ML18208A152

Draft Model Safety Evaluation of TSTF-564, Revision 1, Safety Limit MCPR Using the Consolidated Line Item Improvement Process

ML18208A152
Person / Time
Site:	Technical Specifications Task Force
Issue date:	10/03/2018
From:	NRC/NRR/DSS
To:
Honcharik M, NRR/DSS, 301-415-1774
Shared Package
ML18207A380	List: ML18208A146 ML18208A152
References
CAC MG0161, EPIC L-2017-PMP-0007, TSTF-564, Rev 1
Download: ML18208A152 (10)

Text

1 General Directions: This Model SE provides the format and content to be used when preparing 2 the plant-specific SE of an LAR to adopt traveler TSTF-564, Revision 1. The bolded bracketed 3 information shows text that should be filled in for the specific amendment; individual licensees 4 would furnish site-specific nomenclature or values for these bracketed items. The italicized 5 wording provides guidance on what should be included in each section and should not be 6 included in the SE.

7 8 DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 9

10 RELATED TO TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER 11 12 TSTF-564, REVISION 1, SAFETY LIMIT MCPR, 13 14 USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS 15 16 17 18

1.0 INTRODUCTION AND BACKGROUND

19 20 By application dated [enter date], (Agencywide Documents Access and Management System 21 (ADAMS) Accession No. [MLXXXXXXXXX]), [as supplemented by letters dated [enter 22 date(s))), [name of licensee] (the licensee) submitted a license amendment request (LAR) for 23 [name of facility (abbreviated name), applicable units].

24 25 The LAR proposed to revise the basis, calculational method, and the value of the technical 26 specification (TS) safety limit (SL) 2.1.1.2, which protects against boiling transition on the fuel 27 rods in the core. The current basis ensures that 99.9 percent of the fuel rods in the core are not 28 susceptible to boiling transition. The revised basis will ensure that there is a 95 percent 29 probability at a 95 percent confidence level that no fuel rods will be susceptible to boiling 30 transition using an SL based on critical power ratio (CPR) data statistics. Technical 31 Specification 5.6.3, Core Operating Limits Report [(COLR)], is also modified.

32 33 The proposed changes are based on Technical Specifications Task Force (TSTF) traveler 34 TSTF-564, Revision 1, Safety Limit MCPR [Minimum Critical Power Ratio], dated May 29, 35 2018 (ADAMS Accession No. ML18149A320). The U.S. Nuclear Regulatory Commission (NRC 36 or the Commission) issued a final safety evaluation (SE) approving traveler TSTF-564, 37 Revision 1, on [enter date] (ADAMS Accession No. [MLXXXXXXXXX]).

38 39 [The licensee has proposed several variations from the TS changes described in traveler 40 TSTF-564, Revision 1. The variations are described in Section [2.2] of this SE and 41 evaluated in Section [3.6].]

42 43 [The supplemental letter(s) dated [enter date(s)], provided additional information that 44 clarified the application, did not expand the scope of the application as originally 45 noticed, and did not change the NRC staffs original proposed no significant hazards 46 consideration determination as published in the Federal Register on [enter date] (cite FR 47 reference).]

48 Enclosure 2

1 1.1 Background on Boiling Transition 2

3 During steady-state operation in a boiling-water reactor (BWR), most of the coolant in the core 4 is in a flow regime known as annular flow. In this flow regime, a thin liquid film is pushed up the 5 surface of the fuel rod cladding by the bulk coolant flow, which is mostly water vapor with some 6 liquid water droplets. This provides effective heat removal from the cladding surface; however, 7 under certain conditions, the annular film may dissipate, which reduces the heat transfer and 8 results in an increase in fuel cladding surface temperature. This phenomenon is known as 9 boiling transition or dryout. The elevated surface temperatures resulting from dryout may cause 10 fuel cladding damage or failure.

11 12 1.2 Background on Critical Power Correlations 13 14 For a given set of reactor operating conditions (pressure, flow, etc.), dryout will occur on a fuel 15 assembly at a certain power, known as the critical power. Because the phenomena associated 16 with boiling transition are complex and difficult to model purely mechanistically, 17 thermal-hydraulic test campaigns are undertaken using electrically heated prototypical fuel 18 bundles to establish a comprehensive database of critical power measurements for each BWR 19 fuel product. These data are then used to develop a critical power correlation that can be used 20 to predict the critical power for assemblies in operating reactors. This prediction is usually 21 expressed as the ratio of the actual assembly power to the critical power predicted using the 22 correlation, known as the CPR.

23 24 One measure of the correlations predictive capability is based on its validation relative to the 25 test data. For each point j in a correlations test database, the experimental critical power ratio 26 (ECPR) is defined as the ratio of the measured critical power to the calculated critical power, or:

27 Measured Critical Power 28 ECPR =

Calculated Critical Power 29 30 For ECPR values less than or equal to 1, the calculated critical power is greater than the 31 measured critical power and the prediction is considered to be non-conservative. Because the 32 measured critical power includes random variations due to various uncertainties, evaluating the 33 ECPR for all of the points in the dataset (or, ideally, a subset of points that were not used in the 34 correlations development) results in a probability distribution. This ECPR distribution allows the 35 predictive uncertainty of the correlation to be determined. This uncertainty can then be used to 36 establish a limit above which there can be assumed that boiling transition will not occur (with a 37 certain probability and confidence level).

38 39 1.3 Background on Thermal-Hydraulic Safety Limits 40 41 To protect against boiling transition, BWRs have implemented an SL on the CPR, known as the 42 minimum critical power ratio (MCPR) SL. As discussed in NUREG-1433 and NUREG-1434 for

1 General Electric BWR designs,1 the current basis of the MCPR SL is to prevent 99.9 percent of 2 the fuel in the core from being susceptible to boiling transition. This limit is typically developed 3 by considering various cycle-specific power distributions and uncertainties, and is highly 4 dependent on the cycle-specific radial power distribution in the core. As such, the limit may 5 need to be updated as frequently as every cycle.

6 7 The fuel cladding SL for pressurized-water reactor (PWR) designs, described in the Standard 8 Technical Specifications (STS) for Babcock & Wilcox, Westinghouse, and Combustion 9 Engineering 2 plants in NUREG-1430, NUREG-1431, and NUREG-1432,3 respectively, 10 correspond to a 95 percent probability at a 95 percent confidence level that departure from 11 nucleate boiling will not occur. As a result of the overall approach taken in developing the PWR 12 limits, they are only dependent on the fuel type(s) in the reactor and the corresponding 13 departure from nucleate boiling ratio (DNBR) correlations. The limits are not cycle-dependent 14 and are typically only updated when new fuel types are inserted in the reactor.

15 16 BWRs also have a limiting condition for operation (LCO) that governs MCPR, known as the 17 MCPR operating limit (OL). The OL on MCPR is an LCO which must be met to ensure that 18 anticipated operational occurrences do not result in fuel damage. The current MCPR OL is 19 calculated by combining the largest change in CPR from all analyzed transients, also known as 20 the CPR, with the MCPR SL.

21 22

2.0 REGULATORY EVALUATION

23 24 2.1 Description of TS Sections 25 26 2.1.1 TS 2.1.1, Reactor Core SLs 27 28 Safety limits ensure that specified acceptable fuel design limits are not exceeded during steady 29 state operation, normal operational transients, and anticipated operational occurrences (AOOs).

30 31 [Name of facility] TS 2.1.1.2 currently requires that with the reactor steam dome pressure 32 greater than or equal to () 785 pounds per square inch gauge (psig) and core flow 10 percent 33 rated core flow, MCPR shall be [1.07] for two recirculation loop operation or [1.08] for single 34 recirculation loop operation. The MCPR SL ensures that 99.9 percent of the fuel in the core is 35 not susceptible to boiling transition.

36 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).

1 2.1.2 TS 5.6.3, Core Operating Limits Report [(COLR)]

2 3 [Name of facility] TS 5.6.3 requires core operating limits to be established prior to each reload 4 cycle, or prior to any remaining portion of a reload cycle. These limits are required to be 5 documented in the COLR.

6 7 2.2 Proposed Changes to the TS 8

9 The licensee proposed to revise the MCPR SL to make it cycle-independent, consistent with the 10 method described in traveler TSTF-564, Revision 1.

11 12 The proposed changes to the [name of facility] TS revise the value of the MCPR SL in 13 TS 2.1.1.2 to [proposed value of MCPR SL from LAR], with corresponding changes to the 14 associated bases. The change to TS 2.1.1.2 replaces the existing separate SLs for single- and 15 two-recirculation loop operation with a single limit since the revised SL is no longer dependent 16 on the number of recirculation loops in operation.

17 18 The MCPR99.9% (i.e., the current MCPR SL) is an input to the MCPR operating limit (OL) in 19 limiting condition of operation (LCO) 3.2.2, Minimum Critical Power Ratio (MCPR). While the 20 definition and method of calculation of both the MCPR99.9% and the LCO 3.2.2 MCPR OL 21 remains unchanged, the proposed TS changes include revisions to TS 5.6.3, to require the 22 MCPR99.9% value used in calculating the LCO 3.2.2 MCPR OL to be included in the 23 cycle-specific COLR.

24 25 2.3 Applicable Regulatory Requirements and Guidance 26 27 The regulation at Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36(a)(1),

28 requires an applicant for an operating license to include in the application proposed TSs in 29 accordance with the requirements of 10 CFR 50.36. The applicant must include in the 30 application, a summary statement of the bases or reasons for such specifications, other than 31 those covering administrative controls. However, per 10 CFR 50.36(a)(1), these TS bases 32 shall not become part of the technical specifications.

33 34 As required by 10 CFR 50.36(c), TSs will include items in the following categories: (1) Safety 35 limits, limiting safety system settings, and limiting control settings. As required by 10 CFR 36 50.36(c)(1)(i)(A), safety limits for nuclear reactors are limits upon important process variables 37 that are found to be necessary to reasonably protect the integrity of certain of the physical 38 barriers that guard against the uncontrolled release of radioactivity. If any safety limit is 39 exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the 40 matter, and record the results of the review, including the cause of the condition and the basis 41 for corrective action taken to preclude recurrence. Operation must not be resumed until 42 authorized by the Commission.

43 44 As required by 10 CFR 50.36(c)(2)(i), the TSs will include LCOs, which are the lowest functional 45 capability or performance levels of equipment required for safe operation of the facility. When 46 an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any 47 remedial action permitted by the TSs until the condition can be met.

1 2 [General Design Criterion 10 (GDC), Reactor design, of 10 CFR Part 50 Appendix A, 3 General Design Criteria of Nuclear Power Plants, states:

4 5 The reactor core and associated coolant control and protection systems 6 shall be designed with appropriate margin to assure that specified 7 acceptable fuel design limits are not exceeded during any condition of 8 normal operation, including the effects of anticipated operational 9 occurrences.

10 11 Most plants have a plant-specific design criterion similar to GDC 10. The limit placed on 12 the MCPR acts as a specified acceptable fuel design limit to prevent boiling transition, 13 which has the potential to result in fuel rod cladding failure.]

14 15 The NRC staffs guidance contained in Revision 2 of NUREG-0800, Standard Review Plan for 16 the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP),

17 Section 4.4, Thermal and Hydraulic Design,4 provides the following two examples of 18 acceptable approaches to meeting the SRP acceptance criteria for establishing fuel design 19 limits (as stated in SRP Acceptance Criterion 1):

20 21 A. For departure from nucleate boiling ratio (DNBR), CHFR [critical heat flux ratio]

22 or CPR correlations, there should be a 95-percent probability at the 95-percent 23 confidence level that the hot rod in the core does not experience a DNB or boiling 24 transition condition during normal operation or AOOs.

25 26 B. The limiting (minimum) value of DNBR, CHFR, or CPR correlations is to be 27 established such that at least 99.9 percent of the fuel rods in the core will not 28 experience a DNB or boiling transition during normal operation or AOOs.

29 30

3.0 TECHNICAL EVALUATION

31 32 3.1 Basis for Proposed Change 33 34 As discussed in Section 1.3 of this SE, the current MCPR SL (i.e., the MCPR99.9%), is affected by 35 the plants cycle-specific core design, especially including the core power distribution, fuel 36 type(s) in the reactor, and the power-to-flow operating domain for the plant. As such, it is 37 frequently necessary to change the MCPR SL to accommodate new core designs. Changes to 38 the MCPR SL are usually determined late in the design process and necessitate an accelerated 39 NRC review (i.e., license amendment request) to support the subsequent fuel cycle.

40 41 [Name of licensee] proposed to change the basis for the MCPR SL for [name of facility] so 42 that it is no longer cycle-dependent, reducing the frequency of revisions and eliminating the 43 need for NRCs review on an accelerated schedule. The proposed revised basis for the MCPR 44 SL aligns it with that of the DNBR SL used in PWRs, which, as previously noted in Section 2.3 45 of this SE, provides a 95 percent probability at a 95 percent confidence level that no fuel rods 46 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).

1 2 The intent of the proposed basis for the revised MCPR SL is acceptable to the NRC staff based 3 on the discussion in SRP Section 4.4, SRP Acceptance Criterion 1. The remainder of this SE is 4 devoted to ensuring that the methodology for determining the revised MCPR SL provides the 5 intended result, that the revised MCPR SL can be adequately determined in the core using 6 various types of fuel, that the proposed SL continues to fulfil the necessary functions of an SL 7 without unintended consequences, and that the proposed changes have been adequately 8 implemented in the [name of facility] TS.

9 10 3.2 Revised MCPR SL Definition 11 12 As discussed in Section 1.2 of this SE, a critical power correlations ECPR distribution quantifies 13 the uncertainty associated with the correlation. Traveler TSTF-564, Revision 1, provides a 14 definition for a limit that bounds 95 percent of a correlations ECPR distribution at a 95 percent 15 confidence level, according to the following formula:

16 17 MCPR = +

18 19 where i is the correlations mean ECPR, i is the standard deviation of the correlations ECPR 20 distribution, and i is a statistical parameter chosen to provide 95% probability at 95%

21 confidence (95/95) for the one-sided upper tolerance limit that depends on the number of 22 samples (Ni) in the critical power database. This formula is commonly used to determine a 23 95/95 one-sided upper tolerance limit for a normal distribution, which is appropriate for the 24 situation under consideration. The factor is generally attributed to D. B. Owen5 and was also 25 reported by M. G. Natrella,6 as referenced in traveler TSTF-564, Revision 1. Example values of 26 are provided in Table 2 of traveler TSTF-564, Revision 1. Table 1 of the traveler includes 27 some reference values of the MCPR95/95.

28 29 As discussed by Piepel and Cuta7 for DNBR correlations, the acceptability of this approach is 30 predicated on a variety of assumptions, including the assumptions that the correlation data 31 comes from a common population and that the correlations population is distributed normally.

32 These assumptions are typically addressed generically when a critical power or critical heat flux 33 correlation is reviewed by the NRC staff, who may apply penalties to the correlation to account 34 for any issues identified. The traveler TSTF-564, Revision 1, states that such penalties applied 35 during the NRCs review of the critical power correlation would be imposed on the mean or 36 standard deviation used in the calculating the MCPR95/95 (ADAMS Accession 37 No. ML18149A320). These penalties would also continue to be imposed in the determination of 38 the MCPR99.9%, along with any other penalties associated with the process of (or other inputs 39 used in) determining the MCPR99.9% (e.g., penalties applied to the MCPR99.9% SL for operation in 40 the Maximum Extended Load Limit Line Analysis Plus (MELLLA+) operating domain).

41 42 The NRC staff finds the definition of the MCPR95/95 will appropriately establish a 95/95 upper 43 tolerance limit on the critical power correlation and that any issues in the underlying correlation 44 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.

1 necessary. Therefore, the NRC staff concludes that the MCPR95/95 definition, as proposed, 2 establishes an acceptable fuel design limit and is acceptable.

3 4 3.3 Determination of Revised MCPR SL for Mixed Cores 5

6 Traveler TSTF-564, Revision 1, proposed that a core containing a variety of fuel types would 7 evaluate the MCPR95/95 for all of the fresh and once-burnt fuel in the core and apply the most 8 limiting (i.e., the largest) value of MCPR95/95 for each of the applicable fuel types as the MCPR 9 SL. As stated in Section 3.1 of traveler TSTF-564, Revision 1, this is because bundles that are 10 twice-burnt or more at the beginning of the cycle have significant MCPR margin relative to the 11 fresh and once-burnt fuel. The justification is that the MCPR for twice-burnt and greater fuel is 12 far enough from the MCPR for the limiting bundle that its probability of boiling transition is very 13 small compared to the limiting bundle and it can be neglected in determining the SL. Results of 14 a study provided in the traveler indicate that this is the case even for fuel operated on short 15 (12-month) reload cycles. As discussed in the traveler, twice-burnt or greater fuel bundles are 16 included in the cycle-specific evaluation of the MCPR99.9% and the MCPR OL. If a twice-burnt or 17 greater fuel bundle is found to be limiting, it would be governed by the MCPR OL, which will 18 always be more restrictive than both the MCPR95/95 and the MCPR99.9%. The NRC staff found 19 this justification to be appropriate and determined that it is acceptable to determine the 20 MCPR95/95 SL for the core based on the most limiting value of the MCPR95/95 for the fresh and 21 once-burnt fuel in the core.

22 23 The NRC staff reviewed the information furnished by the TSTF and determined that the process 24 for establishing the revised MCPR SL for mixed cores ensures that the limiting fuel types in the 25 core will be evaluated and the limiting MCPR99.9% will be appropriately applied as the SL. The 26 NRC staff therefore found this process to be acceptable.

27 28 [The size, mean, and standard deviation of the ECPR database may need to be provided 29 by a fuel vendor to determine the MCPR95/95 for a legacy fuel type. The value of 30 depends on the number of samples (Ni) in the critical power database. If the number of 31 data points in the database is not supplied by the vendor, the TSTF response to a 32 request for additional information stated that a value of = 1.8 would be imposed on the 33 MCPR95/95 determination, on the basis that any database used to develop a critical power 34 correlation will need at least 500 points to be acceptable.8 The limiting value from either 35 the new or legacy fuel would then be applied as the SL. The NRC staff finds that there 36 are potential circumstances where the number of data points used in determining the 37 correlations uncertainty may not correspond to a value of 1.8; for example, future 38 correlations may need fewer data points, or the subset of data used to determine a 39 correlations uncertainty may be smaller than the full correlation database. Therefore, 40 the NRC staff determined that a value of 1.8 for legacy fuel types where the number of 41 data points N is not provided may not be acceptable, and the used in determining the 42 MCPR95/95 must be justified to be appropriate or conservative for the fuel type and 43 correlation in question. This determination does not affect the overall acceptability of 44 the process for determining the MCPR95/95 for a mix of fuel types as discussed above.

45 The NRC staff also notes that, as stated in Section 1.0 of the traveler SE, this STS change 46 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.

1 specified in Table 1 of the traveler. Therefore, the use of legacy fuels, for which this 2 determination would be relevant, is outside the scope of a CLIIP application.]

3 4 3.4 Relationship between MCPR Safety and Operating Limits 5

6 As discussed in the traveler TSTF-564, Revision 1, the MCPR99.9% is expected to always be 7 greater than the MCPR95/95 for two reasons. First, because the MCPR99.9% includes 8 uncertainties not factored into the MCPR95/95, and second, because the 99.9 percent probability 9 basis for determining the MCPR99.9% is more conservative than the 95 percent probability at a 10 95 percent confidence level used in determining the MCPR95/95. The level of conservatism in 11 the MCPR95/95 SL is appropriate because the lead fuel rod in the core (i.e., the limiting fuel rod 12 with respect to MCPR) is used to evaluate whether any fuel rods in the core are susceptible to 13 boiling transition, which is also discussed in the traveler). This is consistent with evaluations 14 performed for PWRs using a 95/95 upper tolerance limit on the correlation uncertainty as an SL.

15 16 Traveler TSTF-564, Revision 1, proposed that the MCPR OL defined in LCO 3.2.2 would 17 continue to be evaluated using the MCPR99.9% as an input. The MCPR99.9% will continue to be 18 evaluated in the same way as it is currently, using the whole core.

19 20 Traveler TSTF-564, Revision 1, also changed TS 5.6.3 to require the cycle-specific value of the 21 MCPR99.9% to be included in the COLR. The methods supporting the inclusion of the MCPR99.9%

22 must also therefore, be included in the list of COLR references contained in TS 5.6.3.b. {NOTE:

23 Verify that the licensee calculates MCPR SL and MCPR OL using the methodologies in the TS 24 5.6.3.b COLR reference list.} The changes to TS 5.6.3.b help to ensure that the uncertainties 25 being removed from the MCPR SL are still included as part of the MCPR OL and will continue to 26 appropriately inform plant operation.

27 28 The NRC staff therefore determined that the changes proposed by the licensee will retain an 29 adequate level of conservatism in the MCPR SL in TS 2.1.1.2 while appropriately ensuring that 30 plant- and cycle-specific uncertainties will be retained in the MCPR OL. The NRC staff notes 31 that the MCPR95/95 represents a hard floor on the value of the MCPR99.9%, which should always 32 be higher since it accounts for numerous uncertainties that are not included in the MCPR95/95 (as 33 discussed in Section 3.1 of traveler TSTF-564, Revision 1).

34 35 3.5 Implementation of the Revised MCPR SL in the TSs 36 37 {NOTE: If the licensee is in the midst of a fuel transition, all types of fresh and once-burnt fuel 38 should be evaluated to determine which provides the limiting MCPR95/95, in accordance with the 39 process discussed in traveler TSTF-564, Revision 1.}

40 41 {NOTE: If a fuel type not included in Table 1 of traveler TSTF-564, Revision 1, is loaded as 42 fresh or once-burnt fuel, the value of the MCPR95/95 reported for that fuel type must be 43 calculated using the mean and standard deviation from a critical power correlation found to be 44 acceptable by the NRC staff. This should be evaluated by the NRC staff in this section of the 45 SE.}

46 47 {NOTE: The following text is only applicable if the licensee has a core loaded with the fuel(s) 48 referenced in Table 1 of traveler TSTF-564, Revision 1.}

49 50 The licensee has proposed to change the value of the SL in TS 2.1.1.2 to [value], consistent 51 with the value from Table 1 of the TSTF-564, Revision 1, for the fuel type(s) in use at [name of

1 facility] (i.e., [name of fuel from Table 1 of traveler TSTF-564, Revision 1 and from 2 licensee application]). The licensee has appropriately evaluated the fresh and once-burnt 3 fuels in use at [name of facility] and the NRC staff has determined that the limiting MCPR95/95 4 for these fuels was provided for inclusion in TS 2.1.1.2, consistent with the process described in 5 traveler TSTF-564, Revision 1.

6 7 The value reported in [name of facility] TS 2.1.1.2 was calculated using Equation 1 from 8 traveler TSTF-564, Revision 1, and reported at a precision of two digits past the decimal point 9 with the hundreds digit rounded up.

10 11 [Name of licensee] also modified [name of facility]s TS 5.6.3 to include the value of the 12 MCPR in order to continue to be reported in the COLR. The COLR continues to report the 13 cycle-specific value of the MCPR OL contained in LCO 3.2.2 and [name of facility] TS 5.6.3.b 14 will continue to reference appropriate NRC-approved methodologies for determination of the 15 MCPR99.9% and the MCPR OL.

16 17 The NRC staff reviewed the licensees proposed TS changes and found that the licensee 18 appropriately implemented the revised MCPR SL, as discussed in this SE.

19 20

3.6 NRC Staff Conclusion

21 22 {NOTE: The project manager or reviewer should check the facilitys current licensing basis to 23 determine if GDC 10 is applicable or if an equivalent plant-specific design criterion is used. If the 24 facility licensing basis uses a plant-specific design criterion in lieu of GDC 10, the reference to 25 GDC 10 below should be replaced with a reference to the appropriate design criterion from the 26 facilitys licensing basis.}

27 28 The NRC staff reviewed the licensees proposed TS changes and determined that the proposed 29 SL associated with TS 2.1.1.2 was calculated in a manner consistent with the process described 30 in traveler TSTF-564, Revision 1, and was therefore acceptably modified to suit the revised 31 definition of the MCPR SL. Under the new definition, the MCPR SL will continue to protect the 32 fuel cladding against the uncontrolled release of radioactivity by preventing the onset of boiling 33 transition, thereby fulfilling the requirements of 10 CFR 50.36(c)(1) for SLs. The MCPR OL in 34 LCO 3.2.2 remains unchanged and will continue to meet the requirements of 35 10 CFR 50.36(c)(2) [and GDC 10 or the equivalent plant-specific design criterion] by 36 ensuring that no fuel damage results during normal operation and anticipated operational 37 occurrences. The NRC staff determined that the changes to TS 5.6.3 proposed in the traveler 38 are acceptable; upon adoption of the revised MCPR SL, the COLR will be required to contain 39 the MCPR99.9%, supporting the determination of the MCPR OL using current methodologies.

40 41

4.0 STATE CONSULTATION

42 43 In accordance with the Commission's regulations, the [Name of State] State official was notified 44 of the proposed issuance of the amendment on [enter date]. The State official had [no]

45 comments. [If comments were provided, they should be addressed here].

46 47

5.0 ENVIRONMENTAL CONSIDERATION

48 49 {NOTE: This section is to be prepared by the PM. As needed, the PM should coordinate with 50 NRRs Environmental Review and NEPA Branch (MENB) to determine the need for an EA.

51 Specific guidance on preparing EAs and considering environmental issues is contained in NRR

1 Office Instruction LIC-203, Procedural Guidance for Preparing Categorical Exclusions, 2 Environmental Assessments, and Considering Environmental Issues.}

3 4 The amendment changes requirements with respect to the installation or use of facility 5 components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has 6 determined that the amendment involves no significant increase in the amounts, and no 7 significant change in the types, of any effluents that may be released offsite, and that there is no 8 significant increase in individual or cumulative occupational radiation exposure. The 9 Commission has previously issued a proposed finding that the amendment involves no 10 significant hazards consideration, which was published in the Federal Register on [DATE (XX 11 FR XXX)], and there has been no public comment on such finding. Accordingly, the 12 amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

13 Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment 14 need be prepared in connection with the issuance of the amendment.

15 16

6.0 CONCLUSION

17 18 The Commission has concluded, based on the considerations discussed above, that: (1) there 19 is reasonable assurance that the health and safety of the public will not be endangered by 20 operation in the proposed manner, (2) there is reasonable assurance that such activities will be 21 conducted in compliance with the Commissions regulations, and (3) the issuance of the 22 amendment will not be inimical to the common defense and security or to the health and safety 23 of the public.

24 25 Principal Contributors: R. Anzalone, NRR/DSS 26 C. Tilton, NRR/DSS