ML24250A024
| ML24250A024 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 09/06/2024 |
| From: | Michael Mahoney NRC/NRR/DORL/LPL2-2 |
| To: | Mack J Florida Power & Light Co |
| References | |
| L-2023-LLA-0161 | |
| Download: ML24250A024 (11) | |
Text
From: Michael Mahoney Sent: Friday, September 6, 2024 7:37 AM To: Mack, Jarrett Cc: Perry Buckberg
Subject:
Turkey Point Nuclear Generating Unit Nos. 3 and 4 - Request for Additional Information - Transition to 24-Month Fuel Cycles (L-2023-LLA-0161)
Attachments: RAIs (SNSB and EEEB) - Turkey Point 24-Month Fuel Cycle LAR (L-2023-LLA-0161).docx
Hi Jarrett,
By letter L-2023-078, dated November 15, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession Package No. ML23320A028), 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.
The U.S. Nuclear Regulatory Commission (NRC) staff is reviewing your submittal and has identified areas where additional information is needed to complete its review. Attached are the NRC staffs request for additional information (RAIs).
As discussed, response to the attached RAIs is requested no later than 30 business days from todays date.
The NRC staff considers that timely responses to RAIs help ensure sufficient time is available for staff review and contribute toward the NRCs goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me.
Once this email is added to ADAMS, I will provide the accession number.
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: 2595
Mail Envelope Properties (SA1PR09MB9486402ABA23AB1115D11D93E59E2)
Subject:
Turkey Point Nuclear Generating Unit Nos. 3 and 4 - Request for Additional Information - Transition to 24-Month Fuel Cycles (L-2023-LLA-0161)
Sent Date: 9/6/2024 7:37:09 AM Received Date: 9/6/2024 7:37:11 AM From: Michael Mahoney
Created By: Michael.Mahoney@nrc.gov
Recipients:
"Perry Buckberg" <Perry.Buckberg@nrc.gov>
Tracking Status: None "Mack, Jarrett" <Jarrett.Mack@fpl.com>
Tracking Status: None
Post Office: SA1PR09MB9486.namprd09.prod.outlook.com
Files Size Date & Time MESSAGE 1699 9/6/2024 7:37:11 AM RAIs (SNSB and EEEB) - Turkey Point 24-Month Fuel Cycle LAR (L-2023-LLA-0161).docx 40141
Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
REQUEST FOR ADDITIONAL INFORMATION
BY THE OFFICE OF NUCLEAR REACTOR REGULATION
LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO CHANGE
SURVEILLANCE INTERVALS TO ACCOMMODATE 24-MONTH FUEL CYCLE
INTRODUCTION By letter dated November 15, 2023 (Reference1), in accordance with 10 CFR 50.90, Florida Power and Light (the licensee) submitted a license amendment request (LAR) to the U.S.
Nuclear Regulatory Commission (NRC) for changes to the Turkey Point Nuclear Generating Station, Units 3 and 4 (Turkey Point), technical specifications (TSs). The proposed license amendments would revise the Turkey Point licensing basis by incorporating advanced fuel features (e.g., AXIOM cladding, ADOPT' fuel pellets, and a PRIME' fuel skeleton), extending TS surveillance intervals, modifying TS Allowable Values (AVs) and a Trip Setpoint, and conforming changes to the Updated Final Safety Analysis Report (UFSAR) to fa cilitate a transition to 24-month fuel cycles.
After reviewing the LAR (Reference 1), the Nuclear Systems Performance Branch (SNSB) staff requests response to the request for additional information (RAI) given below.
SNSB-RAI 1 Regulatory Basis:
The applicable 1967 Atomic Energy Commission (AEC) draft General Design Criteria (GDC) 6, Reactor Core Design, states:
The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all offsite power.
RAI:
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 sm all 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 to include 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. Include in the response a table that shows a comparison of all properties and features of the proposed fuel with the currently used fuel.
2
SNSB-RAI 2 Regulatory Basis:
Same as in SNSB-RAI 1 RAI:
For the 12 non-loss of coolant accident (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 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.
SNSB-RAI 3 Regulatory Basis:
Same as in SNSB-RAI 1 RAI:
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 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.
SNSB-RAI 4 Regulatory Basis:
The applicable 1967 AEC draft GDC 10, Containment, states Containment shall be provided.
The containment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain for as long as the situation requires the functional capability to protect the public.
RAI:
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 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.
SNSB-RAI 5 Regulatory Basis:
The applicable 1967 AEC draft GDC 41, Engineered Safety Features Performance Capability, states, Engineered safety features such as emergency core cooling and containment heat removal systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component.
RAI:
A change in the sensible and decay heat in the proposed fuel and the fuel cycle could affect the mass and energy release (M&E) in containment during a large break LOCA. Provide the evaluation and results of LOCA sump temperature response, net positive suction head (NPSH) margin (available NPSH minus required NPSH) evaluation, and changes in the performance of the safety injection (SI) pumps and containment spray system (CSS) that draw water from the containment sump during the recirculation phase of a large break LOCA. Include in the response any impact on the long-term containment and core cooling and if any containment accident pressure (CAP) above the vapor pressure at the transient sump water temperature is needed to have adequate NPSH margin due to the fuel transition and 24-month fuel cycle transition.
SNSB-RAI 6 Regulatory Basis:
Same as in SNSB-RAI 5 RAI:
Refer to LAR, Enclosure 1, Section 4.13. Provide the following for the component cooling water (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.
SNSB-RAI 7 Regulatory Basis:
10 CFR 50.67, Accident source term, requires that the applicant's analysis demonstrates with reasonable assurance the following:
(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 [Sievert] Sv (25 rem) total effective dose equivalent (TEDE).
(ii) An individual located at any point on the outer boundary of the low 4
population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage),
would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).
(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.
RAI:
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.
SNSB-RAI 8 Regulatory Basis:
The applicable 1967 AEC draft GDC 10, Containment, states in part, The containment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain for as long as the situation requires the functional capability to protect the public.
RAI:
A change in the sensible and decay heat in the proposed fuel and the fuel cycle would affect the M&E release in containment during a large break LOCA. Provide the evaluation and results of (a) containment pressure and temperature response, and impacts on: (b) equipment qualification, (c) containment integrated leak test (ILRT) pressure (Pa), and (d) maximum containment wall temperature used for structural design.
5
SNSB-RAI 9 Regulatory Basis:
The applicable 1967 AEC draft GDC 29, Reactivity Shutdown Capability, states: One of the reactivity control systems provided shall be capable of making the core subcritical under any anticipated operating condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margin should assure subcriticality with the most reactive control rod fully withdrawn.
RAI:
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. Provide a discussion on how the new shutdown margin of 1700 pcm will continue to meet the anticipated variance in cycle-to-cycle confirmation for the 24-month fuel cycle design.
SNSB-RAI 10 Regulatory Basis:
The applicable 1967 AEC draft GDC 6, Reactor Core Design, states: The reactor core with its related controls and protection systems shall be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits which have been stipulated and justified. The core and related auxiliary system designs shall provide this integrity under all expected conditions of normal operation with appropriate margins for uncertainties and for specified transient situations which can be anticipated.
RAI:, Section 4.2 of the LAR provides a description of the material changes to mid and intermediate flow mixing (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. Provide comparisons of hydraulic characterization for existing fuel design and the new fuel design including individual spacer loss coefficients and friction factors. Provide impact of the changes on any operating margins or safety analysis.
SNSB-RAI 11 Regulatory Basis:
Same as in SNSB-RAI 10 RAI:, 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 departure from nucleate boiling ratio (DNBR) for the mixed cores during normal operation as well as the following transients:
- locked rotor
- feedwater malfunction
- Rod cluster control assemblies (RCCA) drop/mis-operation
- steam line break accident, and
- uncontrolled RCCA withdrawal from subcritical
6
SNSB-RAI 12 Regulatory Basis:
Same as in SNSB-RAI 10 RAI:, 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 Westinghouse DNBR correlation (WLOP) correlations (WCAP-14565-P-A, Addendum 2-P-A) are used. These references are not included in COLR TS 5.6.3.
Explain why the references for the ABB-NV and WLOP correlations were not included in COLR TS 5.6.3.
SNSB-RAI 13 Regulatory Basis:
Same as in SNSB-RAI 10, &
The applicable 1967 AEC draft GDC 14, Core Protection System, states: Core protection systems, together with associated equipment, shall be designed to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.
RAI:
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.
SNSB-RAI 14 Regulatory Basis:
Regulations in 10 CFR 50.46 state that each pressurized light-water reactor (LWR) fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must perform analysis of core cooling performance under postulated LOCA conditions using an acceptable evaluation model (EM). The acceptable LOCA EM must be used that either applies realistic methods with an explicit accounting for uncertainties or follows the prescriptive, conservative requirements of Appendix K to 10 CFR Part 50 and that the core cooling performance must be analyzed for a number of postulated LOCAs of different sizes, locations, and other characteristics to ensure that the most severe event is calculated.
The applicable 1967 AEC draft GDC 44, Emergency Core Cooling System Capability, states that: An Emergency Core Cooling System with the capability for accomplishing adequate emergency core cooling shall be provided. This core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal water reaction to acceptable amounts for all sizes of breaks in the reactor coolant piping up to the equivalent of a double-ended rupture of the largest pipe. The performance of such emergency core cooling system shall be evaluated conse rvatively in each area of uncertainty.
RAI:
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 FULL SPECTRUM' loss -of-coolant accident 7
(FSLOCA) engineering methodology (EM). Provide a comparison of the plant operating parameters used in the FSLOCA analysis to the TS limits for the same parameters, where applicable. Please refer to the TS limits by their respective LCO numbers.
SNSB-RAI 15 Regulatory Basis:
Same as in SNSB-RAI 14 RAI:
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) evaluatio n EM was again applied to Turkey Point to demonstrate compliance with the Emergency Core Cooling System (ECCS) acceptance criteria. 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.
SNSB-RAI 16 Regulatory Basis:
10 CFR 50.46(b)(5), Long-term cooling, states, After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
RAI:
Section 4.8 of the Enclosure 1 to the LAR states that: The post-LOCA long term cooling (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 [steam generator tube plugging] SGTP from 10% to 15%. Provide the impact of increased SGTP level for implementation of the 24-month fuel cycle on the post-LOCA LTC analysis of record.
After reviewing the LAR (Reference 1), the Electrical Engineering Branch (EEEB) staff has the following RAI.
EEEB-RAI-1 Regulatory Basis:
The following regulatory requirements and General Design Criteria are applicable to the Turkey Point electrical power systems.
- Title 10 of the Code of Federal Regulations (10 CFR) 50.36, "Technical Specifications,"
requires in part, that the operating license of a nuclear production facility include TSs.
Paragraph 56.36(c)(2) of 10 CFR requires that the TSs include limiting conditions for operation (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not 8
met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.
- Appendix A to 10 CFR Part 50, General Design Criterion (GDC) 17, Electric Power Systems, states, in part, that an onsite electric power system and an offsite electric power system be provided to permit functioning of structures, systems, and components (SSCs) important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences, and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents. The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.
- GDC 18, "Inspection and Testing of Electric Power Systems," states, in part, that electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features.
RAI:
In the LAR (Reference 1), the licensee stated that the proposed change follows the guidance of NRC Generic Letter (GL) 91-04 to increase the SR intervals from 18 months to 24 months.
For non-calibration SRs, GL 91-04 recommends, in part, that the licensees should perform the following to support surveillance intervals to accommodate a 24-month fuel cycle:
a) Evaluate of the effect on safety of the change in surveillance intervals to support a conclusion that the effect on safety is small.
b) Confirm that historical maintenance and surveillance data do not invalidate the conclusion that the effect on safety is small.
c) Confirm that the performance of surveillances at the bounding surveillance interval limit provided to accommodate a 24-month fuel cycle would not invalidate any assumption in the plant licensing basis.
For Recommendation (a) above, LAR Attachment 5 states, in part, that:
Each proposed Non-Calibration SR interval change has been evaluated with respect to the effect on plant safety. The methodology utilized to justify the conclusion that changing the SR interval from an 18-month to a 24-month frequency has a minimal effect on safety, is based on whether the associated function/feature is:
- 1. Tested on a more frequent basis during the operating cycle by other plant programs;
- 2. Designed to have redundant counterparts or be single failure proof; or
- 3. Highly reliable.
The staff notes that Attachment 5 of the LAR describes the methodology to justify the conclusion that changing the SR interval from an 18-month to a 24-month frequency has a minimal effect on safety. However, the LAR does not specify the more-frequent testing (Item (1) above). For each 9
of SRs 3.8.1.8, 3.8.1.14, 3.8.1.15, 3.8.4.2, and 3.8.4.3 please specify the more frequent testing(s) to support the justification.
REFERENCES
- 1. Letter from Florida Power and Light Company to NRC, Turkey Point Nuclear Generating Station, Unit 3 and 4, Docket Nos. 50-250 and 50-251, Renewed Facility Operating Licenses DPR-31 and DPR-41, License Amendment Request 278, Incorporate Advanced Fuel Products, Extend Surveillance Intervals and 10 CFR 50.46 Exemption Request to Facilitate Transition to 24-Month Fuel Cycles, November 15, 2023, ADAMS Accession ML23320A027.
- 2. WCAP-10325-P-A, "Westinghouse LOCA Mass and Energy Release Model for Containment Design March 1979 Version," May 1983 (proprietary), ADAMS Accession No. ML080640615.
- 3. NSAL 06-6, "LOCA Mass and Energy Release Analysis," dated June 6, 2006, ADAMS Accession No. ML22195A159.
- 4. NSAL 11-5, "LOCA Mass and Energy Release Calculation Issues," dated July 25, 2011, ADAMS Accession No. ML13239A479.
- 5. NSAL 14-2, "Westinghouse Loss-of-Coolant Accident Mass and Energy Release Calculation Issue for Steam Generator Tube Material Properties," dated March 31, 2014, ADAMS Accession No. ML22195A177
- 6. WCAP-10698-P-A and WCAP-10750-A, SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, August 1987, ADAMS Accession Nos. ML071430455 (Proprietary), ML071430455 (Non-Proprietary).
- 7. WCAP-10698-P-A, Supplement 1 and WCAP-10750-A, Supplement 1, " Evaluation of Offsite Radiation Doses for a Steam Generator Tube Rupture Accident, March 14, 1986, ADAMS Accession Nos. ML19277J334 (Proprietary) and ML20140D480 (Non -Proprietary).