ML24103A020
| ML24103A020 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/22/2024 |
| From: | Richard Guzman NRC/NRR/DORL/LPL1 |
| To: | Carr E Dominion Energy Nuclear Connecticut |
| References | |
| EPID L-2023-LLA-0065 | |
| Download: ML24103A020 (1) | |
Text
April 22, 2024
Eric S. Carr President - Nuclear Operations and Chief Nuclear Officer Dominion Energy Nuclear Connecticut, Inc.
Millstone Power Station Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO. 3 -
SUMMARY
OF REGULATORY AUDIT IN SUPPORT OF LICEN SE AMENDMENT REQUEST TO USE FRAMATOME SMALL BREAK AND REALISTIC LARGE BREAK LOSS-OF-COOLANT ACCIDENT EVALUATION METHODOLOGIES FOR ESTABLISHING CORE OPERATING LIMITS (EPID L-2023-LLA-0065)
Dear Eric Carr:
By letter dated May 2, 2023, as supplemented by letter dated April 1, 2024 (Agencywide Documents Access and Management System Accession (ADAMS) Nos. ML23123A279 and ML24093A216, respectively), Dominion Energy Nuclear Connecticut, Inc. (the licensee) submitted a license amendment request (LAR) for Millstone Power Station, Unit No. 3 (MPS3).
The proposed changes would update the list of approved methodologies in MPS3 TS 6.9.1.6.b to establish the core operating limits included in the Core Operating Limits Report for the GAIA fuel with M5 cladding.
The U.S. Nuclear Regulatory Commission (NRC) staff conducted a virtual audit to support its review of the LAR. The NRC staff reviewed information and interviewed licensee staff. The NRC staff issued its audit plan on August 31, 2023 (ML23258A055). Enclosure 1 of this audit summary lists the individuals that took part in or attended the audit. Enclosure 2 lists the NRC staffs audit questions.
The NRC staff conducted the audit using virtual meetings and an Internet-based portal provided by the licensee. Using the licens ees portal, the NRC staff revi ewed information related to the LAR but not available on the MPS3 dockets. During the audit, the staff also met virtually with the licensee on November 13, 2023, December 20, 2023, and February 29, 2024. The staff used these meetings to confirm its understanding of the LAR, discuss the information in the portal, and decide whether the NRC staff identified any information that needs to be submitted on the docket to complete the staffs safety evaluation.
During the audit, the staff and the licensee discussed the audit items in enclosure 2, and the staff identified information it needed on the docket to support its review. After the audit discussions, the NRC sent the licensee a request for additional information on March 5, 2024 (ML24065A311). The licensee responded to this request on April 1, 2024 (ML24093A216).
E. Carr
The NRCs licensing project manager informed licensee staff by telephone on February 29, 2024, that the NRC staff had completed its audit. There were no open items resulting from the audit.
If you have any questions, please contact me at (301) 415-1030 or by email to Richard.Guzman@nrc.gov.
Sincerely,
/RA/
Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Docket No. 50-423
Enclosures:
- 1. List of Audit Participants
- 2. List of Audited Documents
cc: Listserv List of Audit Participants
U.S Nuclear Regulatory Commission (NRC) Audit Team
NRC Staff Richard Guzman - Senior Project Manager, NRR 1/DORL2/LPL13 Summer Sun - Senior Nuclear Engineer, NRR/DSS 4/SNSB5
Dominion Energy Nuclear Connecticut, Inc. - Millstone Power Station, Unit 3 Team Shayan Sinha - Regulatory Affairs Scott Luchau - Supervisor, Nuclear Safety Analysis Timothy Olsowy - Millstone Station Licensing Brian Mount - Nuclear Safety Analysis Virgilio Esquillo - Nuclear Safety Analysis Nadine Shane - Nuclear Safety Analysis Stephen OHearn - Nuclear Safety Analysis Diana Casique - Millstone Station Licensing
Framatome personnel include:
Bob Clarke - Project Manager Brittany Williams - LOCA Safety Analysis Mireille Cortes - LOCA Safety Analysis Lisa Gerken - LOCA Safety Analysis
1 Office of Nuclear Reactor Regulation 2 Division of Operating Reactor Licensing 3 Plant Licensing Branch 1 4 Division of Safety Systems 5 Nuclear Systems Performance Branch
Enclosure 1 List of Audit Requests
During its review of the license amendment request (LAR) for the Millstone Power Station, Unit No. 3 (MPS3), the U.S. Nuclear Regulator y Commission (NRC) staff provided the licensee with the following audit requests and audited the licensees responses that the licensee posted on its internet-based portal.
Item No. Audit Request 1 Framatome and Westinghouse Loss-of-Coolant Accident (LOCA) Methods in Technical Specification (TS) 6.9.1.6.b
Describe the MPS3 core reload strategy for fuel transition from a full core of Westinghouse fuel to a full core of GAIA fuel. Also, describe the Westinghouse small break (SB) and realistic large break (RLB) LOCA analyses for Westinghouse fuel with mixed core conditions including the Framatome GAIA fuel, and discuss the results of the analyses to show that the applicable acceptance criteria in Title 10 of the Code of Federal Regulations (10 CFR) 50.46(b) are met. Provide the rationale if the Westinghouse LOCA analyses for mixed core conditions including GAIA were not performed.
As stated in Section 3.1 of Attachment 1 to the LAR, the Framatome SBLOCA analysis supporting the GAIA fuel at MPS3 is based on the previously NRC-approved methods, including the methods documents in Topical Report (TR) EMF-2328-P-A Revision 0, Supplement 1-P-A, Revision 0. The NRC safety evaluation (ML15210A257) for the TR EMF-2328-P-A Revision 0, Supplement 1-P-A, Revision 0 imposed modelling requirements in the following areas:
- 1. Spectrum of break sizes
- 2. Breaks in the attached piping
- 3. Delayed reactor coolant pump trip
- 4. Maximum safety injection tank/refueling water tank fluid temperature
- 5. Core bypass-flow path in the reactor vessel
- 6. Reactivity feedback
- 7. Loop seal clearing and cross-over leg modelling
- 8. Core nodalization
Sections 4.1, 4.3, 4.4, and 4.5 of Attachment 3 (ANP-4031P) to the LAR have provided information addressing the modelling requirements in above items 1 through 4. Provide information addressing compliance with the modelling requirements in above item 5 through 8 for the SBLOCA analysis.
3 Acceptance Criteria for SBLOCA Analysis
Section 3.1 of Attachment 3 (ANP-4031P) to the LAR indicates that the SBLOCA analysis is analyzed to meet the first four acceptance criteria in 10 CFR 50.46(b). Specifically, the fourth criterion requires a coolable geometry.
Provide discussion on the SBLOCA analysis for all break sizes to address the compliance with the fourth criterion.
Enclosure 2
4 S-RELAP5 Nodalization Schemes for SBLOCA Analysis
Discuss any modifications to the nodalization schemes used for the MPS3 SBLOCA analysis that were not previously reviewed and approved by the NRC during the review of the TRs, EMF-2328-P-A Revision 0, and EMF-2328-P-A Revision 0, Supplement 1-P-A, Revision 0.
5 Values of Plant Parameters used in the SBLOCA Analysis
Table 3-1 of ANP-4031P in the LAR lists the values of 24 plant parameters used as input for the SBLOCA analysis. Discuss how the following listed values would result in a minimum margin to the applicable 10 CFR 50.46(b) limits, and therefore, is acceptable for the MPS3 SBLOCA analysis.
- 1. The steam generator secondary pres sure is assumed at 937.4 pounds per square inch absolute (psia)
- 2. The assumed auxiliary feedwater (AFW) temperature of 80 degrees Fahrenheit (°F) is based on the average of the maximum and minimum operating temperatures. The assumed initial AFW temperatures does not represent the maximum-allowed temperature and may not be the most conservative condition.
- 3. The nominal main feedwater (MFW) temperature of 447.9 °F is based on the measurement uncertainty recapture conditions. The assumed MFW temperature may not represent its maximum-allowed temperature and may not be the most conservative condition.
- 4. The nominal pressurizer pressure of 2250 psia is assumed at the reactor coolant system (RCS) operating pressure. This value does not include a measurement uncertainty of 50 pounds per square inch and may not be the most conservative condition for the SBLOCA analysis.
6 Safety Injection Flow Rates Assumed for the SB- and RLB-LOCA Analysis
The safety injection (SI) flow rates used for the MPS3 SBLOCA analysis are shown in the following Tables of ANP-4031P in the LAR:
Table 3-2 for Total Intact Loops Flow and Broken Loop Flow from the High Head SI (HHSI) System for Cold-Leg Pump Discharge Break Spectrum; Table 3-3 for Total Intact Loops Flow and Broken Loop Flow from the Intermediate Head SI (IHSI) System for Cold-Leg Pump Discharge Break Spectrum; and Table 3-4 for Total Intact Loops Flow and Broken Loop Flow from the Low Head SI (LHSI) System for Cold-Leg Pump Discharge Break Spectrum.
As shown in Table 4-1 items 3.n, 3.m, and 3.l of Attachment 5 (ANP-4032P) to the LAR, the SI flow rates used in the RLBLOCA analysis are the same for SBLOCA analysis. Provide discussion on how the HHSI, IHSI, and LHSI flow rates in Tables 3-2 to 3-4 are derived. Also, discuss how the listed HHSI, IHSI
and LHSI flow rates assumed in the SBLOCA and RLBLOCA analyses would result in a minimum margin to the applicable 10 CFR 50.46(b) limits, and therefore, are acceptable for the MPS3 SBLOCA analysis.
7 Axial Power Shapes Used in the SBLOCA Analysis
Figure 3-4 of ANP-4031P in the LAR shows the input axial power shape and axial power shape adjusted to the TSs total and radial peaking factors for the SBLOCA analysis. Provide discussion on how the axial power shape assumed in the SBLOCA analysis would result in a minimum margin to the applicable 10 CFR 50.46(b) limits, and therefore, is acceptable. In addition, discuss the values of the total and radial peaking factors used for adjusting the axial power shape to show the peaking factors are consistent with the corresponding TS values.
8 Delayed RCP Trip Study
The SBLOCA analysis assumed that the RCP trip occurred at reactor trip. Section 4.4 of ANP-4031P in the LAR indicates that a delayed RCP trip study was performed to identify the delayed effect of RCP trip on the SBLOCA analysis. The study assumed the delayed RCP trip time of 5 minutes after the specified trip criteria were met for operators to trip all four RCPs. Provide discussion on how the assumed 5-minute RCP trip delayed time is adequate to identify the delayed effect on the peak cladding temperature (PCT) for the SBLOCA analysis, considering that the PCT (Table 4-2 of ANP-4031P) for break sizes from 5.0 to 8.7 inches would occur within 5 minutes from initiation of the LOCA and that for those break size LOCA, the RCP may not trip at the time when the PCT occurs. Also, discuss fr om the human engineering consideration that the available operator action time for operators to trip all four RCPs is adequate.
9 Refueling Water Storage Tank (RWST) Drain-Down Time
Section 4.6 of ANP-4031P in the LAR discusses the RWST drain-down time analysis. Provide discussion on the evaluation performed for MPS3 RWST drain down time and the evaluation performed to determine the potential impact of a higher pumped emergency core cooling system (ECCS) injection source temperature following sump switchover on the SBLOCA results. Discuss how the SBLOCA analysis results are not impacted by a higher ECCS temperature from the switchover.
10 ECCS Temperature Sensitivity Study
Section 4.5 of ANP-4031P in the LAR provides information regarding the ECCS temperature sensitivity study. As shown in item h of Section 2.0 on page 4-4 of ANP-4032P in the LAR, the operating temperature range for water in the accumulator is 75 °F -125 °F. Also, as shown on page 15.6-40 of the MPS3 FSAR, the operating temperature range for the accumulator fluid is 67 °F - 84 °F. The lowest operating temperatures from two sources are lower than 100 °F that were assumed in the sensitivity study. Discuss the use of
100 °F for the accumulator fluid in the ECCS sensitivity study is adequate to cover the lowest operating temperature of 75 °F specified in ANP-4032P, or 67 °F in the final safety analysis report.
11 Thermal Conductivity Degradation (TCD)
NRC Information Notice 2009-23 (ML091550527), Nuclear Fuel TCD, discusses an issue related to the ability of legacy thermal-mechanical fuel modeling codes to predict the exposure-dependent degradation of fuel thermal conductivity accurately. A safety concern with TCD in a LOCA would be that fuel temperatures modeled incorrectly would affect the initial stored energy, resulting in the LOCA evaluation model to underpredict PCTs. Discuss the TCD model included in the SBLOCA and LBLOCA analyses. Provide rationale if the TCD model is not included.
12 LOCA Analyses for Mixed Core Configurations
The LOCA analyses for mixed core configurations is discussed on page 3-8 of ANP-4032P in the LAR. In addition, Limitation 3 in Table 3-1 of ANP- 4032P restricts that the RLBLOCA evaluation methodology (EM) in EMF-2103(P)(A),
Revision 3 is approved based on models that are specific to Framatome proprietary M5 fuel cladding. The application of the model to other cladding types has not been reviewed.
Provide a discussion addressing the compliance with Limitation 3 above for the RLBLOCA analysis with consideration of mixed core configurations. Discuss how the fuel is modeled in the RLBLOCA analyses for mixed core configurations and provide a diagram of core nodalization scheme used in the RLBLOCA analysis.
Also, discuss the results of the RLBLOCA analysis for the mixed core conditions to show that the applicable acceptance criteria in 10 CFR 50.46(b) are met and the analysis is applicable to any core design.
(a) In regard to the Framatome GAIA fuel with M5 cladding being calculated for PCT and maximum local oxidation (MLO) in the MPS3 RLBLOCA analysis, clarify whether an MPS3 RLBLOCA analysis simulating a mixed core was performed. If the analysis was performed, confirm that the analysis results are included in ANP- 4032P of the LAR. If the analysis was not performed, provide rationale supporting that the analysis is not needed.
(b) In regard to the modeling approach that allows flow diversion from the hot assembly to the surrounding assemblies, provide a diagram of core nodalization scheme used in the RLBLOCA analysis. Reference the NRC document approving the core nodalization scheme.
13 S-RELAP5 Nodalization Schemes for RLBLOCA Analysis
Discuss any modifications to the nodalization schemes used for the MPS3 RLBLOCA that were not previously reviewed and approved by the NRC during the review of the TR, EMF-2103-P-A Revision 3.
14 RLBLOCA Analysis - Plant Initial Operating Conditions
Sections 2.0 and 3.0 of Table 4-1 in ANP-4032P of the LAR list the plant initial operating conditions for the RLBLOCA analysis. Provide discussion for each initial condition that the values used in the RLBLOCA analysis in ANP-4032P would result in a minimum margin to the applicable 10 CFR 50.46(b) limits, and therefore, are acceptable. Provide additional information in the portal to show that the use of the operating ranges or bounding TS values for the subject plant parameters is consistent with the Framatome methodologies in EMF-2103-P-A, Revision 3. Reference the related section or page number in EMF-2103-P-A, Revision 3.
15 Acceptance Criteria for the RLBLOCA Analysis
Section 3.1 in ANP-4032P of the LAR indicates that the RLBLOCA analysis is analyzed to meet the first three criteria regarding PCT, MLO, and core wide oxidation in 10 CFR 50.46(b). The final two criteria regarding coolable geometry and long-term cooling are treated in separate plant-specific evaluations. Provide a discussion on the analytical methods used in the evaluations for meeting the final two criteria to assure that none of the proposed LOCA methods in the LAR will be used. If the coolable geometry analysis and long-term cooling analysis involve the proposed LOCA methods, provide the analyses for the NRC staff to review and approve.
16 Compliance with the Limitations for use of GALILEO Code
The licensee addressed in Table 3-2 of ANP-4032P its compliance with the limitations for use of the GALILEO code in the RLBLOCA analysis. As stated in Section 3.3 of ANP-4031P, the GALILEO code is also used in the SBLOCA analysis. Provide a discussion confirming that the information in Table 3-2 of ANP-4032P for the RLBLOCA analysis is applicable to the SBLOCA analysis in ANP-4031P to satisfy the limitations for use of GALILEO.
17 (Follow-up Audit Item) Regarding Audit Item 1, the response indicated that even for larger SBLOCA transients, the thermal-hydraulic response is quasi-stable, with stratified flow conditions eventually developing throughout the core and the RCS. Under these conditions, core flow rate is relatively low, which provides enough time to maintain a near-equilibrium flow between fuel assemblies (that is, cross flow is not a factor).
Provide evidence from the results of a larger SBLOCA analysis applicable to MPS3 to show that: (1) the stratified flow conditions eventually develop throughout the core and the RCS, and (2) the core flow rate is relatively low, providing enough time to maintain a near-equilibrium flow between fuel assemblies.
- 1. The response indicated that based on experience with Surry fuel transition to Westinghouse 15x15 upgrade, a pressure drop between two fuel designs of ~4 percent existed and led to LBLOCA transient core impact of +14 °F.
Discuss the methodologies and flow conditions (such as core volumetric flow rate, temperature, and pressure of the core flow) used in determining the pressure drop difference of ~4 percent between two fuel designs. Address if the methodologies and conditions used would result in a maximum pressure drop between the fuel designs that would lead to a maximum flow diversion from Westinghouse fuel assemblies, resulting in a maximum penalty PCT on Westinghouse fuel, and is conservative. If the pressure drop difference of
~4 percent is not a maximum value, justify the acceptance of the pressure drop used to determine PCT penalty of 14 °F.
Also, discuss the methodologies and assumptions used in determination of the mixed core PCT penalty of 14 °F from based on a pressure drop between two fuel designs of ~4 percent. Reference the NRC document approving the PCT penalty of 14 °F. If the cited PCT penalty was not previously reviewed and approved by the NRC, justify the acceptance of the specific PCT penalty.
- 2. The response indicated that for fuel transition at MPS3, the fuel assembly pressure drop is ~10 percent higher than the GAIA fuel assembly pressure drop. The estimated impact to PCT is +35 °F, based on a linear extrapolation from the PCT penalty of 14 RLBRLB°F at Surry (i.e., 14 °F
- 10 percent / 4 percent).
Discuss applicability of the Surry data, which is based on Surry fuel transition to Westinghouse 15x15 assembly upgrade, to MPS3 with GAIA and Westinghouse 7x17 fuel assemblies. Discuss that the use of a linear extrapolation approach, based on a PCT penalty of 14 °F and pressure drop value of 4 percent at Surry, to determine PCT penalty of 35 °F on Westinghouse fuel at MPS3, is acceptable.
18 (Follow-up Audit Item) Regarding Audit Item 12, the response indicated that the MPS3 RLBLOCA analysis calculated PCT and MLO only for Framatome GAIA fuel with M5 Framatome cladding. Clarify whether an MPS3 RLBLOCA analysis simulating a mixed core, consisting of the Framatome GAIA fuel and the resident Westinghouse fuel, was performed or not. If the analysis was performed, confirm that the analysis results are included in ANP-4032P of the LAR. If the analysis was not performed, provide rationale supporting that the analysis is not needed.
- 1. The response indicated that the fuel assembly with the highest pressure drop (resident fuel assembly) was modeled as the hot assembly. The surrounding central core was modeled with the lowest pressure drop fuel assembly (Framatome fuel assembly). This modeling approach allows flow diversion from the hot assembly to the surrounding assemblies.
Provide a diagram of core nodalization sc heme showing the central resident fuel assembly and surrounding Framatome fuel assemblies used in the RLBLOCA analysis. Reference the NRC document approving the core nodalization scheme.
ML24103A020 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DORL/LPL1/BC NAME RGuzman KEntz HGonzález DATE 4/11/2024 4/17/2024 4/22/2024 OFFICE NRR/DORL/LPL1/BC NAME RGuzman DATE 4/22/2024