ML20269A305
| ML20269A305 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/29/2020 |
| From: | Andrew Hon Plant Licensing Branch II |
| To: | Krakuszeski J Duke Energy Progress |
| Hon A | |
| References | |
| EPID L-2020-LLA-0043 | |
| Download: ML20269A305 (27) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION September 29, 2020 Mr. John A. Krakuszeski Vice President Brunswick Steam Electric Plant Duke Energy Progress, LLC 8470 River Rd. SE (M/C BNP001)
Southport, NC 28461
SUBJECT:
BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 301 AND 329 TO REVISE TECHNICAL SPECIFICATION TO ADOPT TSTF-564 (EPID L-2020-LLA-0043)
The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 301 to Renewed Facility Operating License No. DPR-71 and Amendment No. 329 to Renewed Facility Operating License No. DPR-62 for Brunswick Steam Electric Plant, Units 1 and 2. The amendments consist of changes to the technical specifications (TS) in response to your application dated March 9, 2020. The amendments revise the Technical Specification (TS) safety limit (SL) 2.1.1.2, which protects against boiling transition on the fuel rods in the core. Your proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-564, Revision 2, Safety Limit MCPR [Minimum Critical Power Ratio],
dated October 24, 2018. The proposed amendments revise the TS safety limit on minimum critical power ratio to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for a safety limit. The amendments also revise TS 5.6.5, Core Operating Limits Report (COLR).
The NRC staff has completed its review of the information provided by the licensee. provides the staffs safety evaluation (SE). The staff has determined that it contains proprietary information pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 2.390, Public Inspections, Exemptions, Requests for Withholding.
Accordingly, the NRC staff has prepared a redacted nonproprietary version (Enclosure 4). The NRC staff will delay placing the nonproprietary SE in the public document room for a period of 10 working days from the date of this letter to allow you to comment on any proprietary aspects.
If you believe that any information in Enclosure 4 is proprietary, please identify such information line by line and define the basis pursuant to the criteria of 10 CFR 2.390. After 10 working days, the nonproprietary SE will be made publicly available.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions biweekly Federal Register Notice.
Sincerely,
/RA/
Andrew Hon, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-325 and 50-324
Enclosures:
- 1. Amendment No. 301 to License No. DPR-71
- 2. Amendment No. 329 to License No. DPR-62
- 3. Safety Evaluation (Proprietary Information)
- 4. Safety Evaluation (Nonproprietary Information) cc: w/Enclosures 1, 2, and 4:
Listserv (10 working days after issuance of the amendments to the licensee)
OFFICIAL USE ONLY PROPRIETARY INFORMATION
DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 301 Renewed License No. DPR-71
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by Duke Energy Progress, LLC (the licensee), dated March 9, 2020, 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.
Enclosure 1
- 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-71 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 301, are hereby incorporated in the license. Duke Energy Progress, LLC 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 startup from the 2022 Unit 1 refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Undine S. Undine S. Shoop Date: 2020.09.29 Shoop 12:01:59 -04'00' Undine Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Attachments:
Changes to the Renewed Operating License, Technical Specifications Date of Issuance: September 29, 2020
ATTACHMENT TO LICENSE AMENDMENT NO. 301 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace page 6 of Renewed Facility Operating License No. DPR-71 with the attached page 6.
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 marginal lines indicating the areas of change.
Remove Pages Insert Pages 2.0-1 2.0-1 5.0-20 5.0-20
(c) Transition License Conditions
- 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2. below, risk-informed changes to the licensees fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.
- 2. The licensee shall implement the modifications to its facility, as described in Table S-1, Plant Modifications Committed, of Duke letter BSEP 14-0122, dated November 20, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by the startup of the second refueling outage for each unit after issuance of the safety evaluation. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.
- 3. The licensee shall complete all implementation items, except item 9, listed in LAR Attachment S, Table S-2, Implementation Items, of Duke letter BSEP 14-0122, dated November 20, 2014, within 180 days after NRC approval unless the 180th day falls within an outage window; then, in that case, completion of the implementation items, except item 9, shall occur no later than 60 days after startup from that particular outage. The licensee shall complete implementation of LAR Attachment S, Table S-2, Item 9, within 180 days after the startup of the second refueling outage for each unit after issuance of the safety evaluation.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions 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 2923 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 301, are hereby incorporated in the license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 203 to Renewed Facility Operating License DPR-71, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 203. For SRs that existed prior to Amendment 203, including SRs with modified acceptance criteria and SRs whose frequency of Renewed License No. DPR-71 Amendment No. 301
SLs 2.0 2.0 SAFETY LIMITS (SLS) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%
rated core flow:
THERMAL POWER shall be 23% RTP.
2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10%
rated core flow:
MCPR shall be 1.05.
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 1325 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.
Brunswick Unit 1 2.0-1 Amendment No. 301
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 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 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1;
- 2. The MINIMUM CRITICAL POWER RATIO (MCPR) and MCPR99.9% for Specification 3.2.2;
- 3. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.3;
- 4. The Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II),
the modified APRM Simulated Thermal Power - High scram setpoints used in the Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1; and
- 5. The Allowable Values and power range setpoints for Rod Block Monitor Upscale Functions for Specification 3.3.2.1.
- b. 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:
- 1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel."
- 2. XN-NF-81-58(P)(A), RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model.
- 3. XN-NF-85-67(P)(A), Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel.
- 4. EMF-85-74(P) Supplement 1(P)(A) and Supplement 2(P)(A),
RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model.
- 5. ANF-89-98(P)(A), Generic Mechanical Design Criteria for BWR Fuel Designs.
(continued)
Brunswick Unit 1 5.0-20 Amendment No. 301
DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 329 Renewed License No. DPR-62
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by Duke Energy Progress, LLC (the licensee), dated March 9, 2020, 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.
Enclosure 2
- 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-62 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 329, are hereby incorporated in the license. Duke Energy Progress, LLC 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 startup from the 2021 Unit 2 refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Undine S. Undine S. Shoop Date: 2020.09.29 Shoop 12:02:30 -04'00' Undine Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Attachments:
Changes to the Renewed Operating License, Technical Specifications Date of Issuance: September 29, 2020
ATTACHMENT TO LICENSE AMENDMENT NO. 329 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace page 6 of Renewed Facility Operating License No. DPR-62 with the attached page 6.
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 marginal lines indicating the areas of change.
Remove Pages Insert Pages 2.0-1 2.0-1 5.0-20 5.0-20
(c) Transition License Conditions
- 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2. below, risk-informed changes to the licensees fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.
- 2. The licensee shall implement the modifications to its facility, as described in Table S-1, Plant Modifications Committed, of Duke letter BSEP 14-0122, dated November 20, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by the startup of the second refueling outage for each unit after issuance of the safety evaluation. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.
- 3. The licensee shall complete all implementation items, except Item 9, listed in LAR Attachment S, Table S-2, Implementation Items, of Duke letter BSEP 14-0122, dated November 20, 2014, within 180 days after NRC approval unless the 180th day falls within an outage window; then, in that case, completion of the implementation items, except item 9, shall occur no later than 60 days after startup from that particular outage. The licensee shall complete implementation of LAR Attachment S, Table S-2, Item 9, within 180 days after the startup of the second refueling outage for each unit after issuance of the safety evaluation.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of 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:
(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2923 megawatts (thermal).
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 329, are hereby incorporated in the license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 233 to Renewed Facility Operating License DPR-62, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 233. For SRs that existed prior to Amendment 233, Renewed License No. DPR-62 Amendment No. 329
SLs 2.0 2.0 SAFETY LIMITS (SLS) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%
rated core flow:
THERMAL POWER shall be 23% RTP.
2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10%
rated core flow:
MCPR shall be 1.05.
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 1325 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.
Brunswick Unit 2 2.0-1 Revision No. 329
OFFICIAL USE ONLY - PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 301 AND 329 TO RENEWED FACILITY OPERATING LICENSES NOS. DPR-71 AND DPR-62 DUKE ENERGY PROGRESS, LLC BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324
1.0 INTRODUCTION AND BACKGROUND
By application dated March 9, 2020, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20070H939), Duke Energy Progress, LLC (Duke Energy or the licensee) submitted a license amendment request (LAR) for Brunswick Steam Electric Plant (BSEP), Units 1 and 2.
The LAR proposes to revise Technical Specification (TS) Safety Limit (SL) 2.1.1.2, the reactor core safety limit for the minimum critical power ratio (MCPR). The MCPR protects against boiling transition on the fuel rods in the core. The current MCPR safety limit for BSEP ensures that 99.9-percent of the fuel rods in the core are not susceptible to boiling transition and is referred to as MCPR99.9%. The revised MCPR safety limit 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 critical power ratio (CPR) data statistics and is referred to as the MCPR95/95. Technical Specification 5.6.5, Core Operating Limits Report (COLR), is also proposed to be modified.
The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-564, Revision 2, Safety Limit MCPR [Minimum Critical Power Ratio], dated October 24, 2018 (ADAMS Accession No. 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 (ADAMS Accession No. ML18299A069).
In the LAR, the licensee proposes variations from the TS changes described in Traveler TSTF-564, Revision 2. The variations are described in Section 2.2 and evaluated in Section 3.5 of this SE. BSEP uses Framatome ATRIUM 10XM and ATRIUM 11 fuel types, which are not explicitly identified in Traveler TSTF-564 Table 1. As addressed in Section 3.5, BSEP followed the methodology described in Traveler TSTF-564 to demonstrate allowance of OFFICIAL USE ONLY - PROPRIETARY INFORMATION Enclosure 4
OFFICIAL USE ONLY PROPRIETARY INFORMATION the fuel. This SE contains proprietary information, which is marked with double brackets and bold font such as (( Example )).
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 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 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).
As discussed in Traveler TSTF-564, Revision 2, other fuel vendors may determine the MCPR95/95 for other fuel designs using the described methodology (References 7.1 and 7.2). In the LAR, the licensee provided the required description of the derivation of the MCPR95/95 for ATRIUM 10XM and ATRIUM 11, which is based on the information contained in each fuel types NRC-approved Critical Power Ratio correlation that is referenced in BSEP TS 5.6.5.b.
Framatome defines ECPR as the ratio of the calculated critical power to the measured critical OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION power (i.e., the inverse of the TSTF-564 definition). The TSTF-564, Revision 2 95/95 formulation presumes a mean ECPR of one. The licensee stated that if ((
))
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 and NUREG-1434, the current standard technical specifications (STS) for General Electric BWR designs1, the current 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 fuel cladding SL for pressurized-water reactor (PWR) designs, described in the STS for Babcock & Wilcox, Westinghouse, and Combustion Engineering 2 plants in NUREG-1430, NUREG-1431, and NUREG-14323, respectively, correspond to a 95-percent probability at a 95-percent confidence level that departure from nucleate boiling will not occur. As a result of the overall approach taken in developing the PWR limits, they are only dependent on the fuel type(s) in the reactor and the corresponding departure from nucleate boiling ratio (DNBR) correlations. The limits are not cycle-dependent and are typically only updated when new fuel types are inserted in the reactor.
The TSs for the licensees facility also have a limiting condition for operation (LCO) that governs MCPR, known as the MCPR operating limit (OL). The OL on MCPR is an LCO which must be met to ensure that anticipated operational occurrences do not result in fuel 1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric Plants BWR/4, NUREG-1433, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, April 2012 (ADAMS Accession Nos. ML12104A192 and ML12104A193).
U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric Plants BWR/6, NUREG-1434, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, April 2012 (ADAMS Accession Nos. ML12104A195 and ML12104A196). (Does not apply to Brunswick - a BWR/4 plant) 2 Denotes applicability to Combustion Engineering plants with digital control systems only.
3 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Babcock and Wilcox Plants, NUREG-1430, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, April 2012 (ADAMS Accession Nos. ML12100A177 and ML12100A178).
U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, April 2012 (ADAMS Accession Nos. ML12100A222 and ML12100A228).
U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Combustion Engineering Plants, NUREG-1432, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, April 2012 (ADAMS Accession Nos. ML12102A165 and ML12102A169).
OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION 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 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 anticipated operational occurrences (AOOs).
BSEP TS 2.1.1.2 currently requires that with the reactor steam dome pressure greater than or equal to 785 pounds per square inch gauge (psig) and core flow 10-percent rated core flow, MCPR shall be 1.07 for two recirculation loop operation or 1.09 for single recirculation loop operation. The MCPR SL (also referred to as the MCPR99.9%) ensures that 99.9-percent of the fuel in the core is not susceptible to boiling transition.
2.1.2 TS 5.6.5, Core Operating Limits Report (COLR)
BSEP 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. These limits are required to be documented in the COLR.
2.2 Proposed Changes to the TS The licensee proposes to revise the MCPR SL to make it cycle-independent, consistent with the method described in Traveler TSTF-564, Revision 2.
The proposed changes to the BSEP TS would revise the value of the MCPR SL in TS 2.1.1.2 to 1.05, with corresponding changes to the associated bases. The change to TS 2.1.1.2 replaces the existing separate SLs of 1.09 and 1.07 for single- and two-recirculation loop operation, respectively, with a single limit since the revised SL is no longer dependent on the number of recirculation loops in operation.
The MCPR99.9% (i.e., the current MCPR SL) is an input to the MCPR operating limit (OL) in limiting condition of operation (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 remain unchanged, the proposed TS changes include revisions to TS 5.6.5, to require the additional MCPR99.9% value used in calculating the LCO 3.2.2 MCPR OL to be included in the cycle-specific COLR.
In the LAR, the licensee proposes variations from the TS changes described in Traveler TSTF-564. BSEP uses Framatome ATRIUM 10XM and ATRIUM 11 fuel types, which are not identified in Traveler TSTF-564 Table 1. As discussed in Traveler TSTF-564, other fuel vendors may determine the MCPR95/95 for other fuel designs using the described methodology. The licensee in the LAR provided the required description of the derivation of the MCPR95/95 for OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION ATRIUM 10XM and ATRIUM 11, which is based on the information contained in each fuel types NRC-approved Critical Power Ratio correlation that is referenced in BSEP TS 5.6.5.b.
The BSEP TS also uses different numbering than the STS on which TSTF-564 was based.
Specifically, BSEP TS 5.6.5, Core Operating Limits Report (COLR) corresponds to STS 5.6.3, Core Operating Limits Report.
2.3 Applicable Regulatory Requirements and Guidance 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 items in the following categories: (1) Safety limits, limiting safety system settings, and limiting control settings. As required by 10 CFR 50.36(c)(1)(i)(A), 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.
General Design Criterion (GDC) 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.
With respect to GDC 10, the LAR stated that the BSEP design was reviewed for construction under the "General Design Criteria for Nuclear Power Plant Construction," issued for comment by the Atomic Energy Commission (AEC) on July 11,1967 (32 FR 10213) (ADAMS Accession No. ML043310029). This difference does not alter the conclusion that the proposed change OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION based on TSTF-564, Revision 2, is applicable to BSEP. Many plants, including BSEP, have a plant-specific design criterion similar to GDC 10.
The limit placed on the MCPR acts as a specified acceptable fuel design limit to prevent boiling transition, which has the potential to result in fuel rod cladding failure. Section 3.1.2.2.1 of the BSEP Updated Final Safety Analysis Report (UFSAR, Reference 7.3) discusses compliance with GDC 10.
The NRC staffs guidance for the review of reactor design is in Section 4.4, Thermal and Hydraulic Design, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP), March 2007 (ADAMS Accession No. ML070550060). It provides the following two examples of acceptable approaches for 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 (NUREG-0800), dated March 2010 (ADAMS Accession No. 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 BSEP is NUREG-1433, Revision 4.0, Standard Technical Specifications, General Electric Plants BWR/4, Volume 1, Specifications, and Volume 2, Bases, April 2012 (ADAMS Accession Nos. ML12104A192 and ML12104A193, respectively).
3.0 TECHNICAL EVALUATION
3.1 Basis for Proposed Change As discussed in Section 1.3 of this SE, the current MCPR SL (i.e., the MCPR99.9%), is dependent on 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 licensee proposes to change the methodology for determining the MCPR SL for BSEP 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 PWRs, which, as previously noted in OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION Section 1.3 of this SE, provides 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 MCPR SL based on the 95/95 criterion is acceptable because it meets SRP Section 4.4, Acceptance Criterion 1. The remainder of this SE is devoted to ensuring that the methodology for determining the revised MCPR SL provides the intended result, that the revised MCPR SL can be adequately determined in the core using various types of fuel, that the proposed SL continues to fulfil the necessary functions of an SL without unintended consequences, and that the proposed changes have been adequately implemented in the BSEP TSs.
3.2 Revised MCPR SL Definition Framatome defines ECPR as the ratio of the calculated critical power to the measured critical power (i.e., the inverse of the TSTF-564 definition). A critical power correlations ECPR distribution quantifies the uncertainty associated with the correlation and Framatomes definition of ECPR, which differs from that provided in TSTF-564. 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) = µi + Kii where i is the correlations mean ECPR and i is the standard deviation of the correlations ECPR distribution. The statistical parameter (Ki) 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. Owen4 and was also reported by M. G. Natrella5, as referenced in Traveler TSTF-564, Revision 2.
In the LAR, the licensee proposes variations from the TS changes described in TSTF 564. That is, BSEP uses Framatome ATRIUM 10XM and ATRIUM 11 fuel types, which are not identified in Traveler TSTF-564 Table 1. As discussed in TSTF-564, other fuel vendors may determine the MCPR95/95 for other fuel designs using the methodology. The licensee in the LAR provided the required description of the derivations of the MCPR95/95 for ATRIUM 10XM and ATRIUM 11, which is based on the information contained in each fuel type's NRC-approved CPR correlation that is referenced in BSEP TS 5.6.5.b. The staff finds that the difference is within the scope of the Traveler TSTF-564 approval and does not affect the applicability of Traveler TSTF-564 to the BSEP TS.
As discussed by Piepel and Cuta6 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.
4 D. B. Owen, Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, Sandia Corporation, SCR-607, March 1963, ADAMS Accession No. ML14031A495.
5 M. G. Natrella, Experimental Statistics, National Bureau of Standards, National Bureau of Standards Handbook 91 August 1963.
6 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.
OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION These assumptions are typically addressed generically when a critical power or critical heat flux (CHF) correlation is reviewed by the NRC staff, who may apply penalties to the correlation to account for any issues identified. The Traveler TSTF-564, Revision 2, 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 (ADAMS Accession No. ML18149A320). 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, Revision 2, 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 BSEP TSs.
3.3 Determination of Revised MCPR SL for Mixed Cores The Traveler TSTF-564, Revision 2, 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 Traveler 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 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 letter from the TSTF dated May 29, 2018 (ADAMS Accession No. ML18149A320) indicate that this is the case even for fuel operated on short (12-month) reload cycles. As discussed in the May 29, 2018 letter from the TSTF, 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 of the Traveler TSTF-564, Revision 2, the NRC staff found 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.
In the SE of the Traveler TSTF-564, Revision 2, the NRC staff also reviewed the information furnished by the TSTF and determined that the process for establishing the revised MCPR SL (MCPR95/95) for mixed cores ensures that the limiting fuel types in the core will be evaluated and that 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 BSEP TSs.
3.4 Relationship between MCPR Safety and Operating Limits As discussed in the SE of the Traveler TSTF-564, Revision 2, the MCPR99.9% is expected to always be greater than the MCPR95/95 for two reasons. First, because the MCPR99.9% includes uncertainties not factored into the MCPR95/95, and second, because the 99.9-percent probability OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION 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. This is consistent with evaluations performed for PWRs using a 95/95 upper tolerance limit on the correlation uncertainty as an SL.
Consistent with Traveler TSTF-564, Revision 2, 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 BSEP.
Consistent with Traveler TSTF-564, Revision 2, the licensee proposed to revise the COLR TS (5.6.5 in the BSEP TSs) 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 in the BSEP TS. The changes to TS 5.6.5.b in the BSEP TSs 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 MCPR95/95 represents a lower limit 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 Traveler TSTF-564, Revision 2).
3.5 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 ATRIUM 11 to 1.05. The value reported in BSEP TS 2.1.1.2 was calculated using Equation 1 from Traveler TSTF-564, Revision 2, and reported at a precision of two digits past the decimal point with the hundreds digit rounded up.
Consistent with TSTF-564, Revision 2, the licensee also proposes to modify BSEP 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 BSEP 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.
In the LAR, the licensee provides the details of the calculation of the MCPR95/95 for ATRIUM 10XM using the statistics from the ACE/ATRIUM 10XM CPR correlation database contained in ANP-10298P-A, Revision 1 (Reference 7.4) for TS 5.6.5.b.21. The licensee also provides the details of the calculation of the MCPR95/95 for ATRIUM 11 using the statistics from the OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION ACE/ATRIUM 11 CPR correlation database contained in ANP-10335P-A, Revision 0 (Reference 7.5) for TS 5.6.5.b.25.
MCPR99.9% will continue to be calculated using Framatomes SAFLIM-3D methodology. The NRC staff assessed the licensees deviations for ATRIUM 10XM and ATRIUM 11 and determined that they are consistent with the process described in Traveler TSTF-564, Revision 2. In Enclosure 7 to the LAR, the licensee stated the following:
Framatome defines ECPR as the ratio of the calculated critical power to the measured critical power (i.e., the inverse of the TSTF-564 definition). The TSTF-564 MCPR95/95 formulation ((
)).
In the above quoted paragraph, the licensee states that using the inverse is acceptable ((
)), however, NRC staff finds that this is not the only necessary condition.
Mathematically, the standard deviation of a set of data (ECPR values) is not the same as the standard deviation of the inverse of the set. In addition to requiring (( )),
the NRC staff finds that two other conditions are necessary to achieve the same SL MCPR using either the ECPR or its inverse: a normal distribution and a small standard deviation. NRC staff calculations have shown that when the standard deviation increases above ~5%, the resulting SL MCPR can be different (to the hundredths digit) when using either the Traveler TSTF-564 definition of ECPR or its inverse. Given that the licensee stated in Enclosure 7 to the LAR that the distribution is normal and the standard deviations of ECPR are ((
)), the NRC staff finds the licensees definition of ECPR as the ratio of the calculated critical power to the measured critical power (i.e., the inverse of the TSTF-564 definition) to be acceptable.
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 Traveler TSTF-564, Revision 2.
The staff notes that BSEP TSs have a different numbering than STS for the Core Operating Limits Report; specifically, BSEP TS 5.6.5 versus STS 5.6.3. The NRC staff confirmed that the licensee made appropriate conforming changes in its proposal to adopt this TSTF traveler. .
As addressed earlier, the BSEP was not licensed to GDC 10, but instead was licensed to the applicable AEC preliminary general design criteria. The NRC staff determined that this difference does not affect the applicability of Traveler TSTF-564 for the proposed amendments OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION to the BSEP TSs. The NRC staff reviewed the licensees proposed TS changes and found that the licensee proposed MCPR SL is appropriate, as discussed in this SE.
3.6 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 Traveler TSTF-564, Revision 2, and was 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 of BSEPs applicable design criterion that is similar to GDC 10, by ensuring that no fuel damage results during normal operation and anticipated operational occurrences. 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 appropriate official for the State of North Carolina was notified of the NRCs proposed issuance of the amendments on August 10, 2020.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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 (85 FR 22185, dated April 21, 2020). 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 OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY - PROPRIETARY INFORMATION amendments will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
7.1 Letter from Brian R. Moore, Global Nuclear Fuel, to U.S. NRC, Information Supporting TSTF-564 Safety Limit Minimum Critical Power Ratio, June 16, 2017, ADAMS Accession No. ML17167A108.
7.2 Letter from James A. Gresham, Westinghouse Electric Company, to U.S. NRC, Submittal of 'Calculation for Technical Specification SLM CPR Values Applying to Westinghouse Fuel in Support of TSTF-564', May 16, 2017, ADAMS Accession No. ML17142A319.
7.3 Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Updated Final Safety Analysis Report, Revision 26, August 13, 2019, ADAMS Accession No. ML18249A165.
7.4 ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, Revision 1, March 2014, ADAMS Accession No. ML14183A734.
7.5 ANP-10335P-A, ACE/ATRIUM 11 Critical Power Correlation, Revision 0, May 2018, ADAMS Accession No. ML18207A408.
Principal Contributor: Chris Jackson, NRR/DSS Date: September 29, 2020 OFFICIAL USE ONLY - PROPRIETARY INFORMATION
OFFICIAL USE ONLY - PROPRIETARY INFORMATION
SUBJECT:
BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 301 AND 329 TO REVISE TECHNICAL SPECIFICATION TO ADOPT TSTF-564 (EPID L-2020-LLA-0043)
DATED SEPTEMBER 29, 2020 DISTRIBUTION:
PUBLIC (Letter, approval, change pages, redacted SE) (10 working days after issuance)
NON-PUBLIC (SE)
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