ML18299A076
| ML18299A076 | |
| Person / Time | |
|---|---|
| Site: | Technical Specifications Task Force |
| Issue date: | 11/16/2018 |
| From: | Victor Cusumano Office of Nuclear Reactor Regulation |
| To: | Technical Specifications Task Force |
| Honcharik M | |
| Shared Package | |
| ML18299A048 | List: |
| References | |
| CAC MG0161, EPID L-2017-PMP-0007, TSTF-564 | |
| Download: ML18299A076 (10) | |
Text
General Directions: This Model SE provides the format and content to be used when preparing the plant-specific SE of an LAR to adopt traveler TSTF-564, Revision 2. The bolded bracketed information shows text that should be filled in for the specific amendment; individual licensees would furnish site-specific nomenclature or values for these bracketed items. The italicized wording provides guidance on what should be included in each section and should not be included in the SE.
FINAL MODEL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER TSTF-564, REVISION 2, SAFETY LIMIT MCPR, USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS
1.0 INTRODUCTION AND BACKGROUND
By application dated [enter date], (Agencywide Documents Access and Management System (ADAMS) Accession No. [MLXXXXXXXXX]), [as supplemented by letters dated [enter date(s))), [name of licensee] (the licensee) submitted a license amendment request (LAR) for
[name of facility (abbreviated name), applicable units].
The LAR proposed to revise the basis, calculational method, and the value of the technical specification (TS) safety limit (SL) 2.1.1.2, which protects against boiling transition on the fuel rods in the core. The current basis ensures that 99.9 percent of the fuel rods in the core are not susceptible to boiling transition. The revised basis 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. Technical Specification 5.6.3, Core Operating Limits Report [(COLR)], is also modified.
The proposed changes are based on Technical Specifications Task Force (TSTF) traveler TSTF-564, Revision 2, Safety Limit MCPR [Minimum Critical Power Ratio], 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 [enter date] (ADAMS Accession No. ML18299A069).
[The licensee has proposed several 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 [3.6].]
[The supplemental letter(s) dated [enter date(s)], provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on [enter date] (cite FR reference).]
Enclosure 2
1.1 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.
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:
Measured Critical Power ECPR =
Calculated Critical Power 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).
1.3 Background on Thermal-Hydraulic Safety Limits To protect against boiling transition, BWRs have implemented an SL on the CPR, known as the minimum critical power ratio (MCPR) SL. As discussed in NUREG-1433 and NUREG-1434 for
General Electric BWR designs,1 the current basis 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 fuel cladding SL for pressurized-water reactor (PWR) designs, described in the Standard Technical Specifications (STS) for Babcock & Wilcox, Westinghouse, and Combustion Engineering 2 plants in NUREG-1430, NUREG-1431, and NUREG-1432,3 respectively, correspond to a 95 percent probability at a 95 percent confidence level that departure from nucleate boiling will not occur. As a result of the overall approach taken in developing the PWR limits, they are only dependent on the fuel type(s) in the reactor and the corresponding departure from nucleate boiling ratio (DNBR) correlations. The limits are not cycle-dependent and are typically only updated when new fuel types are inserted in the reactor.
BWRs 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 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.
2.0 REGULATORY EVALUATION
2.1 Description of TS Sections 2.1.1 TS 2.1.1, Reactor Core SLs Safety limits ensure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).
[Name of facility] TS 2.1.1.2 currently requires that [with the reactor steam dome pressure greater than or equal to () 785 pounds per square inch gauge (psig) and core flow 10 percent rated core flow, MCPR shall be [1.07] for two recirculation loop operation or [1.08] for single recirculation loop operation.] The MCPR SL ensures that 99.9 percent of the fuel in the core is not susceptible to boiling transition.
1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric Plants BWR/4, NUREG-1433, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, April 2012 (ADAMS Accession Nos. ML12104A192 and ML12104A193).
U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric Plants BWR/6, NUREG-1434, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, April 2012 (ADAMS Accession Nos. ML12104A195 and ML12104A196).
2 Denotes applicability to Combustion Engineering plants with digital control systems only.
3 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Babcock and Wilcox Plants, NUREG-1430, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, April 2012 (ADAMS Accession Nos. ML12100A177 and ML12100A178).
U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, April 2012 (ADAMS Accession Nos. ML12100A222 and ML12100A228).
U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Combustion Engineering Plants, NUREG-1432, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, April 2012 (ADAMS Accession Nos. ML12102A165 and ML12102A169).
2.1.2 TS 5.6.3, Core Operating Limits Report [(COLR)]
[Name of facility] 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.
2.2 Proposed Changes to the TS The licensee proposed to revise the MCPR SL to make it cycle-independent, consistent with the method described in traveler TSTF-564, Revision 2.
The proposed changes to the [name of facility] TS revise the value of the MCPR SL in TS 2.1.1.2 to [proposed value of MCPR SL from LAR], with corresponding changes to the associated 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 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 operating limit (OL) in limiting condition of operation (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.
2.3 Applicable Regulatory Requirements and Guidance The regulation at 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. The applicant must 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), TSs will include items in the following categories: (1) Safety limits, limiting safety system settings, and limiting control settings. As required by 10 CFR 50.36(c)(1)(i)(A), safety limits 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 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 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.
[General Design Criterion 10 (GDC), 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.
Most plants have a plant-specific design criterion similar to GDC 10. The limit placed on the MCPR acts as a specified acceptable fuel design limit to prevent boiling transition, which has the potential to result in fuel rod cladding failure.]
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,4 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 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.
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%), 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., license amendment request) to support the subsequent fuel cycle.
[Name of licensee] proposed to change the basis for the MCPR SL for [name of facility] 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 revised basis for the MCPR SL aligns it with that of the DNBR SL used in PWRs, which, as previously noted in Section 2.3 of this SE, provides a 95 percent probability at a 95 percent confidence level that no fuel rods will experience departure from nucleate boiling.
4 U.S. Nuclear Regulatory Commission, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants:
LWR [Light-Water Reactor] Edition, NUREG-0800, Section 4.4, Thermal and Hydraulic Design, Revision 2, March 2007 (ADAMS Accession No. ML070550060).
The intent of the proposed basis for 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 fulfil the necessary functions of an SL without unintended consequences, and that the proposed changes have been adequately implemented in the [name of facility] TS.
3.2 Revised MCPR SL Definition As discussed in Section 1.2 of this SE, a critical power correlations ECPR distribution quantifies the uncertainty associated with the correlation. Traveler TSTF-564, Revision 2, 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:
MCPR = +
where i is the correlations mean ECPR, i is the standard deviation of the correlations ECPR distribution, and i is a statistical parameter chosen 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 formula is commonly used 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. Owen5 and was also reported by M. G. Natrella,6 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 Cuta7 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. The traveler TSTF-564, Revision 2, 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 (ADAMS Accession No. ML18149A320). 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).
The NRC staff finds 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 5
D. B. Owen, Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, Sandia Corporation, SCR-607, March 1963, ADAMS Accession No. ML14031A495.
6 M. G. Natrella, Experimental Statistics, National Bureau of Standards, National Bureau of Standards Handbook 91 August 1963.
7 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.
necessary. Therefore, the NRC staff concludes that the MCPR95/95 definition, as proposed, establishes an acceptable fuel design limit and is acceptable.
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 the traveler 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%. The NRC staff found this justification 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.
The NRC staff reviewed the information furnished by the TSTF and determined 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 MCPR99.9% will be appropriately applied as the SL. The NRC staff therefore found this process to be acceptable.
[The size, mean, and standard deviation of the ECPR database may need to be provided by a fuel vendor to determine the MCPR95/95 for a legacy fuel type. The value of depends on the number of samples (Ni) in the critical power database. If the number of data points in the database is not supplied by the vendor, the TSTF response to a request for additional information stated that a value of = 1.8 would be imposed on the MCPR95/95 determination, on the basis that any database used to develop a critical power correlation will need at least 500 points to be acceptable.8 The limiting value from either the new or legacy fuel would then be applied as the SL. The NRC staff finds that there are potential circumstances where the number of data points used in determining the correlations uncertainty may not correspond to a value of 1.8; for example, future correlations may need fewer data points, or the subset of data used to determine a correlations uncertainty may be smaller than the full correlation database. Therefore, the NRC staff determined that a value of 1.8 for legacy fuel types where the number of data points N is not provided may not be acceptable, and the used in determining the MCPR95/95 must be justified to be appropriate or conservative for the fuel type and correlation in question. This determination does not affect the overall acceptability of the process for determining the MCPR95/95 for a mix of fuel types as discussed above.
The NRC staff also notes that, as stated in Section 1.0 of the traveler SE, this STS change is only available to licensees through the CLIIP when using the fuel bundle types 8
The NRC staff notes that a value of 1.8 corresponds to N = 300 data points, as provided in Table T-11b of NUREG-1475, Applying Statistics, Revision 1. This is more conservative than the for N = 500 data points, which would be 1.763.
specified in Table 1 of the traveler. Therefore, the use of legacy fuels, for which this determination would be relevant, is outside the scope of a CLIIP application.]
3.4 Relationship between MCPR Safety and Operating Limits As discussed in the traveler TSTF-564, Revision 2, 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 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.
Traveler TSTF-564, Revision 2, proposed that 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.
Traveler TSTF-564, Revision 2, also changed 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. {NOTE:
Verify that the licensee calculates MCPR SL and MCPR OL using the methodologies in the TS 5.6.3.b COLR reference list.} 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 therefore determined that the changes proposed by the licensee will retain an adequate level of conservatism in the MCPR SL in TS 2.1.1.2 while appropriately ensuring that plant- and cycle-specific uncertainties will be retained in the MCPR OL. The NRC staff notes that the MCPR95/95 represents a hard floor 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 traveler TSTF-564, Revision 2).
3.5 Implementation of the Revised MCPR SL in the TSs
{NOTE: If the licensee is in the midst of a fuel transition, all types of fresh and once-burnt fuel should be evaluated to determine which provides the limiting MCPR95/95, in accordance with the process discussed in traveler TSTF-564, Revision 2.}
{NOTE: If a fuel type not included in Table 1 of traveler TSTF-564, Revision 2, is loaded as fresh or once-burnt fuel, the value of the MCPR95/95 reported for that fuel type must be calculated using the mean and standard deviation from a critical power correlation found to be acceptable by the NRC staff. This should be evaluated by the NRC staff in this section of the SE.}
{NOTE: The following text is only applicable if the licensee has a core loaded with the fuel(s) referenced in Table 1 of traveler TSTF-564, Revision 2.}
The licensee has proposed to change the value of the SL in TS 2.1.1.2 to [value], consistent with the value from Table 1 of the TSTF-564, Revision 2, for the fuel type(s) in use at [name of
facility] (i.e., [name of fuel from Table 1 of traveler TSTF-564, Revision 2 and from licensee application]). The licensee has appropriately evaluated the fresh and once-burnt fuels in use at [name of facility] and the NRC staff has 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 value reported in [name of facility] TS 2.1.1.2 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 hundreds digit rounded up.
[Name of licensee] also modified [name of facility]s TS 5.6.3 to include the value of the MCPR99.9% in order to continue to be reported in the COLR. The COLR continues to report the cycle-specific value of the MCPR OL contained in LCO 3.2.2 and [name of facility] TS 5.6.3.b will continue to reference appropriate NRC-approved methodologies for determination of the MCPR99.9% and the MCPR OL.
The NRC staff reviewed the licensees proposed TS changes and found that the licensee appropriately implemented the revised MCPR SL, as discussed in this SE.
3.6 NRC Staff Conclusion
{NOTE: The project manager or reviewer should check the facilitys current licensing basis to determine if GDC 10 is applicable or if an equivalent plant-specific design criterion is used. If the facility licensing basis uses a plant-specific design criterion in lieu of GDC 10, the reference to GDC 10 below should be replaced with a reference to the appropriate design criterion from the facilitys licensing basis.}
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 or the equivalent plant-specific design criterion] by ensuring that no fuel damage results during normal operation and anticipated operational occurrences. The NRC staff determined that the changes to TS 5.6.3 proposed in the traveler 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 [Name of State] State official was notified of the proposed issuance of the amendment on [enter date]. The State official had [no]
comments. [If comments were provided, they should be addressed here].
5.0 ENVIRONMENTAL CONSIDERATION
{NOTE: This section is to be prepared by the PM. As needed, the PM should coordinate with NRRs Environmental Review and NEPA Branch (MENB) to determine the need for an EA.
Specific guidance on preparing EAs and considering environmental issues is contained in NRR
Office Instruction LIC-203, Procedural Guidance for Preparing Categorical Exclusions, Environmental Assessments, and Considering Environmental Issues.}
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 [DATE (XX FR XXX)], 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
{NOTE: This section is to be prepared by the PM.}
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.
{NOTE: These are the principal contributors for the model SE of the traveler. Replace these names with those who prepared the plant-specific SE.}
Principal Contributors: Reed A. Anzalone, NRR/DSS Caroline E. Tilton, NRR/DSS Date: November 16, 2018