PLA-7962, Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit MCPR PLA-7962

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Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit MCPR PLA-7962
ML22005A183
Person / Time
Site: Susquehanna  Talen Energy icon.png
Issue date: 01/05/2022
From: Cimorelli K
Susquehanna, Talen Energy
To:
Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML22005A182 List:
References
PLA-7962, TSTF-564
Download: ML22005A183 (52)
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Text

Attachment 5 Contains Proprietary Information - Withhold Pursuant to 10 CFR 2.390 TALEN~

Kevin Cimorelli Susquehanna Nuclear, LLC Site Vice President 769 Salem Boulevard Berwick, PA 18603 ENERGY Tel. 570.542.3795 Fax 570.542.1504 Kevin.Cimorelli@TalenEnergy.com January 5, 2022 Attn: Document Control Desk 10 CFR 50.90 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED AMENDMENT TO LICENSES NPF-14 AND NPF-22: APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-564, SAFETY LIMIT MCPR Docket No. 50-387 PLA-7962 and 50-388 Pursuant to 10 CFR 50.90, Susquehanna Nuclear, LLC (Susquehanna), is submitting a request for an amendment to the Technical Specifications (TS) for the Susquehanna Steam Electric Station (SSES), Units 1 and 2, Facility Operating License Numbers NPF-14 and NPF-22.

Susquehanna requests the adoption of TSTF-564, Safety Limit MCPR," Revision 2, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the SSES, Units 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 an SL. provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked to show the proposed changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides existing TS Bases pages marked to show the proposed changes for information only. contains a Framatome report which provides the calculation determining the appropriate safety limit MCPR (i.e., SLMCPR95/95) for the ATRIUM 10 and ATRIUM 11 fuel types. This attachment contains information considered proprietary to Framatome, denoted by brackets. As owner of the proprietary information, Framatome has executed the affidavit contained in Attachment 7 which identifies the information as proprietary, is customarily held in confidence, and should be withheld from public disclosure pursuant to 10 CFR 2.390. contains the non-proprietary version of this information.

Document Control Desk PLA-7962 Copy: NRC Region I Mr. C. Highley, NRC Senior Resident Inspector Ms. A. Klett, NRC Project Manager Mr. M. Shields, PA DEP/BRP

Attachment 1 to PLA-7962 Description and Assessment

1. DESCRIPTION
2. ASSESSMENT 2.1 Applicability of Safety Evaluation 2.2 Variations
3. REGULATORY ANALYSIS 3.1 Precedent 3.2 No Significant Hazards Consideration Analysis 3.3 Conclusion
4. ENVIRONMENTAL CONSIDERATION
5. REFERENCES

Attachment 1 to PLA-7962 Page 1 of 5 SUSQUEHANNA ASSESSMENT

1. Description Susquehanna Nuclear, LLC (Susquehanna), requests adoption of TSTF-564, "Safety Limit MCPR," Revision 2, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Susquehanna Steam Electric Station (SSES), Units 1 and 2 Technical Specifications (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 an SL.
2. Assessment 2.1 Applicability of Safety Evaluation Susquehanna has reviewed the safety evaluation for TSTF-564 provided to the Technical Specifications Task Force in a letter dated May 13, 2020. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-564. Susquehanna has concluded that the justifications presented in TSTF-564 and the safety evaluation prepared by the NRC staff are applicable to SSES, Units 1 and 2 and justify this amendment for the incorporation of the changes to the SSES TS.

The SSES Unit 1 reactor is currently fueled with Framatome ATRIUM 10 fuel bundles and ATRIUM 11 fuel bundles will be introduced during the spring 2022 refueling outage. The SSES Unit 2 reactor is currently fueled with Framatome ATRIUM 10 and ATRIUM 11 fuel bundles. The proposed Safety Limit in SL 2.1.1.2 is 1.05 which is the most limiting value for ATRIUM 10 and ATRIUM 11 fuel types. ATRIUM 11 is identified in the TS Bases as the fuel type the SL is based upon since it is most limiting and will be the dominant fuel type at SSES going forward. contains Framatome report ANP-3857P, Revision 2, Design Limits for Framatome Critical Power Correlations. This report provides details of the calculation of the MCPR95/95 for ATRIUM 10 using the statistics from the SPCB/ATRIUM 10 CPR correlation database contained in Reference 1. This report 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 Reference 2.

Susquehanna is currently informed of any error that impacts the SPCB/ACE CPR correlations in accordance with Framatomes 10 CFR 50 Appendix B program. This existing process ensures

Attachment 1 to PLA-7962 Page 2 of 5 that any potential errors in the SPCB/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 Reference 3.

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

The SSES TS specify a different reactor steam dome pressure value (575 psig) in Specifications 2.1.1.1 and 2.1.1.2, rather than the value of 785 psig specified in the ISTS. This is a plant-specific value and does not affect the applicability of TSTF-564 to SSES.

SSES uses Framatome ATRIUM 10 and ATRIUM 11 fuel types which are not identified in TSTF-564 Table 1. However, as discussed in TSTF-564, other fuel vendors may determine the MCPR95/95 for other fuel designs using methodology described in TSTF-564. Framatome licensing report ANP-3857P (Attachment 5) includes the required derivation of the MCPR95/95 for the ATRIUM 10 and ATRIUM 11 fuel types based on the information contained in the respective NRC approved CPR correlations (Reference 1 and Reference 2). The conclusions in that the TSTF-564 MCPR95/95 formulation remains applicable with Framatome's definition of Experimental CPR (ECPR) have been accepted by Susquehanna. This difference is within the scope of the TSTF-564 approval and does not affect the applicability of TSTF-564 to the SSES TS.

The SSES TS utilize different numbering than the ISTS on which TSTF-564 was based.

Specifically, SSES TS 5.6.5, Core Operating Limits Report (COLR) corresponds to ISTS 5.6.3, Core Operating Limits Report. This difference is administrative and does not affect the applicability of TSTF-564 to the SSES TS.

3. Regulatory Analysis 3.1 Precedent The NRC has approved a similar amendment request for implementation of TSTF-564 for another station utilizing Framatome ATRIUM Fuel. Duke Energy Progress, LLC, submitted on March 9, 2020, an application to revise the Brunswick Steam Electric Plant, Unit Nos. 1 and 2,

Attachment 1 to PLA-7962 Page 3 of 5 Technical Specifications requesting the adoption of TSTF-564, Safety Limit MCPR, (Reference 4). The Brunswick nuclear power plants are fueled with Framatome ATRIUM 10XM fuel type and the ATRIUM 11 fuel type was indicated as expected to be introduced in the spring of 2020. This was approved by the NRC on September 29, 2020 (Reference 5).

3.2 No Significant Hazards Considerations Analysis Susquehanna Nuclear, LLC (Susquehanna), requests adoption of TSTF-564, "Safety Limit MCPR," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the SSES Unit 1 and 2 Technical Specifications (TS). The proposed change revises the 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 cycle specific SLMCPR value (i.e., MCPR99.9%) to be included in the COLR.

Susquehanna 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 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.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Attachment 1 to PLA-7962 Page 4 of 5

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident 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, Susquehanna 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,

Attachment 1 to PLA-7962 Page 5 of 5 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. Environmental Consideration 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 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.

5. References
1. AREVA Topical Report EMF-2209(P)(A), SPCB Critical Power Correlation.
2. Framatome Topical Report ANP-10335P-A, ACE/ATRIUM-11 Critical Power Correlation.
3. AREVA Topical Report ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors.
4. Duke Energy Progress, LLC, Application to Revise Technical Specification to Adopt TSTF 564, Safety Limiy MCPR, Dated March 9, 2020 (ADAMS Accession No. ML20070H939)
5. NRC letter to Duke Energy Progress LLC, 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-1010-LLA-0043). Dated September 29, 2020 (ADAMS Accession No. ML20269A305).

Attachment 2 of PLA-7962 Marked-Up Technical Specification Changes Revised Technical Specifications Pages Unit 1 TS Pages 2.0-1, 5.0-21 Unit 2 TS Pages 2.0-1, 5.0-21

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SL 2.1.1.1 With the reactor steam dome pressure < 575 psig or core flow

< 10 million lbm/hr:

THERMAL POWER shall be 23% RTP.

2.1.1.2 With the reactor steam dome pressure 575 psig and core flow 10 million lbm/hr:

MCPR shall be 1.051.09 for two recirculation loop operation or 1.12 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.

SUSQUEHANNA - UNIT 1 2.0-1 Amendment 178, 186, 199, 216, 227 231, 246, 261, 278, XXX

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 Not Used 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 for Specification 3.2.1;
2. The Minimum Critical Power Ratio (MCPR) and MCPR99.9% for Specification3.2.2;
3. The Linear Heat Generation Rate for Specification 3.2.3;
4. The Shutdown Margin for Specification 3.1.1;
5. Oscillation Power Range Monitor (OPRM) Trip Setpoints, for Specification 3.3.1.1; and
6. The Allowable Values and power range setpoints for Rod Block Monitor Upscale Functions for Specification 3.3.2.1, Table 3.3.2.1-1.
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC.

SUSQUEHANNA - UNIT 1 5.0-21 Amendment 178, 186, 194, 215, 217 230, 233, 242, 246, XXX

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 < 575 psig or core flow

< 10 million lbm/hr:

THERMAL POWER shall be 23% RTP.

2.1.1.2 With the reactor steam dome pressure 575 psig and core flow 10 million lbm/hr:

MCPR shall be 1.051.08 for two recirculation loop operation or 1.11 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.

SUSQUEHANNA - UNIT 2 2.0-1 Amendment 151, 154, 164, 184, 191 194, 218, 224, 230, 242, 260, XXX

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 Not Used 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 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 for Specification 3.2.3;
4. The Shutdown Margin for Specification 3.1.1;
5. Oscillation Power Range Monitor (OPRM) Trip setpoints, for Specification 3.3.1.1; and
6. The Allowable Values and power range setpoints for Rod Block Monitor Upscale Functions for Specification 3.3.2.1, Table 3.3.2.1-1.
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC.

SUSQUEHANNA - UNIT 2 5.0-21 Amendment 151, 154, 169, 190, 192 207, 210, 220, 224, XXX

Attachment 3 of PLA-7962 Revised (Clean) Technical Specification Pages Revised Technical Specifications Pages Unit 1 TS Pages 2.0-1, 5.0-21 Unit 2 TS Pages 2.0-1, 5.0-21

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SL 2.1.1.1 With the reactor steam dome pressure < 575 psig or core flow

< 10 million lbm/hr:

THERMAL POWER shall be 23% RTP.

2.1.1.2 With the reactor steam dome pressure 575 psig and core flow 10 million lbm/hr:

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.

SUSQUEHANNA - UNIT 1 2.0-1 Amendment 178, 186, 199, 216, 227 231, 246, 261, 278, XXX

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 Not Used 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 for Specification 3.2.1;
2. The Minimum Critical Power Ratio (MCPR) and MCPR99.9% for Specification3.2.2;
3. The Linear Heat Generation Rate for Specification 3.2.3;
4. The Shutdown Margin for Specification 3.1.1;
5. Oscillation Power Range Monitor (OPRM) Trip Setpoints, for Specification 3.3.1.1; and
6. The Allowable Values and power range setpoints for Rod Block Monitor Upscale Functions for Specification 3.3.2.1, Table 3.3.2.1-1.
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC.

SUSQUEHANNA - UNIT 1 5.0-21 Amendment 178, 186, 194, 215, 217 230, 233, 242, 246, XXX

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 < 575 psig or core flow

< 10 million lbm/hr:

THERMAL POWER shall be 23% RTP.

2.1.1.2 With the reactor steam dome pressure 575 psig and core flow 10 million lbm/hr:

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.

SUSQUEHANNA - UNIT 2 2.0-1 Amendment 151, 154, 164, 184, 191 194, 218, 224, 230, 242, 260, XXX

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 Not Used 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 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 for Specification 3.2.3;
4. The Shutdown Margin for Specification 3.1.1;
5. Oscillation Power Range Monitor (OPRM) Trip setpoints, for Specification 3.3.1.1; and
6. The Allowable Values and power range setpoints for Rod Block Monitor Upscale Functions for Specification 3.3.2.1, Table 3.3.2.1-1.
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC.

SUSQUEHANNA - UNIT 2 5.0-21 Amendment 151, 154, 169, 190, 192 207, 210, 220, 224, XXX

Attachment 4 of PLA-7962 Marked-Up Technical Specification Bases Changes Revised Technical Specifications Bases Pages Unit 1 TS Bases Pages 2.0-1, 2.0-2, 2.0-3, 2.0-5, 3.2-5, 3.2-6, 3.2-8, 3.2-9 Unit 2 TS Bases Pages 2.0-1, 2.0-2, 2.0-3, 2.0-5, 3.2-5, 3.2-6, 3.2-9 (Provided for Information Only)

Rev. 7 Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and 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 significant 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 for ATRIUM 10 and ATRIUM 11 fuel. 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 at 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.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient. Inside the SUSQUEHANNA - UNIT 1 2.0-1

Rev. 7 Reactor Core SLs B 2.1.1 BASES BACKGROUND in oxidation of the fuel cladding to a structurally weaker form. This weaker (continued) form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

APPLICABLE The fuel cladding must not sustain damage as a result of normal operation SAFETY and AOOs. The Technical Specification SL is set generically on a fuel ANALYSES 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 use of the SPCB (Reference 4) correlation is valid for critical power calculations with ATRIUM 10 fuel at pressures 571.4 psia (conservatively bounded by 575 psig) and bundle mass fluxes > 0.087 x 106 lb/hr-ft2.

The use of the ACE/ATRIUM 11 (Reference 6) correlation is valid for critical power calculations with ATRIUM 11 fuel at pressures 588.8 psia (conservatively bounded by 575 psig) with no minimum bundle mass flux.

For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a relatively high power and potentially can approach a critical heat flux condition. For ATRIUM 10 and ATRIUM 11 fuel, the minimum bundle flow is > 28 x 103 lb/hr, and the coolant minimum bundle flow and maximum area are such that the mass flux is always > 0.24 x 106 lb/hr-ft2. Full scale critical power test data taken from various fuel designs at pressures from 14.7 psia to 1400 psia indicate that the fuel assembly critical power at 0.24 x106 lb/hr-ft2 is approximately 3.35 MWt. At 23% RTP, a bundle power of SUSQUEHANNA - UNIT 1 2.0-2

Rev. 7 Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Cladding Integrity (continued)

SAFETY ANALYSES pressures < 575 psig is conservative and for conditions of lesser power (continued) would remain conservative.

2.1.1.2 MCPR The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e., MCPR = 1.00) and the MCPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty in the critical power correlation. References 2, 4, 5, and 6 describe the methodology used in determining the MCPR SL.The fuel cladding integrity SL is set such that no significant 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 fuel cladding integrity 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 MCPR95/95 values for the fuel product lines that are fresh or once-burnt at the start of the cycle. References 4, 6, and 7 described the methodology used in determining the SLMCPR95/95.

The SPCB and ACE/ATRIUM 11 critical power correlations are based on a significant body of practical test data. As long as the core pressure and flow are within the range of validity of the correlations (refer to Section B.2.1.1.1), the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. These conservatisms and the SUSQUEHANNA - UNIT 1 2.0-3

Rev. 7 Reactor Core SLs B 2.1.1 BASES SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS radioactive releases in excess of regulatory limits. Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, (as identified in the COLR).Not used.
3. Not used.
4. EMF-2209(P)(A), "SPCB Critical Power Correlation, AREVA NP, (as identified in the COLR).
5. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/Microburn-B2, (as identified in the COLR).Not used.
6. ANP-10335P-A, ACE/ATRIUM 11 Critical Power Correlation, (as identified in the COLR).

6.7. ANP-3857P, Design Limits for Framatome Critical Power Correlations, Revision 2.

SUSQUEHANNA - UNIT 1 2.0-5

Rev. 5 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 establish the operating limit MCPR are presented in References 2, 3, 5, 7, ANALYSES and 10 for ATRIUM 10 fuel design analysis and references 2, 3, 5, 7, 10, and 12 through 15 for ATRIUM 11 fuel designs. To ensure that the MCPR Safety Limit (SL) is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been 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 combined withadded to the MCPR99.9% SL, the required operating limit MCPR 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 SUSQUEHANNA - UNIT 1 3.2-5

Rev. 5 MCPR B 3.2.2 used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved Critical Power correlations.

SUSQUEHANNA - UNIT 1 3.2-5

Rev. 5 MCPR B 3.2.2 BASES APPLICABLE Details of the MCPR99.9% calculation are given in References 7, 15, and 16.

SAFETY References 7 and 15 also include a tabulation of the uncertainties and the ANALSYSES nominal values of the parameters used in the MCPR99.9% statistical (continued) analysis.

The MCPR operating limits are derived from the MCPR99.9% value and the transient analysis, and are dependent on the operating core flow and power state to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency. These analyses may also consider other combinations of plant conditions (i.e., control rod scram speed, bypass valve performance, EOC-RPT, cycle exposure, etc.). Flow dependent MCPR limits are determined by analysis of slow flow runout transients.

The MCPR satisfies Criterion 2 of the NRC Policy Statement (Ref. 11).

LCO The MCPR operating limits 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 flow dependent MCPR and power dependent MCPR 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 MCPR SL is not exceeded even if a limiting transient occurs.

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.

SUSQUEHANNA - UNIT 1 3.2-6

Rev. 5 MCPR B 3.2.2 BASES SURVEILLANCE SR 3.2.2.2 (continued)

REQUIREMENTS (continued) based on the LCO 3.1.4 "Control Rod Scram Times" and the realistic scram times, both of which are used in the transient analysis. If the average measured scram times are greater than the realistic scram times then the MCPR operating limits corresponding to the Maximum Allowable Average Scram Insertion Time must be implemented. Determining MCPR operating limits based on interpolation between scram insertion times is not permitted. The average measured scram times and corresponding MCPR operating limit must be determined once within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each set of scram time tests required by SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3 and SR 3.1.4.4 because the effective scram times may change during the cycle. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is acceptable due to the relatively minor changes in average measured scram times expected during the fuel cycle.

REFERENCES 1. NUREG-0562, June 1979.

2. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors, Exxon Nuclear Company, March 1983.
3. XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.
4. Not used.
5. XN-NF-80-19 (P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986.
6. Not used.
7. EMF-2209(P)(A), SPCB Critical Power Correlation, AREVA NP, (as identified in the COLR).
8. Not used.
9. Not used
10. ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, (as identified in the SUSQUEHANNA - UNIT 1 3.2-8

Rev. 5 MCPR B 3.2.2 BASES REFERENCES 11. Final Policy Statement on Technical Specifications Improvements, (continued) July 22, 1993 (58 FR 39132).

12. ANP-10300P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Transient and Accident Scenarios, (as identified in the COLR).
13. BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, (as identified in the COLR)
14. BAW-10247P-A, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors Supplement 2: Mechanical Methods, (as identified in the COLR).
15. ANP-10335P-A, ACE/ATRIUM-11 Critical Power Correlation, (as identified in the COLR).

15.16. ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, (as identified in the COLR).

SUSQUEHANNA - UNIT 1 3.2-9

Rev. 6 Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and 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 significant 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 for ATRIUM 10 and ATRIUM 11 fuel. 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 at 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.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.

SUSQUEHANNA - UNIT 2 2.0-1

Rev. 6 Reactor Core SLs B 2.1.1 BASES BACKGROUND reaction results in oxidation of the fuel cladding to a structurally weaker (continued) form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

APPLICABLE The fuel cladding must not sustain damage as a result of normal operation SAFETY and AOOs. The Technical Specification SL is set generically on a fuel ANALYSES 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 use of the SPCB (Reference 4) correlation is valid for critical power calculations with ATRIUM 10 fuel at pressures 571.4 psia (conservatively bounded by 575 psig) and bundle mass fluxes

> 0.087 x 106 lb/hr-ft2.

The use of the ACE/ATRIUM 11 (Reference 6) correlation is valid for critical power calculation with ATRIUM 11 fuel at pressures 588.8 psia (conservatively bounded by 575 psig) with no minimum bundle mass flux.

For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a relatively high power and potentially can approach a critical heat flux condition. For ATRIUM 10 and ATRIUM 11 fuel, the minimum bundle flow is > 28 x 103 lb/hr and the coolant minimum bundle flow and maximum area are such that the mass flux is always > 0.24 x 106 lb/hr-ft2. Full scale critical power test data taken from various fuel designs at pressures from 14.7 psia to 1400 psia indicate that the fuel assembly critical power at SUSQUEHANNA - UNIT 2 2.0-2

Rev. 6 Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Cladding Integrity (continued)

SAFETY ANALYSES expected peaking factor. Thus, a THERMAL POWER limit of 23% RTP for (continued) reactor pressures < 575 psig is conservative and for conditions of lesser power would remain conservative.

2.1.1.2 MCPR The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,

MCPR = 1.00) and the MCPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty in the critical power correlation. References 2, 4, 5, and 6 describe the methodology used in determining the MCPR SL.The fuel cladding integrity SL is set such that no significant 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 MCPR95/95 values for the fuel product lines that are fresh or once-burnt at the start of the cycle. References 4, 6, and 7 described the methodology used in determining the SLMCPR95/95.

The SPCB and ACE/ATRIUM 11 critical power correlations are based on a significant body of practical test data. As long as the core pressure and flow are within the range of validity of the correlation (refer to Section B 2.1.1.1), the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. These conservatisms and the SUSQUEHANNA - UNIT 2 2.0-3

Rev. 6 Reactor Core SLs B 2.1.1 BASES SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS radioactive releases in excess of regulatory limits. Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, (as identified in the COLR)Not used.
3. Not used.
4. EMF-2209(P)(A), SPCB Critical Power Correlation, AREVA NP, (as identified in the COLR).
5. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4 /

MICROBURN-B2, (as identified in the COLR)Not used.

6. ANP-10335P-A, ACE/ATRIUM 11 Critical Power Correlation, (as identified in the COLR).

6.7. ANP-3857P, Design Limits for Framatome Critical Power Correlations, Revision 2.

SUSQUEHANNA - UNIT 2 2.0-5

Rev. 5 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 establish the operating limit MCPR are presented in References 2, 3, 5, 7, ANALYSES and 10 for ATRIUM 10 fuel design analysis and References 2, 3, 5, 7, 10, and 12 through 15 for ATRIUM 11 fuel designs. To ensure that the MCPR Safety Limit (SL) is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been 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 combined with added to the MCPR99.9% SL, the required operating limit MCPR 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 SUSQUEHANNA - UNIT 2 3.2-5

Rev. 5 MCPR B 3.2.2 that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the SUSQUEHANNA - UNIT 2 3.2-5

Rev. 5 MCPR B 3.2.2 BASES APPLICABLE occurrence of boiling transition is determined using the approved Critical SAFETY Power correlations. Details of the MCPR99.9% calculation are given in ANALYSES References 7, 15, and 16. References 7 and 15 also include a tabulation (continued) of the uncertainties and the nominal values of the parameters used in the MCPR99.9% statistical analysis The MCPR operating limits are derived from the MCPR99.9% value and the transient analysis, and are dependent on the operating core flow and power state to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency. These analyses may also consider other combinations of plant conditions (i.e., control rod scram speed, bypass valve performance, EOC-RPT, cycle exposure, etc.). Flow dependent MCPR limits are determined by analysis of slow flow runout transients.

The MCPR satisfies Criterion 2 of the NRC Policy Statement (Ref. 11).

LCO The MCPR operating limits 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 flow dependent MCPR and power dependent MCPR 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 MCPR SL is not exceeded even if a limiting transient occurs.

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.

SUSQUEHANNA - UNIT 2 3.2-6

Rev. 5 MCPR B 3.2.2 BASES REFERENCES 11. Final Policy Statement on Technical Specifications Improvements, July (continued) 22, 1993 (58 FR 39132).

12. ANP-10300P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Transient and Accident Scenarios, (as identified in the COLR).
13. BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, (as identified in the COLR).
14. BAW-10247P-A Supplement 2P-A, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors Supplement 2:

Mechanical Methods, (as identified in the COLR).

15. ANP-10335P-A, ACE/ATRIUM-11 Critical Power Correlation, (as identified in the COLR).

15.16. ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, (as identified in the COLR).

SUSQUEHANNA - UNIT 2 3.2-9

Attachment 6 of PLA-7962 Framatome Report ANP-3857NP, Revision 2 Design Limits for Framatome Critical Power Correlations (Non-proprietary Version)

For Information Only framatome Design Limits for Framatome ANP-3857NP Revision 2 Critical Power Correlations July 2020 (c) 2020 Framatome Inc.

0414-12-F04 (Rev. 004, 04/27/2020)

For Information Only ANP-3857NP Revision 2 Copyright © 2020 Framatome Inc.

All Rights Reserved 0414-12-F04 (Rev. 004, 04/27/2020)

For Information Only Framatome Inc. ANP-3857NP Revision 2 Design Limits for Framatome Critical Power Correlations Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Corrected the reference numbers in Table 1.

For Information Only Framatome Inc. ANP-3857NP Revision 2 Design Limits for Framatome Critical Power Correlations Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1 2.0 DEFINITIONS .................................................................................................... 2-1 3.0 CRITICAL POWER CORRELATION STATISTICS ........................................... 3-1 4.0 DESIGN LIMITS ................................................................................................ 4-1

5.0 REFERENCES

.................................................................................................. 5-1

For Information Only Framatome Inc. ANP-3857NP Revision 2 Design Limits for Framatome Critical Power Correlations Page iii List of Tables Table 1 MCPR95/95 Limits for Framatome Fuel ............................................................ 4-1

For Information Only Framatome Inc. ANP-3857NP Revision 2 Design Limits for Framatome Critical Power Correlations Page 1-1

1.0 INTRODUCTION

Design limits are provided for Boiling Water Reactor (BWR) critical power correlations.

These limits are based solely on the critical power correlation uncertainty determined from benchmarking the correlation to experimental data. Pressurized Water Reactor (PWR) Departure from Nucleate Boiling (DNB) correlation design limits are typically determined from the correlation uncertainty. The statistical expectation for the design limit is provided in Reference 1 where it states, For departure from nucleate boiling ratio (DNBR), CHFR or CPR correlations, there should be a 95-percent probability at the 95-percent confidence level that the hot rod in the core does not experience a DNB or boiling transition condition during normal operation or AOOs Design limits for PWR are typically reported and reviewed in the correlation topical reports (for example Reference 2) and are associated with the fuel, independent of the reactor. In this report, comparable limits are determined for BWR critical power correlations.

For Information Only Framatome Inc. ANP-3857NP Revision 2 Design Limits for Framatome Critical Power Correlations Page 2-1 2.0 DEFINITIONS With respect to the limit on Critical Heat Flux (CHF), the Safety Limit (SL) in the BWR is defined in the technical specifications as the lowest allowable Critical Power Ratio (CPR) in the reactor core. The CPR is defined in Reference 3.

The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

The SL on MCPR (sometimes referred to as SLMCPR) includes the influence of plant dependent and fuel dependent uncertainties.

In this report, a design limit is established from the critical power correlation and its uncertainty. This limit determined to be the value of CPR such that, at the 95%

confidence level, there is 95% probability that dryout is avoided.

For Information Only Framatome Inc. ANP-3857NP Revision 2 Design Limits for Framatome Critical Power Correlations Page 3-1 3.0 CRITICAL POWER CORRELATION STATISTICS In the context of the CPR defined above, an Experimental Critical Power Ratio (ECPR) is defined Measured Critical Power ECPR = (1)

Calculated Critical Power According to this definition, an ECPR that is less than 1.0 is non-conservative and a value greater than 1.0 is conservative. For best estimate critical power correlations, the mean value of ECPR is 1.0 or very close to 1.0. The representation of the ECPR as a normal distribution is addressed directly in the critical power correlation topical report.

With the conclusion that the distribution is represented by a normal distribution, the design limit is calculated 1+ k 95 95 s MCPR95 95 = (2)

ECPR where k is the one sided tolerance limit factor generally attributed to D. B. Owen and given by Reference 4 or 5 and s is the sample standard deviation.

The ECPR definition applied by Framatome for statistical analysis of the critical power correlations is the inverse of the definition shown in Equation (1).

Calculated Critical Power ECPR = (3)

Measured Critical Power According to this definition of the ECPR, the design limit is calculated MCPR95= 95 ECPR x (1 + k 95 95 s ) (4)

It is observed that when the ECPR is equal to 1.0, Equations (2) and (4) become equal.

For Information Only Framatome Inc. ANP-3857NP Revision 2 Design Limits for Framatome Critical Power Correlations Page 5-1

5.0 REFERENCES

1. Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition - Reactor, NUREG-0800, Chapter 4.4, Revision 2, page 4.4-5.
2. EMF-92-153(P)(A), Revision 1, HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, Siemens Power Corporation, Nuclear Division, January 2005.
3. Standard Technical Specifications General Electric BWR/4 Plants Volume 1, NUREG-1433 Volume 1.0, Revision 4.0, ADAMS Ascension No. ML12104A192, page 1.1-4.
4. D. B. Owen, Factors for One-sided Tolerance Limits and for Variables Sampling Plans, Sandia Corporation Report SCR-607, March 1963.
5. M. G. Natrella, Experimental Statistics, National Bureau of Standards Handbook 91, August 1963.
6. EMF-2209(P)(A), Revision 3, SPCB Critical Power Correlation, AREVA NP Inc.,

September 2009.

7. ANP-10249P-A, Revision 2, ACE/ATRIUM-10 Critical Power Correlation, AREVA Inc.,

March 2014.

8. ANP-10298P-A, Revision 1, ACE/ATRIUM 10XM Critical Power Correlation, AREVA Inc., March 2014.
9. ANP-10335P-A, Revision 0, ACE/ATRIUM 11 Critical Power Correlation, Framatome Inc., May 2018.

Attachment 7 of PLA-7962 Framatome Affidavit Affidavit for ANP-3857P, Revision 2, Design Limits for Framatome Critical Power Correlations

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 the report ANP-3857P, Revision 2 "Design Limits for Framatome Critical Power Correlations," dated July 2020 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.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on: September 23, 2021 MEG IN NIS AIan Digitally signed by MEGINNIS Alan Date: 2021.09.23 14:31 :30 -07'00' Alan B. Meginnis

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