ML24138A057
| ML24138A057 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 06/26/2024 |
| From: | Shilpa Arora Plant Licensing Branch III |
| To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
| Arora, S | |
| References | |
| EPID L-2023-LLA-0120 | |
| Download: ML24138A057 (1) | |
Text
June 26, 2024 David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 - ISSUANCE OF AMENDMENT NOS. 285 AND 278 RE: APPLICATION TO ADOPT TSTF-564, SAFETY LIMIT MCPR (EPID L-2023-LLA-0120)
Dear David P. Rhoades:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 285 to Renewed Facility Operating License No. DPR-19 and Amendment No. 278 to Renewed Facility Operating License No. DPR-25 for Dresden Nuclear Power Station, Units 2 and 3, respectively. These amendments consist of changes to the technical specifications (TSs) in response to your application dated August 30, 2023.
The amendments adopt Technical Specification Task Force (TSTF) Traveler-564 (TSTF-564),
Revision 2, Safety Limit MCPR [Minimum Critical Power Ratio]. The adoption of TSTF-564 revises the TS safety limit on MCPR.
A copy of the safety evaluation is also enclosed. A Notice of Issuance will be included in the Commission's monthly Federal Register notice.
Sincerely,
/RA/
Surinder S. Arora, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-237 and 50-249
Enclosures:
- 1. Amendment No. 285 to DPR-19
- 2. Amendment No. 278 to DPR-25
- 3. Safety Evaluation cc: Listserv
CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 285 Renewed License No. DPR-19
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Constellation Energy Generation, LLC (the licensee), dated August 30, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-19 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 285, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented with the Dresden Nuclear Power Station, Unit 3, fall 2024 refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION Jeffrey A. Whited, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: June 26, 2024 Jeffrey A.
Whited Digitally signed by Jeffrey A. Whited Date: 2024.06.26 13:51:28 -04'00'
CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 278 Renewed License No. DPR-25
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Constellation Energy Generation, LLC (the licensee), dated August 30, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B. of Renewed Facility Operating License No. DPR-25 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 278, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented with the Dresden Nuclear Power Station, Unit 3, fall 2024 refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION Jeffrey A. Whited, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: June 26, 2024 Jeffrey A.
Whited Digitally signed by Jeffrey A. Whited Date: 2024.06.26 13:51:02 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NOS. 285 AND 278 DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 RENEWED FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 DOCKET NOS. 50-237 AND 50-249 Replace the following pages of the Renewed Facility Operating License with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
REMOVE INSERT License DPR-19 License DPR-19 Page 3 Page 3 License DPR-25 License DPR-25 Page 4 Page 4 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
REMOVE INSERT 2.0-1 2.0-1 5.6-2 5.6-2 Renewed License No. DPR-19 Amendment No. 285 (2)
Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear materials as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3)
Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct special nuclear materials as may be produced by the operation of the facility.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2957 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 285, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Operation in the coastdown mode is permitted to 40% power.
Renewed License No. DPR-25 Amendment No. 278
- f.
Surveillance Requirement 4.9.A.10 - Diesel Storage Tank Cleaning (Unit 3 and Unit 2/3 only)
Each of the above Surveillance Requirements shall be successfully demonstrated prior to entering into MODE 2 on the first plant startup following the fourteenth refueling outage (D3R14).
- 3.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
A.
Maximum Power Level The licensee is authorized to operate the facility at steady state power levels not in excess of 2957 megawatts (thermal), except that the licensee shall not operate the facility at power levels in excess of five (5) megawatts (thermal), until satisfactory completion of modifications and final testing of the station output transformer, the auto-depressurization interlock, and the feedwater system, as described in the licensees telegrams; dated February 26, 1971, have been verified in writing by the Commission.
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 278, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
C.
Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.
D.
Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.
E.
Restrictions Operation in the coastdown mode is permitted to 40% power.
SLs 2.0 Dresden 2 and 3 2.0-1 Amendment No. 285/278 263/256 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 685 psig or core flow < 10% rated core flow:
THERMAL POWER shall be 25% RTP.
2.1.1.2 With the reactor steam dome pressure 685 psig and core flow 10% rated core flow:
MCPR shall be 1.07.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be 1345 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.
Reporting Requirements 5.6 Dresden 2 and 3 5.6-2 Amendment No. 285/278 214/206 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued)
(ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.
5.6.3 Radioactive Effluent Release Report
NOTE------------------------------
A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
The Radioactive Effluent Release Report covering the operation of the unit shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
5.6.4 (Deleted) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
1.
The APLHGR for Specification 3.2.1.
2.
The MCPR and MCPR99.9% for Specification 3.2.2.
(continued)
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 285 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-19 AND AMENDMENT NO. 278 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-25 CONSTELLATION ENERGY GENERATION, LLC DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-237 AND 50-249
1.0 INTRODUCTION
By letter dated August 30, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23242A044), Constellation Energy Generation, LLC (CEG or the licensee), submitted a license amendment request (LAR) for the Dresden Nuclear Power Station (DNPS), Units 2 and 3.
The LAR proposed to revise the value of the Technical Specification (TS) 2.1.1.2 reactor core safety limit (SL) minimum critical power ratio (MCPR), which protects against boiling transition on the fuel rods in the core, and to also delete reference to single-loop and two-loop operation.
The current MCPR value is dependent on the number of recirculation loops in operation and ensures that 99.9 percent of the fuel rods in the core are not susceptible to boiling transition, which is referred to as the MCPR99.9%. The revised MCPR would ensure that there is a 95 percent probability at a 95 percent confidence level that no fuel rods will be susceptible to boiling transition using an SL based on critical power ratio (CPR) data statistics, which is referred to as the MCPR95/95. The MCPR95/95 is not dependent on the number of recirculation loops in operation. Additionally, TS 5.6.5, Core Operating Limits Report (COLR), would be modified to require that the COLR include the cycle-specific value for MCPR99.9%, which would still be used to calculate the MCPR operating limit (OL).
The proposed changes are based on Technical Specifications Task Force (TSTF) traveler TSTF-564, Revision 2, Safety Limit MCPR, dated October 24, 2018 (ML18297A361). The U.S. Nuclear Regulatory Commission (NRC or the Commission) issued a final safety evaluation (SE) approving traveler TSTF-564, Revision 2, on November 16, 2018 (ML18299A069).
The LAR proposes variations from the TS changes described in traveler TSTF-564, Revision 2.
The variations are described in Section 2.2 of the LAR and evaluated in section 3.6 of this SE.
In addition, DNPS is currently transitioning from Framatome ATRIUMTM 10XM fuel, to Global Nuclear Fuel (GNF)3 fuel. ATRIUM 10XM is not identified in TSTF-564, Table 1, while GNF3 is identified. As addressed in section 3.5 of this SE, DNPS followed the methodology described in traveler TSTF-564 to demonstrate the MCPR95/95 for the Framatome ATRIUM 10XM fuel is acceptable.
1.1 Background on Boiling Transition During steady-state operation in a boiling-water reactor (BWR), most of the coolant in the core is in a flow regime known as annular flow. In this flow regime, a thin liquid film is pushed up the surface of the fuel rod cladding by the bulk coolant flow, which is mostly water vapor with some liquid water droplets. This provides effective heat removal from the cladding surface; however, under certain conditions, the annular film may dissipate, which reduces the heat transfer and results in an increase in fuel cladding surface temperature. This phenomenon is known as boiling transition or dryout. The elevated surface temperatures resulting from dryout may cause fuel cladding damage or failure.
1.2 Background on Critical Power Correlations For a given set of reactor operating conditions (pressure, flow, etc.), dryout will occur on a fuel assembly at a certain power, known as the critical power. Because the phenomena associated with boiling transition are complex and difficult to model purely mechanistically, thermal-hydraulic test campaigns are undertaken using electrically heated prototypical fuel bundles to establish a comprehensive database of critical power measurements for each BWR fuel product. These data are then used to develop a critical power correlation that can be used to predict the critical power for assemblies in operating reactors. This prediction is usually expressed as the ratio of the actual assembly power to the critical power predicted using the correlation known as the CPR.
One measure of the correlations predictive capability is based on its validation relative to the test data. For each point j in a correlations test database, the experimental critical power ratio (ECPR) is defined as the ratio of the measured critical power to the critical power calculated from the correlation, or:
ECPRj Measured Critical Powerj Critical Power from Correlationj Because the measured critical power includes random variations due to various uncertainties, evaluating the ECPR for all the points in the dataset (or, ideally, a subset of points that were not used in the correlations development) results in a probability distribution. This ECPR distribution allows the predictive uncertainty of the correlation to be determined. This uncertainty can then be used to establish a limit above which there can be assumed that boiling transition will not occur (with a certain probability and confidence level).
As stated in a letter from the TSTF to the NRC1, GNF and Framatome define ECPR as the ratio of the calculated critical power to the measured critical power (i.e., the inverse of the TSTF-564 definition). The letter states that the incorrect formula in the TSTF-564 definition does not affect the model application or TSs. The SLMCPR calculations performed by vendors for their fuel types remain correct. As discussed in TSTF-564, other fuel vendors may determine the 1 Correlation correction information provided in Technical Specifications Task Force (TSTF) letter TSTF-22-09, "Notification of an Error in Approved Traveler TSTF-564, 'Safety Limit MCPR'," dated October 14, 2022 (ML22290A167).
MCPR95/95 for other fuel designs using the methodology described in TSTF-564. This LAR provided the necessary detail of derivation of the MCPR95/95 for ATRIUM 10XM fuel using Equation 1 in section 3.1 of TSTF-564. The methodology provided is based on NRC-approved CPR correlations for each fuel type referenced in DNPS TS 5.6.5.b.
1.3 Background on Thermal-Hydraulic Safety Limits To protect against boiling transition, BWRs have implemented an SL on the CPR, known as the SLMCPR. As discussed in Section 4.4.1.3.1.2, Fuel Cladding Integrity Safety Limit MCPR, of the DNPS Updated Final Safety Analysis Report (UFSAR) and in the Bases for Section 2.1.1, Reactor Core SLs, of the Standard Technical Specifications (STSs) for General Electric BWR plant designs in NUREG-14332, the current calculation of the SLMCPR is to prevent 99.9 percent of the fuel in the core from being susceptible to boiling transition. This limit is typically developed by considering various cycle-specific power distributions and uncertainties and is highly dependent on SL as frequently as every cycle.
The TSs for DNPS also have a limiting condition for operation (LCO) in TS 3.2.2 that governs MCPR, known as the MCPR OL. The OL on MCPR is an LCO which must be met to ensure that anticipated operational occurrences (AOOs) do not result in fuel damage. The current MCPR OL is calculated by combining the largest change in CPR from all analyzed transients, also known as the CPR, with the SLMCPR.
2.0 REGULATORY EVALUATION
2.1 Description of TS Sections 2.1.1 TS 2.1.1, Reactor Core SLs The SLs ensure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and AOOs.
DNPS TS 2.1.1.2 currently requires that when operating with the reactor steam dome pressure greater than or equal to () 685 pounds per square inch gauge (psig) and core flow 10 percent rated core flow, the MCPR shall be 1.08 for two recirculation loop operation or 1.10 for single recirculation loop operation. The SLMCPR (also referred to as MCPR99.9%) ensures that 99.9 percent of the fuel rods in the core are not susceptible to boiling transition.
2.1.2 TS 5.6.5, Core Operating Limits Report (COLR)
DNPS TS 5.6.5 requires core operating limits to be established prior to each reload cycle or prior to any remaining portion of a reload cycle. This TS requires that these limits be documented in the COLR.
2U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric Plants BWR/4, NUREG-1433, Volume 1, Specifications, and Volume 2, Bases, Revision 5.0, September 2021 (ML21272A357 and ML21272A358).
2.2 Description of Proposed Changes to the TSs The licensee proposed to revise the SLMCPR in TS 2.1.1.2 to make it cycle-independent, consistent with the method described in TSTF-564, Revision 2.
The proposed changes to the DNPS TSs revise the value of the SL MCPR in TS 2.1.1.2 to greater than or equal to 1.07, with corresponding changes to the associated TS bases. The change to TS 2.1.1.2 replaces the existing separate SLs for single-loop and two-recirculation loop operation with a single limit since the revised SL (also referred to as the MCPR95/95 SL) is not dependent on the number of recirculation loops in operation.
The MCPR99.9% (i.e., the current MCPR SL) is an input to the MCPR OL in LCO 3.2.2, Minimum Critical Power Ratio (MCPR). While the definition and method of calculation of both the MCPR99.9% and the LCO 3.2.2 MCPR OL remains unchanged, the proposed TS changes include revisions to TS 5.6.5, to require the MCPR99.9% value used in calculating the LCO 3.2.2 MCPR OL to be included in the cycle-specific COLR.
The licensee also stated that both units at DNPS are currently fueled with Framatome ATRIUM 10XM fuel bundles which are not within the scope of TSTF-564, as shown in Table 1 of the approved TSTF-564. As discussed in TSTF-564, other fuel vendors may determine the MCPR95/95 for other fuel designs using the described methodology. The LAR provided the required description of the derivation of the MCPR95/95 for ATRIUM 10XM, which is based on the information contained in each fuel types NRC-approved CPR correlation that is referenced in DNPS TS 5.6.5.b. In addition, the licensee stated that they are transitioning to Global Nuclear Fuels Americas, LLC (GNF) GNF3 fuel at the fall 2023 refueling outage for Unit 2 and the fall 2024 refueling outage for Unit 3.
2.3 Applicable Regulatory Requirements and Guidance Title 10 of the Code of Federal Regulations (10 CFR), section 50.36(a)(1), requires an applicant for an operating license to include in the application proposed TSs in accordance with the requirements of 10 CFR 50.36. The applicant must also include in the application, a summary statement of the bases or reasons for such specifications, other than those covering administrative controls. However, per 10 CFR 50.36(a)(1), these TS bases shall not become part of the technical specifications.
As required by 10 CFR 50.36(c)(1), TSs will include safety limits, limiting safety system settings, and limiting control settings. The regulation, 10 CFR 50.36(c)(1)(i)(A), states, in part:
Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be resumed until authorized by the Commission.
As required by 10 CFR 50.36(c)(2)(i), the TSs will include 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 met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met. Additionally, as required by 10 CFR 50.36(c)(5), TSs must include administrative controls, which are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
Criterion 10, Reactor design, of 10 CFR, part 50, appendix A, General Design Criteria of Nuclear Power Plants, states:
The reactor core and associated coolant control and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
The limits placed on the MCPR act as a specified acceptable fuel design limit to prevent boiling transition, which has the potential to result in fuel rod cladding failure and are used to meet Criterion 10.
DNPS was not licensed to the 10 CFR 50, appendix A, criteria. With respect to General Design Criterion (GDC) 10, Section 2.2 of the LAR states, in part, that:
DNPS UFSAR, Section 3.1.1, "Compliance with Draft Design Criteria, "provides an assessment against the 70 draft GDC published in 1967. The design basis of Unit 2 was later evaluated against the final "General Design Criteria for Nuclear Power Plants," published as 10 CFR 50, Appendix A in July 1971. This evaluation is presented in UFSAR Section 3.1.2, "Compliance with Final Design Criteria," and concluded that DNPS, Unit 2 satisfies and is in compliance with the intent of the General Design Criteria. This evaluation was performed specifically for Unit 2 and may not fully apply to Unit 3; however, the high degree of similarity between the design of Units 2 and 3 indicates that Unit 3 also conforms to the intent of the General Design Criteria.
The NRC staffs guidance contained in Revision 2 of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]
Edition (SRP), Section 4.4, Thermal and Hydraulic Design provides the following two examples of acceptable approaches to meet SRP Acceptance Criterion 1 for establishing fuel design limits:
A. For departure from nucleate boiling ratio (DNBR), CHFR [critical heat flux ratio] or CPR correlations, there should be a 95-percent probability at the 95-percent confidence level that the hot rod in the core does not experience a DNB or boiling transition condition during normal operation or AOOs.
B. The limiting (minimum) value of DNBR, CHFR, or CPR, correlations is to be established such that at least 99.9 percent of the fuel rods in the core will not experience a DNB or boiling transition during normal operation or AOOs.
The NRC staffs guidance for the review of TSs is in Chapter 16.0, Revision 3, Technical Specifications, of the SRP, dated March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STSs for each of the LWR nuclear designs. Accordingly, the NRC staffs review considers whether the proposed changes are consistent with the applicable reference STSs (i.e., the current STSs), as modified by NRC-approved travelers. The STS applicable to DNPS is NUREG-1433, Revision 5.0, Standard Technical Specifications, General Electric Plants BWR/4, Volume 1, Specifications, and Volume 2, Bases, September 2021 (ML21272A357 and ML21272A358, respectively).
3.0 TECHNICAL EVALUATION
3.1 Basis for Proposed Change As discussed in section 1.3 of this SE, the current SLMCPR (i.e., the MCPR99.9% SL) is dependent on the cycle-specific core design, especially including the core power distribution, fuel type(s) in the reactor, and the power-to-flow operating domain for the plant. As such, it is frequently necessary to change the MCPR SL to accommodate new core designs. Changes to the SLMCPR are usually determined late in the design process and necessitate an accelerated NRC review (i.e., LAR) to support the subsequent fuel cycle.
The licensee proposed to change the calculation for determining the SLMCPR for DNPS so that it is no longer cycle-dependent, reducing the frequency of revisions and eliminating the need for NRCs review on an accelerated schedule. The proposed methodology for determining the SLMCPR aligns it with the DNBR SL used in Pressurized Water Reactors, which ensures a 95 percent probability at a 95 percent confidence level that no fuel rods will experience DNB.
The NRC staff finds that calculating the revised SLMCPR based on the 95/95 criterion is acceptable because it meets SRP Section 4.4, Acceptance Criterion 1. The remainder of this SE evaluates whether the methodology for determining the revised SLMCPR provides the intended result and documents the review to ensure that the revised SLMCPR can be adequately determined in the core using various types of fuel, whether the proposed SL continues to fulfil the necessary functions of an SL without unintended consequences, and whether the proposed changes have been adequately implemented in the DNPS, Units 2 and 3, TSs.
3.2 Revised SLMCPR Definition As discussed in section 1.2 of this SE, a critical power correlations ECPR distribution quantifies the uncertainty associated with the correlation. Traveler TSTF-564, Revision 2, provides a definition for a limit that bounds 95 percent of a correlations ECPR distribution at a 95 percent confidence level, according to the following formula:
MCPR9595(i) = i + i i where i is the mean ECPR and i is the standard deviation of the ECPR distribution. The statistical parameter (i) is selected, based on the number of samples in the critical power database, to provide 95% probability at 95% confidence (95/95) for the one-sided upper tolerance limit that depends on the number of samples (Ni) in the critical power database. This is a commonly used statistical formula to determine a 95/95 one-sided upper tolerance limit for a normal distribution, which is appropriate for the situation under consideration. The factor is generally attributed to D. B. Owen3 and was also reported by M. G. Natrella4, as referenced in traveler TSTF-564, Revision 2. Example values of are provided in Table 2 of traveler TSTF-564, Revision 2. Table 1 of the traveler includes some reference values of the MCPR95/95.
DNPS, Units 2 and 3, are transitioning from ATRIUM 10XM to GNF3 fuel. GNF3 fuel is identified in Table 1 of TSTF-564. ATRIUM 10XM is not identified in Table 1 of TSTF-564. As discussed in TSTF-564, other fuel vendors may determine the MCPR95/95 for other fuel designs using the methodology. The NRC staff verified the MCPR95/95 value for ATRIUM 10XM that was calculated by the licensee and confirmed that the MCPR95/95 value for GNF3 bounds the MCPR95/95 value for ATRIUM 10XM. Therefore, the NRC staff finds the calculation and use of the MCPR95/95 values acceptable for application to DNPS, Units 2 and 3.
As discussed by Piepel and Cuta5 for DNBR correlations, the acceptability of this approach is predicated on a variety of assumptions, including the assumptions that the correlation data comes from a common population and that the correlations population is distributed normally.
These assumptions are typically addressed generically when a critical power or critical heat flux correlation is reviewed by the NRC staff, and possible penalties to the correlation in order to account for uncertainties are identified. The TSTF letter dated May 29, 2018 (ML18149A320),
states that such penalties applied during the NRCs review of the critical power correlation would be imposed on the mean or standard deviation used in the calculating the MCPR95/95.
These penalties would also continue to be imposed in the determination of the MCPR99.9%, along with any other penalties associated with the process of (or other inputs used in) determining the MCPR99.9% (e.g., penalties applied to the MCPR99.9% SL for operation in the Maximum Extended Load Limit Line Analysis Plus (MELLLA+) operating domain).
In the SE for TSTF-564, Revision 2, the NRC staff found that the definition of the MCPR95/95 appropriately establishes a 95/95 upper tolerance limit on the critical power correlation and that any issues in the underlying correlation will be addressed through penalties on the correlation mean and standard deviation, as necessary. Therefore, the NRC staff concludes that the method for determining MCPR95/95 can be used to establish acceptable fuel design limits in the DNPS, Unit 2 and 3 TSs.
3.3 Determination of Revised SLMCPR for Mixed Cores Both DNPS, Units 2 and 3, are transitioning from Framatome ATRIUM 10XM fuel to GNF3.
Based on the information provided in the LAR, the licensee began loading GNF3 fuel into the DNPS, Unit 2, core during the fall 2023 refueling outage and will begin loading GNF3 fuel into DNPS, Unit 3, core during fall 2024 refueling outage.
Revision 2 of TSTF-564, proposed that a core containing a variety of fuel types would evaluate the MCPR95/95 for all of the fresh and once-burnt fuel in the core and apply the most limiting (i.e.,
the largest) value of MCPR95/95 for each of the applicable fuel types as the MCPR SL. As stated in Section 3.1 of TSTF-564, Revision 2, this is because bundles that are twice-burnt or more at the beginning of the cycle have significant MCPR margin relative to the fresh and once-burnt 3 D. B. Owen, Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, Sandia Corporation, SCR-607, March 1963, (ML14031A495).
4 M. G. Natrella, Experimental Statistics, National Bureau of Standards, National Bureau of Standards Handbook 91 August 1963.
5 G. F. Piepel and J. M. Cuta, Statistical Concepts and Techniques for Developing, Evaluating, and Validating CHF Models and Corresponding Fuel Design Limits, SKI Technical Report, 93:46, 1993.
fuel. The justification is that the MCPR for twice-burnt and greater fuel is far enough from the MCPR for the limiting bundle that its probability of boiling transition is very small compared to the limiting bundle and it can be neglected in determining the SL. Results of a study provided in the TSTF letter dated May 29, 2018, indicate that this is the case even for fuel operated on short (12-month) reload cycles. As discussed in the traveler, twice-burnt or greater fuel bundles are included in the cycle-specific evaluation of the MCPR99.9% and the MCPR OL. If a twice-burnt or greater fuel bundle is found to be limiting, it would be governed by the MCPR OL, which will always be more restrictive than both the MCPR95/95 and the MCPR99.9%. In the SE for TSTF-564, Revision 2, the NRC staff found the justification for applying the most limiting (i.e., the largest) value of MCPR95/95 for each of the applicable fuel types as the SLMCPR to be appropriate and determined that it is acceptable to determine the MCPR95/95 SL for the core based on the most limiting value of the MCPR95/95 for the fresh and once-burnt fuel in the core.
In the SE for TSTF-564, Revision 2, the NRC staff also reviewed the information furnished by the TSTF and found that the process for establishing the revised SLMCPR (i.e., MCPR95/95 SL) for mixed cores ensures that the limiting fuel types in the core will be evaluated and the limiting MCPR95/95 will be appropriately applied as the SL. Therefore, the NRC staff finds it acceptable to determine the MCPR95/95 SL for the core based on the most limiting MCPR95/95 value for fresh and once-burnt fuel in the core for the DNPS, Units 2 and 3, TSs.
The LAR proposed a SL value in SL 2.1.1.2 of greater than or equal to 1.07, which is the most limiting value for ATRIUM 10XM and GNF3 fuel types. The LAR stated that GNF3 is identified as the fuel type that the SL is based upon since it is the most limiting value, and it is consistent with the TSTF 564 guidance to use the largest MCPR95/95 value for the transition cores containing both ATRIUM 10XM and GNF3. The NRC staff finds the justification provided for use of the SL value for the mixed core at DNPS, Units 2 and 3 to be consistent with the guidance provided in TSTF 564 and is, therefore, acceptable.
3.4 Relationship Between MCPR Safety and Operating Limits As discussed in the TSTF letter dated May 29, 2018, the MCPR99.9% SL is expected to always be greater than the MCPR95/95 SL for two reasons: (1) the MCPR99.9% includes uncertainties not factored into the MCPR95/95, and (2) the 99.9 percent probability basis for determining the MCPR99.9% is more conservative than the 95 percent probability at a 95 percent confidence level used in determining the MCPR95/95. The level of conservatism in the MCPR95/95 SL is appropriate because the lead fuel rod in the core (i.e., the limiting fuel rod with respect to MCPR) is used to evaluate whether any fuel rods in the core are susceptible to boiling transition, which is also discussed in the traveler.
Consistent with TSTF-564, the MCPR OL defined in LCO 3.2.2 will continue to be evaluated using the MCPR99.9% as an input. The MCPR99.9% will continue to be evaluated in the same way as it is currently, using the whole core. The LAR is not proposing a change to LCO 3.2.2 and will continue to determine the MCPR operating limits for LCO 3.2.2 at DNPS using the MCPR99.9%
as an input.
Consistent with TSTF-564, Revision 2, the licensee proposed to revise DNPS TS 5.6.5.a.2 to require inclusion of the cycle-specific value of the MCPR99.9% in the COLR to ensure that the uncertainties being removed from the SLMCPR are still included as part of the MCPR OL.
The methods used for determining MCPR99.9% are included in the list of COLR references contained in TS 5.6.5.b. Specifically, DNPS TS 5.6.5.b states that the analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents. The use of the NRC-approved methodologies listed in TS 5.6.5.b help to ensure that the uncertainties being removed from the MCPR SL are still included as part of the MCPR OL and will continue to appropriately inform plant operation.
The NRC staff finds that the changes proposed by the licensee will retain an adequate level of conservatism in the SLMCPR in TS 2.1.1.2 while appropriately ensuring that plant-and cycle-specific uncertainties will be retained in the MCPR OL. The MCPR95/95 represents a lower limit on the value of the MCPR99.9%, because the MCPR99.9% should always be higher since it accounts for numerous uncertainties that are not included in the MCPR95/95 (as discussed in section 3.1 of TSTF-564, Revision 2).
3.5 Implementation of the Revised SLMCPR in the TSs The licensee proposed to change the value of the SL in TS 2.1.1.2 ATRIUM 10XM and GNF3 to 1.07. The value reported in the DNPS, Units 2 and 3, TS 2.1.1.2 will be the GNF3 MCPR95/95 value proposed by the vendor as shown in Table 1 of TSTF 564. The value was reported at a precision of two digits past the decimal point. The proposed new SLMCPR (MCPR95/95SL) in Specification 2.1.1.2 is consistent with the TSTF-564 guidance to use the largest MCPR95/95 value for the transition cores containing both ATRIUM 10XM and GNF3.
Consistent with TSTF-564, the LAR also proposes to modify DNPS TS 5.6.5 to include the value of the MCPR99.9% to ensure that the cycle-specific MCPR99.9% value will continue to be determined for LCO 3.2.2 and reported in the COLR. As previously approved by the NRC6 for DNPS, the MCPR99.9% will determined using: 1) the GEXL98 correlation in NEDC-33930P, Revision 0, "GEXL98 Correlation for ATRIUM 10XM Fuel," February 2021,7 for ATRIUM10 fuel and 2) NEDC-33880, Revision 0, GEXL21 Correlation for GNF3 Fuel, March 2017, (ML17096A520). Both reports are included in DNPS TS 5.6.5.b. The GEXL 98 report is item 21 in TS 5.6.5.b. The GEXL21 report is included in General Electric topical report NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR II, Main), November 2020 (ML20330A199), which is listed in the NRC approved COLR methodologies in TS 5.6.5.b as Item 1. The COLR, therefore, will continue to report the cycle-specific value of the MCPR OL contained in LCO 3.2.2, and DNPS TS 5.6.5.b will continue to reference appropriate NRC-approved methodologies for determination of the MCPR99.9% and the MCPR OL. Therefore, the NRC staff finds the proposed change to TS 5.6.5 to be acceptable.
The DNPS reactors are transitioning from ATRIUM 10XM fuel assemblies provided by Framatome, Inc., to GNF3 fuel assemblies provided by Global Nuclear Fuel - Americas, LLC.
The GNF3 fuel type is identified in Table 1 of TSTF-564 but the ATRIUM 10 XM fuel type is not. The new SLMCPR (MCPR95/95 SL) in TS 2.1.1.2 is consistent with the TSTF-564 guidance to use the largest MCPR95/95 value for the transition cores containing both ATRIUM 10XM and GNF3. The MCPR95/95 SL is only fuel type dependent and not plant and/or cycle dependent.
6 U.S. Nuclear Regulatory Commission, Dresden Nuclear Power Station, Units 2 And 3 - Issuance of Amendment Nos. 281 and 274 Re: Transition to GNF3 Fuel (EPID L-2022-LLA-0121), July 6, 2023 (ML23144A314).
7A non-proprietary version was included as Attachment 4 to the Quad Cities Nuclear Power Station license amendment request dated September 14, 2021 (ML21257A420).
Therefore, applying the TSTF-564 approach for additional fuel types is within the scope of the TSTF-564 approval and does not affect the applicability of the TSTF to the DNPS TS.
The NRC staff, therefore, finds the proposed change to the SL in TS 2.1.1.2 acceptable. The licensee derived the SL consistent with the process described in traveler TSTF-564.
3.6 Evaluation of Proposed Variations The NRC staff notes that DNPS, Units 2 and 3, TSs have different numbering from STS for the COLR; specifically, DNPS, TS 5.6.5 versus STS 5.6.3. The NRC staff finds that the different TS numbering is acceptable because it does not alter TS requirements.
The licensee stated in the LAR that the DNPS, Units 2 and 3, TSs specify a different steam dome pressure value (685 psig) in the TS 2.1.1.1 and 2.1.1.2 rather than the value specified in STS (785 psig). The NRC staff finds that this plant-specific value does not affect applicability of TSTF 564. The acceptability of the proposed changes in TSTF-564 are not predicated on or related to steam dome pressure. Therefore, this variation is acceptable.
As addressed earlier in section 2.3 of the SE, DNPS was not licensed under GDC 10, but instead was licensed under the applicable AEC preliminary general design criteria. The DNPS UFSAR, Section 3.1.1, "Compliance with Draft Design Criteria," provides an assessment against the 70 draft GDC published in 1967. The licensee states in the LAR that the design basis of DNPS, Unit 2, was evaluated against the final General Design Criteria for Nuclear Power Plants, published as 10 CFR 50, appendix A, in July 1971 and found it to be in compliance with the intent of the General Design Criteria (presented in DNPS UFSAR, Section 3.1.2). The licensee further states that while the evaluation was performed specifically for Unit 2 and may not fully apply to Unit 3, the high degree of similarity between the design of Units 2 and 3 indicates that Unit 3 also conforms to the intent of the GDC, and that this difference does not alter the conclusion that the proposed change is applicable to DNPS. The NRC staff finds the licensee performed evaluation to be acceptable based on the similarity between the design of the two Units. Due to this similarity, the NRC staff determined that this difference does not affect the applicability of TSTF-564 for the proposed amendments to the DNPS TS.
3.7 NRC Staff Conclusion
The NRC staff reviewed the licensees proposed TS changes and determined that the proposed SL associated with TS 2.1.1.2 was calculated in a manner consistent with the process described in TSTF-564, Revision 2, and was, therefore, acceptably modified to suit the revised definition of the SLMCPR. Under the new definition, the SLMCPR will continue to protect the fuel cladding against the uncontrolled release of radioactivity by preventing the onset of boiling transition, thereby fulfilling the requirements of 10 CFR 50.36(c)(1) for SLs.
The MCPR OL in LCO 3.2.2 remains unchanged and will continue to meet the requirements of 10 CFR 50.36(c)(2) and DNPSs plant-specific design criterion similar to GDC 10 as discussed in section 2.3 of this SE by ensuring that no fuel damage results during normal operation and AOOs. The NRC staff determined that the proposed changes to DNPS, Units 2 and 3, TS 5.6.5 are acceptable. Upon adoption of the revised SLMCPR, the COLR will be required to contain the MCPR99.9%, supporting the determination of the MCPR OL using current methodologies.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendment on May 13, 2024. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change the requirements with respect to installation or use of a facilitys components located within the restricted area as defined in 10 CFR, part 20, or a change to SRs. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (88 FR 74529). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: R. Elliott, NRR F. Forsaty, NRR Date: June 26, 2024
ML24138A057 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/SNSB/BC NRR/DSS/STSB/BC NAME SArora SRohrer PSahd SMehta DATE 5/16/2024 5/21/2024 4/5/2024 5/20/2024 OFFICE OGC - NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME MWright JWhited SArora DATE 6/3/2024 6/26/2024 6/26/2024