ML24102A169

NRR E-mail Capture - Audit Plan - Turkey Point - 18 to 24 Month Fuel Cycle License Amendment Request (LAR) (L-2023-LLA-0161)

ML24102A169
Person / Time
Site:	Turkey Point
Issue date:	04/11/2024
From:	Michael Mahoney NRC/NRR/DORL/LPL2-2
To:	Mack J Florida Power & Light Co
References
L-2023-LLA-0161
Download: ML24102A169 (17)

Text

From:

Michael Mahoney Sent:

Thursday, April 11, 2024 11:26 AM To:

Mack, Jarrett

Subject:

Audit Plan - Turkey Point - 18 to 24 Month Fuel Cycle LAR (L-2023-LLA-0161)

Attachments:

Audit-Plan-Turkey-Point-24-Month-Fuel-Cycle-LAR_Final Clean.docx Jarrett, By letter L-2023-078, dated November 15, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession Package No. ML23320A028), as supplemented by letter dated February 9, 2024 (ML24040A189), Florida Power and Light Company (FPL, the licensee) submitted a license amendment request (LAR) to for the Turkey Point Nuclear Generating Unit Nos. 3 and 4 (Turkey Point).

The LAR proposes to revise the Turkey Point licensing basis by incorporating advanced fuel features (e.g., AXIOM cladding, ADOPT' fuel pellets, and a PRIME' fuel skeleton),

extending Technical Specification (TS) surveillance intervals, modifying TS Allowable Values (AVs) and a Trip Setpoint, and conforming changes to the Updated Final Safety Analysis Report to facilitate a transition to 24-month fuel cycles. Note, FPLs letter dated November 15, 2023, also includes an exemption request, and NRCs review of that exemption request will be handled by separate correspondence.

To improve the efficiency of the NRC staffs review, FPLs representatives and the NRC staff have discussed the performance of an NRC staff audit using an online reference portal that would allow the NRC staff limited, read-only access to the information identified in Section 5.0 of the attached audit plan. The NRC staff plans to conduct a desk audit to review the documentation provided on the portal. The online reference portal would allow the NRC staff to audit internal licensee information to confirm that the information support statements were made in the LAR and to determine whether the information included in the documents is necessary to reach a safety conclusion on the application. Any audit information that the NRC staff determines to be necessary to support the development of the NRC staffs safety evaluation will be requested to be formally submitted on the docket. The audit may also include interactions (e.g., teleconferences or webinars) on a mutually agreeable schedule sufficient to understand or resolve issues associated with the information made available on the online reference portal.

Use of the online reference portal is acceptable, as long as the following conditions are met:

The online reference portal will be password-protected, and passwords will be assigned to those directly involved in the review on a need-to-know basis.

The online reference portal will be sufficiently secure to prevent NRC staff from printing, saving, or downloading any documents; and Conditions of the use of the online reference portal will be displayed on the login screen and with concurrence by each user.

These conditions associated with the online reference portal must be maintained throughout the duration of the audit process.

The NRC staff would like to request that the portal be populated with the information identified in section 5.0 of attached audit plan. The NRC staff may request additional documents during the review, which will be transmitted to you via email.

This email will be added to ADAMS (public), and I will provide you with the accession number.

If you have any questions, please contact me.

Thanks Mike Mahoney Project Manager, LPL2-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Desk: (301)-415-3867 Mobile: (301)-250-0450 Email: Michael.Mahoney@nrc.gov

Hearing Identifier:

NRR_DRMA Email Number:

2467 Mail Envelope Properties (SA1PR09MB948643A1BD8FAD9B3A405AC1E5052)

Subject:

Audit Plan - Turkey Point - 18 to 24 Month Fuel Cycle LAR (L-2023-LLA-0161)

Sent Date:

4/11/2024 11:25:43 AM Received Date:

4/11/2024 11:25:48 AM From:

Michael Mahoney Created By:

Michael.Mahoney@nrc.gov Recipients:

"Mack, Jarrett" <Jarrett.Mack@fpl.com>

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Enclosure AUDIT PLAN REGARDING LICENSE AMENDMENT REQUEST TO INCORPORATE ADVANCED FUEL PRODUCTS AND EXTEND SURVEILLANCE INTERVALS TO FACILITATE TRANSITION TO 24-MONTH FUEL CYCLES FLORIDA POWER AND LIGHT COMPANY TURKEY POINT NUCLEAR GENERATING STATION, UNITS 3 AND 4 DOCKET NOS. 50-250 AND 50-251 EPID NO. L-2023-LLA-0161

1.0 BACKGROUND

By letter L-2023-078, dated November 15, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession Package No. ML23320A028), as supplemented by letter dated February 9, 2024 (ML24040A189), Florida Power and Light Company (FPL, the licensee) submitted a license amendment request (LAR) to for the Turkey Point Nuclear Generating Unit Nos. 3 and 4 (Turkey Point).

The LAR proposes to revise the Turkey Point licensing basis by incorporating advanced fuel features (e.g., AXIOM cladding, ADOPT' fuel pellets, and a PRIME' fuel skeleton),

extending Technical Specification (TS) surveillance intervals, modifying TS Allowable Values (AVs) and a Trip Setpoint, and conforming changes to the Updated Final Safety Analysis Report to facilitate a transition to 24-month fuel cycles. Note, FPLs letter dated November 15, 2023, also includes an exemption request, and NRCs review of that exemption request will be handled by separate correspondence.

The NRC staffs review of the LAR has commenced in accordance with the Office of Nuclear Reactor Regulation (NRR) Office Instruction LIC-101, License Amendment Review Procedures. The NRC staff has determined that a regulatory audit should be conducted in accordance with the NRR Office Instruction LIC-111, Revision 1, Regulatory Audits, dated October 31, 2019 (ML19226A274), for the NRC staff to gain a more detailed understanding of the licensees proposed license amendment.

A regulatory audit is a planned, license-related or regulation-related activity that includes the examination and evaluation of primarily non-docketed information. A regulatory audit is conducted with the intent to gain understanding, to verify information, and/or to identify information that will require docketing to support the basis of the licensing or regulatory decision. Performing a regulatory audit of the licensees information is expected to assist the NRC staff in efficiently conducting its review or gain insights on the licensees processes or procedures. Information that the NRC staff relies upon to make the safety determination must be submitted on the docket. However, there may be supporting information retained as records under Title 10 of the Code of Federal Regulations (10 CFR) 50.71, Maintenance of records, making of reports, and/or 10 CFR 54.37, Additional records and record-keeping requirements, which although not required to be submitted as part of the licensing action, would help the NRC staff better understand the licensees submitted information.

2.0 REGULATORY AUDIT BASIS An audit was determined to be the most efficient approach toward a timely resolution of questions associated with this LAR review, because the NRC staff will have an opportunity to minimize the potential for further rounds of requests for additional information (RAIs) and ensure no unnecessary burden will be imposed by requiring the licensee to address issues that are no longer necessary to make a safety determination. The NRC staff is requesting an initial set of internal licensee information to be reviewed by the staff using an online reference portal. Upon completion of this audit, the NRC staff is expected to achieve the following.

1. Confirm licensee information which supports statements made in the LAR.

2. Determine whether the information included in the documents is necessary to be submitted to support a safety conclusion.

The audit information that the NRC staff determines to be necessary to support the development of the NRC staffs safety evaluation will be requested to be submitted on the docket.

3.0 REGULATORY AUDIT SCOPE OR METHOD The purpose of the remote audit is to gain a more detailed understanding of licensees proposed LAR. The areas of focus for the regulatory audit are the information contained in the licensees November 15, 2023, submittal as supplemented, and the proposed audit questions in section 5.0 of this audit plan.

4.0 AUDIT TEAM The audit will be conducted by NRC staff from the NRR Division of Safety Systems (DSS),

Containment and Plant Systems Branch (SCPB), Nuclear System Performance Branch (SNSB),

Nuclear Methods and Fuel Analysis Branch (SFNB), and Technical Specification Branch (STSB) as well as Division of Engineering and External Hazards (DEX), Electrical Engineering Branch (EEEB) and Instrumentation and Controls Branch (EICB). The audit will be led by staff from the NRR Division of Operating Reactor Licensing (DORL). NRC staff from other organizations may be assigned to the team as appropriate and others may participate as observers. Observers at the audit may include other NRR Project Managers and various Regional staff.

The following are members of the NRC audit team:

Team Member Title Organization Mike Mahoney Project Manager NRR/DORL/LPLII-2 Ahsan Sallman Senior Nuclear Engineer NRR/DSS/SNSB Santosh Bhatt Nuclear Engineer NRR/DSS/SNSB Rick Scully Safety And Plant Systems Engineer NRR/DSS/SCPB Thang Thawn Safety And Plant Systems Engineer NRR/DSS/SCPB Khoi Nguyen Electrical Engineer NRR/DEX/EEEB Joseph Messina General Engineer NRR/DSS/SFNB Richard Fu Nuclear Engineer NRR/DSS/SFNB David Rahn Senior Electronics Engineer NRR/DEX/ELTB William Roggenbrodt Electronics Engineer NRR/DEX/EICB Clint Ashley Safety And Plant Systems Engineer NRR/DSS/STSB Tarico Sweat Reactor Systems Engineer NRR/DSS/STSB Matthew Hamm Safety And Plant Systems Engineer NRR/DSS/STSB 5.0 PROPOSED AUDIT QUESTIONS AND INFORMATION REQUEST Audit Questions SFNB

1.

WCAP-15806-P-A, Westinghouse Control Rod Ejection Accident Analysis Methodology Using Multi-Dimensional Kinetics (ADAMS Accession No. ML033350179) was approved before the current control rod ejection guidance in RG 1.236, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents (ML20055F490) was issued. Please discuss how the fuel burnup and cladding corrosion phenomena (e.g., cladding hydrogen uptake) are modeled in the control rod ejection analysis in order to meet the acceptance criteria in RG 1.236.

2. to the LAR details changes to the technical specifications, including to Technical Specification 5.6.3, Core Operating Limits Report. Please justify the lack of inclusion of WCAP-15806-P-A and WCAP-18482-P-A, Westinghouse Advanced Doped Pellet Technology (ADOPTTM) Fuel, in the core operating limits report (COLR) references.

3., Section 4.4 of the LAR states that: As expected, the increase in cycle length, change in void volume, and implementation of new fuel products impact the operating margin available in the fuel rod design evaluation. Rod internal pressure (RIP),

transient clad strain, cladding fatigue, clad stress, and fuel centerline melt criteria have all had a loss of margin. Although margin is reduced for the aforementioned criteria, all limits were still met in the 24-Month Fuel Cycle Transition analyses. Please provide the evaluation performed along with the associated results.

4., Section 4.5 of the LAR states that: There is a slightly lower pressure drop associated with the 15 Upgrade PRIME fuel design, which results in slightly higher best estimate RCS flow. There is also a slight change to the fuel assembly weight (approximately 1% increase) as a result of the transition to ADOPT and GAD fuel pellets. These changes both could have an effect on the FA top nozzle holddown force..

Please provide the evaluation performed along with the associated results.

SNSB

1.

In addition to a fuel cycle transition from 18-months to 24-months, the LAR proposes a change in the fuel from the currently used Westinghouse Vantage to Westinghouse 15x15 Upgrade PRIME assembly with ADOPT fuel pellets (a modified UO2 pellet doped with small amounts of chromia and alumina) and AXIOM cladding. Based on the changes to the overall fuel assembly, pellet, and cladding design, as discussed in various sub-sections of Section 4.0 of the LAR, the NRC staff considers the proposed change includes a fuel transition and therefore requests the licensee to clearly state in its response that the LAR includes a fuel transition to mixed cores (new fuel design plus existing fuel) leading to a 100% new fuel design core.

2.

For the 12 non-LOCA events listed in the LAR, Enclosure 1, Section 4.6, Non-LOCA, that have been reanalyzed to incorporate changes associated with the 24-month fuel cycle transition, provide the following:

a. The NRC-approved TS 5.6.3 COLR/ UFSAR methodology used for analyzing each of these events.

b. The analysis assumptions and input values that differ from the previous analysis; rationale for the differences; and reasons if the conservatism is reduced.

c. Acceptance criteria used for the analysis results.

d. Results; evaluation of results against the acceptance criteria; and the graphs showing the transient response.

3.

For the 6 non-LOCA events listed in the LAR, Enclosure 1, Section 4.6, Non-LOCA, that have been evaluated and determined to be valid, provide the following:

a. The NRC-approved TS 5.6.3 COLR/ UFSAR methodology used for evaluating each of these events.

b. The evaluation assumptions and input values that differ from the previous evaluation; rationale for the differences; and reasons if the conservatism is reduced.

c. Acceptance criteria used for the analysis results.

d. Results; evaluation of results against the acceptance criteria; and the graphs showing the transient response.

e. Describe any impact on the UFSAR Section 14.1.13, Turbine Generator Design Analysis.

4.

Section 4.9 in Enclosure 1 to the LAR discusses the use of Westinghouse WCAP-10325-P-A (Reference 2) methodology for LOCA containment mass and energy (M&E) release analysis. Westinghouse has issued the following Nuclear Safety Advisory Letters (NSALs) which reported errors in this methodology:

NSAL 06-6, "LOCA Mass and Energy Release Analysis" (Reference 3)

NSAL 11-5, "LOCA Mass and Energy Release Calculation Issues" (Reference 4)

NSAL 14-2, "Westinghouse Loss-of-Coolant Accident Mass and Energy Release Calculation Issue for Steam Generator Tube Material Properties," (Reference 5)

Confirm that the corrected WCAP-10325-P-A methodology with the errors reported in the above NSALs removed was used for the M&E release analysis. Provide justification if the methodology was not corrected.

5.

UFSAR, Section 6.5 includes the current net positive suction head (NPSH) evaluation of the residual heat removal (RHR) system pumps and core spray (CS) system pumps during accident conditions. UFSAR, Figures 6.5-1A and 6.5-1B provide the NPSH requirements for these pumps. Provide the results of the NPSH evaluation and changes in the performance of these pumps, including containment accident pressure (CAP) necessary to have adequate NPSH margin (available NPSH minus required NPSH) due to the fuel transition and 24-month fuel cycle transition.

6.

Refer to LAR, Enclosure 1, Section 4.13. Provide the following for the CCW thermal performance analysis for LOCA and SLB events for the transition.

a. Results of the extended power uprate (EPU) or the current analysis (if any) performed after EPU analysis.

b. Key inputs and assumptions, including their changes from the current analysis with justification.

c. Acceptance criteria and results.

7.

UFSAR, Section 14.2.4 describes the current steam generator (SG) tube rupture (SGTR) steam release analysis for dose calculation. The analysis examines a complete tube break adjacent to a tube sheet. The LAR, Enclosure 1, Section 4.15 provides a discussion of the proposed SGTR steam release analysis for dose calculation for the 24-month cycle transition with the new fuel. The proposed analysis is performed using hand calculation for steam release and LOFTTR2 methodology to analyze the thermal-hydraulic margin to overfill (MTO). The NRC staff requests to provide the following information:

a. The proposed analysis is performed for the same type of SG tube break as in the current analysis.

b. Comparison of the assumptions (inputs) used in the hand calculation with the assumptions for the current calculation documented in UFSAR, Table 14.2.4-1. Justify if any of the assumptions for the proposed analysis differ from the assumptions in current analysis.

c. The proposed hand calculated steam release results and their comparison with the current analysis results documented in UFSAR, Table 14.2.4-2.

d. The NRC safety evaluation for WCAP-10698-P-A (Reference 2 in UFSAR, Section 14.2.4.2) and WCAP-10750-A and their supplement 1 (References 6 and 7 below) provides NRC approval of LOFTTR1 code instead of LOFTTR2. Provide differences between the LOFTTR1 and LOFTTR2 codes with justification on the use of LOFTTR2 for analyzing the SG thermal-hydraulic MTO and the confirmatory steam release calculation.

8.

Section 4.1 of the enclosure to the LAR states that the proposed fuel utilizing the 15 Upgrade PRIME fuel design with AXIOM fuel cladding, ADOPT fuel pellets, and the introduction of the burnable absorber Gadolinia (GAD) will allow the safety analysis evaluations to be completed prior to the cycle specific design analysis. Please provide:

a. A list of plant-specific safety analysis to be completed before the cycle-specific design analysis.

b. A plant-specific demonstration analyses performed for the fuel transition for the mixed cores as well as the full new fuel core.

9.

Section 4.1 of the enclosure to the LAR states that due to the aggressive nature of the Turkey Point 24-month cycle designs, the 1770 pcm shutdown margin could not be met and therefore a new shutdown margin of 1700 pcm is proposed. The licensee states that the new value of shut-down margin allows ample margin to accommodate the anticipated variance in cycle-to-cycle confirmation. The licensee discusses various calculations in this section to show that the new shutdown margin will continue to meet the 24-month fuel cycle design. Please provide the calculations performed.

10., Section 4.2 of the LAR provides a description of the material changes to mid and IFM grids to Low Tin Zirlo from Zirlo as well as bottom nozzle flow hole geometry changes which result in flow loss changes in the fuel assembly inlet region as well as overall fuel assembly loss coefficients. Please provide comparisons of hydraulic characterization for existing fuel design and the new fuel design including individual spacer loss coefficients, friction factors.

11., Section 4.2 of the LAR states that the 15 Upgrade PRIME fuel would see a higher average flow through them (in a transition core) than they would in a full core situation. Please provide the impact of such flow redistribution on the core DNBR for the mixed cores during normal operation as well as transients such as locked rotor, feedwater malfunction, RCCA drop/mis-operation, steam line break accident, and uncontrolled RCCA withdrawal from subcritical.

12., Section 4.3 of the LAR states that the thermal-hydraulic analysis uses the WRB-1 departure from nucleate boiling (DNB) correlation (WCAP-8742-P-A) using the VIPRE-01 code (WCAP-14565-P-A). For analyses which are outside of the range of applicability of the WRB-1 correlation, the ABB-NV and WLOP correlations (WCAP-14565-P-A, Addendum 2-P-A) are used. These references are not included in COLR TS 5.6.3. Section 4.3 also states that the cycle-specific evaluations will be performed in accordance with WCAP-9272-P-A which is included in COLR TS 5.6.3.

Explain why the thermal-hydraulic analysis is not performed using the COLR TS 5.6.3 reference WCAP-9272-P-A and why the references were not included in COLR TS 5.6.3.

13.

Please provide the DNB analyses performed to support the 24-month cycle extension for the locked rotor, feedwater malfunction, RCCA drop/mis-operation, steam line break accident, and uncontrolled RCC withdrawal from subcritical events mentioned in Section 4.3 of the enclosure 1 of the LAR.

14., Section 4.17 of the LAR states that the RPV neutron exposure analysis described demonstrates that the current Subsequent License Renewal (SLR) fast neutron (E > 1.0 MeV) fluence projections for the RPV and RPV welds are greater than or essentially equal to the corresponding 24-month cycle ones. Please provide the calculations performed in accordance with methods described in references 7.33 and 7.34 of the Enclosure 1 to LAR.

15.

Table 4.8-2 of the Enclosure 1 to the LAR provides contains the plant operating ranges and key parameters used in the analysis with the FSLOCA EM. Please specify how these compare to TS limits, where applicable.

16.

Section 4.8 of the Enclosure 1 to the LAR states that: To support a 24-month fuel cycle transition with advanced fuel features including AXIOM cladding, ADOPT fuel pellets, and the PRIME fuel design, the FULL SPECTRUM' loss-of-coolant accident (FSLOCA) evaluation model (EM) was again applied to Turkey Point to demonstrate compliance with the Emergency Core Cooling System (ECCS) acceptance criteria.

Please provide the following:

a. The evaluation performed along with assumptions and input values that differ from the previous evaluation, rationale for the differences, and reasons if any conservatism is reduced.

b. Results of the evaluation including summary of break spectrum results, results for key analysis parameters for different break sizes and graphs showing the transient response.

c. Any considerations for the mixed core with co-resident fuel assemblies that have different form loss coefficients, both in the inlet region as well as for the overall fuel assemblies, for the FSLOCA evaluation.

17.

Section 4.8 of the Enclosure 1 to the LAR states that: The post-LOCA LTC analyses were evaluated for Turkey Point Units 3 & 4 for the implementation of the fuel features corresponding to the 24-month fuel cycle and increased maximum SGTP from 10% to 15%. Please provide the evaluation performed along with the associated results.

EEEB

1.

Section 3.8, Electrical Power Systems, of NUREG-1431, Standard Technical Specifications _ Westinghouse Plants, Volume 1, Revision 5 describes SRs for the electrical power systems including the following SRs that have the frequency of 18 months:

SR 3.8.1.8: Verify [automatic [and] manual] transfer of AC power sources from the normal offsite circuit to each alternate [required] offsite circuit.

SR 3.8.1.9: Verify each DG rejects a load greater than or equal to its associated single largest post-accident load, and:

a.

Following load rejection, the frequency is [ ] Hz,

b.

Within [ ] seconds following load rejection, the voltage is [ ] V and [ ] V, and

c.

Within [ ] seconds following load rejection, the frequency is [ ] Hz and [ ] Hz.

SR 3.8.1.10: Verify each DG does not trip and voltage is maintained [ ] V during and following a load rejection of [ ] kW and [ ] kW.

SR 3.8.1.11: Verify on an actual or simulated loss of offsite power signal:

a.

De-energization of emergency buses,

b.

Load shedding from emergency buses,

c.

DG auto-starts from standby condition and:

1.

Energizes permanently connected loads in [ ] seconds,

2.

Energizes auto-connected shutdown loads through [automatic load sequencer],

3.

Maintains steady state voltage [ ] V and [ ] V,

4.

Maintains steady state frequency [ ] Hz and [ ] Hz, and

5.

Supplies permanently connected [and auto-connected] shutdown loads for 5 minutes.

SR 3.8.1.12: Verify on an actual or simulated Engineered Safety Feature (ESF) actuation signal each DG auto-starts from standby condition and:

a.

In [ ] seconds after auto-start and during tests, achieves voltage [ ] V and frequency [ ] Hz,

b.

Achieves steady state voltage [ ] V and [ ] V and frequency [ ] Hz and [ ]

Hz,

c.

Operates for 5 minutes,

d.

Permanently connected loads remain energized from the offsite power system, and

e.

Emergency loads are energized [or auto-connected through the automatic load sequencer] from the offsite power system.

SR 3.8.1.13: Verify each DG's noncritical automatic trips are bypassed on [actual or simulated loss of voltage signal on the emergency bus concurrent with an actual or simulated ESF actuation signal].

SR 3.8.1.14: Verify each DG operates for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

a.

For [ ] hours loaded [ ] kW and [ ] kW and

b.

For the remaining hours of the test loaded [ ] kW and [ ] kW.

SR 3.8.1.15: Verify each DG starts and achieves:

a.

In [ ] seconds, voltage [ ] V and frequency [ ] Hz and

b.

Steady state voltage [ ] V, and [ ] V and frequency [ ] Hz and [ ] Hz.

SR 3.8.1.16: Verify each DG:

a.

Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power,

b.

Transfers loads to offsite power source, and

c.

Returns to ready-to-load operation SR 3.8.1.17: Verify, with a DG operating in test mode and connected to its bus, an actual or simulated ESF actuation signal overrides the test mode by:

a.

Returning DG to ready-to-load operation and

b.

[Automatically energizing the emergency load from offsite power].

SR 3.8.1.18: Verify interval between each sequenced load block is within +/- (( ]% of design interval] for each emergency [and shutdown] load sequencer.

SR 3.8.1.19: Verify on an actual or simulated loss of offsite power signal in conjunction with an actual or simulated ESF actuation signal:

a.

De-energization of emergency buses,

b.

Load shedding from emergency buses, and

c.

DG auto-starts from standby condition and:

1. Energizes permanently connected loads in [ ] seconds,

2. Energizes auto-connected emergency loads through load sequencer,

3. Achieves steady state voltage [ ] V and [ ] V,

4. Achieves steady state frequency [ ] Hz and [ ] Hz, and

5. Supplies permanently connected [and auto-connected] emergency loads for 5 minutes.

1. SR 3.8.4.2: Verify each battery charger supplies [ ] amps at greater than or equal to the minimum established float voltage for [ ] hours.

OR Verify each battery charger can recharge the battery to the fully charged state within [ ]

hours while supplying the largest combined demands of the various continuous steady state loads, after a battery discharge to the bounding design basis event discharge state.

2. SR 3.8.4.3: Verify battery capacity is adequate to supply, and maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test.

In the LAR, the licensee requested the change in SR interval from 18 months to 24 months for SR 3.8.1.8, SR 3.8.1.14, SR 3.8.1.15, SR 3.8.4.2, and SR 3.8.4.3. Confirm that Turkey Point intends to perform the other TS 3.8.1 SRs online (i.e., remained as 18-month frequency).

2.

In the letter dated September 27, 2023, (ML23234A192), the U.S. Nuclear Regulatory Commission (NRC) has issued the Amendment No. 297 to Subsequent Renewed Facility Operating License No. DPR-31 and Amendment No. 290 to Subsequent Renewed Facility Operating License No. DPR-41 for the Turkey Point. These amendments revise the TS in response to Turkey Points application dated September 22, 2021, supplemented by letters dated January 19, March 30, and December 2, 2022, and April 4 and May 24, 2023. In the letter, SR 3.8.4.2 is described, in part, as:

SR 3.8.4.2 Verify each battery charger supplies 400 amps (battery chargers associated with battery banks 3A and 4B) and > 300 amps (battery chargers associated with battery banks 3B and 4A) at greater than or equal to the minimum established float voltage for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

In Attachment 5 of the LAR, the licensee described SR 3.8.4.2 is described, in part, as:

SR 3.8.4.2 Verify each battery charger supplies 400 amps (battery chargers associated with battery banks 3A and 4B) and > 300 amps (battery chargers associated with battery banks 3B and 4A) at greater than or equal to the minimum established float voltage for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The NRC staff notes that there is a discrepancy between the two documents regarding the statement Verify each battery charger supplies 400 amps (battery chargers associated with battery banks 3A and 4B) and [> or >] 300 amps. Clarify this statement.

EICB

1.

Please explain how WCAP-18888 is being incorporated into the plants licensing basis.

Also, please explain the relationship between WCAP-18888 and WCAP-12745 in the licensing basis. Additionally, explain all setpoint methodologies used by Turkey Point for technical specification controlled setpoints and allowable values (AVs), (i.e., limiting safety system settings) in Turkey Points licensing basis, including describing eachs application and use.

2.

As some of the allowable values (AVs) are adjusted due to this refueling cycle extension to 24 months, please provide those values and method for determining them? An evaluation of the method applied needs to occur. Provide the related setpoint calculations for these instruments and the other instruments within scope of WCAP 18888-P for review and explain whether the limiting values were modified or not.

3.

Explain the use of the term as-found setpoint, in Enclosure 1. To what does this terminology refer; the as-found measured value or the as-found tolerance?

4.

In Attachment 2 - Turkey Point Units 3&4 Technical Specification Bases (TSB) in several locations (e.g., Pages B 3.3.1-1) the modified text reads, The Allowable Value

[Trip Setpoint] specified in Table 3.3.1-1 is a predetermined limit for a protection channel chosen to ensure automatic actuation prior to the process variable reaching the analytical limit and thus ensuring the SL would not be exceeded. This change occurs multiple times in the marked-up TSBs.

This statement appears to be a true statement for the AV, but the staff does NOT understand how this statement continues to be true for the Trip Setpoint. Explain the basis for such a change. The setting of the Trip Setpoint ensures the as found tolerance (AFT) is not exceeded, not that the safety limit is not exceeded.

5.

In TSB Markup 3.3.1-2 it states the acceptance criteria band is based on WCAP-17070.

Will this language be updated to reflect the use of the new WCAP-18888P report provided in the supplement dated February 9, 2024? If not, please explain.

6.

In reviewing the marked-up pages of the TSB, in several locations, the licensee proposes to delete references made to the Allowable Value and replace the referenced term with the

'Trip Setpoint'.

For example, in TSB 3.3.1-5, the licensee proposes the following changes (changed in bold and strike-though or underlined.

The Trip Setpoint is the value at which the bistable is set and is the expected value to be achieved during calibration. The Allowable Value is the LSSS Trip Setpoint and ensures the safety analysis limits are met during the surveillance interval selected when a channel is adjusted based on stated channel uncertainties. Any bistable is considered to be properly adjusted when the "as-left" Trip Setpoint value is within the calibration tolerance for CHANNEL CALIBRATION uncertainty allowance (i.e., + rack calibration and comparator setting uncertainties).

Trip Setpoints, in conjunction with the use of calibration tolerances, together with the requirements of the Allowable Value ensure that SLs are not violated during AOOs (and that the consequences of DBAs will be acceptable, providing the unit is operated from within the LCOs at the onset of the AOO or OBA and the equipment functions as designed).

How is this acceptable given that the Allowable Value is a column in the TS Tables, specifically Table 3.3.1-1 and Table 3.3.2-1, that provides the actual limiting value along with the calculated margin in a channel before the channels safety limit is reached, as is currently described in the TS Bases. The setting of the Trip Setpoint ensures the AFT is not exceeded, not that the safety limit is not exceeded.

7.

In supplement dated February 9, 2024, Table C1 (Enclosure 1, Page 4 of 6) the second row describes the surveillance requirement as "SR 3.3.1.10" and the Tech. Spec. Function as Table 3.3.1-1 Function 10 as "Perform CHANNEL Calibration, Pressurizer Pressure-High. However, in letter dated November 15, 2023, the Function Number in TS 3.3.1-1 (page 2 of 10) for Pressurizer Pressure-High is Function 8, please explain the discrepancy.

8.

TSB B.3.3.15 states, A summary description of and reference to the calibration tolerance methodology is provided in UFSAR Section 7.2 (Ref. 2) (Related to Item #1)

To what exact section of the UFSAR and what exact text is the LAR referring.

9. TSB 3.3.1-43, for example, states The Note (c) requires that the nominal Trip Setpoint methodologies for calculating the as-left and as-found tolerances be specified in [UFSAR Chapter 7].

Why are the term 'methodologies used when only WCAP-18888-P appears to be the basis document as described in the supplement. Will other setpoint methodologies still be in use and if so, where are they described in the UFSAR?

10. Which methodology is one obligated to use when updating any given instruments setpoint which setpoint calculation is based on one of the older setpoint methodologies (i.e.,

WCAP-12745-P and/or WCAP-17070-P)? Is the new methodology (i.e., WCAP-18888) always used? Please explain this process.

11. The new setpoint methodology (WCAP-18888-P) has been developed by Westinghouse.

Using that document as a basis, please explain the acceptance process that NextEra undertakes to ensure that the setpoint methodology is sound, accurate, and complete prior to it being accepted by the licensee.

12. Please explain the process of developing a setpoint from the development of a setpoint uncertainty calculation to the manner by which that information is processed and entered into the acceptance criteria band for a surveillance requirement functional test as a setpoint and an accompanying tolerance limit.

13. The staff was unable to locate any information related to offsets applied to instruments during calibration that would account for instrument offset due to various factors (e.g., instrument location versus location of the process piping, etc.).

What procedures govern how to completely account for offset factors and how this information is applied when determining calibration points for a given instrument during a surveillance calibration or functional test. (This question works in tandem with EICB question #12.)

14. The staff noted the licensee states that drift evaluations have been conducted for devices impacted devices by the proposed change in the fuel cycle, presumably to account for the difference in time from when during an 18-month fuel cycle a device (e.g., a transmitter) and loop would be calibrated (i.e., 22.5 months) and a 24-month fuel cycle, (i.e., 30 months).

Based on a review of several of the setpoint summary drift values, the sensor drift and rack drift values were not impacted by this extension of calibration period. Provide the drift evaluation(s) explaining why the drift terms were not impacted by the extension of the calibration period.

15. The staff requests the licensee have its setpoint expert(s) walk the staff through the development of a setpoint uncertainty calculation from its raw data input to the values displayed in the surveillance calibration or functional test procedure.

Questions 13 through 24 are related to WCAP-18888.

16. In Section 2.3, Rack Allowances, several statements are made in relation to what Rack Calibration Accuracy (RCA) is. However, diagrammatically, on Page 17 shows RCA as the upper or lower limits, known as the 'As Left/As-Found Tolerance' (ALT and AFT), (since Turkey Point does not differentiate between the ALT and AFT).

Specifically, the 'difference between the as left value and the desired value' would typically refer to the as found measured value, not the ALT/AFT.

Additionally, RCA is defined as: the two-sided (+/-) calibration tolerance about the nominal trip setpoint (NTS) of the process racks. Please explain.

17. In Section 3.2, Definitions for Protection System Setpoint Tolerances, it's unclear whether this NTS as defined in the technical report, is equal to the limiting safety system setting (LSSS) value described in Title 10 of the Code of Federal Regulation (CFR) 50.36, (e.g., the limiting trip setpoint) at the time of as-left conditions or whether it is the Nominal Trip Setpoint meaning a setpoint value more conservative than the limiting trip setpoint (at the time of the as-left calibration) to add margin between the limiting trip setpoint and the

'nominal' value.

Please explain the meaning and application of the term nominal trip setpoint as defined in WCAP-18888-P and its reference to International Society of Automation (ISA) 67.04.01-2006 standard.

18. Define the term process rack calibration accuracy as it used only once in the document (page 132). Describe how the process rack accuracy will be applied, altered, or otherwise impacted based on the results of the probability distribution function characteristics over several surveillance intervals.

19. Using the text on Page 133 from WCAP-18888-P, provided below, as a reference, please explain that statement described how operability requirements would initially be satisfied, and how those operability requirements would be satisfied in the 'rare occasion when a channel is found outside the AFT', the basis for that position given the information in RIS 2006-17.

As noted on the previous page, with respect to the Westinghouse Setpoint Methodology, operability of the process racks is defined as the ability to be calibrated about the NTS (ALT about the NTS) and subsequent surveillance should find the channel within the AFT = ALT about the NTS. On those rare occasions that the channel is found outside of the AFT = ALT, then operability requirements would be initially satisfied via recalibration, or reset, about the NTS."

20. On Page 133 of the WCAP-18888-P document it states, "Operability defined as conservative with respect to a zero margin LSP is a concept that is insufficient for the Westinghouse Setpoint Methodology and is inconsistent with its basic assumption of the AFT = ALT = RCA definition. In order to have confidence (statistical or otherwise) of appropriate operation of the process racks, it is necessary that the process racks operate within the two-sided limits defined about the NTS."

Explain this statement in both written and diagrammatic form.

21. On Page 134 of the WCAP-18888-P setpoint methodology report it states, "An alternative for the process racks where Westinghouse would use a fixed magnitude, two-sided AFT about the NTS."

What instruments within the scope of WCAP-18888-P use, or potentially will use, the alternative method or is the information in the document purely as a reference?

If this alternative were to be used, please explain the process.

22. Text from Page 134 of WCAP-18888-P reads, "A process loop found inside the RCA tolerance (ALT) on an indicated basis is considered to be operable. A channel found outside the RCA tolerance (ALT) is evaluated and recalibrated. The channel must be returned to within the ALT, for the channel to be considered operable."

Describe the required actions that will be taken by the licensee when the as-found measured value of a controlled instrument is found outside the ALT (a.k.a. the AFT or the RCA) and provide the procedural reference for these actions.

23. On Page 134 of WCAP-18888-P, it states, A process loop found inside the RCA tolerance (ALT) on an indicated basis is considered to be operable. A channel found outside the RCA tolerance (ALT) is evaluated and recalibrated. The channel must be returned to within the ALT, for the channel to be considered operable.

Describe what Turkey Point procedures related to this occurrence direct the engineer, technician, operator, etc., to take what required steps related to how to manage such a situation and the relevant portions of these procedures or policies that apply.

24. On Page 134 of WCAP-18888-P it reads, "This criterion is incorporated into plant, function specific calibration and drift procedures as the defined ALT about the NTS. At a later date, once the as found data is compiled, the relative drift (as found-as left) can be calculated and compared against the RD value. This comparison can then be utilized to ensure consistency with the assumptions of the uncertainty calculations documented in Tables 3-1 through 3-26. A channel found to exceed this criterion multiple times should trigger a more comprehensive evaluation of the operability of the channel."

To which criterion is the text referring? Is this criterion the ALT?

25. It states on Page 134 of WCAP-18888-P:

This criterion is incorporated into plant, function specific calibration and drift procedures as the defined ALT about the NTS. At a later date, once the as found data is compiled, the relative drift (as found-as left) can be calculated and compared against the RD value. This comparison can then be utilized to ensure consistency with the assumptions of the uncertainty calculations documented in Tables 3-1 through 3-26. A channel found to exceed this criterion multiple times should trigger a more comprehensive evaluation of the operability of the channel.

Do you mean to say that if ALT is exceeded multiple times, then the channel should be evaluated in a more comprehensive manner to ensure you have the correct value of ALT.

(For example, a channel with an increasing trip setpoint, if the ALT was found high out of tolerance of the ALT, the ALT would need to be evaluated and possibly increased?)

Please explain.

26. On Page 135 of the WCAP-18888-P report it states, "Therefore, the AVs for the Turkey Point Units 3 and 4 Technical Specifications are performance based and are determined by adding (or subtracting) the calibration accuracy (RCA=ALT) of the device tested during the Channel Operational Test to the NTS in the non-conservative direction (i.e., toward or closer to the SAL) for the application."

Describe and discuss what is meant by performance based Allowable Values. Is this explanation documented in any Turkey Point licensee controlled documents? If so, please provide those document(s).

Information Requests Please make the following information available for the NRC staff to audit:

1.

Analysis on surveillance historical failures in support of a 24-Month fuel cycle license amendment request.

2.

Calculational details regarding how fuel burnup and cladding corrosion phenomena are captured in the application of WCAP-15806-P-A, Westinghouse Control Rod Ejection Accident Analysis Methodology Using Multi-Dimensional Kinetics (ADAMS Accession No. ML033350179) for the control rod ejection analysis.

3.

Reference 7.4 of the enclosure 1 to the November 15, 2023, LAR - WCAP-17070-P, Revision 3, Westinghouse Setpoint Methodology for Protection Systems Turkey Point Units 3 and 4 (Power Uprate to 2644 MWt - Core Power), January 2021 6.0 LOGISTICS The audit will be started once an electronic reference portal is set up and the documentation is made available to the NRC staff. The initial desk audit will be conducted over several weeks.

The licensee will be kept informed on a regular basis during periodic discussions with the project manager regarding the progress. The audit may include interactions (e.g.,

teleconferences or webinars) on a mutually agreeable schedule sufficient to understand or resolve issues associated with the information made available.

7.0 DELIVERABLES An audit summary will be prepared within 90 days of the completion of the audit. If the NRC staff identifies information during the audit that is needed to support its regulatory decision, the staff will issue RAIs to the licensee.