RA-19-0431, Application to Revise Technical Specifications to Adopt TSTF 564, Safety Limit MCPR

Application to Revise Technical Specifications to Adopt TSTF 564, Safety Limit MCPR

ML20070H939
Person / Time
Site:	Brunswick
Issue date:	03/09/2020
From:	Krakuszeski J Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
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ML20070H937	List: RA-19-0431, Application to Revise Technical Specifications to Adopt TSTF 564, Safety Limit MCPR
References
RA-19-0431
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Text

John A. Krakuszeski Vice President Brunswick Nuclear Plant 8470 River Rd SE Southport, NC 28461 o: 910.832.3698 Enclosure 7 Contains Proprietary Information Withhold in Accordance with 10 CFR 2.390 March 9, 2020 Serial: RA-19-0431 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Application to Revise Technical Specifications to Adopt TSTF 564, "Safety Limit MCPR" Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy) is submitting a request for an amendment to the Technical Specifications (TS) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.

Duke Energy requests adoption of TSTF-564, "Safety Limit MCPR," Revision 2, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the BSEP, Unit Nos. 1 and 2 TS. The proposed amendment revises 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 a SL. provides a description and assessment of the proposed changes. Enclosure 2 and 3 provide the existing BSEP Unit 1 and 2 TS pages, respectively, marked to show the proposed changes. Enclosures 4 and 5 provides revised (clean) TS pages for BSEP Unit 1 and 2, respectively. Enclosure 6 provides Unit 1 TS Bases pages marked to show the proposed changes for information only. contains the calculation of the MCPR95/95 for the ATRIUM 10XM and ATRIUM 11 fuel types. This enclosure contains information considered proprietary to Framatome, denoted by brackets. As owner of the proprietary information, Framatome has executed the affidavit contained in Enclosure 9 which identifies the information as proprietary, is customarily held in confidence, and should be withheld from public disclosure in accordance with 10 CFR 2.390. contains the non-proprietary version of this information.

Approval of the proposed amendment is requested within one year of completion of the NRCs acceptance review. Once approved, the Unit 2 amendment shall be implemented prior to startup from the 2021 Unit 2 refueling outage and the Unit 1 amendment shall be implemented prior to startup from the 2022 Unit 1 refueling outage.

U.S. Nuclear Regulatory Commission Page 2 of 3 In accordance with 10 CFR 50.91, Duke Energy is providing a copy of the proposed license amendment to the designated representative for the State of North Carolina.

This document contains no new regulatory commitments. Please refer any questions regarding this submittal to Mr. Art Zaremba, Director- Nuclear Fleet Licensing, at (980) 373-2062.

I declare, under penalty of perjury, that the foregoing is true and correct. Executed on March 9, 2020.

Sincerely, 9....0- A - ~ .

John A. Krakuszeski MAT/mat

Enclosures:

1. Description and Assessment

2. Proposed Technical Specification Changes (Mark-Up) - Unit 1

3. Proposed Technical Specification Changes (Mark-Up) - Unit 2

4. Revised Technical Specification Pages - Unit 1

5. Revised Technical Specification Pages - Unit 2

6. Proposed Technical Specification Bases Changes (Mark-Up) - Unit 1 (For Information Only)

7. ATRIUM 10XM and ATRIUM 11 MCPR9s19s Derivation (Proprietary information -

Withhold from Public Disclosure in Accordance with 10 CFR 2.390)

8. ATRIUM 10XM and ATRIUM 11 MCPR9s19s Derivation (Non-proprietary)

9. Affidavit for ATRIUM 10XM and ATRIUM 11 MCPR9s19s Derivation

U.S. Nuclear Regulatory Commission Page 3 of 3 cc:

U. S. Nuclear Regulatory Commission, Region II ATTN: Ms. Laura Dudes, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U.S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon 11555 Rockville Pike Rockville, MD 20852-2738 U. S. Nuclear Regulatory Commission ATTN: Mr. Gale Smith, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 Chair - North Carolina Utilities Commission (Electronic Copy Only) 4325 Mail Service Center Raleigh, NC 27699-4300 swatson@ncuc.net Mr. W. Lee Cox, III, Section Chief (Electronic Copy Only)

Radiation Protection Section North Carolina Department of Health and Human Services 1645 Mail Service Center Raleigh, NC 27699-1645 lee.cox@dhhs.nc.gov

RA-19-0431 Enclosure 1 Page 1 of 4 Description and Assessment

1.0 DESCRIPTION

Duke Energy Progress, LLC (Duke Energy) requests adoption of TSTF-564, "Safety Limit MCPR," Revision 2, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the (BSEP), Unit Nos. 1 and 2 Technical Specifications (TSs). The proposed amendment revises 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 a SL.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation Duke Energy has reviewed the safety evaluation for TSTF-564 provided to the Technical Specifications Task Force in a letter dated November 16, 2018. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-564. Duke Energy has concluded that the justifications presented in TSTF-564 and the safety evaluation prepared by the NRC staff are applicable to BSEP, Unit Nos. 1 and 2 and justify this amendment for the incorporation of the changes to the BSEP TS.

The BSEP, Unit Nos. 1 and 2 reactors are currently fueled with Framatome ATRIUM 10XM fuel assemblies and will be introducing ATRIUM 11 during the Spring 2020 Unit 1 refueling outage and the Spring 2021 Unit 2 refueling outage. The proposed Safety Limit in SL 2.1.1.2 is 1.05.

This is the value calculated for both the ATRIUM 10XM and ATRIUM 11 fuel types using the Equation 1 in Section 3.1 of TSTF-564. ATRIUM 11 is identified in the TS Bases as the fuel type the SL is based upon since it will be the dominant fuel type at BSEP going forward. 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 (i.e., TS 5.6.5.b.21). Enclosure 7 also provides the details of the calculation of the MCPR95/95 for ATRIUM 11 using the statistics from the ACE/ATRIUM 11 CPR correlation database contained in ANP-10335P-A, Revision 0 (i.e., TS 5.6.5.b.25).

Duke Energy is currently informed of any error that impacts the ACE CPR correlations in accordance with Framatome's 10 CFR 50 Appendix B program. This existing process ensures that any potential errors in the ACE CPR correlations will be addressed as they relate to the MCPR95/95 for SL 2.1.1.2 following implementation of TSTF-564.

The MCPR value calculated as the point at which 99.9% of the fuel rods would not be susceptible to boiling transition (i.e., reduced heat transfer) during normal operation and anticipated operational occurrences is referred to as MCPR99.9%. Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," is revised to require the MCPR99.9% value to be included in the cycle-specific COLR. MCPR99.9% will continue to be calculated using Framatomes SAFLIM-3D methodology (i.e., TS 5.6.5.b.11).

2.2 Variations Duke Energy is proposing the following variations from the TS changes described in TSTF 564 or the applicable parts of the NRC staffs safety evaluation.

RA-19-0431 Enclosure 1 Page 2 of 4 BSEP uses Framatome ATRIUM 10XM and ATRIUM 11 fuel types which are not identified in TSTF-564 Table 1. However, as discussed in TSF-564, other fuel vendors may determine the MCPR95/95 for other fuel designs using the methodology described in TSTF-564. Duke Energy has included the required description of the derivation of the MCPR95/95 for ATRIUM 10XM and ATRIUM 11 in Enclosure 7 based on the information contained in each fuel type's NRC-approved CPR correlation that is referenced in BSEP TS 5.6.5.b. Duke Energy has confirmed in Enclosure 7 that the TSTF-564 MCPR95/95 formulation remains applicable with Framatome's definition of Experimental CPR (ECPR). This difference is within the scope of the TSTF-564 approval and does not affect the applicability of TSTF-564 to the BSEP TS.

The BSEP TS utilize different numbering than the Standard Technical Specifications on which TSTF-564 was based. Specifically, TS 5.6.5 Core Operating Limits Report (COLR) corresponds to STS 5.6.3 Core Operating Limits Report. This difference is administrative and does not affect the applicability of TSTF-564 to the BSEP TS.

The BSEP design was reviewed for construction under the "General Design Criteria for Nuclear Power Plant Construction," issued for comment by the AEC in July 1967. This difference does not alter the conclusion that the proposed change is applicable to BSEP.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis Duke Energy Progress, LLC (Duke Energy) requests adoption of TSTF-564, "Safety Limit MCPR," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the into the (BSEP), Unit Nos. 1 and 2 Technical Specifications (TS). The proposed change revises the Technical Specifications (TS) safety limit on minimum critical power ratio (SLMCPR). The revised limit calculation method is based on using the Critical Power Ratio (CPR) data statistics and is revised from ensuring that 99.9% of the rods would not be susceptible to boiling transition to ensuring that there is a 95% probability at a 95% confidence level that no rods will be susceptible to transition boiling. A single SLMCPR value will be used instead of two values applicable when one or two recirculation loops are in operation. TS 5.6.5, "Core Operating Limits Report (COLR)," is revised to require the current SLMCPR value to be included in the COLR.

Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the Core Operating Limits Report (COLR). The SLMCPR is not an initiator of any accident previously evaluated. The revised safety limit values continue to ensure for all accidents previously evaluated that the fuel cladding will be protected from failure due to transition boiling. The proposed change does not affect plant operation or any procedural or administrative controls on plant operation that affect the functions of preventing or mitigating any accidents previously evaluated.

RA-19-0431 Enclosure 1 Page 3 of 4 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?

Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. The proposed change will not affect the design function or operation of any structures, systems or components (SSCs). No new equipment will be installed. As a result, the proposed change will not create any credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. This will result in a change to a safety limit, but will not result in a significant reduction in the margin of safety provided by the safety limit. As discussed in the application, changing the SLMCPR methodology to one based on a 95% probability with 95% confidence that no fuel rods experience transition boiling during an anticipated transient instead of the current limit based on ensuring that 99.9%

of the fuel rods are not susceptible to boiling transition does not have a significant effect on plant response to any analyzed accident. The SLMCPR and the TS Limiting Condition for Operation (LCO) on MCPR continue to provide the same level of assurance as the current limits and do not reduce a margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Duke Energy concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.2 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an

RA-19-0431 Enclosure 1 Page 4 of 4 inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

RA-19-0431 Enclosure 2 Proposed Technical Specification Changes (Mark-Up) -

Unit 1

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 1.051.07 for two recirculation loop operation or 1.09 for single recirculation loop operation.

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. 272

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. 285

RA-19-0431 Enclosure 3 Proposed Technical Specification Changes (Mark-Up) -

Unit 2

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 1.051.07 for two recirculation loop operation or 1.09 for single recirculation loop operation.

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. 300

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 2 5.0-20 Amendment No. 313

RA-19-0431 Enclosure 4 Revised Technical Specification Pages - Unit 1

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. 272

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. 285

RA-19-0431 Enclosure 5 Revised Technical Specification Pages - Unit 2

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. 300

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 2 5.0-20 Amendment No. 313

RA-19-0431 Enclosure 6 Proposed Technical Specification Bases Changes (Mark-Up) - Unit 1 (For Information Only)

Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLS)

B 2.1.1 Reactor Core SLs BASES BACKGROUND SLs ensure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2.

MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e.,

MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as the SLMCPR95/95, which corresponds to a 95% probability of a 95% confidence level (the 95/95 MCPR criterion) that transition boiling will not occur. The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core do not experience transition boiling.

(continued)

Brunswick Unit 1 B 2.1.1-1 Revision No. 31

Reactor Core SLs B 2.1.1 BASES BACKGROUND Operation above the boundary of the nucleate boiling regime could result (continued) in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

The reactor vessel water level SL ensures that adequate core cooling capability is maintained during all MODES of reactor operation.

Establishment of Emergency Core Cooling System initiation setpoints higher than this safety limit provides margin to ensure the safety limit will not be reached or exceeded such that fuel damage would occur.

APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and AOOs. The Technical Specification SL is set generically on a fuel product MCPR correlation basis as the MCPR which corresponds to a 95% probability at a 95% confidence level that transition boiling will not occur, referred to as SLMCPR95/95. The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.

The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR limit.

2.1.1.1 Fuel Cladding Integrity The critical power correlations are valid for critical power calculations at pressures 600 psia and bundle mass fluxes 0.09E+6 lbm/hr-ft2 (References 1, 4, and 6). For operation at low pressures or low flows, another basis is used, as follows:

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be > 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia (0 psig) to 800 psia (785 psig) indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER > 46% RTP.

Thus, a THERMAL POWER limit of 23% RTP for reactor pressure

< 785 psig is conservative.

Brunswick Unit 1 B 2.1.1-2 Revision No. 109

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR SAFETY ANALYSES (continued) The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. The Technical Specification SL value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent. The value is based on the Critical Power Ratio (CPR) data statistics and a 95% probability with 95% confidence that rods are not susceptible to boiling transition, referred to as MCPR95/95.

The SL is based on ATRIUM 11 fuel. For cores with a single fuel product line, the SLMCPR95/95 is the MCPR95/95 for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR values for the fuel product lines that are fresh or once-burnt at the start of the cycle.

However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved AREVA critical power correlations. One specific uncertainty included in the SL is the uncertainty inherent in the critical power correlation. References 1, 2, 3,4 and 6 describe the uncertainties and methodology used to determine the MCPR SL.

2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active irradiated fuel to provide core cooling capability. In conjunction with LCOs, the limiting safety system settings, defined in LCO 3.3.1.1 as the Allowable Values, establish the threshold for protective system action to prevent exceeding acceptable limits, including this reactor vessel water level SL, during Design Basis Accidents. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to Brunswick Unit 1 B 2.1.1-4 3 Revision No. 109

MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs) and that 99.9% of the fuel rods are not susceptible to boiling transition if the limit is not violated.

Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref.1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE The analytical methods and assumptions used in evaluating the AOOs to SAFETY ANALYSES establish the operating limit MCPR are presented in References 2, 3, 4, 5, 7, 8, and 9. To ensure that 99.9% of the fuel rods avoid boiling transition the MCPR Safety Limit (SL) is not exceeded during any transient that occurs with moderate frequency, limiting transients are analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (CPR). When the largest CPR is added to combined with the SL MCPR99.9% SL, the required operating limit MCPR (OLMCPR) is obtained.

MCPR99.9% is determined to ensure more than 99.9% of the fuel rods in the core are not susceptible to boiling transition using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved Critical Power correlations. Details of the MCPR99.9% calculation are given in Reference 2. The cycle-specific Reload Safety Analysis Report (as referenced in the COLR) includes a tabulation of the fuel- and plant-related uncertainties for the parameters used in the MCPR99.9% statistical analysis.

Brunswick Unit 1 B 3.2.2-1 Revision No. 108

MCPR B 3.2.2 The MCPR operating limits are derived from the MCPR99.9% value and the transient analysis are dependent on the operating core flow and power state (MCPRf and MCPRp respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency as identified in the UFSAR Chapter 15 (Reference 5).

Flow dependent MCPR limits are determined using steady state thermal hydraulic methods (Reference 7) to analyze slow flow runout transients.

The MCPRf limits are dependent on the maximum core flow runout capability of the Recirculation System.

(continued)

Brunswick Unit 1 B 3.2.2-2 Revision No. 108

MCPR B 3.2.2 BASES APPLICABLE Power dependent MCPR limits (MCPRp) are determined on a cycle-SAFETY ANALYSES specific basis using the methodologies presented in References 8 and 9.

(continued) The MCPRp limits are established for a set of exposure intervals. Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which turbine stop valve and/or turbine control valve fast closure and the associated scrams occur, high and low flow MCPRp operating limits may be provided for the following core power level ranges: 23 to 26% RTP and 26 to 50% RTP. The 26% RTP is the previously mentioned bypass power level and 50% RTP is the power level below which the power load unbalance unit (PLU) may not generate a turbine control valve fast closure on a generator loss of load event.

The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 11).

LCO The MCPR operating limits, as a function of core flow, core power, and cycle exposure, specified in the COLR (MCPR99.9% value, MCPRf values, and MCPRp values) are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the larger of the MCPRf and MCPRp limits, which are based on the MCPR99.9%

limit specified in the COLR.

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 23% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 23% RTP is unnecessary due to the large inherent margin that ensures that the MCPRSL is not exceeded even if a limiting transient occurs.

Statistical analyses indicate that the nominal value of the initial MCPR expected at 23% RTP is > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 23% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels

< 23% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.

(continued)

Brunswick Unit 1 B 3.2.2-3 Revision No. 108

RA-19-0431 Enclosure 8 ATRIUM 10XM and ATRIUM 11 MCPR95/95 Derivation (Non-proprietary)

ATRIUM 10XM and ATRIUM 11 MCPR95/95 Derivation Equation 1 of TSTF-564 (Reference 1) is used to determine the MCPR95/95 that will replace the current definition of the Safety Limit MCPR (MCPR99.9%) in TS 2.1.1.2. The equation requires the CPR correlation mean Experimental CPR (ECPR), the correlation standard deviation and the 95/95 multiplier based on the size of the critical power correlation database. These correlation parameters are contained in References 2 and 3 for ATRIUM 10XM and ATRIUM 11, respectively.

As required by TSTF-564, the 95/95 multiplier is to be calculated using formulas attributed to Natrella (1963) as recommended in Reference 4. The equations for the calculation of a one-sided tolerance interval for a normal distribution are provided in Section 7.6.2 of Reference 4 in the "Tolerance Intervals for a Normal Distribution" section.

Technical Specification SLMCPR = +

Product Line ATRIUM 10XM ATRIUM 11 Correlation Reference 2 3 Distribution Normal Normal Mean of ECPR ( ) [ ]

Standard Deviation of ECPR ( )% [ ]

Number of Data Points [ ]

Statistical Factor for 95/95 ( ) [ ]

TS SLMCPR 1.05 1.05 Note: 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 [

].

References:

1. TSTF-564-A, Revision 2, "Safety Limit MCPR," dated November 16, 2018 (ADAMS Accession No. ML18299A048)

2. ANP-10298P-A, "ACE/ATRIUM 10XM Critical Power Correlation," Revision 1

3. ANP-10335P-A, "ACE/ATRIUM 11 Critical Power Correlation," Revision 0

4. NIST/SEMATECH e-Handbook of Statistical Methods, http://www.itl.nist.gov/div898/handbook/, last updated 10/30/2013

RA-19-0431 Enclosure 9 Affidavit for ATRIUM 10XM and ATRIUM 11 MCPR95/95 Derivation

AFFIDAVIT

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for Framatome Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.

3. I am familiar with the Framatome information contained in Enclosure 7 to the Letter RA-19-0431, 'Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Application to Revise Technical Specifications to Adopt TSTF 564, "Safety Limit MCPR",' and referred to herein as "Document." Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:

(a) The information reveals details of Framatome's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability.

(e) The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.

7. In accordance with Framatome's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

Alan Meginnis 13~--

STATE OF WASHINGTON )

) ss.

COUNTY OF BENTON )

SUBSCRIBEDbeforemethis '2.4th dayof Febr({<<Y-,Y , 2020.

Katherine Kerr NOTARY PUBLIC, STATE OF WASHINGTON MY COMMISSION EXPIRES: 9/12/2022