ML20302A026
| ML20302A026 | |
| Person / Time | |
|---|---|
| Site: | Columbia  (NPF-021) |
| Issue date: | 12/15/2020 |
| From: | Mahesh Chawla Plant Licensing Branch IV |
| To: | Sawatzke B Energy Northwest |
| Chawla M | |
| References | |
| EPID L-2020-LLA-0015 | |
| Download: ML20302A026 (17) | |
Text
December 15, 2020 Mr. Bradley J. Sawatzke Chief Executive Officer Energy Northwest Mail Drop 1023 76 North Power Plant Loop P.O. Box 968 Richland, WA 99352
SUBJECT:
COLUMBIA GENERATING STATION - ISSUANCE OF AMENDMENT NO. 262 TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-564, SAFETY LIMIT MCPR REVISION 2 (EPID L-2020-LLA-0015)
Dear Mr. Sawatzke:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 262 to Renewed Facility Operating License No. NPF-21 for Columbia Generating Station (Columbia). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated January 27, 2020.
The amendment revises the Columbia TSs to adopt Technical Specifications Task Force (TSTF) traveler TSTF-564, Safety Limit MCPR [Minimum Critical Power Ratio], Revision 2, dated October 24, 2018. The amendment revises the Columbia TS Safety Limit 2.1.1.2 and TS 5.6.3, Core Operating Limits Report (COLR).
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Mahesh L. Chawla, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosures:
- 1. Amendment No. 262 to NPF-21
- 2. Safety Evaluation cc: Listserv
ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERATING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 262 License No. NPF-21
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Energy Northwest (the licensee), dated January 27, 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.
- 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. NPF-21 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 262 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented after startup from the next refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-21 and the Technical Specifications Date of Issuance: December 15, 2020 Jennifer L.
Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2020.12.15 14:49:49 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 262 RENEWED FACILITY OPERATING LICENSE NO. NPF-21 COLUMBIA GENERATING STATION DOCKET NO. 50-397 Replace the following pages of Renewed Facility Operating License No. NPF-21 and the Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Renewed Facility Operating License REMOVE INSERT Technical Specification REMOVE INSERT 2.0-1 2.0-1 5.6-2 5.6-2
Renewed License No. NPF-21 Amendment No. 262 (2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 262 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- a. For Surveillance Requirements (SRs) not previously performed by existing SRs or other plant tests, the requirement will be considered met on the implementation date and the next required test will be at the interval specified in the Technical Specifications as revised in Amendment No. 149.
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SLs 2.0 Columbia Generating Station 2.0-1 Amendment No. 169,186 225 234, 242 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 686 psig or core flow < 10%
rated core flow:
THERMAL POWER shall be 25% RTP.
2.1.1.2 With the reactor steam dome pressure 686 psig and core flow 10%
rated core flow:
The MCPR shall be 1.07.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be 1325 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.
262
Reporting Requirements 5.6 Columbia Generating Station 5.6-2 Amendment No. 190 225 226 253 5.6 Reporting Requirements 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; 4.
Deleted; 5.
The Oscillation Power Range Monitor (OPRM) Instrumentation for Specification 3.3.1.1; and 6.
The Rod Block Monitor Instrumentation for Specification 3.3.2.1.
b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1.
XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company 2.
XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company 3.
EMF-85-74(P) Supplement 1(P)(A) and Supplement 2(P)(A),
"RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model,"
Siemens Power Corporation 4.
ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation 5.
XN-NF-80-19(P)(A) Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis,"
Exxon Nuclear Company 6.
XN-NF-80-19(P)(A) Volume 4, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company 262
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 262 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-21 ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397
1.0 INTRODUCTION
By application dated January 27, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20027D541, Energy Northwest (the licensee) submitted a license amendment request (LAR) for Columbia Generating Station (Columbia), which is a boiling-water reactor (BWR).
The LAR proposed to revise the value of the Technical Specification (TS) 2.1.1.2 reactor core safety limit (SL) minimum critical power ratio (MCPR). The MCPR protects against boiling transition on the fuel rods in the core and deletes reference to single-and two-loop operation.
The current MCPR value is dependent on the number of recirculation loops in operation and ensures that 99.9 percent of the fuel rods in the core are not susceptible to boiling transition, which is referred to as the MCPR99.9%. The revised MCPR would 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, which is referred to as the MCPR95/95. The MCPR95/95 is not dependent on the number of recirculation loops in operation. Additionally, TS 5.6.3, Core Operating Limits Report (COLR), would be modified to require that the COLR include the cycle-specific value for MCPR99.9%, which would still be used to calculate the MCPR operating limit (OL).
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). The U.S. Nuclear Regulatory Commission (NRC or the Commission) issued a final safety evaluation (SE) approving traveler TSTF-564, Revision 2, on November 16, 2018 (ADAMS Accession No. ML18299A069). The NRC staff issued a revised final model SE for traveler TSTF-564, Revision 2, on February 25, 2020 (ADAMS Accession No. ML19248C823).
The licensee has proposed two variations from the TS changes described in traveler TSTF-564, Revision 2. The variations are described in Section 2.2 of this SE and evaluated in Section 3.5 of this SE.
1.1 Background on Boiling Transition During steady-state operation in a BWR, most of the coolant in the core is in a flow regime known as annular flow. In this flow regime, a thin liquid film is pushed up the surface of the fuel rod cladding by the bulk coolant flow, which is mostly water vapor with some liquid water droplets. This provides effective heat removal from the cladding surface; however, under certain conditions, the annular film may dissipate, which reduces the heat transfer and results in an increase in fuel cladding surface temperature. This phenomenon is known as boiling transition or dryout. The elevated surface temperatures resulting from dryout may cause fuel cladding damage or failure.
1.2 Background on Critical Power Correlations For a given set of reactor operating conditions (pressure, flow, etc.), dryout will occur on a fuel assembly at a certain power, known as the critical power. Because the phenomena associated with boiling transition are complex and difficult to model purely mechanistically, thermal-hydraulic test campaigns are undertaken using electrically heated prototypical fuel bundles to establish a comprehensive database of critical power measurements for each BWR fuel product. These data are then used to develop a critical power correlation that can be used to predict the critical power for assemblies in operating reactors. This prediction is usually expressed as the ratio of the actual assembly power to the critical power predicted using the correlation, known as the CPR.
One measure of the correlations predictive capability is based on its validation relative to the test data. For each point j in a correlations test database, the experimental critical power ratio (ECPR) is defined as the ratio of the measured critical power to the calculated critical power, or:
ECPRMeasured Critical Power Calculated Critical Power For ECPR values less than or equal to 1, the calculated critical power is greater than or equal to () 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 it can be assumed that boiling transition will not occur (with a certain probability and confidence level).
1.3 Background on Thermal-Hydraulic Safety Limits To protect against boiling transition, BWRs have implemented an SL on the CPR, known as the MCPR SL. As discussed in Chapter 4, Reactor, of the Columbia Final Safety Analysis Report (FSAR) (ADAMS Accession No. ML19343C115) and the Standard Technical Specifications
(STSs) for General Electric BWR plant designs in NUREG-14341, the current calculation of the MCPR SL 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 Columbia also have a limiting condition for operation (LCO) that governs MCPR, known as the MCPR OL. The OL on MCPR is an LCO that 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. The MCPR OL is already a COLR parameter and as such, the methodology to calculate it is included in TS 5.6.3.b. The MCPR SL (i.e., the MCPR99.9%) is calculated using the same methodology as the MCPR OL.
2.0 REGULATORY EVALUATION
2.1 Description of TS Sections 2.1.1 TS 2.1.1, Reactor Core SLs The SLs ensure that specified acceptable fuel design limits are not exceeded during steady-state operation, normal operational transients, and AOOs.
Columbia TS 2.1.1.2 currently requires that the reactor steam dome pressure 686 pounds per square inch gauge and core flow 10 percent rated core flow, the MCPR shall be 1.10 for two recirculation loop operation or 1.13 for single recirculation loop operation. The MCPR SL (also referred to as MCPR99.9%) ensures that 99.9 percent of the fuel rods in the core are not susceptible to boiling transition.
2.1.2 TS 5.6.3, Core Operating Limits Report (COLR)
Columbia 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. This TS requires that these limits be documented in the COLR.
2.2 Proposed Changes to the TSs The licensee proposed to revise the MCPR SL in TS 2.1.1.2 to make it cycle-independent, consistent with the method described in traveler TSTF-564, Revision 2.
The proposed changes to the Columbia TSs revise the value of the MCPR SL in TS 2.1.1.2 to 1.07, with corresponding changes to the associated TS bases. The change to TS 2.1.1.2 replaces the existing separate SLs for single-and two-recirculation loop operation with a single limit since the revised SL (also referred to as the MCPR95/95 SL) is not dependent on the number of recirculation loops in operation.
1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric BWR/6 Plants, NUREG-1434, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, dated April 2012 (ADAMS Accession Nos. ML12104A195 and ML12104A196, respectively)
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 remains unchanged, the proposed TS changes include revisions to TS 5.6.3, to require the MCPR99.9% value used in calculating the LCO 3.2.2 MCPR OL to be included in the cycle-specific COLR.
In addition, the licensee proposed variations from the TS changes described in TSTF-564 or the applicable parts of the NRC staffs SE. The proposed variations are described below:
Columbia is a BWR/5 plant. Columbia TSs 2.1.1 and 5.6.3 are aligned with the BWR/6 STSs in NUREG-1434.
Columbias TSs contain a requirement that differs from the STSs on which TSTF-564 was based, such as reactor steam dome pressure in SL 2.1.1.2.
2.3 Applicable Regulatory Requirements and Guidance Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) requires an applicant for an operating license to include in the application proposed TSs in accordance with the requirements of 10 CFR 50.36, Technical specifications. The applicant must also include in the application, a summary statement of the bases or reasons for such specifications, other than those covering administrative controls.... However, per 10 CFR 50.36(a)(1), these TS bases shall not become part of the technical specifications.
As required by 10 CFR 50.36(c)(1), TSs will include safety limits, limiting safety system settings, and limiting control settings. The regulation, 10 CFR 50.36(c)(1)(i)(A), states, in part:
Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be resumed until authorized by the Commission.
As required by 10 CFR 50.36(c)(2)(i), the TSs will include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications 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.
The limits placed on the MCPR act as a specified acceptable fuel design limit to prevent boiling transition, which has the potential to result in fuel rod cladding failure, and are used to meet 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 [Light-Water Reactor]
Edition (SRP), Section 4.4, Thermal and Hydraulic Design, Revision 2, dated March 2007 (ADAMS Accession No. ML070550060), provides the following two examples of acceptable approaches to meet SRP Acceptance Criterion 1 for establishing fuel design limits:
A.
For departure from nucleate boiling ratio (DNBR), CHFR [critical heat flux ratio] or CPR correlations, there should be a 95-percent probability at the 95-percent confidence level that the hot rod in the core does not experience a DNB [departure from nucleate boiling] or boiling transition condition during normal operation or AOOs.
B.
The limiting (minimum) value of DNBR, CHFR, or CPR correlations is to be established such that at least 99.9 percent of the fuel rods in the core will not experience a DNB or boiling transition during normal operation or AOOs.
The NRC staffs guidance for the review of TSs is in Chapter 16.0, Revision 3, Technical Specifications, of the SRP, dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STSs for each of the LWR nuclear designs. Accordingly, the NRC staffs review considers whether the proposed changes are consistent with the applicable reference STSs (i.e., the current STSs), as modified by NRC-approved travelers. The STSs applicable to Columbia is NUREG-1434, Revision 4.0, Volume 1 and Volume 2.
3.0 TECHNICAL EVALUATION
3.1 Basis for Proposed Change As discussed in Section 1.3 of this SE, the current MCPR SL (i.e., the MCPR99.9% SL) is dependent on the cycle-specific core design, especially including the core power distribution, fuel type(s) in the reactor, and the power-to-flow operating domain for the plant. As such, it is frequently necessary to change the MCPR SL to accommodate new core designs. Changes to the MCPR SL are usually determined late in the design process and necessitate an accelerated NRC review (i.e., LAR) to support the subsequent fuel cycle.
The licensee proposed to change the calculation for determining the MCPR SL for Columbia 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 the DNBR SL used in PWRs, which ensures a 95 percent probability at a 95 percent confidence level that no fuel rods will experience DNB.
The NRC staff finds that calculating the revised MCPR SL based on the 95/95 criterion is acceptable because it meets SRP Section 4.4, Acceptance Criterion 1. The remainder of this SE evaluates whether the methodology for determining the revised MCPR SL provides the intended result and documents the review to ensure that the revised MCPR SL can be
adequately determined in the core using various types of fuel, that the proposed SL continues to fulfil the necessary functions of an SL without unintended consequences, and that the proposed changes have been adequately implemented in the Columbia TSs.
3.2 Revised MCPR SL Definition As discussed in Section 1.2 of this SE, the ECPR distribution quantifies the uncertainty associated with the correlation. Traveler TSTF-564, Revision 2, provides the following formula:
where i is the mean ECPR and i is the standard deviation of the 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. Owen2 and was also reported by M. G. Natrella,3 as referenced in traveler TSTF-564, Revision 2. Example values of are provided in Table 2 of traveler TSTF-564, Revision 2. Table 1 of the traveler includes some reference values of the MCPR95/95.
As discussed by Piepel and Cuta4 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), states that such penalties applied during the NRCs review of the critical power correlation would be imposed on the mean or standard deviation used in the calculating the MCPR95/95. 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 operating domain).
In the SE of traveler TSTF-564, Revision 2, the NRC staff found that the definition of the MCPR95/95 appropriately establishes 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 can be used to establish acceptable fuel design limits in Columbia TSs.
2 D. B. Owen, Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, Sandia Corporation, SCR-607, March 1963 (ADAMS Accession No. ML14031A495).
3 M. G. Natrella, Experimental Statistics, National Bureau of Standards, National Bureau of Standards Handbook 91 August 1963.
4 G. F. Piepel and J. M. Cuta, Statistical Concepts and Techniques for Developing, Evaluating, and Validating CHF Models and Corresponding Fuel Design Limits, SKI Technical Report, 93:46, 1993.
3.3 Determination of Revised MCPR SL for Mixed Cores Traveler TSTF-564, Revision 2, proposed that a core containing a variety of fuel types would evaluate the MCPR95/95 for all of the fresh and once-burnt fuel in the core and apply the most limiting (i.e., the largest) value of MCPR95/95 for each of the applicable fuel types as the MCPR SL. As stated in Section 3.1 of traveler TSTF-564, Revision 2, this is because bundles that are twice-burnt or more at the beginning of the cycle have significant MCPR margin relative to the fresh and once-burnt fuel. The justification is that the MCPR for twice-burnt and greater fuel is far enough from the MCPR for the limiting bundle that its probability of boiling transition is very small compared to the limiting bundle and it can be neglected in determining the SL. Results of a study provided in TSTF letter dated May 29, 2018, indicate that this is the case even for fuel operated on short (12-month) reload cycles. As discussed in the traveler, twice-burnt or greater fuel bundles are included in the cycle-specific evaluation of the MCPR99.9% and the MCPR OL.
If a twice-burnt or greater fuel bundle is found to be limiting, it would be governed by the MCPR OL, which will always be more restrictive than both the MCPR95/95 and the MCPR99.9%.
In the SE of traveler TSTF-564, Revision 2, the NRC staff found the justification for applying the most limiting (i.e., the largest) value of MCPR95/95 for each of the applicable fuel types as the MCPR SL to be appropriate and determined that it is acceptable to determine the MCPR95/95 SL for the core based on the most limiting value of the MCPR95/95 for the fresh and once-burnt fuel in the core. In the SE of traveler TSTF-564, Revision 2, the NRC staff also found that the process for establishing the revised MCPR SL for mixed cores ensures that the limiting fuel types in the core will be evaluated and the limiting MCPR95/95 will be appropriately applied as the SL. Therefore, the NRC staff finds it acceptable to determine the MCPR95/95 SL for the core based on the most limiting MCPR95/95 value for fresh and once-burnt fuel in the core for the Columbia TSs.
3.4 MCPR Safety and Operating Limits As discussed in the TSTF letter dated May 29, 2018, the MCPR99.9% SL is expected to always be greater than the MCPR95/95 SL 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 SL is appropriate because the lead fuel rod in the core (i.e., the limiting fuel rod with respect to MCPR) is used to evaluate whether any fuel rods in the core are susceptible to boiling transition, which is also discussed in the traveler. This is consistent with evaluations performed for PWRs using a 95/95 upper tolerance limit on the correlation uncertainty as an SL.
Consistent with traveler TSTF-564, Revision 2, the licensee is not proposing any change to LCO 3.2.2 and will continue to calculate the MCPR OL using the MCPR99.9%. The MCPR99.9%
will continue to be calculated in the same way as it is currently, using the whole core.
Consistent with traveler TSTF-564, Revision 2, the licensee proposed to revise TS 5.6.3 to require inclusion of the cycle-specific value of the MCPR99.9% in the COLR to ensure that the uncertainties being removed from the MCPR SL are still included as part of the MCPR OL. The methods used for determining MCPR99.9% are included in the list of COLR references contained in TS 5.6.3.b. The changes to TS 5.6.3.b help to ensure 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.
The NRC staff finds that the changes proposed by the licensee will retain an adequate level of conservatism in the MCPR SL in TS 2.1.1.2 and that plant-and cycle-specific uncertainties will be retained in the MCPR OL as specified in the COLR. The MCPR95/95 represents a lower limit on the value of the MCPR99.9%, because the MCPR99.9% should always be higher since it accounts for numerous uncertainties that are not included in the MCPR95/95 (as discussed in Section 3.1 of traveler TSTF-564, Revision 2).
3.5 Implementation of the Revised MCPR SL in the TSs The licensee proposed to change the value of the SL in TS 2.1.1.2 from 1.10 for two recirculation loop operation or 1.13 for single recirculation loop operation, to 1.07, consistent with the value from Table 1 of the TSTF-564, Revision 2, for the fuel type in use at Columbia (i.e., Global Nuclear Fuel 2). As noted in Section 3.3 above, the licensee appropriately evaluated the fresh and once-burnt fuels in use at Columbia and the NRC staff determined that the limiting MCPR95/95 for these fuels was provided for inclusion in TS 2.1.1.2, consistent with the process described in traveler TSTF-564, Revision 2.
The NRC staff finds that the proposed MCPR value of 1.07 in Columbia TS 2.1.1.2 is acceptable because it was calculated using Equation 1 from traveler TSTF-564, Revision 2, and reported at a precision of two digits past the decimal point with the hundredths digit rounded up.
Thus, the proposed TS change is acceptable.
The licensee also proposed that Columbia TS 5.6.3 requires the MCPR99.9% value used to calculate the LCO 3.2.2 MCPR limit be specified in the COLR. Thus Columbia TS 5.6.3.b will continue to reference appropriate NRC-approved methodologies for determination of the MCPR99.9% and the MCPR OL, which will ensure that cycle-specific parameters are determined such that applicable limits are met. Therefore, the NRC staff finds the proposed change acceptable.
The licensee proposed the following variations from the TS changes described in TSTF-564 or the applicable parts of the NRC staff's safety evaluation:
Columbia is a BWR/5 plant. Columbias TS 2.1.1 and 5.6.3 are aligned with BWR/6 STSs (NUREG-1434). This difference is administrative and does not affect the applicability of TSTF-564 to the Columbia TS.
Columbias TSs contain a requirement that differs from the STSs on which TSTF-564 was based, such as reactor steam dome pressure in SL 2.1.1.2. This difference does not affect the applicability of the TSTF-564 to the Columbia TS.
The NRC staff reviewed the licensees proposed TS changes and finds that the licensee appropriately implemented the revised MCPR SL, as discussed in this SE.
3.6 NRC Staff Conclusion
The NRC staff reviewed the licensees proposed TS changes and determined that the proposed SL associated with TS 2.1.1.2 was calculated in a manner consistent with the process described in traveler TSTF-564, Revision 2, and was therefore acceptably modified to suit the revised definition of the MCPR SL. Under the new definition, the MCPR SL will continue to protect the fuel cladding against the uncontrolled release of radioactivity by preventing the onset of boiling transition, thereby fulfilling the requirements of 10 CFR 50.36(c)(1) for SLs. The MCPR OL in
LCO 3.2.2 remains unchanged and will continue to meet the requirements of 10 CFR 50.36(c)(2) and GDC 10 by ensuring that no fuel damage results during normal operation and AOOs. The NRC staff determined that the proposed changes to Columbia TS 5.6.3 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 Washington State official was notified of the proposed issuance of the amendment on August 4, 2020. On August 13, 2020, the State official confirmed that the State of Washington had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. 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, which was published in the Federal Register on April 7, 2020 (85 FR 19510), and there has been no public comment on such finding. 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: M. Razzaque Date: December 15, 2020
- via e-mail
- via memorandum OFFICE NRR/DORL/LPL4/PM*
NRR/DORL/LPL4/LA*
NRR/DSS/STSB/BC*
NRR/DSS/SNSB/BC**
NAME MChawla PBlechman VCusumano SKrepel DATE 11/03/2020 11/03/2020 08/03/2020 07/09/2020 OFFICE OGC - NLO*
NRR/DORL/LPL4/BC*
NRR/DORL/LPL4/PM*
NAME CCarson JDixon-Herrity MChawla DATE 12/01/2020 12/15/2020 12/15/2020