ML23340A155
| ML23340A155 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 03/15/2024 |
| From: | Robert Kuntz Plant Licensing Branch III |
| To: | Rhoades D Constellation Energy Generation |
| Kuntz R | |
| References | |
| EPID L-2023-LLA-0031 | |
| Download: ML23340A155 (22) | |
Text
March 15, 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:
QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 299 AND 295 RE: ADOPTION OF TSTF-564, SAFETY LIMIT MCPR (EPID L-2023-LLA-0031)
Dear David Rhoades:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 299 to Renewed Facility Operating License No. DPR-29 and Amendment No. 295 to Renewed Facility Operating License No. DPR-30 for the Quad Cities Nuclear Power Station, Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated March 3, 2023.
The amendments adopt Technical Specifications 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 related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Robert F. Kuntz, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-254 and 50-265
Enclosures:
- 1. Amendment No. 299 to DPR-29
- 2. Amendment No. 295 to DPR-30
- 3. Safety Evaluation cc: Listserv
CONSTELLATION ENERGY GENERATION, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50 254 QUAD CITIES NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 299 Renewed License No. DPR-29
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Constellation Energy Generation, LLC (the licensee) dated March 3, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions 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-29 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 299, 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 prior to the start-up from the Spring 2024 Quad Cities Nuclear Power Station, Unit 2 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: March 15, 2024 Jeffrey A.
Whited Digitally signed by Jeffrey A. Whited Date: 2024.03.15 13:20:19 -04'00'
CONSTELLATION ENERGY GENERATION, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 295 Renewed License No. DPR-30
- 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 March 3, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions 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-30 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 295, 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 prior to the start-up from the Spring 2024 Quad Cities Nuclear Power Station, Unit 2 outage.
FOR THE NUCLEAR REGULATORY COMMISSION Jeffery 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: March 15, 2024 Jeffrey A.
Whited Digitally signed by Jeffrey A. Whited Date: 2024.03.15 13:19:46 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NOS. 299 AND 295 RENEWED FACILITY OPERATING LICENSE NOS. DPR-29 AND DPR-30 QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert License No. DPR-29 License No. DPR-29 Page 4 Page 4 License No. DPR-30 License No. DPR-30 Page 4 Page 4 TSs TSs 2.0-1 2.0-1 5.6-2 5.6-2 Renewed License No. DPR-29 Amendment No. 299 B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 299, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
C.
The licensee shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:
The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.
D.
Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.
E.
The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined sets of plans1, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2, submitted by letter dated May 17, 2006.
Constellation Energy Generation, LLC shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).
The CSP was approved by License Amendment No. 249 as modified by License Amendment No. 259.
1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.
Renewed License No. DPR-30 Amendment No. 295 B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 295, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
C.
The licensee shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:
The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.
D.
Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.
E.
The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2, submitted by letter dated May 17, 2006.
Constellation Energy Generation, LLC shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The CSP was approved by License Amendment No. 244 and modified by License Amendment No. 254.
1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.
SLs 2.0 Quad Cities 1 and 2 2.0-1 Amendment No. 299/295 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 5.6 Reporting Requirements Quad Cities 1 and 2 5.6-2 Amendment No. 299/295 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. 299 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT NO. 295 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-30 CONSTELLATION ENERGY GENERATION, LLC AND MIDAMERICAN ENERGY COMPANY QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265
1.0 INTRODUCTION
By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated March 3, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23062A450), Constellation Energy Generation, LLC (the licensee) submitted a license amendment request (LAR) to change the technical specifications (TSs) for the Quad Cities Nuclear Power Station (Quad Cities), Units 1 and 2.
The proposed amendment changes the TS safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL based upon NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-564, Revision 2, Safety Limit MCPR (TSTF-564, ML18297A361),
and the associated NRC staff safety evaluation (SE) (ML18299A069).
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 critical power ratio (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 calculated critical power, or:
Measured Critical Powerj ECPRj =
Calculated Critical Powerj For ECPR values less than or equal to 1, the calculated critical power is greater than or equal to the measured critical power and the prediction is considered to be non-conservative. Because the measured critical power includes random variations due to various uncertainties, evaluating the ECPR for all of 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).
Fuel vendors may determine the MCPR95/95 for other fuel designs using the methodology described in TSTF-564. The licensee 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 information provided is based on NRC-approved CPR correlations for each fuel type.
1.3 Background on Thermal Hydraulic Safety Limits To protect against boiling transition, BWRs have implemented an SL on the CPR, known as the MCPR SL. As discussed in NUREG-1433, Revision 4 (ML12104A192) for the TSs and ML12104A193 for the bases) the STSs at the time TSTF-564 was approved for the applicable BWR design, the basis of the MCPR SL for the licensees facility 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 the cycle-specific radial power distribution in the core. As such, the limit may need to be updated as frequently as every cycle.
The TSs for Quad Cities, Units 1 and 2, also have a limiting condition for operation (LCO) that governs MCPR, known as the MCPR OL (operating limit). The OL on MCPR is an LCO which must be met to ensure that anticipated operational occurrences 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 MCPR SL.
2.0 REGULATORY EVALUATION
2.1 Description of Proposed Change TSTF-564 modifies Standard Technical Specification (STS) 2.1.1, Reactor Core SLs. SLs ensure that specified acceptable fuel design limits are not exceeded during steady-state operation, normal operational transients, and anticipated operational occurrences (AOOs).
Quad Cities, Units 1 and 2, TS 2.1.1.2 currently requires that:
With the reactor steam dome pressure 685 psig [pounds per square inch gauge] and core flow 10% rated core flow:
For two recirculation loop operation, MCPR shall be 1.08, or for single recirculation loop operation, MCPR shall be 1.10.
The MCPR SL (MCPR99.9%) ensures that 99.9-percent of the fuel in the core is not susceptible to boiling transition.
TSTF-564 also modifies STS 5.6.3 for core operating limits report (COLR). STS 5.6.3 corresponds to Quad Cities, Units 1 and 2, TS 5.6.5.
Quad Cities, Units 1 and 2, TS 5.6.5 currently requires that:
- 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 [average planar linear heat generation rate] for Specification 3.2.1.
- 2. The MCPR for Specification 3.2.2.
The licensee proposes to revise the MCPR SL, consistent with the method described in TSTF-564 to make it cycle-independent.
The proposed changes to the Quad Cities, Units 1 and 2, TSs would revise the value of the MCPR SL in TS 2.1.1.2 to 1.07.
The change to TS 2.1.1.2 replaces the existing separate MCPR SL for single-recirculation loop operation and two-recirculation loop operation, with a single limit since the revised SL is no longer dependent on the number of recirculation loops in operation.
The current MCPR SL (i.e., MCPR99.9%) is an input to the MCPR OL in LCO 3.2.2, Minimum Critical Power Ratio (MCPR). 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 definition and method of calculation of both the MCPR99.9% and the LCO 3.2.2 MCPR OL remains unchanged.
2.2 Applicable Regulatory Requirements The regulation at 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 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), TSs will include SLs, limiting safety system settings, and limiting control settings. As required by 10 CFR 50.36(c)(1)(i)(A), SLs 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.
The original design of Quad Cities, Units 1 and 2, was reviewed and approved against the draft General Design Criteria (GDC) issued in July 1967. As stated in Section 3.1 of the licensees Updated Final Safety Analysis Report (ML24051A049), based on the licensees understanding of the intent of the proposed criteria, Quad Cities station fully satisfies the intent of the criteria.
As the GDC were finalized, the requirements were placed in Appendix A, General Design Criteria for Nuclear Power Plants to 10 CFR Part 50.
Based on its review, the staff finds that design Criterion 6, used at Quad Cities, Units 1 and 2, is analogous to and meets the intent of GDC 10.
GDC 10, Reactor Design, of 10 CFR Part 50 Appendix A 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 GDC 10.
2.3 Applicable Regulatory Guidance 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, (ML070550060) provides the following two examples of acceptable approaches to meeting the SRP acceptance criteria for establishing fuel design limits (as stated in SRP Acceptance Criterion 1):
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 [departure from nucleate boiling] 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 (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 STSs applicable to Quad Cities, Units 1 and 2, 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).
Revision 5 to the STSs incorporates NRC-approved changes from TSTF-564.
3.0 TECHNICAL EVALUATION
3.1 Background
Quad Cities, Units 1 and 2, are currently using Framatome ATRIUM 10XM fuel which is not explicitly identified in TSTF-564 Table 1. The Quad Cities, Units 1 and 2, TSs also specify a different reactor steam dome pressure value compared to the value specified in the STSs on which TSTF-564 was based. In addition, the Quad Cities, Units 1 and 2 TSs uses different numbering than the STSs for the COLR; specifically, Quad Cities uses TS 5.6.5 versus STS 5.6.3. These variations are described and evaluated in section 3.7 of this SE.
Quad Cities, Unit 1, began loading Global Nuclear Fuels Americas, LLC (GNF) GNF3 fuel during the spring 2023 refueling outage, and the Unit 2 fuel transition will begin in the spring 2024 refueling outage. Based on TSTF-564, for transition cores loaded with a mix of fuel types, the SLMCPR95/95 is based on the largest MCPR95/95 value for the fuel types used. During the initial phases of loading GNF fuel at each unit, the core configuration will include once-burnt ATRIUM 10XM and fresh GNF3 fuel.
The LAR proposes to revise TS SL 2.1.1.2, the reactor core SL for the MCPR. The MCPR ensures protection against boiling transition on the fuel rods in the reactor core. The current MCPR SL for Quad Cities ensures that 99.9-percent of the fuel rods would not be susceptible to boiling transition (i.e., reduced heat transfer) during normal operation and anticipated operational occurrences and is referred to as MCPR99.9%. The revised MCPR SL will 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 CPR data statistics and is referred to as the MCPR95/95. The revised MCPR would also delete reference to single-and two-loop operation because MCPR95/95 is not dependent on the number of recirculation loops in operation. The proposed amendment reduces the need for cycle-specific changes to the SL value while still meeting the regulatory requirement for an SL. Quad Cities TS 5.6.5, COLR subsection a.2, is also proposed to be modified to require the MCPR (existing requirement) and MCPR99.9% (new requirement) value to be included in the cycle-specific COLR.
3.2 Basis for Proposed Changes The current MCPR SL (i.e., the MCPR99.9%), is affected by the plants 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 MCPR SL are usually determined late in the design process and necessitate an accelerated NRC review (i.e., LAR) to support the subsequent fuel cycle.
The LAR proposed to change the basis for the MCPR SL for Quad Cities, Units 1 and 2, 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 MCPR SL aligns it with that of the DNBR SL used in pressurized water reactors, which provides a 95 percent probability at a 95 percent confidence level that no fuel rods will experience departure from nucleate boiling.
The NRC staff finds the proposed basis for the revised MCPR SL calculation is acceptable based on the discussion in SRP, section 4.4, SRP Acceptance Criterion 1. The remainder of this SE will ensure that the methodology for determining the revised MCPR SL provides the intended results and that the revised MCPR SL can be adequately determined in the core using various types of fuel. The SE will also evaluate whether the proposed SL continues to fulfil the necessary functions of a SL without unintended consequences, and the proposed changes have been adequately implemented in the Quad Cities, Units 1 and 2, TSs.
3.3 Revised MCPR SL Definition A critical power correlations ECPR distribution quantifies the uncertainty associated with the correlation. TSTF-564 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:
MCPR95/95(i) = µi + Kii Where µi is the correlations mean ECPR, i is the standard deviation of the correlations ECPR distribution, and Ki is a statistical parameter chosen to provide 95% probability at 95%
confidence (95/95) for the one-sided upper tolerance limit that depends on the number of samples ( ) in the critical power database. This formula is commonly used 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. Owen (ML14031A495) and was also reported by M. G. Natrella Experimental Statistics, National Bureau of Standards, National Bureau of Standards Handbook 91 August 1963, as referenced in TSTF-564.
Quad Cities, Units 1 and 2, 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 Quad Cities, Units 1 and 2.
As discussed by Piepel and Cuta Statistical Concepts and Techniques for Developing, Evaluating, and Validating CHF Models and Corresponding Fuel Design Limits, SKI Technical Report, 93:46, 1993 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, who may apply penalties to the correlation to account for any issues identified. TSTF-564 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 approving TSTF-564, the NRC staff found that the definition of the MCPR95/95 will appropriately establish 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, as proposed, can be used to establish acceptable fuel design limits in the Quad Cities, Units 1 and 2, TSs.
3.4 Determination of Revised MCPR SL for Mixed Cores Currently, Quad Cities, Units 1 and 2, reactors are fueled with Framatome ATRIUM-10XM fuel.
Based on the information provided in the LAR, the licensee plans to begin loading GNF3 fuel into the Quad Cities, Unit 1, core during the spring 2023 refueling outage and into Quad Cities, Unit 2, core during spring 2024 refueling outage.
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, 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 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 TSTF-564 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%. The NRC staff finds this justification 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.
The LAR proposed a SL value in SL 2.1.1.2 of 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 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 SL value for mixed core consistent with the guidance provided in TSTF-564 and acceptable for the Quad Cities, Units 1 and 2 TSs.
3.5 Relationship Between MCPR Safety and Operating Limits As discussed in TSTF-564, the MCPR99.9% is expected to always be greater (and thus more conservative) than the MCPR95/95 because: (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 as discussed in the traveler).
Consistent with TSTF-564, the MCPR OL defined in LCO 3.2.2 would 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 licensee 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 Quad Cities, Units 1 and 2.
Consistent with TSTF-564, the licensee proposed to revise the COLR TS (Quad Cities, Units 1 and 2, TS 5.6.5) to require the cycle-specific value of the MCPR99.9% to be included in the COLR. The methods supporting the inclusion of the MCPR99.9% must also therefore, be included in the list of COLR references contained in TS 5.6.5.b. The Quad Cities, Units 1 and 2, TS 5.6.5.b, states that methods used to determine the core operating limits shall be those reviewed and approved by the NRC and support 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.
Based on the review, the NRC staff, therefore, finds that the changes proposed by the licensee will retain an adequate level of conservatism in the MCPR SL in TS 2.1.1.2 while appropriately ensuring that plant-and cycle-specific uncertainties will be retained in the MCPR OL. The NRC staff notes that the MCPR95/95 represents a hard floor on the value of the MCPR99.9%, which 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).
3.6 Implementation of the Revised MCPR SL in the TSs The licensee proposes to change the value of the SL in TS 2.1.1.2 for ATRIUM 10XM and GNF3 to 1.07. The value reported in Quad Cities, Units 1 and 2, TS 2.1.1.2 will be the GNF3 MCPR95/95 value proposed by the vendor as mentioned in Table 1 of TSTF-564. The value was reported at a precision of two digits past the decimal point.
Consistent with TSTF-564, the licensee also proposes to modify Quad Cities, Units 1 and 2, 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. The COLR, therefore, will continue to report the cycle-specific value of the MCPR OL contained in LCO 3.2.2, and Quad Cities, Units 1 and 2, TS 5.6.5 will continue to use 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 LAR indicated that a SL value of 1.05 was calculated by Framatome for ATRIUM-10XM fuel using Equation 1 in section 3.1 of TSTF-564 and consistent with amendments that have been previously approved for other facilities (ML17059D146, ML21223A280, and ML22146A207).
The LAR proposed value of 1.07 to be included in TS 2.1.1.2 is consistent with TSTF-564 guidance to use the most limiting value which is the MCPR95/95 value GNF3 fuel.
The NRC staff, therefore, finds the proposed change to the SL in TS 2.1.1.2 is acceptable. The licensee derived the SL consistent with the process described in TSTF-564.
3.7 Variations from TSTF-564 The NRC staff notes that Quad Cities, Units 1 and 2, TSs have a different numbering than STS for the COLR; specifically, Quad Cities, Units 1 and 2, 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 states in the LAR that the Quad Cities, Units 1 and 2, 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 licensee states that this plant-specific value does not affect applicability of TSTF-564. This is not a change in licensee dome pressure value, and the licensee is noting variation from the value noted in STS. Therefore, the NRC staff finds the licensees applicability statement acceptable.
As noted in Section 2.2 of this SE, Quad Cities, Units 1 and 2, were not licensed to the GDC in Appendix A to 10 CFR Part 50. However, based on its review, the NRC staff finds that design Criterion 6, used at Quad Cities, Units 1 and 2, is analogous to and meets the intent of GDC 10.
Therefore, the NRC staff finds that this variation is acceptable.
3.8 NRC Staff Conclusion
The NRC staff reviewed the proposed TS changes in the LAR 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 and is therefore acceptably modified to suit the revised definition of the MCPR SL. Under the new definition, the MCPR SL 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 the intent of GDC 10, by ensuring that no fuel damage results during normal operation and AOOs. The NRC staff determined that the proposed changes to TS 5.6.5 are acceptable; upon adoption of the revised MCPR SL, 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 Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendments on December 5, 2023. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR, part 20, and change surveillance requirements. 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 23695; dated April 18, 2023). 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Noushin Amini, NRR Date of Issuance: March 15, 2024
ML23340A155 NRR-058 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/SNSB/BC NRR/DSS/STSB/BC NAME RKuntz SRohrer PSahd SMehta (A) (CAshley for)
DATE 12/6/2023 12/7/2023 12/12/2023 12/12/2023 OFFICE OGC NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME ANaber JWhited RKuntz DATE 3/15/2024 3/15/2024 3/15/2024