ML18269A210

Requests for Additional Information for the Safety Review of the Turkey Point Subsequent License Renewal Application - Set 4 (EPID L-2018-RNW-0002) - Enclosure

ML18269A210
Person / Time
Site:	Turkey Point
Issue date:	10/01/2018
From:	Lois James NRC/NRR/DMLR/MRPB
To:	Nazar M Florida Power & Light Co
Lois James
Shared Package
ML18269A208	List: ML18269A209 ML18269A210
References
EPID L-2018-RNW-0002
Download: ML18269A210 (21)

Text

TURKEY POINT NUCLEAR GENERATING UNITS 3 AND 4 (TURKEY POINT)

SUBSEQUENT LICENSE RENEWAL APPLICATION (SLRA)

REQUESTS FOR ADDITIONAL INFORMATION (RAIS)

SAFETY - SET 4

1. Leak-Before-Break RCS Pipe, TLAA 4.7 Regulatory Basis:

Title 10 of the Code of Federal Regulations (10 CFR) Section 54.21(c) 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 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 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 current licensing basis. As described in NUREG-2192, Rev. 0, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, dated July 2017 (SRP-SLR), an applicant may demonstrate compliance with 10 CFR 54.21(a)(3) by referencing the applicable aging management programs and activities in the UREG-2191, Rev. 0, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, dated July 2017. 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 analysis (TLAA) that is dispositioned in accordance with this regulation, the applicant must demonstrate that the analysis has been projected to the subsequent end of the period of extended operation.

Subsequent license renewal application (SLRA) Section 4.7.3, Leak-Before-Break Analysis for Reactor Coolant System Piping, identifies the leak-before-break (LBB) analysis in the current licensing basis as an analysis that meets the definitions of a TLAA for the SLRA.

As part of the SLRA, the applicant evaluated the LBB analysis of reactor coolant system (RCS) primary loop piping as documented in WCAP-15354, Revision 1, Technical Justification for Eliminating Primary Loop Pipe Rupture as a Structural Design Basis for Turkey Point Units 3 and 4 Nuclear Power Plants for the Subsequent License Renewal Time-Limited Aging Analysis Program (80 Years) Leak-Before-Break Evaluation, September 2017.

The cast austenitic stainless steel material in RCS piping may be affected by thermal embrittlement during the subsequent period of extended operation. In addition, fatigue crack growth calculation for the reactor coolant system piping is part of the TLAA.

Enclosure

RAI 4.7.3-1 Issue:

In Section 8 of WCAP-15354, Revision 1, the applicant analyzed fatigue crack growth for the postulated circumferential flaw. The staff understands that typically an axial flaw is not limiting in LBB analyses; therefore, fatigue crack growth would not be performed for the axial flaw.

However, WCAP-15354 does not clearly state that an axial flaw is not limiting for the applicants LBB analysis.

Request Discuss whether fatigue crack growth was performed for a postulated axial flaw. In addition, discuss whether an axial flaw is not limiting in terms of pipe rupture such that fatigue crack growth is not needed for the axial flaw.

RAI 4.7.3-2 Issue:

WCAP-14237 is the original LBB analysis for the primary loop piping. Table 7-1 in WCAP-14237 lists fracture toughness values Jlc and Jmax of locations 2 and 11 of the primary loop piping. Table 4-5 of WCAP-15354 also lists Jlc and Jmax of location 2 (hot leg) and location11 (cold leg). The staff noted that there are some discrepancies between WCAP-14237 and WCAP-15354. Specifically, the Jlc values in WCAP-15354 are different than those in WCAP-14237. The Jmax values in WCAP-15353 are also different than that in WCAP-14237.

Request Discuss why there are differences in Jlc and Jmax between WCAP-14237 and WCAP-15353. As part of the response, discuss specifically why Jmax values in WCAP-15353 are different than those in WCAP-14237.

RAI 4.7.3-3 Issue:

On page 8-2 of WCAP-15354, the applicant stated that the calculated fatigue crack growth for semi-elliptic surface flaws of circumferential orientation and various depths is summarized in Table 8-2. However, it is not clear from Table 8-2 whether the final flaw size is in terms of the though-wall crack depth or circumferential crack length.

Request:

Clarify whether the final flaw size derived in Table 8-2 of WCAP-15354 is the depth of the pipe wall thickness, or the length in the circumferential direction.

RAI 4.7.3-4 Issue:

During the staff audit of the applicants documents, the staff noticed that document Action Report (AR) 01610224 is related to errors in pipe stress software.

Request Discuss whether errors in pipe stress software as discussed in AR 01610224 affected the applied loads and stresses used in the LBB analysis of reactor coolant piping.

RAI 4.7.3-5 Issue:

The staff noted that Electric Power Research Institute (EPRI) topical report Materials Reliability Program: Assessment of Residual Heat Removal Mixing Tee Thermal Fatigue in PWR

[Pressurized Water Reactors] Plants (MRP [Materials Reliability Program]-192, Revision 2),

August 2012, and Materials Reliability Program: Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines (MRP-146, Revision 1),

June 2011 are related to thermal fatigue of safety-related piping.

Request:

Discuss whether the RCS primary loop piping at Turkey Point Units 3 and 4 is subject to the thermal fatigue as discussed in MRP-146 and MRP-192. If yes, discuss whether RCS primary loop piping satisfies the LBB screening criteria as specified in Standard Review Plan 3.6.3 of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. Discuss whether thermal fatigue data are included in the fatigue crack growth calculations in WCAP-15354, Revision 1. If yes, provide examples of the thermal fatigue data (transient loading).

RAI 4.7.3-6 Issue:

The elbows in the RCS primary loop piping are made of cast austenitic stainless steel material which is susceptible to thermal embrittlement when the component is placed in a long term service. In SLRA Section 4.7.3, the applicant used the method in NRC document NUREG/CR-4513, Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems, Revision 2, to predict the fully aged fracture toughness values for the elbows at the end of 80 years. SLRA Section 4.7.5 discusses thermal embrittlement of the reactor coolant pump casing which is made of cast austenitic stainless steel material. In SLRA Section 4.7.5, the applicant used NUREG/CR-4513, Revision 2 and Westinghouse report,

WCAP-13045, to predict the fully aged fracture toughness values for the reactor coolant pump casing.

Request:

Discuss whether the fracture toughness data in WCAP-13045 as discussed in Section 4.7.5 are applicable to the elbows in the RCS primary piping as discussed in Section 4.7.3. If yes, discuss whether the fracture toughness values used for the elbows in the RCS primary piping in the LBB analysis in Section 4.7.3 are the lowest values (i.e., most limiting) based on the data in WCAP-13045 and NUREG/CR-4513, Revision 2.

2. Reactor Vessel Neutron Embrittlement Analyses, TLAA 4.2 Regulatory Basis:

Section 54.21(c)(1) of 10 CFR states that a list of time-limited aging analyses, as defined in 10 CFR 54.3, must be provided, and that 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 order to verify that TLAAs for pressurized thermal shock (PTS) and adjusted reference temperature (ART) have been conservatively projected in accordance with 10 CFR 54.21(c)(1)(ii), the staff requires additional information as detailed below.

RAI 4.2-1

Background:

For the pressurized thermal shock (PTS), SRP-SLR Section 4.2.2.1.3 references 10 CFR 50.61 TLAA, which requires that the RTPTS values be updated when there is a change in the expiration date of a plants operating license. Therefore, the SRP-SLR states the RTPTS values must be calculated for the subsequent period of extended operation.

The applicant described its evaluation of the PTS TLAA in SLRA Section 4.2.2. The applicant dispositioned the PTS TLAA as projected through the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii). Among the input data for calculation of RTPTS is the unirradiated reference temperature, RTNDT(u), and the standard deviation of RTNDT(u),

designated U. SLRA Tables 4.2.2-1 and 4.2.2-2 provide the input data and results of the RTPTS calculations for 72 EFPY.

The staff also notes that the same RTNDT(u) and U values are used in the calculation of ART in SLRA Tables 4.2.4-1 and 4.2.4-2, and that this RAI is also applicable to the ART TLAA discussed in SLRA Section 4.2.4.

Issue:

The unirradiated reference temperature (RTNDT(u)) and the standard deviation of RTNDT(u) (U) values for certain reactor pressure vessel (RPV) materials, and the ART values for TP3 and TP4 reported in the SLRA have changed compared to those used in the current PTS analysis of record, both of which are contained in the Turkey Point Units 3 and 4 Extended Power Uprate (EPU) Licensing Report,1 , which is Attachment 4 to the license amendment request for an EPU2. The license amendment request for EPU was approved via a license amendment dated June 15, 2012.3 The tables below compare the changes from the EPU to the SLRA for the two parameters.

The SLRA did not provide references to source documents for the revised values of RTNDT(u) and U. The staff noted that some of the revised values result in lower values of RTPTS and ART. Therefore, it is necessary for the staff to verify the revised RTNDT(u) and U are accurate in order to assess the applicants disposition for the PTS and ART TLAAs in accordance with 10 CFR 54.21(c)(1)(ii).

Request:

For the materials listed in Table 1 and Table 2, the staff requests that the applicant:

a) Justify the discrepancy between the SLRA1 (Table 4.2.2-1 for Unit 3 and Table 4.2.2-2 for Unit 4) and the EPU licensing report2 (Table 2.1.2-3 for Unit 3 and Table 2.1.2-4 for Unit 4) for the RTNDT(u) values and u.

b) Describe how the RTNDT(u) values and u reported in the SLRA were determined, including a description of the data set.

3. Metal Fatigue of Class 1 Components, TLAA 4.3.1 Regulatory Basis:

Section 54.21(c)(1) of 10 CFR states a list of time-limited aging analyses, as defined in Section 54.3, must be provided. 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.

1 Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 4; Licensing Report, December 14, 2010 (ADAMS Accession No. ML103560204) 2 FPL Letter No. L-2010-113 from Michael Kiley to U.S. Nuclear Regulatory Commission, Turkey Point Units 3 and 4, Docket Nos. 50-250 and 50-251, License Amendment Request for Extended Power Uprate (LAR 205),

October 21, 2010, (ML103560167) 3 Jason C. Paige (NRC) letter to Mano Nazar (FPL), "Turkey Point Units 3 and 4 -Issuance of Amendments Regarding Extended Power Uprate," June 15, 2012 (ML11293A365)

RAI 4.3.1-1

Background:

As discussed in SLRA Section 4.3.1, the applicant dispositioned the TLAA for American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Class 1 fatigue calculations, in accordance with 10 CFR 54.21(c)(1)(i), that the analyses remain valid for the subsequent period of extended operation (SPEO). The applicant stated that the results demonstrate that the number of assumed design cycles will not be exceeded in 80 years of plant operation. The applicant also stated that it will monitor design cycles using the Fatigue Monitoring Program and assure that corrective action specified in the program is taken if any of the actual design cycles approach 80 percent of their analyzed numbers during the SPEO.

Issue:

The staff noted that the applicant is relying on its Fatigue Monitoring Program to monitor design cycles to ensure that these fatigue TLAAs remain valid for the SPEO in accordance with 10 CFR 54.21(c)(1)(i). Thus, the staff identified that the applicants disposition for these TLAAs (i.e., 10 CFR 54.21(c)(1)(i)) is not applicable because the Fatigue Monitoring Program is managing fatigue by ensuring these analyses to continually remain valid and that the ASME Code design limit will not be exceeded during the SPEO.

Request:

• If the disposition for these fatigue TLAAs remains in accordance with 10 CFR 54.21(c)(1)(i) - the staff requests the following:

• Based on the weighted projection methodology discussed in the SLRA and Calc 1700109.402P.R4, the staff noted that the weighted projection method was not applied to all transients to determine 80-year cycles.

o Discuss the method used for these transients and justify that it is conservative for determining 80-year projected cycles and supports the disposition of 10 CFR 54.21(c)(1)(i).

• SLRA Table 4.3-2 and SLRA Table 4.3 foot note 12 - no additional design cycles expected o Provide the basis that the Hydrostatic pressure tests (pressurized to 1356 psig) transient in the secondary Coolant system is no longer expected during the SPEO.

• Otherwise, if the Fatigue Monitoring Program is managing fatigue during the SPEO, justify that the disposition for these TLAAs in accordance with 10 CFR 54.21(c)(1)(i) is appropriate when compared to 10 CFR 54.21(c)(1)(iii).

4. Metal Fatigue of Non-Class 1 Components, TLAA 4.3.2 Regulatory Basis:

Section 54.21(c)(1) of 10 CFR states a list of time-limited aging analyses, as defined in Section 54.3, must be provided. 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.

RAI 4.3.2-1

Background:

SLRA Section 4.3.2 states a review of the American National Standards Institute (ANSI) B31.1, Power Piping, piping within the scope of SLR was performed to identify those systems that operate at elevated temperature and to establish a conservative number of projected cycles based on 80 years of operation. The applicant cited EPRI Report TR-104534, Fatigue Material Handbook Volume 2, Section 4, and indicated that carbon steel systems or portions of systems with operating temperatures less than 220°F and stainless-steel systems or portions of systems with operating temperatures less than 270°F may generally be excluded from such concerns, since room temperature represents a practical minimum exposure temperature for most plant systems.

SLRA Section 17.3.3.2 states that any system or portions of systems with operating temperatures less than 220°F were conservatively excluded from further consideration for fatigue. [emphasis added]

Issue:

Analyses that meet the definition of a TLAA in 10 CFR 54.3 are required to be identified in the SLRA in accordance with 10 CFR 54.21(c)(1). Based on SLRA Section 4.3.2 and specifically on Section 17.3.3.2, it appears that the applicant may have implemented the methodology described in EPRI Report TR-104534 to exclude systems or portions of systems from consideration in the SLRA as a TLAA.

Request:

• Confirm that this screening criteria was not used to exclude systems or components designed for fatigue from consideration as a TLAA in the SLRA.

o If it was not used in this way, discuss how the screening criteria was used in the SLRA and explain how this is in accordance with 10 CFR 54.21(c)(1).

o If it was used in this way, justify that exclusion of these systems and components from consideration in the SLRA is in accordance with 10 CFR 54.21(c)(1).

Otherwise, identify the systems or portions of systems designed for fatigue that were excluded and evaluate them in accordance with 10 CFR 54.21(c)(1).

RAI 4.3.2-2

Background:

SLRA Section 4.3.2 states that for the systems that are subjected to elevated temperatures above the fatigue threshold, a calculation was performed to determine a conservative number of projected full temperature cycles for 80 years of plant operation for the piping, tubing and in-line components. SLRA Table 4.3.2-2 provides the 80-year projected number of full temperature cycles for each of the systems evaluated in SLRA Section 4.3.2.

Issue:

The staff noted that the applicant provided the total number of projected transients through the end of the SPEO but did not provide the details on how these projections were determined (e.g.,

which transients are represented by the projections and how the projections were determined).

The applicant did not explain or provide the basis for how these 80-year projections are conservative for 80 years of plant operation and how it supports the disposition of this TLAA in accordance with 10 CFR 54.21(c)(1)(i).

Request:

• Explain how the number of projected full temperature cycles for 80 years of plant operation for the piping, tubing and in-line components identified in SLRA Table 4.3-2 were determined.

• Justify that these methods are conservative for demonstrating that the ASME Section III, Class 3 and ANSI B31.1 allowable stress calculations remain valid for the SPEO and that the number of assumed thermal cycles will not be exceeded in 80 years of plant operation in accordance with 10 CFR 54.21(c)(1)(i).

RAI 4.3.2-2

Background:

SLRA Section 4.3.2 states that the design thermal cycle limit for the reactor coolant system B hot leg tubing (i.e., less than 14,000 full temperature cycles) could be reached at the end of 2018 based on current operation. In order to ensure that the tubing can continue to perform its function for the current period of extended operation (PEO) as well as the SPEO, the applicant identified that one of the four actions in the SLRA can be completed.

The applicant dispositioned the TLAA for the reactor coolant system B hot leg tubing in accordance with 10 CFR 54.21(c)(1)(i) and stated that the results demonstrate that the number of assumed thermal cycles will not be exceeded in 80-years of plant operation.

Issue:

As noted in the SLRA, based on the cycle accumulation for the reactor coolant system B hot leg tubing the applicant expects that the design thermal cycle limit can be reached at the end of 2018. Thus, the staff noted that the applicant has not demonstrated that the TLAA for the

RCS B hot leg sample tubing will remain valid for the SPEO in accordance with 10 CFR 54.21(c)(1)(i) since the number of design cycles will be exceeded prior to the end of the SPEO.

Request:

• Considering that the design cycle limit is expected to be reached by the end of 2018, justify the disposition for the RCS B hot leg sample tubing TLAA in accordance with 10 CFR 54.21(c)(1)(i).

• Otherwise, disposition the TLAA for the RCS B hot leg sample tubing in accordance with 10 CFR 54.21(c)(1)(ii) or (iii) and demonstrate accordingly:

o If 10 CFR 54.21(c)(1)(ii) is selected - Provide the details of the design code allowable stress range and stress range reduction factor. Furthermore, justify that the projected number of cycles through the SPEO will be less than the revised allowable number of equivalent full temperature cycles.

o If 10 CFR 54.21(c)(1)(iii) is selected - Describe and justify the method that will be used to manage fatigue of the RCS B hot leg sample tubing during the SPEO.

5. Environmentally-Assisted Fatigue, TLAA 4.3.3 Regulatory Basis:

Section 54.21(c)(1) of 10 CFR states a list of time-limited aging analyses, as defined in Section 54.3, must be provided. 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.

RAI 4.3.3-1

Background:

During its audit, the staff reviewed the licensees refined calculations for environmentally assisted fatigue (EAF) documented in Enclosure 4 and 5 of the SLRA. The staff noted that the licensee used the methodology in the Draft Report for Comment version of NUREG/CR-6909, Effect of LWR Water Environments on the Fatigue Life of Reactor Materials, Revision 1, dated March 2014.

GALL-SLR AMP X.M1 and the SRP-SLR states, in part, that environmental effects on fatigue for these critical components may be evaluated using the guidance in Regulatory Guide (RG) 1.207, Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components Due to the Effects of the Light-Water Reactor Environment for New Reactors, Revision 1. RG 1.207, Revision 1, which was issued in June 2018, recommends the use of NUREG/CR-6909, Revision 1 (May 2018).

Issue:

The staff noted that due to the timing of the SLRA, the applicant used the most up-to-date methodology available, which was documented in the Draft Report for Comment version of NUREG/CR-6909, Rev. 1, dated March 2014. However, the use of this draft report is not consistent with the recommendations in the GALL-SLR and SRP-SLR.

Request:

o Qualitatively, discuss and justify the impacts, if any, to the refined calculations for EAF due to the recent issuance of RG 1.207, Revision 1.

o Otherwise, justify that the use of the Draft Report for Comment version of NUREG/CR-6909, Rev. 1, dated March 2014, is more conservative when compared to RG 1.207, Revision 1.

RAI 4.3.3-2

Background:

Section 3.4 of Report No. 1700109.401P.R5 indicates that 14 locations have an environmentally adjusted cumulative usage factor (CUFen) value greater than 1.0 when using the ASME Code fatigue curves of record for each location. Of these 14 locations, the following were not addressed in SLRA 4.3.3: the Steam Generator (SG) Tube to Tubesheet weld, RPV Head Flange and SG Primary Chamber, Tubesheet and Stub Barrel Complex.

Per the SLRA, the SG Tube to Tubesheet weld is no longer part of the reactor coolant pressure boundary since the applicant has a permanently approved H* alternate repair criteria for both the hot- and cold-leg side of the steam generator; thus, the staff noted this component would not be subject to further EAF assessment consistent with the SRP-SLR.

Issue:

The rationale for the inconsistency between Report No. 1700109.401P.R5 and the SLRA is not clear for the RPV Head Flange and SG Primary Chamber, Tubesheet and Stub Barrel Complex.

Request:

• Explain and justify the inconsistency between SLRA Section 4.3.3 and Report No. 1700109.401P.R5 for the RPV Head Flange and SG Primary Chamber, Tubesheet and Stub Barrel Complex.

RAI 4.3.3-3

Background:

SLRA Section 4.3.3 states the following with regard to the refined CUFen calculations:

• Reactor Vessel Shell at Core Support Pads - A revised CUFen was calculated by crediting 80-year projected design cycles for the hydrostatic test at 2485 psig pressure and 400oF temperature. CUFen Final = 0.910

• Pressurizer Upper Head - A revised CUFen was calculated by crediting 80-year projected design cycles for plant loading, unloading, and boron concentration equalization transients. CUFen Final = 0.974 The applicant dispositioned these refined CUFen calculations in accordance with 10 CFR 54.21(c)(1)(iii), such that the effects of aging on the intended function(s) of these components will be adequately managed for the period of extended operation.

Issue:

SLRA Section 4.3.3 and Report No. LTR-SDA-II-17-13-P, Revision 2, indicates that the CUFen Final is applicable for both Units 3 and 4, and that 80-year projected cycles were used for certain transients.

Based on SLRA Table 4.3-2 and 4.3-3, the 80-year projected cycles are different between Units 3 and 4; thus, its not clear which 80-year cycles were used or whether the revised CUFen results are applicable to both units. This information is necessary to ensure that the Fatigue Monitoring Program incorporates the appropriate cycle limits and can adequately manage environmentally assisted fatigue during the SPEO Request:

• Confirm that the 80-year projected cycles used for the transients in the refined CUFen analyses for the Reactor Vessel Shell at Core Support Pads and, the Pressurizer Upper Head is the larger number of cycles between the two units. If not, justify that the CUFen value is applicable to Units 3 and 4.

RAI 4.3.3- 4

Background:

SLRA Section 4.3.3 states that for the Pressurizer Spray Nozzle, a revised CUFen was calculated by performing a finite element fatigue calculation using the methodology of Subarticle NB-3200 of Section III of the ASME Code and projected design cycles for plant heatup and cooldown. The applicant dispositioned this refined CUFen calculation in accordance with 10 CFR 54.21(c)(1)(iii), such that the effects of aging on the intended function(s) of these components will be adequately managed for the period of extended operation.

Issue:

During its review of Calculation 1700804.315P it appears that the 80-year projected number of cycles for the inadvertent auxiliary spray was incorporated into the refined CUFen calculation for the pressurizer spray nozzle. This information is necessary to ensure that the Fatigue Monitoring Program incorporates the appropriate cycle limits and can adequately manage environmentally assisted fatigue during the SPEO.

Request:

• Clarify the discrepancy between Calculation 1700804.315P and the SLRA. Identify the transients that used 80-year projected number of cycles in the refined environmentally assisted fatigue evaluation for the Pressurizer Spray Nozzle and confirm that the Fatigue Monitoring Program manages the appropriate number of cycles to ensure the analysis remains valid for the SPEO.

RAI 4.3.3-5

Background:

SLRA Section 4.3.3 states the following as it relates to the Control Rod Drive Mechanism (CRDM) Lower Joint:

• A revised CUFen was calculated by performing a more refined analysis and crediting 80-year projected design cycles for plant heatup, 10 percent step load increases, 50 percent step load decreases, loss of load, loss of AC power, and reactor trips (Unit 4 only). CUFen Final = 0.749 Issue:

During its review of Areva Calculation No. 32-9280202, Revision 1, Turkey Point CRDM Lower Joint Environmentally Assisted Fatigue, December 15, 2017, the staff noted that a unique CUFen value for both components was calculated for Units 3 and 4, which may indicate a possible variation in, but not limited to, component design or geometry, and assumed transients and number of cycles used in the calculation. However, SLRA Section 4.3.3 only indicates one refined CUFen value for the component.

In order to determine if the Fatigue Monitoring Program will adequately manage environmentally assisted fatigue for the CRDM Lower Joint it is necessary to understand whether the calculation for one unit bounds the other or whether the assumptions for both Units 3 and 4 will be incorporated into the Fatigue Monitoring Program.

In addition, as noted in the SLRA and Areva Calculation No. 32-9280202, Revision 1, the transient Rod Trips is only applicable to the refined CUFen calculation for Unit 4; however, the basis for this was not clear.

Request:

• Clarify if a bounding CUFen represents the CRDM Lower Joint for Units 3 and 4.

o If so, justify that the CUFen value selected is applicable and appropriate for both units. Aspects such as, but not limited to transient selection, assumed number of cycles, design loading, material fabrication and geometry, should be addressed, if applicable.

o If not, confirm that the Fatigue Monitoring Program incorporates the appropriate transient cycle limits used in the respective calculation for Units 3 and 4.

• Explain and justify why the refined CUFen calculation for Unit 3 does not incorporate the transient, Rod Trips.

6. Leak-Before-Break Analysis for Class 1 Auxiliary Piping, TLAA 4.7.4, 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 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 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 current licensing basis (CLB). As described in SRP-SLR, an applicant may demonstrate compliance with 10 CFR 54.21(a)(3) by referencing the applicable aging management programs and activities in the GALL-SLR Report. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff finds that additional information is necessary in regard to the matters described below.

Background:

The regulation in 10 CFR 54.21(c)(1)(ii) states that, for a specific 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. Section 4.7.4, Leak-Before-Break Analysis for Class 1 Auxiliary Piping, of the SLRA identifies the LBB analysis in the current licensing basis as a TLAA for the SLRA. The fatigue crack growth calculation is part of the TLAA.

In September 2017, the applicant updated the LBB analysis for the Class 1 auxiliary piping to address the operation during the SPEO as documented in Structural Integrity Associates (SIA)

Engineering Report No. 0901350.401, Revision 3, Leak-Before-Break Evaluation -

Accumulator, Pressurizer Surge, and Residual Heat Removal Lines, Turkey Point Units 3 and 4, 2017.

As part of the LBB analysis, the applicant also submitted SIA Engineering Report No. 0901350.304, Revision 2, Fatigue Crack Growth Evaluation, September 18, 2017.

RAI 4.7.4-1 Issue:

In 2009, the applicant performed a LBB analysis of the reactor coolant system auxiliary piping as part of the EPU application as documented in SIA Engineering Report No. 0901350.401, Revision 0, Leak-Before-Break Evaluation - Accumulator, Pressurizer Surge, and Residual Heat Removal Lines, Turkey Point Units 3 and 4, April 2010.

As discussed above, in September 2017, the applicant updated the LBB analysis for the reactor coolant system auxiliary lines to address operation during the SPEO as documented in SIA Engineering Report No. 0901350.401, Revision 3.

Based on the NRC record, it appears that the applicants LBB analysis for the Class 1 auxiliary lines has not been submitted prior to the submission of the SLRA.

Request:

If the LBB analysis for the Class 1 auxiliary piping has not been submitted for NRC review and approval prior to the SLRA, discuss which time period the LBB application is requested for NRC approval, i.e., (1) 60-year operating license, (2) 80-year operating license, or (c) both a and b.

RAI 4.7.4-2 Issue:

Section 17.3.7.4 of SLRA Appendix A (UFSAR supplement) summarizes the LBB analyses of Class 1 auxiliary piping and stated that An LBB analysis of the Class 1 auxiliary lines was performed for the EPU, during the initial PEO, and it is valid for the current 60-year period of operation Based on the NRCs record, it appears that the NRC has not received the licensees LBB analysis for the Class 1 auxiliary lines for the 40-year license period or 60-year license renewal period.

Request:

Justify the statement in Section 17.3.7.4 of SLRA Appendix A that the LBB analysis of the Class 1 auxiliary lines is valid for the current 60-year period of operation.

RAI 4.7.4-3 Issue:

The bottom of every page of SIA Report No. 0901350.304, Revision 2, is marked with a proprietary statement that reads: Contains Vendor Proprietary Information. Many proprietary

information in SIA Report No. 0901350.304, Revision 2, appear in SIA Report No. 0901350.401, Revision 3, which has no proprietary marking.

Request:

Clarify whether the information in SIA Report No. 0901350.401, Revision 3 that is associated with SIA Report No. 0901350.304, Revision 2 is or is not proprietary. Clarify the discrepancy in the proprietary classification between these two reports.

RAI 4.7.4-4 Issue:

During the staff audit of the applicants documents, the staff noticed that applicants document AR 01610224 is related to errors in pipe stress software.

Request:

Discuss whether errors in pipe stress software as discussed in AR 01610224 affected the loading calculations used in the LBB analysis of Class 1 auxiliary piping in SIA Report No. 0901350.401, Revision 3 and SIA Report No. 0901350.304, Revision 2.

RAI 4.7.4-5 Issue:

The following issues are related to the LBB analysis in SIA Report No. 0901350.401, Revision 3.

(a) Section 3 of SIA Report states that the subject piping has no active degradation mechanisms such as water hammer, corrosion, and high cycle fatigue. However, the applicant did not provide inspection history of the subject piping.

(b) Page v of the SIA report states that The LBB evaluation was performed in accordance with the 10 CFR 50, Appendix A GDC-4 and NUREG-1061, Vol. 3 [6] as supplemented by NUREG-0800, Standard Review Plan 3.6.3 [7]... The NRC staff notes that SRP-SLR Section 3.6.3 has been revised and the latest edition is March 2007.

(c) Page viii of the LBB analysis states that Limit load analysis as outlined in NUREG-0800, SRP-SLR Section 3.6.3, was utilized in this evaluation in order to determine the critical flaw sizes since the materials involved in this evaluation are stainless steel piping Page 1-3 of the SIA report stated that Critical flaw size evaluation, based on elastic-plastic fracture mechanics techniques, is used to determine the length and depth of defects that would be predicted to cause pipe rupture under specific design basis loading conditions, including abnormal conditions such as a seismic event and including appropriate safety margins for each loading condition Section 5-1, Page 5-1, of the SIA report stated that the limit load method was used to calculate the critical crack sizes. It is not clear why the limit load method is discussed on pages viii and 5-1 but page 1-3 discusses the elastic-plastic fracture mechanics method.

(d) Figures 5-1 to 5-6 of the LBB analysis present leakage flaw size versus normal operating stress. However, the leakage flaw size on the Y axis and the normal operating stress values on the X axis are not identified or marked in the figures.

Request:

(a) Describe the inspection history of the subject piping including results from previously performed inspections and inspection frequency since the commercial operation.

(b) Discuss whether the latest SRP-SLR Section 3.6.3 dated March 2007 was used in the LBB analysis.

(c) Clarify why pages viii and 5-1 stated that critical crack size was calculated based on the limited load method whereas page 1-3 stated that critical flaw size was calculated based on the elastic-plastic fracture mechanics analysis.

(d) Provide values of leakage flaw size and normal operating stress in Figures 5-1 to 5-6.

Identify the limiting leakage flaw size for each subject piping in Figures 5-1 to 5-6.

RAI 4.7.4-6 Issue:

Section 3.6.3 of SRP-SLR specifies that piping qualified for LBB be evaluated to determine whether degradation mechanisms of wall thinning, creep and cleavage exist. Section 3 of SIA Report No. 0901350.401, Revision 3, does not discuss these degradation mechanisms.

Request:

Discuss whether wall thinning, creep and cleavage have occurred or will occur in the accumulator lines, residual head removal (RHR) lines and safety injection lines.

RAI 4.7.4-7 Issue:

The staff notes that EPRI topical report, Materials Reliability Program: Assessment of Residual Heat Removal Mixing Tee Thermal Fatigue in PWR Plants (MRP-192, Revision 2, 1024994),

August 2012, and Materials Reliability Program: Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines (MRP-146, Revision 1, 1022564),

June 2011 are related to thermal fatigue of safety-related piping.

Request:

Discuss whether the thermal fatigue issue in MRP-146 and MRP-192 affects the LBB analysis of Class 1 auxiliary piping. If yes, discuss whether the LBB analysis in SIA Report No. 0901350.401, Revision 3 and SIA Report No. 0901350.304, Revision 2, considered the thermal fatigue discussed in these two MRP reports. If the Class 1 auxiliary piping does experience thermal fatigue, discuss how the auxiliary piping satisfies the screening criteria of SRP-SLR Section 3.6.3 which prohibits LBB be applied to piping experiencing fatigue.

RAI 4.7.4-8 Issue:

The reactor coolant system primary loop piping in SLRA Section 4.7.3 contains elbows that are made of cast austenitic stainless steel. Section 4 of SIA Report No. 0901350.401, Revision 3, does not mention any pipe components made of cast austenitic stainless steel in the accumulator, RHR and Surge piping.

Request:

Confirm that the accumulator, RHR and Surge piping does not use any components or fittings that are made of cast austenitic stainless steel.

RAI 4.7.4-9 Issue:

Section 5.4 of SIA Report No. 0901350.401, Revision 3, states that From the BACs

[bounding analysis curves] and load points plotted in Figure 5-7 to Figure 5-12 all the stress points for both Units 3 and 4 are below 10 gpm BACs except for stress point 210M in the Unit 4 Accumulator Line as shown in Figure 5-7. Since the stress point 210M is in the middle of an elbow, removing the conservatism in using the weld material Z factor for pipe/elbow materials, the BAC is plotted using a Z factor of 1.0 as shown in Figure 5-13. Using a Z factor of 1.0, stress point 210M is under the 10 GPM BAC (a) It is not clear what conservatism was included in the original analysis. For example, what is the original Z factor used prior to use 1.0 for the Z factor? Page 5-3 discusses the use of Z factor for welds based on the shield metal arc welding process.

(b) It is not clear why the stress point 210M is located in the middle of an elbow. Figure 4-2 of SIA Report No. 0901350.401, Revision 3, identified only the points associated with the welds at the both ends of the elbow in Loop A of the accumulator line, not in the middle of the elbow.

Request:

(a) Provide the original Z factor used for the 210M stress point in the Unit 4 accumulator pipe in the limit load analysis. Discuss the calculated leak rate using the original Z factor. Discuss the conservatism in the original analysis in terms of the Z factor.

(b) Clarify the exact location of the stress point 210M as modeled in the pipe stress analysis.

RAI 4.7.4-10 Issue:

Figures 5-9 to 5-12 of SIA Report No. 0901350.401, Revision 3, show BACs [bounding analysis curves] and Load Points for pressurizer surge lines with a single stress point whereas Figures 5-7 and 5-8 show many stress points for the accumulator and RHR lines, respectively.

The staff notes that each of Figures 5-9 to 5-12 represents a specific location of the surge line.

Request:

Explain why pressurizer surge lines have only one stress data point in each of Figures 5-9 to 5-12 whereas for the accumulator and RHR piping multiple stress points are indicated in Figures 5-7 and 5-8, respectively.

RAI 4.7.4-11 Issue:

Section 5.3 of SIA Report No. 0901350.401, Revision 3, discusses bounding analysis curves which represent the maximum allowable membrane (pressure) plus bending stress (as determined from piping analysis for the system) as a function of the applied membrane (pressure) plus bending stress during normal plant operation. The latter condition represents the conditions during which leakage would have to be detected Request:

Explain the objective of the bounding analysis curves. It is not clear how the bounding analysis curves demonstrate (a) the margins on the crack size and leakage detection in SRP-SLR Section 3.6.3 have been satisfied, and (b) the crack stability.

RAI 4.7.4-12 Issue:

Page 5-5 of SIA Report No. 0901350.401, Revision 3, states that The maximum allowable bending stress is determined from the curve of critical crack size (a) versus applied bending moment such that acritical = 2aleakage It is not evident that Section 5 of SIA Report No. 0901350.401, Revision 3, provided the curves of critical crack sizes.

Request:

Provide curves of critical crack sizes for each subject piping.

RAI 4.7.4-13 Issue:

Section 6.1 of SIA Report No. 0901350.401, Revision 3, states that the transient information from generic Westinghouse nuclear steam supply system documents is used to perform the crack growth evaluation.

Request:

Discuss whether the generic transient information bounds the plant-specific (a) transients that the accumulator, RHR and surge piping will have experienced at the end of 80 years, (b) transients that are specified in the current licensing design basis, and (c) transients that are predicted to the end of the 80 years.

RAI 4.7.4-14 Issue:

Section 6.3 of SIA Report No. 0901350.401, Revision 3, discusses the derivation of the allowable flaw size. SRP-SLR Section 3.6.3 specifies a leakage flaw size and critical flaw size, not an allowable flaw size.

Request:

Discuss how the allowable flaw size is used in the LBB analysis and the relationship among the allowable flaw size, leakage flaw size and critical flaw size.

RAI 4.7.4-15 Issue:

(a) Section 6.4.4 of SIA Report No. 0901350.401, Revision 3, provides the leakage flaw size for the subject piping. However, the report does not provide specific calculated critical crack size for the subject piping. The report does state that the critical crack size is twice of the calculated leakage flaw size. However, without showing the actual calculated critical crack size based on material fracture toughness, it is not evident that there is a margin of 2 on the crack size as specified by SRP-SLR Section 3.6.3.

(b) The second paragraph on Page 6-7 of SIA Report No. 0901350.401, Revision 3, compares the calculated crack growth to the circumference of the accumulator pipe to demonstrate the crack stability. However, in a typical LBB evaluation, the crack growth is added to the leakage crack size to obtain the final leakage crack size at the end of 80 years. The final leakage crack size is compared to the critical crack size. The final leakage crack size should not exceed the half of the critical crack size in order to satisfy the margin of 2 as specified in SRP-SLR Section 3.6.3.

(c) The second paragraph on Page 6-7 of SIA Report No. 0901350.401, Revision 3, states that For the Accumulator Line, the maximum m+b is 19.02 ksi (including internal pressure), and the bounding leakage flaw size is 2.53 inches with bending stress= 0 for

5 GPM (Figure 5-1) The staff has questions on the use of 5 gpm because the leakage flaw size should provide a leak rate of 10 gpm in order to satisfy the margin of 10 with respect to the RCS detection system capability of 1 gpm.

Request:

(a) Provide the calculated critical crack size based on material fracture toughness for the accumulator, RHR and surge lines.

(b) Provide numbers to show that the leakage flaw size plus the crack growth (i.e., the final leakage crack size at the end of 80 years) still maintain a margin of 2 with respect to the critical crack size for each of the subject piping at the end of 80 years.

(c) Explain the statement: the bounding leakage flaw size is 2.53 inches with bending stress = 0 for 5 GPM Explain why bending stress at a leak rate of 5 gpm is used and not 10 gpm.

RAI 4.7.4-16 Issue:

Reference 51 in Section 8 of SIA Report No. 0901350.401, Revision 3, is titled: SI Report No. 1700109.402, (under Preparation) "Evaluation of Fatigue of ASME Section III, Class 1 Components for Turkey Point Units 3 and 4 for Subsequent License Renewal".

Request:

(a) Justify why data from an incomplete report are valid to be used to perform the fatigue crack growth calculations in a formal submittal to the NRC.

(b) Discuss whether Reference 51 has been published. (b)(1) If yes, discuss whether the data in the incomplete Reference 51 that were used to perform the fatigue crack growth calculations in SIA Report No. 0901350.401, Revision 3 are still valid. (b)(2) If yes, discuss whether No. 0901350.401, Revision 3, needs to be revised to indicate the publication of Reference 51.

(c) If Reference 51 has not been published, discuss when it will be published.

RAI 4.7.4-17 Issue:

The following issues are related to the fatigue crack growth evaluations in SIA Report No. 0901350.304, Revision 2.

(a) Section 1, Page 4, of the fatigue crack growth evaluation discusses correction for the errors documented in Corrective Action Report (CAR)17-012.

(b) Section 2 of the fatigue crack growth evaluation states that the ASME Code Section XI fatigue crack growth law for austenitic stainless steels (for 60 years) and the ASME

Code Case N-809 fatigue crack growth law (for 80 years) were used to perform the crack growth analysis.

Request:

(a) Discuss the errors in CAR 17-012 that caused the update of the fatigue crack growth calculations for the subject piping. Discuss whether the error affects calculations related to other nuclear power plants.

(b) Discuss why two separate documents (methods) were used to calculate the fatigue crack growth.

RAI 4.7.4-18 Issue:

Section 3.2.5 of SIA Report, 0901350.304, Rev. 2, states that 51 cycles of safe shutdown earthquake (SSE) loading (one SSE cycle assumed, along with 50 cycles of operating basis earthquake (OBE)) were used in the fatigue crack growth calculations. The staff notes that the 51 cycles of SSE and OBE loads in calculating the fatigue crack growth for the subject piping are based on generic values, not plant-specific values.

Request:

Demonstrate that the 51 cycles of OBE plus SSE, with associated earthquake loads used in the fatigue crack growth calculations bound the plant-specific transient cycles and earthquake loadings specified in the current licensing basis at Turkey Point Units 3 and 4.