L-MT-20-023, License Amendment Request: Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit MCPR

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License Amendment Request: Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit MCPR
ML20308A826
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 11/03/2020
From: Conboy T
Northern States Power Company, Minnesota, Xcel Energy
To:
Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML20308A825 List:
References
L-MT-20-023
Download: ML20308A826 (40)
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Attachment 5 Contains Proprietary Information Withhold in Accordance with 10 CFR 2.390 fl Xcel Energy 2807 West County Road 75 Monticello, MN 55089 November 3, 2020 L-MT-20-023 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 License Amendment Request: Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit MCPR In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), is submitting a request for an amendment to the Technical Specifications (TS) for the Monticello Nuclear Generating Plant (MNGP).

NSPM requests adoption of TSTF-564, Safety Limit MCPR, Revision 2, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the MNGP TSs. The proposed amendment revises the TS safety limit (SL) minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for a safety limit.

Enclosed is a description and assessment of the proposed TS changes. Attachment 1 to the enclosure provides the existing TS pages marked up to show the proposed change. to the enclosure provides the TS Bases pages marked up to show the associated TS bases changes and is provided for information only.

Attachments 3 and 5 to the enclosure provide a non-proprietary and a proprietary version, respectively, of a Framatome, Inc., licensing report entitled ANP-3857P, Revision 2, Design Limits for Framatome Critical Power Correlations, which provides the SLMCPR95/95 value for the ATRIUM' 10XM fuel type. Framatome, as owner of the proprietary information, has executed an affidavit provided in Attachment 4 to the enclosure which identifies the proprietary information. Attachment 5 to the enclosure is requested to be withheld in accordance with 10 CFR 2.390, Public inspections, exemptions, requests for withholding.

Document Control Desk Page 2 In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), NSPM is notifying the State of Minnesota by providing a copy of this application to the designated official.

NSPM requests approval of this proposed license amendment within 12 months after NRC acceptance, with an implementation period of 90 days.

If there are any questions or if additional information is needed, please contact Mr. Richard Loeffler at (612) 342"8981 or Rick.A.Loeffler@xcelenergy.com.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

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

Thomas A. Conboy Site Vice President, Monticello lear Generating Plant Northern States Power Company- Minnesota Enclosure cc: Administrator, Region Ill, US NRC Project Manager, Monticello, US NRC Resident Inspector, Monticello, US NRC State of Minnesota (non"proprietary portion)

L-MT-20-023 NSPM Enclosure LICENSE AMENDMENT REQUEST APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-564, SAFETY LIMIT MCPR

1.0 DESCRIPTION

In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requests an amendment to the Technical Specifications (TS) for the Monticello Nuclear Generating Plant (MNGP). NSPM requests adoption of TSTF-564, Safety Limit MCPR, Revision 2, (Reference 1), which is an approved change to the Improved Standard Technical Specifications (ISTS), into the MNGP 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 safety limit.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation NSPM has reviewed the final safety evaluation for TSTF-564 provided to the Technical Specifications Task Force (TSTF) in a letter dated November 16, 2018 (Reference 2). This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-564. As described herein, NSPM has concluded that the justifications presented in TSTF-564 and the safety evaluation prepared by the NRC staff are applicable to the MNGP and justify this amendment for the incorporation of the changes to the MNGP TS.

The MNGP reactor is currently fueled with ATRIUM'10XM fuel assemblies provided by Framatome, Inc., and General Electric (GE)14 fuel assemblies provided by Global Nuclear Fuel - Americas, LLC. The new MCPR safety limit (referred to as MCPR95/95) in Specification 2.1.1.3 is 1.05 as specified for the ATRIUM' 10XM fuel design in Framatome licensing report ANP-3857P, Design Limits for Framatome Critical Power Correlations, (Reference 3), provided in non-proprietary and proprietary versions in Attachments 3 and 5, respectively. This report provides details of the calculation of the MCPR95/95 safety limit for the ATRIUM' 10XM fuel type using the statistics from the ACE/ATRIUM 10XM CPR correlation database contained in ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, (Reference 4), which is listed in TS 5.6.3, Core Operating Limits Report (COLR), as Item b.20.

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%. Specification 5.6.3 is revised to require the MCPR99.9% to be included in the cycle-specific COLR. The ATRIUM' 10 XM is identified in the TS Bases as the fuel type the safety limit is based upon since it will be the limiting fuel type in the MNGP core.

Page 1 of 6

L-MT-20-023 NSPM Enclosure 2.2 Variations NSPM is proposing the following variations from the TS changes described in TSTF-564 or the applicable parts of the NRC staffs safety evaluation.

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

Specifically, Specification 2.1.1.2 in the ISTS corresponds to Specification 2.1.1.3 in the MNGP TS. This difference is administrative and does not affect the applicability of TSTF-564 to the MNGP.

The MNGP TS specify a different reactor steam dome pressure value (586 psig) in Specifications 2.1.1.1 and 2.1.1.3, 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 the MNGP.

NSPM currently uses the Framatome ATRIUM' 10XM fuel and GE14 fuel types. The GE14 fuel type is identified in Table 1 of TSTF-564 but the ATRIUM' 10 XM fuel type is not.

Framatome licensing report ANP-3857P presents information on the derivation of the MCPR95/95 safety limit for the ATRIUM' 10XM fuel type, along with other fuel types, and the respective NRC approved CPR correlations. Applying the TSTF-564 approach for additional fuel types is within the scope of the TSTF-564 approval and does not affect the applicability of the TSTF to the MNGP TS.

The Framatome Extended Flow Window (EFW) operating domain stability solution was approved for application at the MNGP in Amendment No. 191 (Reference 5). The EFW and MELLLA+ operating domains are analogous. A MCPR penalty (0.03) is added to the safety limit when the ratio of core power to core flow is greater than or equal to 42 MWt/Mlb/hr in the EFW domain (SL 2.1.1.3.c), to address power distribution uncertainties. The TS Bases provided in Attachment 2 discuss the EFW.

With the redefinition of the MCPR safety limit under this TSTF, it is no longer appropriate to adjust the safety limit value specified in Specification 2.1.1.3 in this manner because the MCPR95/95 safety limit is only fuel type dependent and not plant and/or cycle-dependent.

Critical power correlations are generically reviewed and are independent of any specific plant or cycle. Penalties assessed on the Experimental Critical Power Ratio (ECPR) mean or ECPR standard deviation, used to calculate the MCPR95/95 safety limit, would be applied both in the MCPR95/95 safety limit determination and in the MCPR99.9% safety limit determination. These penalties, if any, are found in the associated critical power correlation topical report. As discussed in Section 3.2 of the NRC safety evaluation for TSTF-564, the 0.03 MCPR safety limit adder corresponds to a penalty applied to the MCPR99.9% safety limit for operation in the Maximum Extended Load Limit Line Analysis Plus (MELLLA+) operating domain - analogous to the EFW, as previously stated. Thus, according to the definition and intent of the MCPR95/95 safety limit in TSTF-564, the 0.03 safety limit adder should not be added to the MCPR95/95 safety limit but should be added to the MCPR99.9% safety limit when operating in the EFW operating domain when the ratio of core power to core flow is greater than or equal to 42 MWt/Mlb/hr.

Page 2 of 6

L-MT-20-023 NSPM Enclosure The traveler and safety evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). The MNGP was not licensed to the 10 CFR 50, Appendix A, GDC. As discussed in the safety evaluation the applicable GDC is GDC 10, Reactor Design, which is listed below.

GDC 10 - Reactor Design The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

The MNGP equivalents to the GDC are contained in the MNGP Updated Safety Analysis Report (USAR), Appendix E, Plant Comparative Evaluation with the Proposed AEC 70 Design Criteria, and are listed below.

Criterion 6 - Reactor Core Design (Category A) The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of off-site power.

Criterion 14 - Core Protection Systems (Category B) Core protection systems together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.

As can be seen, both the GDC and the MNGP-specific design criteria are effectively equivalent in that they both require the reactor core and associated control and protection systems to assure that specified acceptable fuel design limits are not exceeded under the specified conditions. Therefore, this difference does not alter the conclusion that the proposed change is applicable to the MNGP.

3.0 REGULATORY ANALYSIS

3.1 Precedent Duke Energy Progress, LLC, submitted on March 9, 2020, an application to revise the Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Technical Specifications requesting the adoption of TSTF 564, Safety Limit MCPR, (Reference 6). The Brunswick nuclear power plants are fueled with the Framatome ATRIUM' 10XM fuel type and the ATRIUM 11' fuel type was indicated as expected to be introduced in the of spring 2020.

3.2 No Significant Hazards Consideration Analysis Page 3 of 6

L-MT-20-023 NSPM Enclosure Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requests adoption of TSTF-564, Safety Limit MCPR, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the MNGP Technical Specifications (TS). The proposed change revises the TS minimum critical power ratio (MCPR) safety limit. 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 MCPR safety limit value will be used instead of two values applicable when one or two recirculation loops are in operation. TS 5.6.3, Core Operating Limits Report (COLR), is revised to require the MCPR99.9% safety limit value to be included in the COLR.

NSPM has evaluated if 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 MCPR safety limit and the list of core operating limits to be included in the Core Operating Limits Report (COLR). The MCPR safety limit 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.

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 MCPR safety limit 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 accident previously evaluated.

Page 4 of 6

L-MT-20-023 NSPM Enclosure

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

Response: No The proposed amendment revises the TS MCPR safety limit 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 MCPR safety limit 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 MCPR safety limit specification 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, NSPM 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.

3.3 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, Standards for Protection Against Radiation, 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, Criteria for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review, specifically paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed change.

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L-MT-20-023 NSPM Enclosure

5.0 REFERENCES

1. Technical Specifications Task Force (TSTF)-564, Revision 2, Safety Limit MCPR (ADAMS Accession No. ML18297A361)
2. U.S. Nuclear Regulatory Commission, (USNRC), Final Safety Evaluations of Technical Specifications Task Force Traveler TSTF-564, Revision 2, Safety Limit MCPR, Using the Consolidated Line Item Improvement Process (CAC No. MG0161, EPID L-2017-PMP-0007), dated October 24, 2018 (ADAMS Accession No. ML18299A054)
3. ANP-3857P, Revision 2, Design Limits for Framatome Critical Power Correlations, Framatome Inc., dated July 2020
4. ANP-10298P-A, Revision 1, ACE/ATRIUM 10XM Critical Power Correlation, AREVA Inc., dated March 2014
5. Amendment No. 191, Monticello Nuclear Generating Plant - Issuance of Amendment Re: Extended Flow Window (CAC No. MF5002), dated February 23, 2017 (ADAMS Accession No. ML163428311)
6. Duke Energy Progress, LLC, Application to Revise Technical Specifications to Adopt TSTF 564, Safety Limit MCPR, dated March 9, 2020 (ADAMS Accession No. ML20070H939)

Page 6 of 6

ENCLOSURE ATTACHMENT 1 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-564 SAFETY LIMIT MCPR TECHNICAL SPECIFICATION PAGES (MARKED-UP)

(3 pages follow)

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 < 586 psig or core flow

< 10% rated core flow:

THERMAL POWER shall be 25% RTP.

2.1.1.2 (Deleted) 2.1.1.3 With the reactor steam dome pressure 586 psig and core flow 10% rated core flow:

1.05.

a. For operation not in the EFW domain, MCPR shall be 1.08 for two recirculation loop operation, or 1.13 for single recirculation loop operation, or
b. For operation in the EFW domain and the ratio of power to core flow < 42 MWt/Mlb/hr, MCPR shall be 1.08, or
c. For operation in the EFW domain and the ratio of power to core flow 42 MWt/Mlb/hr, MCPR shall be 1.14.

2.1.1.4 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 1332 psig.

Monticello 2.0-1 Amendment No. 201 TBD

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.2 Radiological Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to May 15 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

5.6.3 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
1. The APLHGR for Specification 3.2.1;
2. The MCPR for Specification 3.2.2;
3. The LHGR for Specification 3.2.3; and MCPR99.9%

Monticello 5.6-1 Amendment No. 146 TBD

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued)

AREVA NP, Inc.,

21. ANP-10307P-A Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, June 2011
22. BAW-10255(P)(A) Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP Inc., May 2008 Revision 0 "
23. ANP-10262PA, Enhanced Option III Long Term Stability Solution, "

Revision 0, May 2008

24. (Deleted) ANP-3857P Revision 2, "Design Limits for Framatome Critical Power Correlations," Framatome, Inc., July 2020 The COLR will contain the complete identification for each of the Technical Specification referenced topical reports used to prepare the COLR (i.e.,

report number, title, revision, date, and any supplements).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

Monticello 5.6-4 Amendment No. 201 TBD

ENCLOSURE ATTACHMENT 2 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-564 SAFETY LIMIT MCPR TECHNICAL SPECIFICATION BASES PAGES (MARKED-UP - FOR INFORMATION ONLY)

(12 pages follow)

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

B 2.1.1 Reactor Core SLs BASES BACKGROUND USAR Section 1.2.2 (Ref. 1) requires the reactor core and associated systems to be designed to accommodate plant operational transients or maneuvers that might be expected without compromising safety and without fuel damage. Therefore, 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 and 2.1.1.3. 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.

This is accomplished by Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally having a Safety Limit cumulative and continuously measurable. Fuel cladding perforations, Minimum Critical Power however, can result from thermal stresses, which occur from reactor Ratio (SLMCPR) design operation significantly above design conditions.

basis, referred to as MCPR95/95 safety limit, While fission product migration from cladding perforation is just as which corresponds to a measurable as that from use related cracking, the thermally caused 95% probability at a 95% cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding confidence level (the deterioration. Therefore, the fuel cladding SL is defined with a margin to 95/95 MCPR criterion) the conditions that would produce onset of transition boiling (i.e.,

that transition boiling will MCPR = 1.00). These conditions represent a significant departure from not occur. the condition intended by design for planned operation. 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 steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical Monticello B 2.1.1-1 Revision No. 53 TBD

Reactor Core SLs B 2.1.1 BASES BACKGROUND (continued) 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 SL provides margin such that the SL 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 operation and AOOs. The reactor core SLs are established to preclude ANALYSES violation of the fuel design criterion that a 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 The Technical Protection System (RPS) Instrumentation"), in combination with the other Specification SL is LCOs, are designed to prevent any anticipated combination of transient set generically on a conditions for Reactor Coolant System water level, pressure, and fuel product MCPR THERMAL POWER level that would result in reaching the MCPR Safety Limit.

correlation basis as the MCPR which Framatome critical power correlations (ACE and SPCB) are applicable at corresponds to a reactor steam dome pressures > 586 psig. A Pressure Regulator Failure 95% probability at a Maximum Demand (Open) transient applying Framatome safety analysis 95% confidence methods would not violate Reactor Core Safety Limit 2.1.1.1.

level that transition boiling will not occur, referred to a 2.1.1.1 Fuel Cladding Integrity as MCPR95/95 safety Framatome critical power correlation ACE is applicable at pressures limit. 586 psig and core flows > 10% of rated flow. Framatome critical power correlation SPCB is applicable at pressures 586 psig and bundle inlet mass fluxes of 0.18 Mlb/hr/ft2. The ACE correlation is used for Framatome fuel and the SPCB correlation is used for co-resident fuel.

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 > 4.56 psi. Analyses (Ref. 2) 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 Monticello B 2.1.1-2 Revision No. 53 TBD

Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued)

The Technical a 4.56 psi driving head will be > 28 x 103 lb/hr. Full scale ATLAS test Specification SL value is data taken at pressures from 0 psig to 785 psig indicate that the fuel dependent on the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER product line and the

> 50 % RTP. Thus, a THERMAL POWER limit of 25% RTP for corresponding MCPR reactor pressure < 686 psig or < 10% core flow is conservative.

correlation, which is cycle independent. The value is based on the 2.1.1.2 (Deleted)

Critical Power Ratio (CPR) data statistics and 2.1.1.3 MCPR a 95% probability with 95% confidence that The fuel cladding integrity SL is set such that no significant fuel damage rods are not susceptible is calculated to occur if the limit is not violated. Since the parameters that to boiling transition, result in fuel damage are not directly observable during reactor operation, referred to as MCPR95/95 the thermal and hydraulic conditions that result in the onset of transition safety limit. 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 The SL is based on which boiling transition is calculated to occur has been adopted as a ATRIUM 10XM fuel convenient limit. However, the uncertainties in monitoring the core (Reference 3). For operating state and in the procedures used to calculate the critical power cores with a single fuel result in an uncertainty in the value of the critical power. Therefore, the product line, the fuel cladding integrity SL is defined as the critical power ratio in the MCPR95/95 safety limit is limiting fuel assembly for which more than 99.9% of the fuel rods in the the MCPR95/95 for the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

fuel type. For cores loaded with a mix of The MCPR SL is determined using a statistical model that combines all applicable fuel types, the the uncertainties in operating parameters and the procedures used to MCPR95/95 safety limit is calculate critical power.

based on the largest 3, (i.e., the most limiting) of The probability of the occurrence of boiling transition is determined using the approved Framatome correlations. References 8, 9, 10, and 11 the MCPR values for the describe the uncertainties and methodologies used in determining the fuel product lines that MCPR SL. TS Safety Limit 2.1.1.3.c applies when the ratio of core power are fresh or once-burnt to core flow exceeds 42 MWt/Mlb/hr in the extended operating domain at the start of the cycle. (Extended Flow Window (EFW)). Safety Limit 2.1.1.3.c consists of:

1) the cycle-specific SLMCPR calculated with Framatome methods and uncertainties each reload, and 2) a penalty of 0.03 added due to reduced confidence in power distribution uncertainties when operating at greater than 42 MWt/Mlb/hr in the EFW domain. This threshold is provided in Reference 13, and the basis for the 0.03 penalty is provided in Reference
14. This threshold is appropriate for MNGP because it represents a Monticello B 2.1.1-3 Revision No. 53 TBD

Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) sufficiently high power-flow ratio that is outside the normal range of plant maneuvering. In this way, the SLMCPR adder (0.03) will not adversely affect full power operation. The SLMCPR adder is not imposed on single-loop operation because single-loop operation is prohibited in the EFW region.

2.1.1.4 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. 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 elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to prevent the release of radioactive materials to the environs.

SL 2.1.1.1 and SL 2.1.1.3 ensure that the core operates within the fuel design criteria. SL 2.1.1.4 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.3, and 2.1.1.4 are applicable in all MODES.

SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS radioactive releases in excess of 10 CFR 50.67, Accident source term, limits (Ref. 4). 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.

Monticello B 2.1.1-4 Revision No. 53 TBD

Reactor Core SLs B 2.1.1 BASES REFERENCES 1. USAR, Section 1.2.2.

2. NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (revision specified in Specification 5.6.3)
3. (Deleted) ANP-3857P Revision 2, "Design Limits for Framatome Critical Power Correlations," Framatome, Inc., July 2020
4. 10 CFR 50.67.
5. (Deleted)
6. (Deleted)
7. Amendment No. 185, Issuance of Amendment to Reduce the Reactor Steam Dome Pressure Specified in the Reactor Core Safety Limits, dated November 25, 2014. (ADAMS Accession No. ML14281A318).
8. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.
9. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporations Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.
10. ANP-10298P-A Revision 1, ACE/ATRIUM 10XM Critical Power Correlation, AREVA, March 2014.
11. ANP-10307PA, Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.
12. Amendment No. 188, Issuance of Amendment to Transition to AREVA ATRIUM 10XM Fuel and AREVA Safety Analysis Methods, dated June 5, 2015. (ADAMS Accession Nos. ML15072A141, ML15154A477, and ML15072A135)
13. NRC letter to General Electric - Hitachi, Final Safety Evaluation for GE Hitachi Nuclear Energy Americas Topical Report NEDC-33173P, Revision 2 and Supplement 2, Parts 1-3, Analysis of Gamma Scan Data and Removal of Safety Limit Minimum Critical Power Ratio (SLMCPR) Margin (TAC No. ME1891). (ADAMS Accession No. ML113340215)

Monticello B 2.1.1-5 Revision No. 53 TBD

Reactor Core SLs B 2.1.1 BASES REFERENCES (continued)

14. GE-Hitachi, Final SE for NEDC-33173P, Applicability of GE Methods to Expanded Operating Domains, dated July 21, 2009. (ADAMS Accession No. ML083520464). This SE is an enclosure to NEDC-33173 Revision 4.

Monticello B 2.1.1-6 Revision No. 53 TBD

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 transition boiling to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid transition boiling if the limit is not violated (refer to the Bases for SL 2.1.1). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs). Although fuel damage does not necessarily occur if a fuel rod actually experienced transition boiling and that 99.9% of (Ref. 1), the critical power at which transition boiling is calculated to occur the fuel rods are has been adopted as a fuel design criterion.

not susceptible to boiling transition if The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these the limit is not experimental data, correlations have been developed to predict critical violated. 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.

is combined with the MCPR99.9% safety limit, APPLICABLE The analytical methods and assumptions used in evaluating the AOOs to SAFETY establish the operating limit MCPR are presented in References 2, 3, 4, 5, ANALYSES 6, 7, 8, 9, 13, 14, 15, 16, and 17. To ensure that the MCPR 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 added to the MCPR SL, the required operating limit MCPR is obtained.

INSERT A are MCPR99.9% value and the The MCPR operating limits derived from the transient analysis are (new paragraph) 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 (Refs. 7, 8, 9, and 10). Flow dependent MCPR limits are determined by steady state thermal hydraulic methods using the three-dimensional BWR simulator code (Ref. 14) and the multichannel thermal hydraulics code (Ref. 15).

The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.

Monticello B 3.2.2-1 Revision No. 52 TBD

INSERT A 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 Framatome critical power correlations. Details of the MCPR99.9% calculation are given in References 18, 19, and 20. References 18 and 20 also includes a tabulation of the uncertainties and the nominal values of the parameters used in the MCPR99.9% statistical analysis.

Due to reduced confidence in the power distribution uncertainties when operating at greater than 42 MWt/Mlb/hr in the Extended Flow Window (EFW) operating domain an 0.03 penalty is added to the MCPR99.9% safety limit. This threshold is provided in Reference 21, and the basis for the 0.03 penalty is provided in Reference 22. This threshold is appropriate for MNGP because it represents a sufficiently high power-flow ratio that is outside the normal range of plant maneuvering.

MCPR B 3.2.2 BASES approved transient analysis models.

APPLICABLE SAFETY ANALYSES (continued)

Power dependent MCPR limits (MCPRp) are determined by the three-dimensional BWR simulator code (Ref. 14) and the one-dimensional transient codes (Refs. 16 and 17). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPRp operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level. (MCPR99.9% value, the MCPRf values, and the The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). MCPRp values)

LCO The MCPR operating limits specified in the COLR 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 25% RTP, the reactor is operating at a low recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs. Statistical analyses indicate that the nominal value of the initial MCPR expected at 25% 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 25% 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 monitors provide rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels

< 25% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.

ACTIONS A.1 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met.

Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is normally sufficient to Monticello B 3.2.2-2 Revision No. 52 TBD

MCPR B 3.2.2 BASES ACTIONS (continued) restore the MCPR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.

B.1 If the MCPR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is 25% RTP and periodically thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.2.2.2 Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis.

SR 3.2.2.2 determines actual scram speed distribution and compares it with the assumed distribution. The MCPR operating limit is determined based either on the applicable limit associated with scram times of LCO 3.1.4, "Control Rod Scram Times," or the nominal scram times. The scram speed dependent MCPR limits are contained in the COLR. This determination must be performed 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 control rod scram time tests required by SR 3.1.4.1 and SR 3.1.4.2 because the effective scram speed distribution 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 the actual control rod scram speed distribution expected during the fuel cycle.

Monticello B 3.2.2-3 Revision No. 52 TBD

MCPR B 3.2.2 BASES REFERENCES 1. NUREG-0562, June 1979.

2. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (revision specified in Specification 5.6.3).
3. USAR, Section 3.2.4.
4. USAR, Section 6.2.6.
5. USAR, Chapter 14.
6. USAR, Chapter 14A.
7. NEDE-23785-P (A), Revision 1, "The GESTR-LOCA and SAFER Models for Evaluation of the Loss-of-Coolant Accident (Volume III),

SAFER/GESTR Application Methodology," October 1984.

8. NEDC-30515, "GE BWR Extended Load Line Limit Analysis for Monticello Nuclear Generating Plant, Cycle 11," March 1984.
9. NEDC-31849P, including Supplement 1, "Maximum Extended Load Line Limit Analysis for Monticello Nuclear Generating Plant Cycle 15," June 1992.
10. NEDC-30492-P, "Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Monticello Nuclear Generating Plant," April 1984.
11. NEDO-30130-A, "Steady State Nuclear Methods," May 1985.
12. NEDO-24154, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," October 1978.
13. ANP-10307PA, Revision 0, AREVA MCPR Safety Limits Methodology for Boiling Water Reactors, AREVA NP, June 2011.
14. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation for CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.
15. 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.

Monticello B 3.2.2-4 Revision No. 52 TBD

MCPR B 3.2.2 BASES REFERENCES (continued)

16. ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analysis, Advanced Nuclear Fuels Corporation, August 1990.
17. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.
18. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation,"

AREVA NP, September 2009.

19. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporations Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.
20. ANP-10298P-A Revision 1, ACE/ATRIUM 10XM Critical Power Correlation, AREVA, March 2014 NRC letter to General Electric - Hitachi, Final Safety Evaluation for GE 21.

Hitachi Nuclear Energy Americas Topical Report NEDC-33173P, Revision 2 and Supplement 2, Parts 1-3, Analysis of Gamma Scan Data and Removal of Safety Limit Minimum Critical Power Ratio (SLMCPR) Margin (TAC No. ME1891). (ADAMS Accession No. ML113340215)

22. GE-Hitachi, Final SE for NEDC-33173P, Applicability of GE Methods to Expanded Operating Domains, dated July 21, 2009. (ADAMS Accession No. ML083520464). This SE is an enclosure to NEDC-33173 Revision 4.

Monticello B 3.2.2-5 Revision No. 52 TBD

ENCLOSURE ATTACHMENT 3 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-564 SAFETY LIMIT MCPR ANP-3857NP, REVISION 2 DESIGN LIMITS FOR FRAMATOME CRITICAL POWER CORRELATIONS (10 pages follow)

Controlled Document 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)

Controlled Document ANP-3857NP Revision 2 Copyright © 2020 Framatome Inc.

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

Controlled Document 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.

Controlled Document 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

Controlled Document 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

Controlled Document 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.

Controlled Document 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.

Controlled Document 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.

Controlled Document Controlled Document 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.

ENCLOSURE ATTACHMENT 4 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-564 SAFETY LIMIT MCPR AFFIDAVIT FOR ANP-3857NP, REVISION 2 DESIGN LIMITS FOR FRAMATOME CRITICAL POWER CORRELATIONS (3 pages follow)

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: July 23, 2020

~i~

Alan Meginnis

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