ML21223A280

Issuance of Amendment No. 207 Adoption of TSTF-564 Safety Limit MCPR

ML21223A280
Person / Time
Site:	Monticello
Issue date:	10/15/2021
From:	Robert Kuntz Plant Licensing Branch III
To:	Conboy T Northern States Power Company, Minnesota
Kuntz R
References
EPID L-2020-LLA-0243
Download: ML21223A280 (19)

Text

October 15, 2021 Mr. Thomas A. Conboy Site Vice President Northern States Power Company - Minnesota Monticello Nuclear Generating Plant 2807 West County Road 75 Monticello, MN 55362

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT - ISSUANCE OF AMENDMENT NO. 207 RE: ADOPTION OF TSTF-564 SAFETY LIMIT MCPR (EPID L-2020-LLA-0243)

Dear Mr. Conboy:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 207 to Renewed Facility Operating License No. DPR-22, for the Monticello Nuclear Generating Plant. The amendment consists of changes to the technical specifications (TSs) in response to your application dated November 3, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20308A826).

The amendment revises TS Safety Limit (SL) 2.1.1.3, the reactor core safety limit for the minimum critical power ratio (MCPR). The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-564, Revision 2, Safety Limit MCPR, dated October 24, 2018 (ADAMS Accession No. ML18297A361).

A copy of the Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Robert F. Kuntz, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263

Enclosure:

1. Amendment No. 207 to DPR-22

2. Safety Evaluation cc: Listserv

NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 207 Renewed License No. DPR-22

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Northern States Power Company - Minnesota dated November 3, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-22 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 207, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Joel S.

Joel S. Wiebe Date: 2021.10.15 Wiebe 14:04:28 -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: October 15, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 207 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE DOCKET NO. 50-263 Renewed Facility Operating License No. DPR-22 Replace the following page of the Renewed Facility Operating License No. DPR-22 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

INSERT REMOVE Page 3 Page 3 Technical Specifications Replace the following pages of Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

INSERT REMOVE 2.0-1 2.0-1 5.6-1 5.6-1 5.6-4 5.6-4

2. Pursuant to the Act and 10 CFR Part 70, NSPM to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operations, as described in the Final Safety Analysis Report, as supplemented and amended, and the licensees filings dated August 16, 1974 (those portions dealing with handling of reactor fuel);

3. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required;

4. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

5. Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess, but not separate, such byproduct and special nuclear material as may be produced by operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission, now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

1. Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 2004 megawatts (thermal).

2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 207, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.

3. Physical Protection NSPM shall implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Amendment No. 207

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:

MCPR shall be 1.05.

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. 207

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 and MCPR99.9% for Specification 3.2.2;

3. The LHGR for Specification 3.2.3; Monticello 5.6-1 Amendment No. 207

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

21. ANP-10307P-A Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, Inc., June 2011

22. BAW-10255(P)(A) Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP Inc., May 2008

23. ANP-10262PA Revision 0, Enhanced Option III Long Term Stability Solution, AREVA NP, Inc., May 2008

24. 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. 207

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 207 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263

1.0 INTRODUCTION

By application dated November 3, 2020, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20308A826) Northern States Power Company (the licensee) submitted a license amendment request (LAR) for Monticello Nuclear Generating Plant (MNGP).

The LAR proposes to revise technical specification (TS) Safety Limit (SL) 2.1.1.3, the reactor core SL for the minimum critical power ratio (MCPR). The MCPR protects against boiling transition on the fuel rods in the core. The current MCPR SL for MNGP ensures that 99.9 percent of the fuel rods in the core are not susceptible to boiling transition. The revised MCPR SL will ensure that there is a 95 percent probability at a 95 percent confidence level that no fuel rods will be susceptible to boiling transition using an SL based on critical power ratio (CPR) data statistics. TS 5.6.3, Core Operating Limits Report (COLR), is also proposed to be modified.

The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-564, Revision 2, Safety Limit MCPR [Minimum Critical Power Ratio] (TSTF-564), dated October 24, 2018 (ADAMS Accession No. ML18297A361). The U.S. Nuclear Regulatory Commission (NRC or the Commission) issued a final safety evaluation (SE) approving TSTF-564 on November 16, 2018 (ADAMS Accession No. ML18299A069).

The LAR proposes variations from the TS changes described in TSTF-564. The variations are described in LAR Section 2.2 and evaluated in Section 3.5 of this SE. In addition to the General Electric (GE) 14 fuel type, MNGP also uses Framatome ATRIUMTM 10XM fuel type. The ATRIUMTM 10XM fuel assemblies are not explicitly identified in Table 1 of TSTF-564. As addressed in Section 3.5 of this SE, MNGP followed the methodology described in TSTF-564 to demonstrate allowance of the fuel.

Enclosure 2

2.0 REGULATORY EVALUATION

2.1 Background on Boiling Transition During steady state operation in a boiling-water reactor (BWR), most of the coolant in the core is in a flow regime known as annular flow. In this flow regime, a thin liquid film is pushed up the surface of the fuel rod cladding by the bulk coolant flow, which is mostly water vapor with some liquid water droplets. This provides effective heat removal from the cladding surface; however, under certain conditions, the annular film may dissipate, which reduces the heat transfer and results in an increase in fuel cladding surface temperature. This phenomenon is known as boiling transition or dryout. The elevated surface temperatures resulting from dryout may cause fuel cladding damage or failure.

2.2 Background on Critical Power Correlations For a given set of reactor operating conditions (pressure, flow, etc.), dryout will occur on a fuel assembly at a certain power known as the critical power. Because the phenomena associated with boiling transition are complex and difficult to model purely mechanistically, thermal-hydraulic test campaigns are undertaken using electrically heated prototypical fuel bundles to establish a comprehensive database of critical power measurements for each BWR fuel product. These data are then used to develop a critical power correlation that can be used to predict the critical power for assemblies in operating reactors. This prediction is usually expressed as the ratio of the actual assembly power to the critical power predicted using the correlation, known as the CPR.

One measure of the correlations predictive capability is based on its validation relative to the test data. For each point j in a correlations test database, the experimental critical power ratio (ECPR) is defined as the ratio of the measured critical power to the calculated critical power, or:

Measured Critical Powerj ECPRj = ____________________

Calculated Critical Powerj For ECPR values less than or equal to 1, the calculated critical power is greater than the measured critical power and the prediction is considered to be non-conservative. Because the measured critical power includes random variations due to various uncertainties, evaluating the ECPR for all of the points in the dataset (or, ideally, a subset of points that were not used in the correlations development) results in a probability distribution. This ECPR distribution allows the predictive uncertainty of the correlation to be determined. This uncertainty can then be used to establish a limit above which there can be assumed that boiling transition will not occur (with a certain probability and confidence level).

As discussed in Traveler TSTF-564, other fuel vendors may determine the MCPR95/95 for other fuel designs using the described methodology. The LAR provided the required description of the derivation of the MCPR95/95 for ATRIUMTM 10XM, which is based on the information contained in each fuel types NRC-approved CPR correlation that is referenced in MNGP TS 5.6.3.b. Framatome defines ECPR as the ratio of the calculated critical power to the measured critical power (i.e., the inverse of the TSTF-564 definition). The TSTF-564 95/95 formulation presumes a mean ECPR of one.

2.3 Background on Thermal-Hydraulic Safety Limits To protect against boiling transition, BWRs have implemented an SL on the CPR, known as the MCPR SL. As discussed in NUREG 1433 Standard Technical Specifications [STS], General Electric Plants BWR/4, NUREG 1433, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0 (ADAMS Accession Nos. ML12104A192 and ML12104A193) and 1434 Standard Technical Specifications, General Electric Plants BWR/6, NUREG-1434, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0 (ADAMS Accession Nos. ML12104A195 and ML12104A196), the current STS for GE BWR designs, the current basis of the MCPR SL for MNGP is to prevent 99.9 percent of the fuel in the core from being susceptible to boiling transition. This limit is typically developed by considering various cycle-specific power distributions and uncertainties, and is highly dependent on the cycle-specific radial power distribution in the core. As such, the limit may need to be updated as frequently as every cycle.

The TSs for MNGP also have a limiting condition for operation (LCO) that governs MCPR, known as the MCPR operating limit (OL). The OL on MCPR is an LCO which must be met to ensure that anticipated operational occurrences (AOOs) do not result in fuel damage. The current MCPR OL is calculated by combining the largest change in CPR from all analyzed transients, also known as the CPR, with the MCPR SL.

2.4 Description of TS Sections 2.4.1 TS 2.1.1 Reactor Core SLs SLs ensure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and AAOs.

MNGP TS 2.1.1.3 currently requires that with the reactor steam dome pressure greater than or equal to 586 pounds per square inch gauge (psig) and core flow 10 percent rated core flow, MCPR shall be:

a. For operation not in the Extended Flow Window (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.

The MCPR SL (also referred to as the MCPR99.9%) ensures that 99.9 percent of the fuel in the core is not susceptible to boiling transition.

2.4.2 TS 5.6.3, Core Operating Limits Report (COLR)

MNGP TS 5.6.3 requires core operating limits to be established prior to each reload cycle, or prior to any remaining portion of a reload cycle. These limits are required to be documented in the COLR.

2.5 Proposed Changes to the TSs The LAR proposes to revise the MCPR SL to make it cycle-independent, consistent with the method described in TSTF-564.

The proposed changes to the MNGP TS would revise the value of the MCPR SL in TS 2.1.1.3 to 1.05. The change to TS 2.1.1.3 replaces the existing separate SLs for single- and two-recirculation loop operation, as well as the power-dependent SLs for operations in the EFW domain, respectively, with a single limit since the revised SL is no longer dependent on the number of recirculation loops in operation.

The MCPR99.9% (i.e., the current MCPR SL) is an input to the MCPR OL in LCO 3.2.2, Minimum Critical Power Ratio (MCPR). While the definition and method of calculation of both the MCPR99.9% and the LCO 3.2.2 MCPR OL remain unchanged, the proposed TS changes include revisions to TS 5.6.3. The proposed change to TS 5.6.3 would require the additional MCPR99.9% value used in calculating the LCO 3.2.2 MCPR OL to be included in the cycle-specific COLR by changing the entry for item 5.6.3.a.2 from The MCPR for Specification 3.2.2; to The MCPR and MCPR99.9% for Specification 3.2.2.

The LAR proposes variations from the TS changes described in TSTF-564. MNGP uses Framatome ATRIUMTM 10XM fuel types which are not identified in TSTF-564, Table 1. As discussed in TSTF-564, other fuel vendors may determine the MCPR95/95 for other fuel designs using the described methodology. The LAR provided the required description of the derivation of the MCPR95/95 for ATRIUMTM 10XM, which is based on the information contained in each fuel types NRC-approved CPR correlation that is referenced in MNGP TS 5.6.3.b. Accordingly, the LAR proposed adding the appropriate report to TS 5.6.3.b, item 24.

The MNGP TS uses different numbering than the STS which TSTF-564 was based.

Specifically, MNGP TS 2.1.1.3 corresponds to STS 2.1.1.2.

The LAR also proposed minor changes to the titles of items 21 and 23 of TS 5.6.3.b. Item 21 would have the phrase AREVA NP, INC., added between the title and report date. Item 23 would have the phrase Revision 0 inserted between the report number and title as well as the phrase AREVA NP, INC., added between the title and report date.

2.6 Applicable Regulatory Requirements and Guidance The regulation at Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36(c),

requires that TSs include items in the following categories: SLs, limiting safety system settings, and limiting control settings. As required by 10 CFR 50.36(c)(1)(i)(A), SLs for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. If any SL is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be resumed until authorized by the Commission.

As required by 10 CFR 50.36(c)(2)(i), the TSs will include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met. Additionally, as required by

10 CFR 50.36(c)(5), TSs must include administrative controls, which are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

General Design Criterion (GDC) 10, Reactor design, of 10 CFR Part 50, Appendix A, General Design Criteria of Nuclear Power Plants, states:

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

With respect to GDC 10, the LAR stated that the MNGP equivalents to the GDC are contained in the MNGP Updated Safety Analysis Report (USAR), Appendix E, Plant Comparative Evaluation with the Proposed [Atomic Energy Commission] AEC 70 Design Criteria (ADAMS Accession No. ML20003D166). The plant-specific equivalent criteria are MNGP Criteria 6 and

14. This difference does not alter the conclusion that the proposed change based on TSTF-564 is applicable to MNGP because the plant-specific design criteria are similar to GDC 10.

The NRC staffs guidance contained in Revision 2 of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP),

Section 4.4, Thermal and Hydraulic Design (ADAMS Accession No. ML070550060) provides the following two examples of acceptable approaches to meeting the SRP acceptance criteria for establishing fuel design limits (as stated in SRP Acceptance Criterion 1):

A. For departure from nucleate boiling ratio (DNBR), CHFR [critical heat flux ratio] or CPR correlations, there should be a 95-percent probability at the 95-percent confidence level that the hot rod in the core does not experience a DNB [departure from nucleate boiling] or boiling transition condition during normal operation or AOOs.

B. The limiting (minimum) value of DNBR, CHFR, or CPR correlations is to be established such that at least 99.9 percent of the fuel rods in the core will not experience a DNB or boiling transition during normal operation or AOOs.

The NRC staffs guidance for the review of TSs is in Chapter 16.0, Revision 3, Technical Specifications, of SRP, Section 4.4, dated March 2010 (ADAMS Accession No. ML100351425).

As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STSs for each of the light water reactor (LWR) nuclear designs. Accordingly, the NRC staffs review considers whether the proposed changes are consistent with the applicable reference STSs (i.e., the current STSs), as modified by NRC-approved travelers. The STS applicable to MNGP is found in NUREG-1433, Revision 4.0.

3.0 TECHNICAL EVALUATION

3.1 Basis for Proposed Change As discussed in Section 2.3 of this SE, the current MCPR SL (i.e., the MCPR99.9%), is affected by the plants cycle-specific core design, especially including the core power distribution, fuel type(s) in the reactor, and the power-to-flow operating domain for the plant. As such, it is frequently necessary to change the MCPR SL to accommodate new core designs. Changes to

the MCPR SL are usually determined late in the design process and necessitate an accelerated NRC review (i.e., LAR) to support the subsequent fuel cycle.

The LAR proposed to change the methodology for determining the MCPR SL for MNGP so that it is no longer cycle dependent, reducing the frequency of revisions and eliminating the need for NRCs review on an accelerated schedule. The proposed methodology for determining the MCPR SL aligns it with that of the DNBR SL used in pressurized water reactors, which provides a 95 percent probability at a 95 percent confidence level that no fuel rods will experience departure from nucleate boiling.

The intent of the proposed calculational method for determining the revised MCPR SL is acceptable to the NRC staff based on the discussion in SRP Section 4.4, SRP Acceptance Criterion 1. The remainder of this SE is devoted to ensuring that the methodology for determining the revised MCPR SL provides the intended result, that the revised MCPR SL can be adequately determined in the core using various types of fuel, that the proposed SL continues to fulfill the necessary functions of an SL without unintended consequences, and that the proposed changes have been adequately implemented in the MNGP TSs.

3.2 Revised MCPR SL Definition As discussed in Section 2.2 of this SE, a critical power correlations ECPR distribution quantifies the uncertainty associated with the correlation and Framatomes definition of ECPR, which differs from that provided in TSTF-564. TSTF-564 provides a definition for a limit that bounds 95 percent of a correlations ECPR distribution at a 95 percent confidence level, according to the following formula:

MCPR9595(i) = µi + Kii where i is the correlations mean ECPR and i is the standard deviation of the correlations ECPR distribution. The statistical parameter(i) is selected, based on the number of samples in the critical power database, to provide 95% probability at 95% confidence (95/95) for the one-sided upper tolerance limit that depends on the number of samples (Ni) in the critical power database. This is a commonly used statistical formula to determine a 95/95 one-sided upper tolerance limit for a normal distribution, which is appropriate for the situation under consideration. The factor is generally attributed to D. B. Owen (ADAMS Accession No. ML14031A495) and was also reported by M. G. Natrella (Experimental Statistics, National Bureau of Standards, National Bureau of Standards Handbook 91, August 1963), as referenced in TSTF-564.

The LAR proposed variations from the TS changes described in TSTF 564. The variations include the use Framatome ATRIUMTM 10XM fuel type which is not identified in TSTF-564, Table 1. As discussed in TSTF-564, other fuel vendors may determine the MCPR95/95 for other fuel designs using the methodology. The LAR provided the required description of the derivations of the MCPR95/95 for ATRIUMTM 10XM, which is based on the information contained in each fuel type's NRC-approved CPR correlation that is referenced in MNGP TS 5.6.3.b. The NRC staff agrees that the difference is within the scope of the TSTF-564 approval and does not affect the applicability of TSTF-564 to the MNGP TS.

As discussed by Piepel and Cuta (Statistical Concepts and Techniques for Developing, Evaluating, and Validating CHF Models and Corresponding Fuel Design Limits, SKI Technical Report, 1993) for DNBR correlations, the acceptability of this approach is

predicated on a variety of assumptions, including the assumptions that the correlation data comes from a common population and that the correlations population is distributed normally.

These assumptions are typically addressed generically when a critical power or critical heat flux correlation is reviewed by the NRC staff, who may apply penalties to the correlation to account for any issues identified. TSTF letter dated May 29, 2018 (ADAMS Accession No. ML18149A320) that such penalties applied during the NRCs review of the critical power correlation would be imposed on the mean or standard deviation used in the calculating the MCPR95/95. These penalties would also continue to be imposed in the determination of the MCPR99.9%, along with any other penalties associated with the process of (or other inputs used in) determining the MCPR99.9% (e.g., penalties applied to the MCPR99.9% SL for operation in the Maximum Extended Load Limit Line Analysis Plus (MELLLA+) operating domain).

In the SE approving TSTF-564, the NRC staff found that the definition of the MCPR95/95 will appropriately establish a 95/95 upper tolerance limit on the critical power correlation and that any issues in the underlying correlation will be addressed through penalties on the correlation mean and standard deviation, as necessary. Therefore, the NRC staff concludes that the method for determining MCPR95/95, as proposed, can be used to establish acceptable fuel design limits in the MNGP TSs.

3.3 Determination of Revised MCPR SL for Mixed Cores TSTF-564 proposed that a core containing a variety of fuel types would evaluate the MCPR95/95 for all of the fresh and once-burnt fuel in the core and apply the most limiting (i.e., the largest) value of MCPR95/95 for each of the applicable fuel types as the MCPR SL. As stated in Section 3.1 of TSTF-564, this is because bundles that are twice-burnt or more at the beginning of the cycle have significant MCPR margin relative to the fresh and once-burnt fuel. The justification is that the MCPR for twice-burnt and greater fuel is far enough from the MCPR for the limiting bundle that its probability of boiling transition is very small compared to the limiting bundle and it can be neglected in determining the SL. Results of a study provided in the letter from the TSTF dated May 29, 2018 indicate that this is the case even for fuel operated on short (12-month) reload cycles. As discussed in the May 29, 2018 letter from the TSTF, twice-burnt or greater fuel bundles are included in the cycle-specific evaluation of the MCPR99.9% and the MCPR OL.

If a twice-burnt or greater fuel bundle is found to be limiting, it would be governed by the MCPR OL, which will always be more restrictive than both the MCPR95/95 and the MCPR99.9%. In the SE for TSTF-564 the NRC staff found this justification to be appropriate and determined that it is acceptable to determine the MCPR95/95 for the core based on the most limiting value of the MCPR95/95 for the fresh and once-burnt fuel in the core.

In the SE for TSTF-564 the NRC staff also reviewed the information furnished by the TSTF and determined that the process for establishing the revised MCPR SL (MCPR95/95) for mixed cores ensures that the limiting fuel types in the core will be evaluated and that the limiting MCPR95/95 will be appropriately applied as the SL. Therefore, the NRC staff finds it acceptable to determine the MCPR95/95 for the core based on the most limiting MCPR95/95 value for fresh and once-burnt fuel in the core for the MNGP TS.

3.4 Relationship between MCPR Safety and Operating Limits As discussed in the SE for TSTF-564 the MCPR99.9% is expected to always be greater than the MCPR95/95 for two reasons. First, because the MCPR99.9% includes uncertainties not factored into the MCPR95/95, and second, because the 99.9 percent probability basis for determining the

MCPR99.9% is more conservative than the 95 percent probability at a 95 percent confidence level used in determining the MCPR95/95. The level of conservatism in the MCPR95/95 is appropriate because the lead fuel rod in the core (i.e., the limiting fuel rod with respect to MCPR) is used to evaluate whether any fuel rods in the core are susceptible to boiling transition.

Consistent with TSTF-564, the MCPR OL defined in LCO 3.2.2 would continue to be evaluated using the MCPR99.9% as an input. The MCPR99.9% will continue to be evaluated in the same way as it is currently, using the whole core. The LAR proposed a change to LCO 3.2.2 that will continue to determine the MCPR operating limits for LCO 3.2.2 at MNGP using the MCPR99.9%

as an input.

Consistent with TSTF-564, the LAR proposed to revise TS 5.6.3 to require the cycle-specific value of the MCPR99.9% to be included in the COLR. The methods supporting the inclusion of the MCPR99.9% must also therefore, be included in the list of COLR references contained in TS 5.6.3.b in the MNGP TS. The changes to TS 5.6.3.b in the MNGP TS support that the uncertainties being removed from the MCPR SL are still included as part of the MCPR OL and will continue to appropriately inform plant operation.

Based on the review, the NRC staff therefore determined that the changes proposed by the licensee will retain an adequate level of conservatism in the MCPR SL in TS 2.1.1.3 while appropriately ensuring that plant- and cycle-specific uncertainties will be retained in the MCPR OL. The MCPR95/95 represents a lower limit on the value of the MCPR99.9%, which should always be higher since it accounts for numerous uncertainties that are not included in the MCPR95/95 (as discussed in Section 3.1 of TSTF-564).

3.5 Implementation of the Revised MCPR SL in the TSs The LAR proposed to change the value of the SL in TS 2.1.1.3 for both ATRIUMTM 10XM and GE 14 fuels to 1.05. The value reported in MNGP TS 2.1.1.3 was calculated using Equation 1 from TSTF-564 and reported at a precision of two digits past the decimal point with the hundreds digit rounded up.

MCPR99.9% will continue to be calculated using Framatomes SAFLIM-3D methodology.

Consistent with TSTF-564, the LAR also proposes to modify MNGP TS 5.6.3 to include the value of the MCPR99.9% to ensure that the cycle-specific MCPR99.9% value will continue to be determined for LCO 3.2.2 and reported in the COLR. The COLR, therefore, will continue to report the cycle-specific value of the MCPR OL contained in LCO 3.2.2, and MNGP TS 5.6.3.b will continue to reference appropriate NRC-approved methodologies for determination of the MCPR99.9% and the MCPR OL. Therefore, the NRC staff finds the proposed change to TS 5.6.3 to be acceptable.

The NRC staff assessed the licensees deviations from TSTF-564 for ATRIUMTM 10XM and determined that they are consistent with the process described in TSTF-564.

The MNGP reactor is currently fueled with ATRIUM'10XM fuel assemblies provided by Framatome, Inc., and GE14 fuel assemblies provided by Global Nuclear Fuel - Americas, LLC.

The GE14 fuel type is identified in Table 1 of TSTF-564 but the ATRIUM' 10 XM fuel type is not.

The new MCPR SL (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, (ADAMS Accession No. ML20308A27). This report provides calculation of the MCPR95/95 SL for the ATRIUM' 10XM fuel type using the ACE/ATRIUMTM 10XM CPR correlation database contained in ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, (ADAMS Accession No. ML14183A734), 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 percent of the fuel rods is not susceptible to boiling transition during normal operation and AOOs is referred to as MCPR99.9%. The ATRIUM' 10 XM is identified in the LAR as the fuel type the SLis based on since it will be the limiting fuel type in the MNGP core. Specification 5.6.3 is also revised to require the MCPR99.9%

to be included in the cycle-specific COLR. As discussed in Section 3.2 of the NRC SE for TSTF-564, the 0.03 MCPR SL adder corresponds to a penalty applied to the MCPR99.9% SL when operating in the EFW operating domain and only when the ratio of core power to core flow is greater than or equal to 42 MWt/Mlb/hr.

The MCPR95/95 SL is only fuel type dependent and not plant and/or cycle dependent. Therefore, 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 NRC staff, therefore, finds the proposed change to the SL in TS 2.1.1.3 acceptable. The licensee derived the SL consistent with the process described in Traveler TSTF-564.

The NRC staff notes that MNGP TS have a different numbering than STS 2.1.1.2; specifically, MNGP TS 2.1.1.3 aligns with STS 2.1.1.2. The NRC staff confirmed that the licensee made appropriate conforming changes in its proposal to adopt this TSTF. The NRC staff also confirmed that changes to items 21 and 23 of TS 5.6.3.b are administrative and acceptable.

As addressed earlier, the MNGP was not licensed to GDC 10, but instead was licensed to the applicable AEC preliminary general design criteria. The NRC staff determined that this difference does not affect the applicability of TSTF-564 for the proposed amendments to the MNGP TSs.

The NRC staff reviewed the proposed TS changes and found that the licensee appropriately implemented the revised MCPR SL, as discussed in this SE.

3.6 NRC Staff Conclusion

The NRC staff reviewed the proposed TS changes and determined that the proposed SL associated with TS 2.1.1.2 was calculated in a manner consistent with the process described in TSTF-564 and was, therefore, acceptably modified to suit the revised definition of the MCPR SL. Under the new definition, the MCPR SL will continue to protect the fuel cladding against the uncontrolled release of radioactivity by preventing the onset of boiling transition, thereby, fulfilling the requirements of 10 CFR 50.36(c)(1) for SLs. The MCPR OL in LCO 3.2.2 remains unchanged and will continue to meet the requirements of 10 CFR 50.36(c)(2), and of MNGPs plant-specific design criterion similar to GDC 10 discussed in Section 2.6, by ensuring that no fuel damage results during normal operation and AOOs. The NRC staff determined that the changes to TS 5.6.5 are acceptable. Upon adoption of the revised MCPR SL, the COLR will be required to contain the MCPR99.9%, supporting the determination of the MCPR OL using current methodologies.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendment on August 9, 2021. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or change the surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (86 FR 7888; February 2, 2021). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

Principal Contributor: F. Forsaty, NRR M. Hamm, NRR Date of issuance: October 15, 2021

ML21223A280 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/SNSB/BC NRR/DSS/STSB/BC NAME RKuntz SRohrer SKrepel NJordan(A)

DATE 8/9/2021 8/12/2021 8/12/2021 8/19/2021 OFFICE OGC - NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME KGamin NSalgado (JWiebe for) RKuntz DATE 9/17/2021 10/15/2021 10/15/2021