cracking due to stress corrosion cracking (SCC) by either the AMP XI.M32, "One-Time Inspection;" AMP XI.S6, "Structures Monitoring;" or AMP XI.M36, "External Surfaces Monitoring of Mechanical Components" program. This Table 1 line item is associated with a further evaluation, SRP-SLR Section 3.5.2.2.2.4, which states the acceptance criteria for the review and the recommended actions (including AMP enhancements) when loss of material or cracking has occurred and is sufficient to potentially affect the intended function of these components.

Item 3.5-1, 099 in the subsequent license renewal application (SLRA) Table 3.5-1 was not used by the applicant and states that item 3.5-1, 100 will be use instead to manage loss of material for aluminum and stainless steel support members, welds, bolted connections, and support anchorages using the Structures Monitoring Program. SLRA Table 3.5-1, item 100, associated with SLRA Section 3.5.2.2.2.4, states that this item is consistent with the GALL-SLR Report recommendation. SLRA Section 3.5.2.2.2.4, states that cracking due to SCC and loss of material due to pitting and crevice corrosion are applicable aging effects for these components, and that they will be monitored by the Structures Monitoring Program.

Issue:

The staff notes that line items for some components only address the aging effect of loss of material, and do not address the aging effect of cracking due to SCC as recommended by the GALL-SLR Report. Several items/components in SLRA Tables 3.5.2-X associated with Table 1 Item 3.5-1, 100, are not being managed for cracking due to SCC, consistent with the GALL-SLR Report recommendation. The staff also notes that no technical justification was provided in the application for not requiring management of the aging effect of cracking due to SCC for these components. For those components that will be managed for cracking due to SCC, it is not clear how the proposed AMP was enhanced to adequately manage the aging effect of SCC, as recommended in SRP-SLR Section 3.5.2.2.2.4.

Request:

Clarify why some components in SLRA Tables 3.5.2-X that are associated with Table 1 Item 3.5-1, 100, are not being managed for cracking due to SCC as recommended by the GALL-SLR Report. Otherwise if an exception is taken to the GALL~SLR Report recommendations, provide adequate technical justification for not requiring management of this aging effect.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.100-1 L-2018-193 Attachment 18 Page 2 of 7 For those components that will be managed for SCC, describe how the credited AMP will be capable to adequately manage cracking due to SCC.

FPL Response:

SRP-SLR Table 3.5-1, item 100, recommends that aluminum and stainless steel support members, welds, bolted connections and support anchorage to building structures be managed for loss of material and cracking due to stress corrosion cracking (SCC) by either the AMP XI.M32, "One-Time Inspection;" AMP XI.S6, "Structures Monitoring;" or AMP XI.M36, "External Surfaces Monitoring of Mechanical Components" program. This Table 1 line item is associated with a further evaluation, SRP-SLR Section 3.5.2.2.2.4, which states the acceptance criteria for the review and the recommended actions (including AMP enhancements) when loss of material or cracking has occurred and is sufficient to potentially affect the intended function of these components.

Item 3.5-1, 099 in the subsequent license renewal application (SLRA) Table 3.5-1 was not used by the applicant and states that item 3.5-1, 100 will be used instead to manage loss of material for aluminum and stainless steel support members, welds, bolted connections, and support anchorages using the Structures Monitoring Program. SLRA Table 3.5-1, item 100, associated with SLRA Section 3.5.2.2.2.4, states that this item is consistent with the GALL-SLR Report recommendation. SLRA Section 3.5.2.2.2.4, states that cracking due to SCC and loss of material due to pitting and crevice corrosion are applicable aging effects for these components, and that they will be monitored by the Structures Monitoring Program.

References:

None Associated SLRA Revisions:

SLRA Section B.2.3.35 Structures Monitoring, Table 17-3, Item 35 and the Section 3.5 Aging Management Evaluation tables are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise the "Program Description" in Section B.2.3.35 on page B-255 as follows:

Concrete surfaces are inspected for cracking, scaling, spalling, pitting, erosion, corrosion of reinforcing bars, settlement, deformation, leaching, discoloration, groundwater leakage, rust stains, exposed rebar, rust bleeding, and other surface irregularities. Structural steel is inspected for loss of material (corrosion), deflection, and distortion. Bracing connections associated with masonry walls are also inspected for degradation. The PTN Structures Monitoring AMP inspects accessible sliding surfaces for indication of significant loss of material due to wear or corrosion, and for accumulation of debris or dirt. Additionally, stainless steel and aluminum components are inspected for cracking due to SCC.

Loose, missing, or damaged anchor bolts are visually inspected. High-strength bolts are visually inspected. Each structure's foundation is monitored for overall settlement and

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.100-1 L-2018-193 Attachment 18 Page 3 of 7 differential settlement. Structures are monitored to confirm the absence of water in-leakage or signs of concrete leaching, chemical attack or steel reinforcement degradation.

Elastomers will be inspected for signs of hardening.

Revise the Enhancements in Section B.2.3.35 on page B-257 as follows:

Element Affected Enhancement

3. Parameters Monitored or Uedate the governing AMP erocedure to include Inspected monitoring for cracking due to sec for stainless steel and aluminum comeonents

4. Detection of Aging Effects Uedate the governing AMP erocedure to include guidance on surface examination inseections for cracking due to sec for stainless steel and aluminum comeonents

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.100-1 L-2018-193 Attachment 18 Page 4 of 7 Revise the Table 17-3, Item 39 on page A-107 as follows:

Table 17-3 List of SLR Commitments and Implementation Schedule (Continued)

No.

Aging Management Program NUREG-2191 Commitment Implementation Schedule or Activity (Section)

Section 39 Structures Monitoring XI.S6 h)

• No later than 6 months prior to the (17.2.2.35)

SPEO, i.e.:

i) Revise ins12ection 12rocedures to PTN3: 1/19/2032 include guidance on ins12ection for PTN4: 10/10/2032 cracking due to sec for stainless steel and aluminum com12onents

• SLRA revision associated with response to Set 6 RAI B.2.3.35-2.

Revise Table 3.5-1, Summary of Aging Management Evaluations for the Containment, Structures, and Component Supports, on page 3.5-86 as follows:

Table 3.5-1: Summary of Aging Management Evaluations for the Containment, Structures, and Component Supports Item Number Component Aging Effect I Aging Management Further Discussion Mechanism Programs Evaluation Recommended 3.5-1, 100

Aluminum, Loss of material due AMP XI.M32, Yes (SRP-SLR Consistent with NUREG-2191.

stainless steel to pitting and crevice "One-Time Section The Turkey Point Structures support members; corrosion, cracking Inspection," AMP 3.5.2.2.2.4)

Monitoring AMP will be used to welds; bolted due to sec XI.S6, manage loss of material and connections; "Structures cracking for aluminum and support anchorage Monitoring," or AMP stainless steel support members, to building structure XI.M36, welds, bolted connections, and "External Surfaces support anchorage exposed to Monitoring of uncontrolled indoor air, outdoor Mechanical air, and water - flowing or Components" standino environments.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.100-1 L-2018-193 Attachment 18 Page 5 of 7 Revise the following line item to Table 3.5.2-1, Containment Structure and Internal Structural Components - Summary of Aging Management Evaluation, on page 3.5-92 as follows:

Table 3.5.2-1: Containment Structure and Internal Structural Components - Summary of Aging Management Evaluation Intended Material Environment Aging Effect Aging NUREG-2191 Table 1 Function Requiring Management Item Item Management Program NaTB sump Structural Stainless steel Air-indoor Loss of material Structures III.B5.T-37b 3.5-1,100 fluid pH support uncontrolled Cracking Monitoring control basket Revise the following line items to Table 3.5.2-2, Auxiliary Building - Summary of Aging Management Evaluation, on page 3.5-104 and 3.5-111 as follows:

Table 3.5.2-2: Auxiliary Building - Summary of Aging Management Evaluation Intended Material Environment Aging Effect Aging NUREG-2191 Table 1 Function Requiring Management Item Item Management Program Fan/filter

Shelter, Stainless steel Air - outdoor Loss of material Structures III.B2.T-37b 3.5-1, 100 intake hood protection Cracking Monitoring Stop logs Flood Aluminum Air - outdoor Loss of material Structures III.B5.T-37b 3.5-1, 100 barrier Cracking Monitoring Notes D

Notes D

D

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.100-1 L-2018-193 Attachment 18 Page 6 of 7 Revise the following line items to Table 3.5.2-6, Diesel Driven Fire Pump Enclosure - Summary of Aging Management Evaluation, on page 3.5-126 as follows:

Table 3.5.2-6: Diesel Driven Fire Pump Enclosure - Summary of Aging Management Evaluation Intended Material Environment Aging Effect Aging NUREG-2191 Table 1 Function Requiring Management Item Item Management Program Doors

Shelter, Aluminum Air - outdoor Loss of material Structures 111.B5.T-37b 3.5-1, 100 protection Cracking Monitoring Louvers

Shelter, Aluminum Air - outdoor Loss of material Structures III.B5.T-37b 3.5-1,100 protection Cracking Monitoring Manufactured

Shelter, Aluminum Air - outdoor Loss of material Structures III.B5.T-37b 3.5-1,100 structure protection Cracking Monitoring Revise the following line items to Table 3.5.2-9, Emergency Diesel Generator Buildings - Summary of Aging Management Evaluation, on page 3.5-141 and 3.5-142 as follows:

Table 3.5.2-9: Emergency Diesel Generator Buildings - Summary of Aging Management Evaluation Intended Material Environment Aging Effect Aging NUREG-2191 Table 1 Function Requiring Management Item Item Management Program Electrical and Structural Stainless steel Air-indoor Loss of material Structures III.B3.T-37b 3.5-1, 100 instrument support uncontrolled Cracking Monitoring panels and

Shelter, enclosures protection HVAC roof Structural Stainless steel Air - outdoor Loss of material Structures III.B4.T-37b 3.5-1, 100 hoods support Cracking Monitoring Notes D

D D

Notes D

D

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.100-1 L-2018-193 Attachment 18 Page 7 of 7 Revise the following line items to Table 3.5.2-11, Intake Structure - Summary of Aging Management Evaluation, on page 3.5-156 as follows:

Table 3.5.2-11: Intake Structure - Summary of Aging Management Evaluation Intended Material Environment Aging Effect Aging NUREG-2191 Table 1 Function Requiring Management Item Item Management Program Electrical Structural Stainless steel Air - outdoor Loss of material Structures III.B3.T-37b 3.5-1, 100 enclosures support Cracking Monitoring

Shelter, protection Revise the following line items to Table 3.5.2-16, Turbine Building - Summary of Aging Management Evaluation, on page 3.5-184 as follows:

Table 3.5.2-16: Turbine Building - Summary of Aging Management Evaluation Intended Material Environment Aging Effect Aging NUREG-2191 Table 1 Function Requiring Management Item Item Management Program Perimeter Flood Aluminum Air - outdoor Loss of material Structures III.B5.T-37b 3.5-1,100 stop logs barrier Cracking Monitoring Associated

Enclosures:

None Notes D

Notes D

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.66-1 L-2018-193 Attachment 19 Page 1 of 6 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 3.5.1.66-1

Background:

The GALL-SLR Report, Chapter Ill, "Structures and Component Supports," includes aging management review (AMR) line items for tanks and missile barriers to ensure that the structural aging effects in their foundations, sub-foundations, supports, structural steel and bolting are adequately managed for the extended period of operation, consistent with the 10 CFR 54.21 (a)(3) requirement.

The subsequent license renewal application (SLRA) Section 17.2.2.17 states, in part, that tank's external surfaces, other than the tank-to-concrete interface, are accessible and inspected through the External Surfaces Monitoring of Mechanical Components AMP (SLRA Section 17.2.2.23) or the Structures Monitoring AMP (SLRA Section 17.2.2.35).

Issue:

It is not clear how GALL-SLR Report Group 7 (concrete tanks and missile barriers) and Group 8 (steel tanks and missile barriers) structures are being managed for structural aging effects since no line items in the SLRA associated with tanks or missile bar~iers credits the structural AMP or references associated GALL-SLR structural AMR line items for aging management (e.g. items associated with SRP-SLR Table 3.5-1 items 042, 054, 066, 067, 080).

Request:

For Group 7 and 8 structures subject to AMR, describe how the structural effects of aging (e.g., SRP-SLR Table 3.5-1 items 042, 054, 066, 067, 077, 080, etc.) will be adequately managed. Otherwise, provide adequate technical justification for not requiring,

management on the effects of aging.

If the response identifies alternate AMPs, explain how the identified AMPs will be able to identify and manage the associated structural aging effects. If the Structures Monitoring Program is not being credited for aging management, clarify the statement crediting such in SLRA Section 17.2.2.17 accordingly.

FPL Response:

The intention of the structural line items in tables 3.5.2-9 and 3.5.2-18 was to include the tanks and missile barriers and associated structural aging effects in various group's AMR line items. To make the application more clear, applicable Group 7 and Group 8 AMR line items were added to address the structural aging effects associated with concrete and steel tank components respectively (see Associated SLRA Revisions below).

Missile barriers are addressed in the SLRA AMR tables by including the designation in the component's intended function list rather than including a separate "missile barrier" component. All components with an intended function of missile barrier are grouped in

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRG RAI No. 3.5.1.66-1 L-2018-193 Attachment 19 Page 2 of 6 associated commodities, such as reinforced concrete, and can be identified by the listing in the intended function column. Each of these commodities are aligned to the applicable GALL-SLR line items for managing the associated structural aging effect and no SLRA revisions or clarifications are required. The commodity is managed for aging to ensure there is no loss of intended function.

As shown in the AMR tables, the Structures Monitoring AMP is credited with managing the associated structural aging effects of most tanks and missile barriers consistent with the GALL-SLR. Additionally, aging effects associated with other structural components with the missile barrier intended function such as the containment, components that may be exposed to boric acid leakage, and masonry walls are managed by the ASME Section XI, Subsection IWL AMP, Boric Acid Corrosion AMP, and Masonry Walls AMP, as appropriate, which is consistent with the GALL-SLR recommendations.

References:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.66-1 L-2018-193 Attachment 19 Page 3 of 6 Associated SLRA Revisions:

The SLRA is amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Add the following line items to Table 3.5.2-9, Emergency Diesel Generator Buildings - Summary of Aging Management Evaluation, on page 3.5-140 as follows:

Table 3.5.2-9: Emergency Diesel Generator Buildings - Summary of Aging Management Evaluation Aging Effect Aging Component Intended Requiring Management NUREG-2191 Table 1 Type Function Material Environment Management Program Item Item Reinforced Structural Concrete Air-indoor Cracking Structures III.A7.TP-25 3.5-11 054 concrete: U4 SU(;!(;!Ort uncontrolled Monitoring DOST foundation

{accessible}

Reinforced Structural Concrete Soil Cracking Structures III.A7.TP-27 3.5-11 065 concrete: U4 SU(;!(;!Ort Loss of bond Monitoring DOST Loss of material foundation

{accessible}

Reinforced Structural Concrete Air-indoor Cracking Structures III.A7.TP-26 3.5-11 066 concrete: U4 SU(;!~Ort uncontrolled Loss of bond Monitoring DOST Loss of material foundation

{accessible}

Notes B

B B

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.66-1 L-2018-193 Attachment 19 Page 4 of 6 Table 3.5.2-9: Emergency Diesel Generator Buildings - Summary of Aging Management Evaluation Aging Effect Aging Component Intended Requiring Management Type Function Material Environment Management Program Reinforced Structural Concrete Air-indoor Cracking Structures concrete: U4 SURROrt uncontrolled Monitoring DOST foundation

{inaccessible)

Reinforced Structural Concrete Soil Cracking Structures concrete: U4 SURROrt Loss of bond Monitoring DOST Loss of material foundation

{inaccessible)

Reinforced Structural Concrete Soil Increase in Structures concrete: U4 SURROrt ROrosit~ and Monitoring DOST Rermeabilitv foundation Cracking

{inaccessible)

Loss of material Reinforced Structural Concrete Air-indoor Increase in Structures

. concrete: U4 SURROrt uncontrolled ROrositv and Monitoring DOST Rermeabilitv foundation Cracking (accessible)

Loss of material Steel Structural Carbon steel Air-indoor Loss of material Structures comRonents: U4 SURROrt uncontrolled Monitoring DOST NUREG-2191 Table 1 Item Item Notes 111.A7.TP-204 3.5-11 043 8

III.A7.TP-212 3.5-11 065

.§ 111.A7. TP-29 3.5-11 067 8

III.A7.TP-28 3.5-11 067 8

III.A7.TP-302 3.5-11 077 8

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251' FPL Response to NRC RAI No. 3.5.1.66-1 L-2018-193 Attachment 19 Page 5 of 6 Table 3.5.2-9: Emergency Diesel Generator Buildings - Summary of Aging Management Evaluation Aging Effect Aging Component Intended Requiring Management Type Function Material Environment Management Program Structural Structural Carbon steel Air-indoor Loss of Qreload Structures bolting: U4 SUQQOrt uncontrolled Monitoring DOST Structural Structural Carbon steel Air-indoor Loss of material Structures bolting: U4 SUQQOrt uncontrolled Monitoring DOST NUREG-2191 Table 1 Item Item III.A7. TP-261 3.5-11 088 III.A7. TP-248 3.5-11 080 Add the following line items to Table 3.5.2-18, Yard Structures - Summary of Aging Management Evaluation, on page 3.5-192 as follows:

Table 3.5.2-18: Yard Structures - Summary of Aging Management Evaluation Aging Effect Aging Component Intended Requiring Management NUREG-2191 Table 1 Type Function Material Environment Management Program Item Item Reinforced Structural Concrete Air-outdoor Cracking Structures 111.AS.TP-25 3.5-11 054 concrete: Tank SUQQOrt Monitoring foundations

{accessible}

Reinforced Structural "

Concrete Soil Cracking Structures 111.AS.TP-27 3.5-11 065 concrete: Tank SUQQOrt Loss of bond Monitoring foundations Loss of material

{accessible}

Reinforced Structural Concrete Air-outdoor Cracking Structures Ill.AS. TP-204 3.5-11 043 concrete: Tank SUQQOrt Monitoring foundations

{inaccessible}

Notes

~

~

I Notes B

B B

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.5.1.66-1 L-2018-193 Attachment 19 Page 6 of 6 Table 3.5.2-18: Yard Structures - Summary of Aging Management Evaluation Aging Effect Component Intended Requiring Type Function Material Environment Management Reinforced Structural Concrete Soil Cracking concrete: Tank SUE!E!Ort Loss of bond foundations Loss of material

{inaccessible}

Reinforced Structural Concrete Soil Increase in concrete: Tank SUE!E!Ort E!Orositv and foundations E!ermeabilitv

{inaccessible}

Cracking Loss of material Steel Structural Carbon steel Air-outdoor Loss of material comE!onents:

SUE!E!Ort Tanks Structural Structural Carbon steel Air-outdoor Loss of E!reload bolting: Tanks SUE!t20rt Structural Structural Carbon steel Air-outdoor Loss of material bolting: Tanks SUt2E!Ort Associated

Enclosures:

None Aging Management NUREG-2191 Table 1 Program Item Item Notes Structures III.AB.TP-212 3.5-11 065 B

Monitoring Structures 111.AB.TP-29 3.5-11 067 B

Monitoring Structures Ill.AB. rp:.302 3.5-11 077 B

Monitoring Structures Ill.AB. TP-261 3.5-11 088 B

Monitoring Structures Ill.AB. TP-248 3.5-11 080 B

Monitoring

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.13-1 L-2018-193 Attachment 20 Page 1 of 3 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 1 O. Inspection of Overhead Heavy Load and Light Load (Related to Refueling)

Handling Systems, GALL AMP XI.M23 Regulatory Basis:

Title 10 of the Code of Federal Regulations (10 CFR) Section 54.21(a)(1)(i) and (ii) state that systems, structures and components within the scope of license renewal and subject to an aging management review shall encompass those structures and components that (i) perform an intended function, as described in 10 CFR 54.4, without moving parts or without a change in configuration or properties; and (ii) are not subject to replacement based on a qualified life or specified time period.

r 10 CFR 54.21 (a)(3) requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis for the period of extended operation. As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report and when evaluation of the matter in the GALL-SLR Report applies to the plant.

RAI 8.2.3.13-1

Background:

SLRA Section B.2.3.13, "Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems," states that the Turkey Point Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Program is an existing AMP that will be consistent with enhancements with the GALL-SLR Report AMP XI.M23, "Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems." GALL-SLR Report AMP XI.M23 addresses the inspection and monitoring of crane-related structures and components to provide reasonable assurance that the handling system does not affect the intended function of nearby safety-related equipment. The GALL-SLR Report Volume 1, page VII B-1, states that the primary components that the GALL-SLR Report AMP XI.M23 is concerned with are the "structural girders and beams that make up the bridge and the trolley."

Issue:

SLRA Section B.2.3.13 states that the AMP does not manage aging effects for load handling components such as trolleys and rigging because such components are considered "active" components. The staff position is that, depending on the lifting capacity of a crane, trolleys and rigging may include structural girders and beams that are passive components that should be within the scope of this AMP and subject to AMR in accordance with 10 CFR 54.4 and 54.21. The staff noted that although the SLRA states that trolleys and rigging components will not be managed for aging effects, the SLRA has AM Rs assigned to trolleys (e.g., trolley structures of the turbine gantry crane and polar crane in SLRA Tables 3.5.2-17 and 3.5.2-14) and rigging (e.g., intake cooling water (ICW) valve pit rigging beam, LRA Table 3.5.2-11) components of some of its in-scope cranes.

Furthermore, SLRA Section 4.7 reviews the TLAA of the "Intake cooling water (ICW)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRG RAI No. 8.2.3.13-1 L-2018-193 Attachment 20 Page 2 of 3 valve pit rigging beam." It is not clear whether trolleys and rigging of other cranes have such passive components that need to be included in the "scope of program," program element of SLRA AMP B.2.3.13.

Request:

1. Clarify whether structural component(s) (e.g., girders and beams), of trolleys and rigging used in load handling systems listed in SLRA Section B.2.3.13, are not in the "scope of program" of the Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling AMP and not subject to an AMR.

2. For each such component that is not within the "scope of program" program element of the AMP and not subject to an AMR provide your basis as to how such determination is consistent with the regulatory requirements in 10 CFR 54.4 and 54.21, and the guidance in the GALL-SLR Report AMP XI.M23.

3. If these components are to be age managed in a manner that is not consistent with the guidance in GALL-SLR Report AMP XI.M23, describe the exception(s) to the GALL-SLR Report AMP XI.M23 and provide the basis for such exception(s).

FPL Response:

The responses to the individual requests are as follows:

1. Structural components of trolleys and rigging are within the scope of the Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems Aging Management Program (AMP), are subject to an Aging Management Review (AMR), and have been included in the AMR tables. SLRA Section B.2.3.13, paragraph 5, included unnecessary repetition of the screening methodology addressed in SLRA Section 2.1. Therefore, the statement that identifies items excluded from the AMP scope is deleted. See the "Associated SLRA Revisions" section for the respective SLRA deletion.

2. The structural components of trolleys and rigging are not excluded from the scope of the Inspection of Overhead Heavy Load and Light Load (Related to Refueling)

Handling Systems AMP, therefore, no additional determination is required.

3. As stated above, the structural components of trolleys and rigging are within the scope of the Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems AMP and are therefore managed in a manner that is consistent with GALL-SLR Report AMP XI.M23.

References:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50.,251 FPL Response to NRC RAI No. B.2.3.13-1 L-2018-193 Attachment 20 Page 3 of 3 Associated SLRA Revisions:

SLRA Section B.2.3.13 is amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise SLRA Section B.2.3.13 paragraph 5 as follows:

This AMP does not manage aging effects for the following load handling components:

motors, trolleys, cables, hooks, rigging, etc. Such components are considered "active" components and are not screened into any AMP.

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.3.1.199-1 L-2018-193 Attachment 21 Page 1 of 5 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 3.3.1.199-1

Background:

SLRA Section B.2.3.13 states that the Turkey Point Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Program is an existing AMP that will be consistent with enhancements with the GALl.-SLR Report AMP XI.M23, "Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems." The staff noted that GALL-SLR Report AMR 3.3-1, item 199 recommends that cranes' structural bolting be managed under the GALL-SLR Report AMP XI.M23 for the aging effects for loss of preload, loss of material, and cracking.

Issue:

During its audit review of document O-PMM-009.05, "Intake Area Gantry Crane Inspection and Preventive Maintenance," Revision 0, the staff noted that this document states that structural bolting of the intake crane must be visually inspected for signs of defect. The staff also noted that SLRA AMP B.2.3.13 is enhanced to account for aging effects of bolted connections. However, the components in the following tables do not include structural bolting for cranes:

SLRA Table 3.5.2-11 "Intake Structures" component types intake structures cranes and intake cooling water (ICW) valve pit rigging beam; and SLRA Table 3.5.2-12, "Main Steam and Feedwater," component type main steam platform rails The staff noted that that these crane components may have bolting that should be age managed for loss of preload, loss of material, and cracking aging effects as recommended in the guidance of GALL-SLR AMR 3.3-1, item 199. If these cranes have bolting, it is not clear how the aforementioned aging effects will be managed during the subsequent period of extended operation (SPEO).

Request:

1. Clarify if the intake structures cranes, ICW valve pit rigging beam, and main steam platform rails have bolted components.

2. If so, state how the aging effects for loss of preload, loss of material, and cracking will be managed during the SPEO to be consistent with GALL-SLR.

FPL Response:

The responses to the individual requests are as follows:

1. The intake structure bridge crane, ICW valve pit rigging beam, and main steam platform rails do have bolted components. See the "Associated SLRA Revisions" section for the correction to SLRA Table 3.5.2-11 and Table 3.5.2-12.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.3.1.199-1 L-2018-193 Attachment 21 Page 2 of 5

2. The Inspection of Overhead Heavy Load and Light Load (Related to Refueling)

Handling Systems Aging Management Program (AMP) will manage the aging effects of loss of preload, loss of material, and cracking for these respective bolting items during the SPEO. See the "Associated SLRA Revisions" section for additional details.

References:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.3.1.199-1 L-2018-193 Attachment 21 Page 3 of 5 Associated SLRA Revisions:

SLRA Table 3.5.2-11 and Table 3.5.2-12 are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise SLRA Table 3.5.2-11 as follows:

Table 3.5.2-11: Intake Structure -

Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging NUREG-Table 1 Notes Type Function Requiring Management 2191 Item Item Management Program ICWValve Pit Structural Carbon steel Air - outdoor Loss of material Inspection of Overhead VII.B.A-07 3.3-1, 052 lA rigging beam support Galvanized Deformation Heavy Load and Light steel Cracking Load (Related to Refueling) Handling Systems ICW Valve Pit Structural Carbon steel Air - outdoor Loss of (!reload lns[!ection of VII.B.A-730 3.3-1, 199 A rigging beam SU[![!Ort Loss of Overhead Hea~ Load bolting material and Light Load Cracking

{Related to Refueling}

Handling Sl£stems Intake Structure Structural Carbon steel Air - outdoor Cumulative TLAA-Section 4.7.6, VII.B.A-06 3.3-1, 001 lA Cranes support fatigue damage Crane Load Cycle Limit Intake Structure Structural Carbon steel Air - outdoor Loss of material Inspection of Overhead VII.B.A-07 3.3-1, 052 lA Cranes support Deformation Heavy Load and Light Cracking Load (Related to Refueling) Handling Systems

1-Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.3.1.199-1 L-2018-193 Attachment 21 Page 4 of 5 Table 3.5.2-11: Intake Structure -

Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Type Function Requiring Management Intake Structure Structural Carbon steel Air - outdoor Loss of greload Cranes bolting suggort Loss of material Cracking Revise SLRA Table 3.5.2-12 as follows:

Aging Management Program lnsgection of Overhead Hea~ Load and Light Load

{Related to Refueling}

Handling S:istems Table 3.5.2-12: Main Steam and Feedwater Platforms-Summary of Aging Management Evaluation Aging Effect Component Intended Requiring Aging Management NUREG-Table 1 Notes 2191 Item Item VII.B.A-730 3.3-11 199

~

NUREG-Table 1 Type Function Material Environment Management Program 2191 Item Item Notes Main Steam Structural Carbon steel Air - outdoor Loss of material Inspection of VII.B.A-07 3.3-1, 052

~

platform support Deformation Overhead Heavy monorails Cracking Load and Light Load (Related to Refueling)

Handling Systems Main Steam Structural Carbon steel Air - outdoor Loss of greload lnsgection of VII.B.A-730 3.3-11 199

~

glatform suggort Loss of material Overhead Hea~

monorails Cracking Load and Light Load bolting

{Related to Refueling} Handling S3tstems Miscellaneous HELB Carbon steel Air - outdoor Loss of material Structures Monitoring III.A3.TP-302 3.5-1, 077 B

steel shielding

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 3.3.1.199-1 L-2018-193 Attachment 21 Page 5 of 5 Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.9-1 L-2018-193 Attachment 22 Page 1 of 8 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

11. Bolting Integrity, GALL AMP XI.M18 RAI 8.2.3.9-1

Background:

Section 54.21 (a)(3) 10 CFR requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis for the period of extended operation. As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report and when evaluation of the matter in the GALL-SLR Report applies to the plant.

SLRA Section 8.2.3.9, ""Bolting Integrity" states that the Turkey Point Bolting Integrity Program is an existing AMP that will be consistent with enhancements with the GALL-SLR Report AMP XI.M18, "Bolting Integrity."

Issue:

The "parameters monitored or inspected" and "detection of aging effects" program elements in the GALL-SLR Report AMP XI.M18 recommend that high strength closure bolting (with a yield strength greater than or equal to 150 ksi and a diameter greater than 2 inches) should be monitored for indications of cracking due to stress corrosion cracking (SCC) and subject to volumetric examination in accordance with ASME Code Section XI, Table IWB-2500-1, Examination Category B-G-1, to detect sec. LRA Section B.2.3.9 states that there is currently no high strength bolting within the scope of the Bolting Integrity Program. However, during the In-Office audit the staff noted that specification SPEC-M-004, "Maintenance Bolting Specification for St. Lucie Units 1 & 2 and Turkey Point Units 3 & 4," Revision 15, lists bolting material of American Society for Testing and Materials (ASTM) No. SA/A540, Grade 823 C1.1 as acceptable for use at the site. The applicant's specification also states that this bolting has a yield strength equal to 150 ksi and a diameter of 3 inches or less. Based on specification SPEC-M-004 it appears that bolting material ASTM No. SA/A540, Grade 823 C1.1 can be installed as closure bolting, and the staff cannot confirm the applicant's claim that high strength bolts are not being used in systems, structures, and components (SSCs) within the scope of the Bolting Integrity Program. If this material is used it is not clear how the Bolting Integrity AMP will adequately manage the aging effect of cracking due to SCC of the closure bolting and how the AMP is consistent with the recommendations in the "parameters monitored or inspected" and "detection of aging effects" program elements of GALL-SLR Report AMP XI.M18.

Request:

Clarify whether closure bolting with material ASTM No. SA/A540, Grade 823 C1.1, or any other high strength closure bolting material with a yield strength greater than or equal to

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRG RAI No. B.2.3.9-1 L-2018-193 Attachment 22 Page 2 of 8 150 ksi and a diameter greater than 2 inches, will be used or precluded from use in SSCs within the scope of the Bolting Integrity AMP during the subsequent period of extended operation. If such material will be precluded from use clarify how this will be accomplished given that the specification appears to authorize its use. If such material can be used state whether and how the aging effects of cracking due to stress corrosion cracking will be managed consistent with recommendations in GALL-SLR Report AMP XI.M18.

FPL Response:

Site specifications currently list bolting material ASTM No. SA/A540, Grade 823 Cl.1 as acceptable for use at the site, and therefore, high strength closure bolting is assumed to be in use at Turkey Point Units 3 and 4. The SLRA is revised to include discussion of aging management for high strength closure bolting for the Bolting Integrity AMP.

Additionally, SLRA Section 8.2.3.9, "Bolting Integrity" and Table 17-3, item 13 states, per the recommendation of GALL-SLR, that any replacement or new pressure-retaining bolting will have an actual yield strength less than 150 ksi, which ensures ASTM No.

SA/A540, Grade 823 Cl.1, or any other high strength closure bolting will not be installed at PTN in the future.

References:

None Associated SLRA Revisions:

SLRA Section 8.2.3.9, Section 17.2.2.9, Table 3.2-1, Table 3.3-1, Table 3.4-1, and Table 17-3 are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise SLRA Section 8.2.3.9 on page B-103 as follows:

Element Affected Enhancement

1. Scope of Program Create a new governing procedure and update

4. Detection of Aging existing procedures for this AMP to do the Effects following in accordance with this NUREG-2191 element XI.M18:

• Include submerged pressure-retaining bolting in inspections.

• Include closure bolting for piping systems that contain air or gas for which leakage is difficult to detect.

• Include monitoring of high strength bolting for surface and subsurface discontinuities indicative of cracking. This will be accomQlished bl£ Qerforming volumetric examination 1 in accordance with ASME Code Section Xl 1 Table IWB-2500-1 1 Examination Cateaorv B-G-1 le.a..

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.9-1 L-2018-193 Attachment 22 Page 3 of 8 Element Affected Enhancement acceptance standards1 extent and freguency of examination}. Specified bolting material properties {e.g.1 design and procurement specifications1 fabrication and vendor drawings1 material test reports} may be used to determine if the bolting exceeds the threshold to be classified as hiah strenath.

2. Preventive Actions Create a new governing procedure and update existing procedures for this AMP to do the following in accordance with this NUREG-2191 element XI.M18:

• Ensure any replacement or new pressure-retaining bolting has an actual yield strength less than 150 ksi.

• Ensure that lubricants containing molybdenum disulfide will not be used in conjunction with pressure-retaining bolting.

6. Acceptance Criteria Include appropriate acceptance criteria for submerged pressure-retaining bolting and closure bolting for piping systems that contain gas or air for which leakage is difficult to detect.

Revise SLRA Section B.2.3.9 paragraph 2 on page B-103 as follows:

Per the GALL-SLR, SCC has occurred in high strength bolts used for nuclear steam supply system component supports (EPRI NP-5769). Additionally, operating experience and laboratory examinations show that the use of molybdenum disulfide as a lubricant is a potential contributor to SCC. Based on investigation in response to an NRC request for information supporting license renewal for the current PEO (reported in March 2001 ),

there is currently no high strength bolting within the scope of this program and molybdenum disulfide lubricant is not in use. PTN may have high strength bolting installed. If high strength bolting is found1 it shall be monitored for surface and subsurface discontinuities indicative of cracking. The existing activities of this AMP will be enhanced to ensure that no new high strength bolting within the scope of this program will be installed and molybdenum disulfide will not be used as a lubricant.

Insert the following paragraph into SLRA Section 17.2.2.9 after the 1st paragraph on page A-18 as follows:

If closure bolting greater than 2 inches in diameter {regardless of code classification} with actual yield strength greater than or egual to 150 ksi {1 1034 MPa} is found and for closure bolting for which yield strength is unknown1 volumetric examination in accordance to that of ASME Code Section Xl1 Table IWB-2500-11 Examination Category B-G-1 1 is performed. Specified bolting material properties may be used to determine if the bolting exceeds the threshold to be classified as high strength.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.9-1 L-2018-193 Attachment 22 Page 4 of 8 Revise SLRA Table 3.2-1, item 012 as follows:

Table 3.2-1: Summary of Aging Management Evaluations for the Engineered Safety Features Further Item Aging Effect/

Aging Management Evaluation Number Component Mechanism Programs Recommended 3.2-1, 012 High-strength steel Cracking due to sec;

~MP XI.M18, No closure bolting exposed cyclic loading "Bolting Integrity" o air, soil, underground Discussion I

L --.-1" -' r Th--- ;~ -- h'-L.. -*----*L.. -*--*

..-..1--

ti"- 1--IL:--

j...,.f.h......

r-_ -'- -

_,,-,. ~-

t:'..

• -*1 J

Consistent with NUREG-2191.

Hiah-strenath closure boltina mav be used in the Enaineered Safetv Features.

If used the Boltina lntearitv

~MP is used to manaae crackina.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.9-1 L-2018-193 Attachment 22 Page 5 of 8 Revise SLRA Table 3.3-1, item 010 as follows:

Table 3.3-1: Summary of Aging Management Evaluations for the Auxiliary Systems Item Aging Effect/

Aging Management Number Component Mechanism Programs 3.3-1, 010 High-strength steel Cracking due to SCC;

~MP XI.M18, closure bolting exposed cyclic loading "Bolting Integrity"

~o air, soil, underground l

Further Evaluation Recommended Discussion No I

-.-1" -L.1-r *-

~* --,_ --....

~-l~C-- -----C-L-..1.. :LL LL-I\\... :,: __ <',*-L---

J -

Consistent with NUREG-2191.

Hiah strenath boltina mav be used in the Auxiliarv Svstems. If found the Boltina lntearitv AMP is used to manaae crackina.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.9-1 L-2018-193 Attachment 22 Page 6 of 8 Revise SLRA Table 3.4-1, item 007 as follows:

Table 3.4-1: Steam and Power Conversion Systems Item Aging Effect I Number Component Mechanism 3.4-1, 007 High-strength steel Cracking due to sec; closure bolting exposed cyclic loading o air, soil, underground Further Aging Management Evaluation Programs Recommended AMP XI.M18, No "Bolting Integrity" Discussion I" -

ITh--- ;~ -- L"~h

-*-~1 11-IL:-,_

~.... 4,t.,..-

. ~

Ir - *-

-J-Consistent with NUREG-2191.

Hiah-strenath boltina mav be used in the Steam and Power Conversion Svstems. If found

• he Boltinci lnteciritv AMP is used to manaae crackina.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.9-1 L-2018-193 Attachment 22 Page 7 of 8 Revise SLRA Table 17-3, Item 13 as follows:

No.

Aging Management NUREG-Program or Activity 2191 (Section)

Section 13 Bolting Integrity XI.M18 (17.2.2.9)

Commitment Implementation Schedule Continue the existing PTN Bolting Integrity No later than 6 months AMP, including enhancement to:

prior to the SPEO, i.e.:

a) Inspect submerged pressure-retaining PTN3: 1/19/2032 bolting when submerged portions of PTN4: 10/10/2032 components (e.g., pump casings) are overhauled or replaced during maintenance activities; b) Evaluate closure bolting for piping systems that contain air or gas, for which leakage is difficult to detect, on a case-by-case basis through -

• Visual inspection during maintenance activities;

• Visual inspection for discoloration of nearby external surfaces;

• Monitoring and Trending of pressure decay within an isolated boundary;

• Soap bubble testing; or

• Thermography when fluid temperature is higher than ambient.

c) Ensure any replacement or new pressure-retaining bolting has an actual yield strength less than 150 ksi; d) Ensure that lubricants containing molybdenum disulfide will not be used in conjunction with pressure retaining boltinq;

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.9-1 L-2018-193 Attachment 22 Page 8 of 8 No.

Aging Management NUREG-Program or Activity 2191 (Section)

Section Associated

Enclosures:

None Commitment Implementation Schedule e) Include appropriate acceptance criteria for submerged pressure-retaining bolting and closure bolting for piping systems that contain gas or air for which leakage is difficult to detect.

f) Include monitoring high strength bolting for surface and subsurface discontinuities indicative of cracking.

This will be accomRlished bl£ Rerforming volumetric examination 1 in accordance with ASME Code Section Xl 1 Table IWB-2500-1 1 Examination Categor)l B-G-1 {e.g. 1 acceRtance standards 1 extent and freguencll of examination}. SRecified bolting material RrORerties {e.g. 1 design and Rrocurement SRecifications 1 fabrication and vendor drawings 1 material test reRorts} mall be used to determine if the bolting exceeds the threshold to be classified as high strenath.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.6-1 L-2018-193 Attachment 23 Page 1 of 5 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

12. Fatigue Analysis of Cranes (crane cycle limits), TLAA 4.7.6 Regulatory Basis:

Title 10 of the Code of Federal Regulations (10 CFR), Section 54.21 (c)(1) requires the applicant to evaluate time limited aging analyses (TLAAs). 10 CFR 54.21 (a)(3) requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis for the period of extended operation. As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report and when evaluation of the matter in the GALL-SLR Report applies to the plant. The SRP-SLR and GALL-SLR documents provide guidance for SLR applicants that voluntary choose and plan to demonstrate consistency with the GALL-SLR Report.

RAI 4.7.6-1

Background:

The staff's guidance in SRP-SLR Section 4.7.3.1.1 states, in part, the following:

Justification provided by the applicant is reviewed to verify that the existing analysis remains valid for the subsequent period of extended operation [(SPEO)].

[... ] The applicant describes the TLAA with respect to the objectives of the analysis, assumptions used in the analysis, conditions, acceptance criteria, relevant aging effects, and intended function(s). For those TLAAs that consider cyclic loading, each load should be identified along with the corresponding number of total cycles assumed in the analysis and the number of cycles that are anticipated to occur through the SPEO.

Turkey Point's SLRA Section 4.7.6, "Crane Load Cycle Limit," states that the EOCl-61, "Electric Overhead Crane Institute," compliant cranes are acceptable for 2,000,000 cycles and as such for these cranes to exceed this cycle limit through the extended period of operation they have to experience 68 cycles per day which is far more than what the cranes would experience during the SPEO. The staff noted that the applicant established a limit of 2,000,000 load cycles for these cranes using Section 1. 7.3 of the American Institute of Steel Construction (AISC) Manual (6th Edition). The staff also noted that the applicant claims that these cranes will be managed consistent with the guidance in the GALL-SLR Report.

Issue:

Section 1.7.3 of the AISC Manual (6th Edition) states that for cranes that meet its stress criteria the load cycle limits range from 100,000 to 2,000,000. It is not clear on what basis the applicant concluded in SLRA Section 4.7.6 that the load cycle limit for these cranes is the upper limit of 2,000,000 instead of the more conservative lower limit of 100,000, or

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.6-1 L-2018-193 Attachment 23 Page 2 of 5 any other value of load cycles within that range. It is also noted that the SLRA does not identify the derivation of the related loads and the corresponding estimated number of load cycles for each of the EOCl-61 cranes through the SPEO. To complete its review and determine the adequacy of the associated TLAAs the staff needs a description of the applied loads and the derivation of the number of cycles expected for each of these cranes through the SPEO.

Request:

For the TLAAs associated with each of the EOCl-61 cranes listed in SLRA Section 4.7.6:

1. Provide the basis for the selection of 2,000,000 load cycles as the design limit.

2. Provide a description of the design loads and the corresponding number of load cycles estimated to occur through the end of the SPEO. Include a description of conditions and assumptions made to reach the conclusion for the estimated number of cycles through the end of the SPEO.

FPL Response:

The responses to the individual requests are as follows:

1. The 2,000,000-load cycle design limit for cranes within the scope of license renewal (except the spent fuel bridge cranes) was identified as part of the TLAA evaluation performed in support of the original Turkey Point (PTN) license renewal application.

Additionally, estimated crane cycles were originally compared to the 2,000,000-cycle limit within the response to RAI 4.7.4-1 included in Attachment 1 to Reference 1. The original license renewal evaluation and supporting calculation determined that a design limit of 2,000,000 load cycles was appropriate based on the original design codes and loading conditions.

The applicable industry 'Standards in place at the time the cranes were designed and manufactured formed the basis of the evaluation. Structural design for the subject cranes was in accordance with EOCI Specification #61; requirements providing for reduction in strength due to fatigue were summarized in the 1963 AISC design specification (included in Part 5 of the AISC Manual of Steel Construction, Sixth Edition, Third Revised Printing). Section 1.7.3 of the AISC specification requires that members, connection material and fasteners subject to not more than 2,000,000 applications of maximum design loading be proportioned to unit stresses allowed by Sections 1.5 and 1.6 of the AISC specification. It also stipulates that these allowable stresses be compared to the algebraic difference between the maximum computed stress and the minimum computed stress, but not be less than those required to support either the maximum or minimum computed stress. The cranes are configured so that the minimum computed stress is effectively zero. In other words, there is no stress reversal. Therefore, the maximum loading controls the design of the structural elements (max. stress).

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.6-1 L-2018-193 Attachment 23 Page 3 of 5 Because the equivalent unit stresses (tension, compression due to bending, and shear) imposed during each cycle of maximum loading on structural elements of cranes whose design complies with EOCl-61 are either the same or lower (more conservative) than the stresses allowed by the AISC specification for A7 steel structural components, acceptability of the 2,000,000-cycle limit is inherent in the original design of the subject cranes at PTN.

2. As confirmed in Attachment 13 to Reference 2, the following cranes within the scope of the PTN crane fatigue TLAA calculation were designed to meet pertinent requirements of EOCl-61:

a. Reactor building polar cranes

b. Spent fuel cask crane

c. Intake structure bridge crane

d. Turbine gantry crane

e. Reactor cavity manipulator cranes According to the response to original license renewal response to RAI 4.7.4-1 included in Attachment 1 to Reference 1, the PTN cranes are used primarily during refueling outages. Occasionally, cranes make lifts at or near their rated capacity (e.g., the turbine gantry crane lifting a turbine rotor). Usually, these cranes make lifts substantially less than their rated capacity.

The end-of-SPEO crane cycle estimates and respective loading assumptions for the above cranes are listed in the bullets below. Note that the spent fuel cask crane, intake structure bridge crane, and turbine gantry crane have each had major component replacements/upgrades installed during 2011, and for conservatism, their respective crane cycles prior to 2011 have been back-calculated and included in the total number of cycles.

a. The Unit 3 and Unit 4 reactor building polar cranes each have a 205-ton capacity.

They are required to lift their respective 151-ton reactor pressure vessel (RPV) head twice (2 crane cycles) every refueling outage.

By the end of the 2019 spring refueling outage, 30 refueling cycles will have occurred at each unit, and likewise, the following number of crane cycles:

60 RPV head lifts (crane cycles) per unit.

60 upper reactor internals lifts (crane cycles) (every 18 months at 2 crane cycles each) per unit.

30 additional crane cycles (including*the 10-year core barrel lifts) per unit.

The total of number of critical lifts (crane cycles) with main hooks on both units would be assumed at 150 crane cycles per unit by the end of the 2019 spring refueling outage. By doubling the 45-year estimate for 90 years (10 years longer than expected plant life), each unit would experience 300 crane cycles.

b. The spent fuel cask crane has a 125-ton capacity. Since 2011, the crane has lifted 38 spent fuel canisters that weigh near the crane capacity. Assuming 3 crane cycles per canister lift yields a total of 114 crane cycles in 7 years. This is approximately 17 crane cycles per year.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.6-1 L-2018-193 Attachment 23 Page 4 of 5 For the 35 years through the end of the SPEO (in 2053) the crane the crane is expected to experience about the same amount of use for each ?-year period.

Likewise, this rate is conservative for the years prior to 2011. The rate of 17 crane cycles per year can be multiplied by 73 years to project 1,241 crane cycles for all of those years. Therefore, this value for the entire 80-year plant life yields an estimate of 1,355 crane cycles by the end of the SPEO.

c. The intake structure bridge crane has a 25-ton capacity. The expected average load for this crane is 17.5 tons. The crane is expected to perform 40 crane cycles per year from 2011 through 2053 (42 years) for a total of 1,680 crane cycles.

Conservatively assuming the same rate (40 crane cycles per year) for the previous 38 years yields 1,520 crane cycles. Therefore, this value for the entire 80-year plant life yields an estimate of 3,200 crane cycles by the end of the SPEO.

d. The turbine gantry crane has a 170-ton capacity. The maximum load is expected to be the new low-pressure turbines and rigging at 167 tons. The average main hook load lifts are expected to be approximately 100 tons.

Based on an assumed inspection cycle for the low-pressure turbines, high-pressure turbines, and turbine generators, the expected crane lifts from 2011 through 2053 (42 years) are expected to be approximately 417 lifts from the installed hoist, with an additional 200 lifts due to EPU and low-pressure turbine upgrades. This yields a total of 617 lifts per unit, multiplied by 2 crane cycles per lift, for a total of 1,234 crane cycles over 42 years. Conservatively assuming the same rate (30 crane cycles per year) for the previous 38 years yields 1, 140 crane cycles. Therefore, this value for the entire 80-year plant life yields an estimate of 2,374 crane cycles by the end of the SPEO.

e. The reactor cavity manipulator cranes each have a 1-ton capacity hoist. The typical load is 1439 pounds. For every 18-month refueling outage, 157 fuel assemblies are moved at each unit. Since there are two crane cycles per fuel assembly, each refueling cycle requires a total of 314 crane cycles at each unit.

By spring 2019, 30 refueling outages will have occurred at each unit, and therefore, approximately 9,420 crane cycles will have been experienced on each manipulator crane (30 outages x 314 cycles= 9,420 cycles). Another 23 refueling outages until the end of the 80-year license would assume an additional 7,222 crane cycles per unit (23 outages x 314 cycles= 7,222 cycles). Therefore, it is estimated that each reactor cavity manipulator crane will have performed 16,642 crane cycles by the end of the SPEO.

The estimated crane cycles for each of the in-scope cranes listed above is well below the 2,000,000-crane cycle limit. Therefore, there is reasonable assurance that the cranes will not exceed their design or suffer fatigue-induced failure.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.6-1 L-2018-193 Attachment 23 Page 5 of 5

References:

1. FPL Letter L-2001-75 to NRC dated April 19, 2001, Response to Request for Additional Information for the Review of the Turkey Point Units 3 and 4 License Renewal Application (ADAMS Accession No. ML011170195)

2. FPL Letter L-2018-177 to NRC dated October 9, 2018, Turkey Point Units 3 and 4 Subsequent License Renewal Application, Safety Review Requests for Confirmation of Information (RCI) Responses Associated SLRA Revisions:

None Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.6-2 L-2018-193 Attachment 24 Page 1 of 3 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 4.7.6-2

Background:

The staff's guidance in SRP-SLR Section 4.7.3.1.1 states, in part, the following:

Justification provided by the applicant is reviewed to verify that the existing analysis remains valid for the subsequent period of extended operation [(SPEO)].

[... ] The applicant describes the TLAA with respect to the objectives of the analysis, assumptions used in the analysis, conditions, acceptance criteria, relevant aging effects, and intended function(s). For those TLAAs that consider cyclic loading, each load should be identified along with the corresponding number of total cycles assumed in the analysis and the number of cycles that are anticipated to occur through the SPEO.

SLRA Section 4.7.6 states that the spent fuel bridge cranes design "was in accordance with CMAA-70 [... ],with added seismic requirements," and is acceptable for up to 200,000 load cycles of maximum loads. The SLRA also states:

For original license renewal, the projected number of cycles for these cranes was 16,000. Applying a simple 80/60 multiplier, the total number of cycles for SLR would be conservatively estimated to be 22,000. This is well below the design cycles of 200,000.

The staff noted that its review of the spent fuel bridge crane TLAA for the original license renewal of Turkey Point is documented in Section 4.7.4 of NUREG-1759, "Safety Evaluation Report Related to the License Renewal of Turkey Point Nuclear Plant, Units 3 and 4," (ADAMS Accession No. ML021280532). The staff also noted that its conclusion for safety determination for the spent fuel bridge crane TLAA was based on its evaluation of the applicant's response to RAI 4.7.4-1 (ADAMS Accession No. ML011170195).

Issue:

It is not clear whether the added crane seismic requirements referenced in SLRA Section 4.7.6 are part of the spent fuel bridge cranes TLAAs and have been included in the applicant's estimate for the SPEO. It is also not clear whether the conditions and assumptions described by the applicant in response to RAI 4.7.4-1 dated April 19, 2001, are still valid and applicable with regards to load (lift) cycles and usage of the spent fuel bridge cranes for the estimated number of load cycles during the SPEO.

Request:

For the TLAAs associated with the spent fuel bridge cranes:

1. Clarify whether the seismic requirements (including seismic load cycles and their accountability in stresses) are part of the cranes TLAAs. If so, in accordance with 1 O CFR 54.21 (c)(1 ), account for their inclusion in the TLAAs through the SPEO.

2. Clarify whether the conditions and assumptions {e.g., type of loads (fuel assemblies, fuel assembly shuffles, etc.,) and number of load cycles} described in

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.6-2 L-2018-193 Attachment 24 Page 2 of 3 response to RAI 4.7.4-1 dated April 19, 2001, are still valid for the Turkey Point SLRA and applicable through the SPEO. If the conditions and assumptions are no longer valid and applicable provide the revised conditions and assumptions made to determine the validity of the disposition of the TLAA as 10 CFR 54.21(c)(1)(i) for the loads and the corresponding number of load cycles experienced thus far and estimated through the end of the SPEO.

FPL Response:

The responses to the individual requests are as follows:

1. Seismic forces were addressed in the design of the spent fuel bridge cranes, which were replaced in 1990. The "added seismic requirements" statement, mentioned in SLRA Section 4.7.6 for the spent fuel bridge cranes, refers to the acceleration response spectra specified by FPL to be included by the crane manufacturer in analyzing various dead load combinations. The "added seismic requirements" statement is not relevant in the context of the crane cycle fatigue TLAA and is deleted.

2. In the response to RAI 4.7.4-1 (included in Attachment 1 of the referenced FPL letter),

400 spent fuel bridge crane lifts (crane cycles) were assumed per refueling cycle. This total includes lifting 60 new fuel assemblies, a full core offload of 157 fuel assemblies, a full core reload of 157 fuel assemblies, and 24 miscellaneous fuel assembly shuffles, as well as rounding up by 2 (400=60+157+157+24+2).

Starting in 2011, implementing campaigns involving the PTN independent spent fuel storage installation (ISFSI) introduced new service requirements for the spent fuel bridge cranes. Current plans call for 600 additional crane cycles every three years (two refueling cycles) for an average of 300 more crane cycles each refueling cycle.

Combining the two estimates above yields a 700-crane cycle average per 18-:month refueling cycle since 2011. This estimate factors in previous statements of outages, fuel receipts, fuel inspections, insert shuffles, and cleaning. For the 25 refueling cycles from original plant startup to the beginning of 2011, each PTN spent fuel bridge crane has performed approximately 10,000 crane cycles (25 outages x 400 cycles).

Applying the revised conditions and assumptions to the remaining 28 refueling cycles anticipated between 2011 and the end of the SPEO in 2053 yields 19,600 total crane cycles (28 outages x 700 cycles). The total crane cycles per unit at the end of the SPEO is estimated to be 29,600 crane cycles (10,000 cycles+ 19,600 cycles= 29,600 cycles). This estimate is below the design maximum of 200,000 crane cycles.

With incorporation of the ISFSI campaigns into the estimate and no current plans to alter the 18-month refueling cycle timeline, the revised conditions and assumptions are expected to remain valid through the SPEO. Therefore, there is reasonable assurance that the actual number of spent fuel bridge crane cycles will not exceed the respective maximum allowed number of cycles.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.6-2 L-2018-193 Attachment 24 Page 3 of 3

References:

FPL Letter L-2001-75 to NRC dated April 19, 2001, Response to Request for Additional Information for the Review of the Turkey Point Units 3 and 4 License Renewal Application (ADAMS Accession No. ML011170195)

Associated SLRA Revisions:

SLRA Section 4.7.6 is amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revision.

Revise SLRA Section 4.7.6 paragraph 9 (paragraph under "Spent Fuel Bridge Cranes")

as follows:

The Spent Fuel Bridge Cranes were replaced in 1990. Design of the Spent Fuel Bridge Cranes was in accordance with CMAA-70 (Reference 4.7.7.17), with added seismic requirements.

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-1 L-2018-193 Attachment 25 Page 1 of 3 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

13. PWR Reactor Vessel Internals, GALL AMP XI.M16A Regulatory Basis:

Section 54.21 (a)(3) of states that for each structure and component identified in paragraph (a)(1) of this section, the applicant shall demonstrate that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the CLB for the subsequent period of extended operation.

RAI 8.2.3. 7-1

Background:

For each structure and component identified in 10 CFR 54.21 (a)(1 ), the applicant for subsequent license renewal (SLR) has the option to demonstrate compliance with 10 CFR 54.21 (a)(3), by including in the SLR application (SLRA) an aging management program (AMP) that is consistent with the applicable AMP described in NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, July 2017.

The Turkey Point SLRA Section B.2.3. 7 states that the reactor vessel internals (RVI) AMP with enhancements will be consistent with GALL-SLR Report AMP XI.M16A. (Note: The AMP enhancements are unrelated to this RAI.) The GALL-SLR Report AMP XI.M16A specifies that for existing RVI AMPs that are based on implementation of MRP-227-A inspection and evaluation guidelines, the guidelines are supplemented through a "gap analysis" that identifies changes to the AMP that are needed to address an 80-year operating period. Further, the GALL-SLR Report AMP "Scope of Program" element specifies that if the SLRA AMP is based on MRP-227-A with a gap analysis, the scope of the program focuses on identification and justification of the following:

a. RVI components that screen in for additional aging degradation mechanisms (DMs) when assessed for the 60-to-80-year operating period (SPEO);

b. RVI components that previously screened in for certain DMs, and the severity of these 60-year DMs could significantly increase for the 60-to-80 year SPEO;

c. Changes to the existing MRP-227-A program characteristics, including but not limited to changes in inspection categories, inspection criteria, or primary-to-expansion component criteria and relationships.

Issue:

Two Class 1 RVI components listed in SLRA Table 4.3-1, deep beam and lower support plate, are not listed in the SLRA gap analysis summary table, nor in SLRA Table 3.1.2-4 AMR results. One component, lower support plate to core barrel weld, shows an EPU CUF value that is inconsistent with the CUF for the lower core barrel flange weld in the gap analysis summary table. ("Note 5" in Attachment 4 of the gap analysis states that the

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-1 L-2018-193 Attachment 25 Page 2 of 3 lower core barrel flange weld and the support plate to core barrel weld are the same component.)

Request:

Please either justify or correct these apparent inconsistencies.

FPL Response:

The deep beam component listed in PTN SLRA Table 4.3-1 is a welded attachment to the upper support plate and is treated as a single component in Table 3.1.2-4 as well as throughout Appendix C. This is consistent with the component listing endorsed in the staff assessment (Reference 1 ). The lower support plate is a generic term which encompasses multiple Westinghouse RVI designs for a single component. The lower support forging is a specific term for the lower support plate. To clarify the component names and create consistency throughout the PTN SLRA, Table 4.3-1 and Appendix Care revised.

FPL submitted an affidavit on April 10, 2018 (Reference 2) for the proprietary information contained in the 'Associated SLRA Revisions' section below; accordingly, FPL continues to rely on that affidavit for this response.

References:

1. Turkey Point Nuclear Generating Unit Nos. 3 and 4 - Staff Assessment of License Renewal Commitment for Reactor Vessel Internals Implementation Report and Inspection Plan (CAC Nos. MF1485 and MF1486), December 18, 2015, ADAMS Accession No. ML15336A046

2. FPL Letter L-2018-082 to NRC dated April 10, 2018, Turkey Point Units 3 and 4 Subsequent License Renewal Application - Revision 1: Enclosure 2, Attachment 1 (ADAMS Accession No. ML18113A142)

Associated SLRA Revisions:

SLRA Table 4.3-1 and Appendix C Attachment 1 are amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

WITHHELD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-1 L-2018-193 Attachment 25 Page 3 of 3 Revise SLRA Table 4.3-1 as follows:

Table 4.3-1 PTN Unit 3 and Unit 4 60-Year Fatigue Cumulative Usage Factors for Reactor Coolant System Components Component Cumulative Usage Allowable Reactor Vessel Internals Upper support plat#!

[

](2) 1.0 Deep beamMl

[

](2) 1.0 Lower support plate!fil

[

](2) 1.0 Notes for Table 4.3-1

1. From UFSAR Table 4.3-2.

2. From fatigue analysis performed for the EPU (Reference 4.3.6.2) but not included in the UFSAR.

3. From UFSAR Table 4.3-2a.

4. The upper support plate and deep beam are treated as a single component for aging management purposes.

5. The lower support plate is also referred to as the lower support forging.

Revise "EPU CUF" value for the lower support plate in SLRA Appendix C Attachment 1 as follows:

COMPONENT MATERIAL Effect Structural Wear CUF2:

EPU Stress 2:

Weld Potential 0.1 CUF Threshold (30ksi)

Lower 304 ss No No No No I

I Support Forging Associated

Enclosures:

Westinghouse Letter CAW-17-4686 dated December 14, 2017, Application for Withholding Proprietary Information from Public Disclosure Preload Req'd No

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-1 L-2018-193 Attachment 25 Enclosure Page 1 of 8 Enclosure Westinghouse Letter CAW-17-4686 dated December 14, 2017 Application for Withholding Proprietary Information from Public Disclosure Westinghouse Affidavit CAW-17-4686 Proprietary Information Notice and Copyright Notice LTR-MRCDA-17-81-P, Rev. 3, "Requested Cumulative Fatigue Usage Factors from Turkey Point Unit 3 and Unit 4 EPU Licensing Report" (Proprietary)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Westinghouse Non-Proprietary Class 3 L-2018-193 Attachment 25 Enclosure Page2of8

@Westinghouse Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Document Control Desk Directtel: {412)374-4643 Direct fax: {724) 940-8542 11555 Rockville Pike Rockville, MD 20852 e-mail: greshaja@westinghouse.com CA W-17-4686 December 14, 2017 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

LTR-MRCDA-17-81-P, Rev. 3, "Requested Cumulative Fatigue Usage Factors from Turkey Point Unit 3 and Unit 4 EPU Licensing Report" (Proprietary)

The Application for Withholding Proprietary Infonnation from Public Disclosure is submitted by Westinghouse Electric Company LLC ("Westinghouse"), pursuant to the provisions of paragraph (b )( 1) of Section 2390 of the Nuclear Regulatory Commission's ("Commission's) regulations. It contains commerc~al strategic infonnation proprietary to Westinghouse and customarily held in confidence.

The proprietary infonnation for which withholding is being requested in the above-referenced report is further identified in Affidavit CA W-17-4686 signed by the owner of the proprietary infonnation, Westinghouse. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Florida Power & Light Company.

Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-17-4686 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 2 Suite 259, Cranberry Township, Pennsylvania 16066.

James A. Gresham, Manager Regulatory Compliance

© 2017 Westinghouse Electric Company LLC. All Rights Reserved.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 COMMONWEALTH OF PENNSYLVANIA:

COUNTY OF BUTLER:

AFFIDAVIT ss L-2018-193 Attachment 25 Enclosure Page 3 of 8 CAW-17-4686 L James A. Gresham, am authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC ("Westinghouse") and declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

Executed on: iA(r~}7

~~

Regulatory Compliance

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 3

L-2018-193 Attachment 25 Enclosure Page 4 of 8 CAW-17-4686 (1)

I am Manager, Regulatory Compliance, Westinghouse Electric Company LLC ("Westinghouse"),

and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2)

I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3)

I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or :financial information.

( 4)

Pursuant to the provisions of paragraph (b )( 4) of Section 2.3 90 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i)

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii)

The information is of a type customarily held in confide~ce by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that systt:m and the substance of that system constitute Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

( a)

The information reveals the distinguishing aspects of a process ( or component, structure, tool, method, etc.) where prevention of its use by any of

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 4

L-2018-193 Attachment 25 Enclosure Page5of8 CAW-17-4686 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b)

It consists of supporting data, including test data, relative to a process ( or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage ( e.g., by optimiz.ation or improved marketability).

( c)

Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

( d)

It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

( e)

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f)

It contains patentable ideas, for which patent protection may be desirable.

(iii)

There are sound policy reasons behind the Westinghouse system which include the following:

(a)

The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b)

It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

( c)

Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 5

L-2018-193 Attachment 25 Enclosure Page 6 of 8 CA W-17-4686

( d)

Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e)

Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f)

The Westinghouse capacity to invest corporate asse~s in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iv)

The information is being transmitted to the Commission in confidence and, under the provisions of IO CFR Section 2.390, is to be received in confidence by the Commission.

(v)

The information sought to be protected is not available in public sources or available infonnation has not been previously employed in the same original manner or method to the best of our knowledge and belie£ (vi)

The proprietary information sought to be withheld in this submittal is that which is appropriately marked in LTR-l\\.1R.COA-17-81-P, Rev. 3, "Requested Cumulative Fatigue Usage Factors from Turkey Point Unit 3 and Unit 4 EPU Licensing Report

(Proprietary), dated December 13, 2017 for submittal to the Commission, being transmitted by Florida Power & Light Company letter. The proprietary information as submitted by Westinghouse is that associated with Westinghouse's request for NRC approval ofLlll.:.~CDA-}7:-81-I>, and may be used only for that purpose.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 6

L-2018-193 Attachment 25 Enclosure Page 7 of 8 CAW-17-4686 (a)

This infonnation is part of that which will enable Westinghouse to provide a technical justification for acceptability of environmental assisted fatigue for various components for Turkey Point Units 3 and 4 in support of their subsequent license renewal program.

(b)

Further, this infonnation has substantial commercial value as follows:

(i)

Westinghouse plans to sell the use of similar information to its customers for the purpose of supporting other subsequent license renewal programs.

(ii)

Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications.

(iii)

The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluationjustifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 PROPRIETARY INFORMATION NOTICE L-2018-193 Attachment 25 Endosure Page8of8 Transmitted herewith are proprietaiy and non-proprietary versions of a document, furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary infonnation has been deleted in the non-proprietary versions, only the brackets remain (the infonnation that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means oflower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of infonnation being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(l).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessmy for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of IO CFR 2.390 regarding restrictions on public disclosure to the extent such infonnation has been identified as proprietmy by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietmy versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-2 L-2018-193 Attachment 26 Page 1 of 2 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI B.2.3.7-2 Issue:

Most of the RVI components evaluated in the gap analysis are not designated as "Class 1" RVI components, and as such they do not show a specific EPU CUF value. Many of these were generically screened as not susceptible to fatigue for the 60-year AMP based on the MRP-191 generic assessments because their CUF was generically determined to be less than 0.1. These components remain screened out for fatigue for 80-years. The gap analysis provides no apparent update to the EPU CUF screening results for these non-Class 1 RVI components to address an 80-year period.

Request:

For the RVI components evaluated in the gap analysis that screened out for fatigue for 60 and 80 years, and that do not have a specific EPU CUF listed in the gap analysis summary table, please address how the MRP-191 generic fatigue screening results for the 60-year AMP were determined to remain valid for 80 years.

FPL Response:

As di$cussed in PTN SLRA Section 4.3.1, the maximum number of fatigue cycles that Class 1 components, including the RVI, are exposed to during the 80-year SPEO is the same as the fatigue cycles assumed for the 60-year PEO. Since the RVI will not be exposed to fatigue cycles in excess of what was assumed for the 60-year PEO, the MRP-191 generic fatigue screening results for the 60-year AMP remain valid for 80 years.

The PTN Fatigue Monitoring program is credited with managing fatigue of Class 1 components by ensuring that the number of occurrences and severity of each design transient remains within the limits of the components design cycle limit. The PTN Fatigue Monitoring program provides for corrective actions when any applicable transient cycle count comes within 80 percent of the design or projected cycle limit. As stated above, the 80-year RVI fatigue cycles will not exceed those assumed for the 60 year PEO.

Additionally, the MRP-191 screening estimates of potential fatigue usage considered an environmental correction factor of 10 which is explicitly not required for RVI components per NUREG-2192 Section 4.3.3.1.2 as they are not a part of the reactor coolant pressure boundary nor are they included in NUREG/CR-6260. With the significant conservatism applied by considering an environmental correction factor of 10, the components which were previously identified as not susceptible to fatigue remain not susceptible to fatigue.

References:

None Associated SLRA Revisions:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-2 L-2018-193 Attachment 26 Page 2 of 2 Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-3 L-2018-193 Attachment 27 Page 1 of 2 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 8.2.3. 7-3 Issue:

SLRA Gap Analysis, Attachment 4, Primary Components - Control Rod Guide Tube (CRGT) Assembly Guide Cards: SLRA Gap Analysis, Attachment 4, including "Note 7" for Turkey Point Unit 4, appears to propose long term inspection of guide cards for Turkey Point 3 and Turkey Point 4 in accordance with MRP-227-A. It is not apparent to the NRC staff how the proposed program addresses interim inspection guidance for guide cards contained MRP Letter 2014-006, PWROG Letter OG-14-55, and WCAP-17451-P, Revision 1 that were issued under NEI 03-08.

\\

Request:

CRGT Guide Cards (Primary): Please address how the above SLRA Gap Analysis guide card inspections for Turkey Point 3 and Turkey Point 4, including Turkey Point 4 "Note 7," are consistent with the above-cited interim guidance, revise the gap analysis to be consistent with the guidance, or justify an alternate approach.

FPL Response:

Interim inspection guidance for the CRGT guide cards issued under NEI 03-08 is currently being managed under the existing Reactor Vessel Internals AMP as stated in PTN SLRA Section B.2.3.7. MRP 2016-006 contains both PWROG Letter OG-14-55 and a summary of WCAP-17451 as enclosures. These three documents are included as addendums in MRP-227-A, which establishes the baseline for the PTN SLRA Appendix C Gap Analysis and Reactor Vessel Internals AMP. The Unit 3 inspections showed satisfactory results which allow for the inspection frequency to remain unchanged. At this time, there is not sufficient examination evidence to define a long-term inspection plan for PTN Unit 4 CRGT guide cards. However, the PTN CRGT guide cards inspection interval is currently augmented, as stated in Note 7 of PTN SLRA Appendix C Attachment 4, consistent with the guidance ofWCAP-17451-P. Note 7 of PTN SLRA is revised to clarify the applicablility of WCAP-17451-P to both Units 3 and 4. PTN will implement an inspection interval for the CRGT guide cards informed by inspection results following completion of the augmented inspections. The PTN SLRA Appendix C is revised to clarify the inspection plan.

References:

None Associated SLRA Revisions:

The following changes to SLRA Appendix C will be made in a future SLRA revision as indicated by text deletion (strikethrough) and text addition (red underlined font).

See RAI B.2.3.7-4 for revisions to SLRA Appendix C.

l

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-3 L-2018-193 Attachment 27 Page 2 of 2 Revise SLRA Appendix C Attachment 4 Notes as follows:

Notes:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Attachment 6.

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

3. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Attachment 6, must be examined for inspection credit.

4. A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Attachment 6, must be examined from either the inner or outer diameter for inspection credit.

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

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

7. Unit 4 inspection frequency changed by corrective action to approximately 9 years based on 8.8 EFPY accumulated and from 20% to 100% examination of thes number of CRGT assemblies based on the inspections performed spring 2016 in accordance with WCAP-17451-P (AR 02124311 ). After the expanded inspection coverage is complete, the inspection frequency and examination coverage will revert to the 10 year interval and 20% inspection coverage, barring further corrective actions. This does not apply to Unit 3. The existing Unit 3 inspection frequency is adequate based on satisfactory inspection results.

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-4 L-2018-193 Attachment 28 Page 1 of 2 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI 8.2.3. 7-4 Issue:

SLRA Gap Analysis, Attachment 4, Primary Components - Baffle-Former Bolts: of the gap analysis shows that inspection criteria for baffle-former bolts are the same as those in the original MRP-227-A. It is not apparent to the NRC staff how the proposed program addresses interim inspection guidance for baffle-former bolts contained in MRP Letter 2017-009 and issued under NEI 03-08.

Request:

Baffle-Former Bolts (Primary): Please address how the above SLRA Gap Analysis baffle-former bolt inspections are consistent with the above-cited interim guidance, revise the gap analysis to be consistent with the guidance, or justify an alternate approach.

FPL Response:

Interim inspection guidance for baffle-former bolts contained in MRP Letter 2017-009 and issued under NEI 03-08 is currently being managed under the corrective action program as stated in PTN SLRA Section B.2.3.7. At this time, there is not sufficient evidence to define a long-term inspection plan which deviates from the existing guidance in MRP-227-A. However, the inspection schedule was augmented by the corrective action program as described in PTN SLRA Section B.2.3.7. The corrective action program will implement a long-term inspection interval for the baffle-former bolts informed by inspection results following completion of the augmented inspections.

References:

None Associated SLRA Revisions:

SLRA Appendix C is amended as indicated by the following text deletion (strikethrough) and text addition (red underlined font) revisions.

Revise SLRA Appendix C Section C.2.7 as follows C.2.7 Categorize for Inspection (Primary, Expansion, Existing, No Additional Measures) and Aging Management Strategy The seventh step is to assign components into Primary, Expansion, Existing Programs, and No Additional Measures groups. These assignments are based on the results of all preceding efforts. While five additional components screened in for fatigue and five additional components screened in for fluence related degradation, this only impacted the FMECA score for one component (upper support plate), and did not impact the severity categorization for any components.

The industry OE concerning observed wear of the fuel alignment pins and the resulting increase in FMECA and severity rankings is considered significant enough to elevate the

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRG RAI No. B.2.3.7-4 L-2018-193 Attachment 28 Page 2 of 2 fuel alignment pins from No Additional Measures to Existing Programs Components.

Visual (VT-3) inspections will be conducted under the ASME Section XI Program during the 80-year period of operation.

&#l-t!he clevis insert and baffle bolting, for which there has been industry OE, wefeWas categorized as Existing Programs Components for the 60-year period of operation. This categorization remains unchanged for the 80-year period of operation and the componentsclevis insert bolting will continue to be inspected under the Section XI Program.

The CRGT guide cards and baffle-former bolting inspection frequency is currently augmented as stated in Section 8.2.3.7. The long-term inspection frequency of the CRGT guide cards is defined as 20% examination coverage on a ten year interval.

The long-term inspection frequency of the baffe-former bolts is 100% of accessible bolting on a ten year interval. However, both of these RVI component inspection frequencies are subject to change pending the results of the corrective action program inspections to create an inspection interval informed by plant-specific evaluations.

The table in Attachment 2 outlines the new component categorization and aging management strategy. With the exception of the fuel alignment pins, the table is consistent with the component categorization and aging management strategy approved in Reference C.4.6. The details regarding the expansion links, examination method, and examination frequency are presented in Attachments 3,4, and 5 for Existing, Primary, and Expansion categories respectively.

Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-5 L-2018-193 Attachment 29 Page 1 of 2 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI B.2.3.7-5 Issue:

SLRA Gap Analysis, Attachment 3, Existing Programs Components - Alignment and Interfacing Components, Clevis Insert Bolts: The clevis insert bolt inspection criteria are listed in Attachment 3 of the gap analysis. This includes some new "Note 2" inspection guidance to look for degradation at clevis insert and radial key interfaces and evidence of degraded bolting components. However, SLRA Section B.2.3. 7 states that the clevis insert bolts are already categorized as a "Primary" component.

Request:

Clevis Insert Bolts (Existing Programs): Given the industry OpE with clevis insert bolt degradation and the statement in SLRA Section B.2.3. 7 that they are "Primary" components, please address whether EVT-1 or UT exams for this bolting should be included with the Primary Components inspections in Attachment 4 of the gap analysis, or address whether the current "Note 2" guidance in Attachment 3 is adequate for the bolting. Please amend the clevis insert bolting OpE discussion in SLRA Section B.2.3.7 and/or the SLRA Gap Analysis Attachment 3 inspection table to ensure they are consistent.

FPL Response:

The clevis insert bolts are categorized as an "Existing Programs" component in PTN SLRA Section C.2.7. SLRA Section B.2.3.7 is revised to be consistent with Appendix C.

As such, the current PTN SLRA Appendix C Attachment 3 examination method and frequency is adequate for the clevis insert bolts.

References:

None Associated SLRA Revisions:

The "Clevis Insert Bolting" discussion on page B-90 of SLRA Section B.2.3.7 is amended as indicated by the following text deletion (strikethrough).

Clevis Insert Bolting EPRI released industry letter MRP-2017-024, Clevis Insert Bolt OE. This letter identified degradation of the clevis insert bolting at Salem Unit 2 during its spring 2017 outage. This issue was documented and evaluated as having no immediate impact to PTN in the CAP.

This operating experience was considered in the reactor vessel internals gap analysis based on the evidence of an active degradation mechanism. However, the clevis insert bolts are already categorized as a "Primary" component, and recognized to have a high failure likelihood. At this point, there are no further steps to be taken to address clevis insert bolts. Any unsatisfactory inspection results will be addressed through the CAP.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-5 L-2018-193 Attachment 29 Page 2 of 2 Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-7 L-2018-193 Attachment 30 Page 1 of 2 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018 RAI B.2.3.7-7 Issue:

SLRA Gap Analysis, Attachment 1, 60-to-80 Year Results, No Additional Measures Components: Attachment 1 provides a summary of the 60-to-80 year OM screening, FMECA scoring, and categorization of the Turkey Point RVI components. Several of the components with new DMs that screen in for 80 years are to remain "No Additional Measures" components for SPEO. The staff noted, in particular, that the upper support column bolting shows new 80- year DMs of IASCC and IE, which is in addition to the 60-year DMs of wear, fatigue, and irradiation-induced stress relaxation (ISR).

Request:

Upper Support Column Bolting (No Additional Measures Components): Considering the new DMs of IASCC and IE that screen in for 80 years, please address why the upper support column bolting shows no change in the FMECA score ("L,M, 1 ") and Severity Category of "A" for 60-to80 years and why no additional action (inspections) are required, or revise the analysis to account for the additional potential for degradation.

FPL Response:

In accordance with PTN SLRA Table 3.1.2-4, cracking and loss of material for the upper support column bolting are managed by the ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD and the Water Chemistry AMPs. The inspection history as an ASME Section XI component paired with a lack of any observed degradation in domestic Westinghouse designed PWRs is determined to be sufficient evidence to maintain a low failure likelihood for the bolting. Consistent with the Categorization and Ranking process defined by the referenced MRP-191 document, a low failure likelihood and medium failure consequence places upper support column bolting in the "No Additional Measures" inspection category. However, as this component will continue to be inspected within the interval as a part of the ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD AMP, plant specific operating experience will continue to be gathered. Any plant specific or industry operating experience is tracked and addressed through PTNs corrective action program. If future plant operating experience or industry operating experience do not support the current low failure likelihood of the upper support column bolting, the corrective actions program will address the inconsistency to ensure the bolting remains appropriately managed in the Reactor Vessel Internals AMP.

References:

EPRI Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191), Rev. 1, October 2016

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. B.2.3.7-7 L-2018-193 Attachment 30 Page 2 of 2 Associated SLRA Revisions:

None Associated

Enclosures:

None

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-193 Proprietary Attachments 12P, 13P and 25P End of Non-Proprietary Attachments Proprietary Attachments 12P, 13P and 25P are Inserted Beginning on the Following Page

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.5-9 L-2018-193 Attachment 12P Page 1 of 3 NRC RAI Letter Nos. ML18269A227 and ML18269A228 Dated October 04, 2018

9. Reactor Coolant Pump Integrity Analysis, GALL TLAA 4.7 Regulatory Basis:

Section 54.21 (c) of 10 CFR requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis (CLB) for the subsequent period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to the managing the effects of aging during the subsequent period of extended operation (SPEO) on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the subsequent renewed license will continue to be conducted in accordance with the CLB. As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21 (a)(3) by referencing the GALL-SLR Report. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

Background:

The regulation in 10 CFR 54.21 (c)(1 )(ii) states that, for a specific time limited aging analyses (TLAA) that is dispositioned in accordance with this regulation, the applicant must demonstrate that the analysis has been projected to the end of the SPEO.

Subsequent license renewal application (SLRA) Section 4.7.5, "Code Case N-481 Reactor Coolant Pump Integrity Analysis," identifies the examination reactor coolant pump (RCP) casing in the current licensing basis as a TLAA item.

In 2000, the applicant submitted for NRC review and approval the 60-year license renewal application. As part of that application, the applicant performed a reactor coolant pump (RCP) integrity analysis for Turkey Point Units 3 and 4 as documented in Westinghouse topical reports, WCAP-13045 and WCAP-15355. To demonstrate continued compliance during SPEO, the Pressurized Water Reactor Owner's Group (PWROG) re-evaluated WCAP-13045 associated with the application of Code Case N-481 to the RCP casing during the SPEO as documented in PWROG-17033, Revision 0.

The applicant submitted the topical report PWROG-17033, Revision Oas part of the SLRA.

RAI 4.7.5-9 Issue:

Section 2.2 of PWROG-17033 discusses fracture toughness calculation based on NUREG/CR4513, Revision 2. The NRC staff notes that Aging Management Program XI.M12, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel, in Generic Aging

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.5-9 L-201 S-193 Attachment 12P Page 2 of 3 Lessons Learned for Subsequent License Renewal (GALL-SLR) report, NUREG-2191, Volume 2, discusses fracture toughness values based on the prediction method in NUREG/CR-4513, Revision 1. The GALLSLR report does not reference Revision 2 of NUREG/CR-4513.

Request:

Discuss whether the saturated fracture toughness value used in the crack stability analysis of pump casing at Turkey Point would still be limiting and bounding if the method of predicting fracture toughness in accordance with NUREG/CR-4513, Revision 1 was used. That is, regardless whether the method in revision 1 or revision 2 of NUREG/CR-4513 was used, the fracture toughness value u_sed in the crack stability analysis at Turkey Point would still be limiting.

FPL Response:

Section 4.2 of WCAP-15355 provides the saturated fracture toughness values (see Table

4. 7.5-9.1 below for convenience) that were derived using the methodology of NUREG/CR-4513, Revision 1, considering the limiting Turkey Point CFS material chemistry. The NUREG/CR-4513, Revision 1 values in Table 4.7.5-9.1 can be used for comparison to the generic fracture toughness values used in the stability criteria per Table 5-2 ofWCAP-15355 (provided in Table 4.7.5-9.2 below for convenience) for the Turkey Point pump casings. The fracture toughness values in Table 4.7.5-9.2 were reported in WCAP-13045 for the Model 93 pump casing considering CFS material.

Finally, Table 4.7.5-9.3 provides the fracture toughness values reported in the response to RAI 4.7.5-3(d) for the Model 93 pump casing using the methodology of NUREG/CR-4513, Revision 2, also considering the limiting Turkey Point CFS material chemistry. As shown by comparing Tables 4.7.5-9.1, 4.7.5-9.2, and 4.7.5-9.3 below, the saturated fracture toughness values derived using NUREG/CR-4513, Revision 1 (Table 4.7.5-9.1) and Revision 2 (Table 4.7.5-9.3) are higher than the generic CFS material fracture toughness values reported in WCAP-13045 (Table 4.7.5-9.2) and used for Turkey Point.

Therefore, the fracture toughness values used in the crack stability analyses of the pump casing for Turkey Point, as reported in WCAP-15355, are limiting and more bounding than the fracture toughness values determined by either Revision 1 or Revision 2 of NUREG/CR-4513 for the Turkey Point CFS material chemistry.