TSTF-18-12, TSTF Comments on Draft Safety Evaluation for Traveler TSTF-564, Revision 1, Safety Limit MCPR

TSTF Comments on Draft Safety Evaluation for Traveler TSTF-564, Revision 1, Safety Limit MCPR

ML18297A361
Person / Time
Site:	Technical Specifications Task Force, 99902042
Issue date:	10/24/2018
From:	Gullott D, Joyce R, Miksa J, Sparkman W, Vaughan J Technical Specifications Task Force
To:	Document Control Desk
References
EPID L-2017-PMP-0007, TSTF-18-12
Download: ML18297A361 (79)

Text

TECHNICAL SPECIFICATIONS TASK FORCE TSTF A JOINT OWNERS GROUP ACTIVITY October 24, 2018 TSTF-18-12 PROJ0753 Attn: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

TSTF Comments on Draft Safety Evaluation for Traveler TSTF-564, Revision 1, "Safety Limit MCPR"

REFERENCE:

Letter Victor Cusumano (NRC) to the TSTF, "Draft Safety Evaluation of Technical Specifications Task Force Traveler TSTF-564, Revision 1, 'Safety Limit MCPR [Minimum Critical Power Ratio]'," dated October 4, 2018 (ADAMS Accession No. ML18207A380).

On May 29, 2018, the TSTF submitted traveler TSTF-564, Revision 1, "Safety Limit MCPR

[Minimum Critical Power Ratio]," to the Nuclear Regulatory Commission (NRC) for review (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18149A320). In the referenced letter, the NRC provided the draft Safety Evaluations for TSTF-564 for comment.

Attachment 1 contains a summary table providing the TSTF's comments on the draft Safety Evaluations. Attachment 2 contains a mark-up reflecting the TSTF's comments.

The TSTF identified additional minor changes to TSTF-564, Revision 1, following its submittal.

These changes reflect a modification of the rounding methodology incorporated in Revision 1.

The Safety Limit for Global Nuclear Fuel (GNF) GE14 fuel is revised from 1.05 to 1.06 and the referenced GNF methodology document is updated. These changes have been previously discussed with the NRC staff. Revision 2 of TSTF-564 is enclosed.

11921 Rockville Pike, Suite 100, Rockville, MD 20852 Phone: 301-984-4400, Fax: 301-984-7600 Administration by EXCEL Services Corporation

TSTF-18-12 October 24, 2018 Page 2 Should you have any questions, please do not hesitate to contact us.

James P. Miksa (PWROG/CE) Ryan M. Joyce (BWROG)

David M. Gullott (PWROG/W) Jordan L. Vaughan (PWROG/B&W)

Wesley Sparkman (APOG) TSTF Comments on the TSTF-564 Draft Safety Evaluations TSTF Markup of Draft Safety Evaluations Enclosure TSTF-564, Revision 2.

cc: Michelle Honcharik, Technical Specifications Branch, NRC Victor Cusumano, Technical Specifications Branch, NRC

Attachment 1 TSTF Comments on the TSTF-564 Draft Safety Evaluations Comments on the TSTF-564 Traveler Draft Safety Evaluation Page(s) Line(s)1 Comment 1 20 The term "technical specifications" should be revised to be plural to be consistent with 10 CFR 50.36 and the Standard Technical Specifications (STS).

1 34-35 For completeness, we recommend adding a reference to the Revision 0 submittal of TSTF-564.

1 36 Editorial suggestion. There was only one request for additional information.

Comments on the TSTF-564 Draft Model Safety Evaluation Page(s) Line(s)1 Comment 3 31-34 Recommend placing the description of the current TS 2.1.1.2 in brackets and in bold to accommodate various plant TS without invoking a variation from the SE.

4 14-16 Recommend placing the description of the current TS 2.1.1.2 in brackets and in bold to accommodate various plant TS without invoking a variation from the SE.

9 12 For clarity, the SE should state that the MCPR99.9% value will be included in the COLR instead of the MCPR value.

1 Line numbers correspond to the attached proposed revision, not to the documents provided by the NRC.

Page 3

Attachment 2 TSTF Markup of Draft Safety Evaluations

1 DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 2

3 RELATED TO TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER 4

5 TSTF-564, REVISION 1, SAFETY LIMIT MCPR 6

7 USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS 8

9 (EPID L-2017-PMP-0007) 10 11 12

1.0 INTRODUCTION AND BACKGROUND

13 14 By letter dated May 29, 2018 (Agencywide Documents Access and Management System 15 (ADAMS) Accession No. ML18149A320), the Technical Specifications Task Force (TSTF) 16 submitted Traveler TSTF-564, Revision 1, Safety Limit MCPR [Minimum Critical Power Ratio].

17 Traveler TSTF-564, Revision 1, proposed changes to the Standard Technical Specifications 18 (STS) for boiling-water reactor (BWR) designs.1 These changes will be incorporated into future 19 revisions of NUREG-1433 and NUREG-1434. Associated changes were also made to the 20 technical specifications (TS) Bases.

21 22 The proposed changes revise the basis, calculational method, and the value of the TS safety 23 limit (SL) 2.1.1.2, which protects against boiling transition on the fuel rods in the core. The 24 current basis ensures that 99.9 percent of the fuel rods in the core are not susceptible to boiling 25 transition. The revised basis will ensure that there is a 95 percent probability at a 95 percent 26 confidence level that no fuel rods will be susceptible to boiling transition using an SL based on 27 critical power ratio (CPR) data statistics. Technical Specification 5.6.3, Core Operating Limits 28 Report [(COLR)], is also modified.

29 30 This STS change will be made available to licensees through the consolidated line item 31 improvement process (CLIIP) and is applicable to licensees utilizing those vendor-specific and 32 fuel bundle types which are specified in Table 1 of the traveler.

33 34 By letter dated August 28, 2017 (ADAMS Accession No. ML17240A265), the TSTF submitted 35 Traveler TSTF-564, Revision 0. The U.S. Nuclear Regulatory Commission (NRC) staff 36 transmitted a requests for additional information (RAIs) to the TSTF by letter dated April 12, 37 2018 (ADAMS Accession No. ML18095A229). Responses to these RAIs were transmitted from 38 the TSTF by letter dated May 29, 2018 (ADAMS Accession No. ML18149A320).

39 40 1.1 Background on Boiling Transition 41 42 During steady-state operation in a BWR, most of the coolant in the core is in a flow regime 43 known as annular flow. In this flow regime, a thin liquid film is pushed up the surface of the fuel 44 rod cladding by the bulk coolant flow, which is mostly water vapor with some liquid water 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).

Enclosure 1

1 droplets. This provides effective heat removal from the cladding surface; however, under 2 certain conditions, the annular film may dissipate, which reduces the heat transfer and results in 3 an increase in fuel cladding surface temperature. This phenomenon is known as boiling 4 transition or dryout. The elevated surface temperatures resulting from dryout may cause fuel 5 cladding damage or failure.

6 7 1.2 Background on Critical Power Correlations 8

9 For a given set of reactor operating conditions (pressure, flow, etc.), dryout will occur on a fuel 10 assembly at a certain power, known as the critical power. Because the phenomena associated 11 with boiling transition are complex and difficult to model purely mechanistically, 12 thermal-hydraulic test campaigns are undertaken using electrically heated prototypical fuel 13 bundles to establish a comprehensive database of critical power measurements for each BWR 14 fuel product. These data are then used to develop a critical power correlation that can be used 15 to predict the critical power for assemblies in operating reactors. This prediction is usually 16 expressed as the ratio of the actual assembly power to the critical power predicted using the 17 correlation, known as the CPR.

18 19 One measure of the correlations predictive capability is based on its validation relative to the 20 test data. For each point j in a correlations test database, the experimental critical power ratio 21 (ECPR) is defined as the ratio of the measured critical power to the calculated critical power,2 22 or:

23 Measured Critical Power 24 ECPR =

Calculated Critical Power 25 26 For ECPR values less than or equal to 1, the calculated critical power is greater than the 27 measured critical power and the prediction is considered to be non-conservative. Because the 28 measured critical power includes random variations due to various uncertainties, evaluating the 29 ECPR for all of the points in the dataset (or, ideally, a subset of points that were not used in the 30 correlations development) results in a probability distribution. This ECPR distribution allows the 31 predictive uncertainty of the correlation to be determined. This uncertainty can then be used to 32 establish a limit above which there can be assumed that boiling transition will not occur (with a 33 certain probability and confidence level).

34 35 1.3 Background on Thermal-Hydraulic Safety Limits 36 37 To protect against boiling transition, BWRs have implemented an SL on the CPR, known as the 38 minimum critical power ratio (MCPR) SL. As discussed in NUREG-1433 and NUREG-1434 for 39 General Electric BWR designs, the current basis of the MCPR SL is to prevent 99.9 percent of 40 the fuel in the core from being susceptible to boiling transition. This limit is typically developed 41 by considering various cycle-specific power distributions and uncertainties, and is highly 42 dependent on the cycle-specific radial power distribution in the core. As such, the limit may 43 need to be updated as frequently as every cycle.

44 2

Consistent with the definition used in Section 3.1 of the revised TSTF traveler (ADAMS Accession No. ML18149A320) and associated RAI response (i.e., RAI 1) (ADAMS Accession No. ML18149A320).

1 The fuel cladding SL for pressurized-water reactor (PWR) designs, described in the STS for 2 Babcock & Wilcox, Westinghouse, and Combustion Engineering3 plants in NUREG-1430, 3 NUREG-1431, and NUREG-1432,4 respectively, correspond to a 95 percent probability at a 4 95 percent confidence level that departure from nucleate boiling will not occur. As a result of 5 the overall approach taken in developing the PWR limits, they are only dependent on the fuel 6 type(s) in the reactor and the corresponding departure from nucleate boiling ratio (DNBR) 7 correlations. The limits are not cycle-dependent and are typically only updated when new fuel 8 types are inserted in the reactor.

9 10 BWRs also have a limiting condition for operation (LCO) that governs MCPR, known as the 11 MCPR operating limit (OL). The OL on MCPR is an LCO which must be met to ensure that 12 anticipated operational occurrences do not result in fuel damage. The current MCPR OL is 13 calculated by combining the largest change in CPR from all analyzed transients, also known as 14 the CPR, with the MCPR SL.

15 16

2.0 REGULATORY EVALUATION

17 18 2.1 Description of STS Sections 19 20 2.1.1 TS 2.1.1, Reactor Core SLs 21 22 Safety limits ensure that specified acceptable fuel design limits are not exceeded during steady 23 state operation, normal operational transients, and anticipated operational occurrences (AOOs).

24 25 Technical Specification 2.1.1.2 currently requires that with the reactor steam dome pressure 26 greater than or equal to () 785 pounds per square inch gauge (psig) and core flow 10 percent 27 rated core flow, MCPR shall be [1.07] for two recirculation loop operation or [1.08] for single 28 recirculation loop operation. The value in brackets represents plant-specific parameters. The 29 MCPR SL ensures that 99.9 percent of the fuel in the core is not susceptible to boiling transition.

30 31 2.1.2 TS 5.6.3, Core Operating Limits Report [(COLR)]

32 33 Technical Specification 5.6.3 requires core operating limits to be established prior to each 34 reload cycle, or prior to any remaining portion of a reload cycle. These limits are required to be 35 documented in the COLR.

36 37 2.2 Proposed Changes to the STS 38 39 Traveler TSTF-564, Revision 1, proposed a method for determining a revised, 40 cycle-independent MCPR SL for any BWR fuel applicable to all BWR designs. Though the 41 process for determining a revised MCPR SL is broadly applicable to any BWR fuel, the traveler 42 provides a table of sample limits for fuels from Global Nuclear Fuel and Westinghouse Electric 3

Denotes applicability to Combustion Engineering plants with digital control systems only.

4 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 Company. The original MCPR SL, referred to in traveler TSTF-564, Revision 1, as the 2 MCPR99.9% SL, ensures that 99.9 percent of the fuel in the core is not susceptible to boiling 3 transition. The revised MCPR SL, referred to in traveler TSTF-564, Revision 1, as the 4 MCPR95/95 SL, ensures there is a 95 percent probability at a 95 percent confidence level that no 5 fuel rods will be susceptible to transition boiling. Additional changes to the STS and TS Bases 6 proposed in the traveler support the revision of the SL.

7 8 The proposed changes to the STS revise the value of the MCPR SL in TS 2.1.1.2, with 9 corresponding changes to the associated bases. The change to TS 2.1.1.2 replaces the 10 existing separate SLs for single- and two-recirculation loop operation with a single limit since the 11 revised SL is no longer dependent on the number of recirculation loops in operation. In 12 addition, the current MCPR SL is renamed MCPR99.9% and the new MCPR SL is named 13 MCPR95/95. Corresponding changes are made to the associated TS Bases.

14 15 The MCPR99.9% (i.e., the current MCPR SL) is an input to the MCPR operating limits (OL) in 16 limiting condition of operation (LCO) 3.2.2, Minimum Critical Power Ratio (MCPR). While the 17 definition and method of calculation of both the MCPR99.9% and the LCO 3.2.2 MCPR OL 18 remains unchanged, the proposed STS changes include revisions to TS 5.6.3, to require the 19 MCPR99.9% value used in calculating the LCO 3.2.2 MCPR OL to be included in the 20 cycle-specific COLR. Corresponding TS Bases changes for LCO 3.2.2 and TS 5.6.3 support 21 the proposed STS changes.

22 23 2.3 Applicable Regulatory Requirements and Guidance 24 25 Section IV, The Commission Policy, of the Final Policy Statement on Technical Specifications 26 Improvements for Nuclear Power Reactors, published in the Federal Register on July 22, 1993 27 (58 FR 39132), states, in part:

28 29 The purpose of Technical Specifications is to impose those 30 conditions or limitations upon reactor operation necessary to 31 obviate the possibility of an abnormal situation or event giving rise 32 to an immediate threat to the public health and safety by 33 identifying those features that are of controlling importance to 34 safety and establishing on them certain conditions of operation 35 which cannot be changed without prior Commission approval.

36 37 [T]he Commission will also entertain requests to adopt portions 38 of the improved STS [(e.g., TSTF-563)], even if the licensee does 39 not adopt all STS improvements In accordance with this Policy 40 Statement, improved STS have been developed and will be 41 maintained for each NSSS [nuclear steam supply system] owners 42 group. The Commission encourages licensees to use the 43 improved STS as the basis for plant-specific Technical 44 Specifications. [I]t is the Commission intent that the wording 45 and Bases of the improved STS be used to the extent 46 practicable.

47 48 As described in the Commissions Final Policy Statement on Technical Specifications 49 Improvements for Nuclear Power Reactors, NRC and industry task groups for new STS 50 recommended that improvements include greater emphasis on human factors principles in order 51 to add clarity and understanding to the text of the STS, and provide improvements to the Bases

1 of STS, which provides the purpose for each requirement in the specification. The improved 2 vendor-specific STS were developed and issued by the NRC in September 1992.

3 4 As required by 10 CFR 50.36(c), TSs will include items in the following categories: (1) Safety 5 limits, limiting safety system settings, and limiting control settings. As required by 10 CFR 6 50.36(c)(1)(i)(A), safety limits for nuclear reactors are limits upon important process variables 7 that are found to be necessary to reasonably protect the integrity of certain of the physical 8 barriers that guard against the uncontrolled release of radioactivity. If any safety limit is 9 exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the 10 matter, and record the results of the review, including the cause of the condition and the basis 11 for corrective action taken to preclude recurrence. Operation must not be resumed until 12 authorized by the Commission.

13 14 As required by 10 CFR 50.36(c)(2)(i), the TSs will include LCOs, which are the lowest functional 15 capability or performance levels of equipment required for safe operation of the facility. When 16 an LCO is not met, the licensee shall shut down the reactor or follow any remedial action 17 permitted by the TSs until the condition can be met.

18 19 General Design Criterion 10 (GDC), Reactor design, of 10 CFR Part 50 Appendix A, General 20 Design Criteria of Nuclear Power Plants, states:

21 22 The reactor core and associated coolant control and protection systems shall be 23 designed with appropriate margin to assure that specified acceptable fuel design 24 limits are not exceeded during any condition of normal operation, including the 25 effects of anticipated operational occurrences.

26 27 Most plants have a plant-specific design criterion similar to GDC 10. The limit placed on the 28 MCPR acts as a specified acceptable fuel design limit to prevent boiling transition, which has 29 the potential to result in fuel rod cladding failure.

30 31 The NRC staffs guidance contained in Revision 2 of NUREG-0800, Standard Review Plan for 32 the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP),

33 Section 4.4, Thermal and Hydraulic Design,5 provides the following two examples of 34 acceptable approaches to meeting the SRP acceptance criteria for establishing fuel design 35 limits (as stated in SRP Acceptance Criterion 1):

36 37 A. For departure from nucleate boiling ratio (DNBR), CHFR [critical heat flux ratio]

38 or CPR correlations, there should be a 95-percent probability at the 95-percent 39 confidence level that the hot rod in the core does not experience a DNB or boiling 40 transition condition during normal operation or AOOs.

41 42 B. The limiting (minimum) value of DNBR, CHFR, or CPR correlations is to be 43 established such that at least 99.9 percent of the fuel rods in the core will not 44 experience a DNB or boiling transition during normal operation or AOOs.

45 5

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

3.0 TECHNICAL EVALUATION

2 3 3.1 Basis for Proposed Change 4

5 As discussed in Section 2.3 of traveler TSTF-564, Revision 1, and Section 1.3 of this safety 6 evaluation, the current MCPR SL (i.e., the MCPR99.9%) is affected by each plants cycle-specific 7 core design, especially including the core power distribution, fuel type(s) in the reactor, and the 8 power-to-flow operating domain for the plant. As such, it is frequently necessary to change the 9 MCPR SL to accommodate new core designs. Changes to the MCPR SL are usually 10 determined late in the design process and necessitate an accelerated NRC review (i.e., license 11 amendment request) to support the subsequent fuel cycle.

12 13 Traveler TSTF-564, Revision 1, proposes to change the basis for the MCPR SL so that it is no 14 longer cycle-dependent, reducing the frequency of revisions and eliminating the need for NRC 15 review on an accelerated schedule. The proposed revised basis for the MCPR SL aligns it with 16 that of the DNBR SL used in PWRs, which, as previously noted in Section 2.3 of this safety 17 evaluation, provides a 95 percent probability at a 95 percent confidence level that no fuel rods 18 will experience departure from nucleate boiling.

19 20 The intent of the proposed basis for the revised MCPR SL is acceptable to the NRC staff based 21 on the discussion in SRP Section 4.4, SRP Acceptance Criterion 1. The remainder of this 22 safety evaluation is devoted to ensuring that the methodology for determining the revised MCPR 23 SL provides the intended result, that the revised MCPR SL can be adequately determined in 24 cores using various types of fuel, that the proposed SL continues to fulfill the necessary 25 functions of an SL without unintended consequences, and that the proposed changes have 26 been adequately implemented in the STS and associated TS Bases.

27 28 3.2 Revised MCPR SL Definition 29 30 As discussed in Section 1.2 of this safety evaluation, a critical power correlations ECPR 31 distribution quantifies the uncertainty associated with the correlation. The TSTF traveler 32 provides a definition for a limit that bounds 95 percent of a correlations ECPR distribution at a 33 95 percent confidence level, according to the following formula:

34 35 MCPR = +

36 37 where i is the correlations mean ECPR, i is the standard deviation of the correlations ECPR 38 distribution, and i is a statistical parameter chosen to provide 95% probability at 95%

39 confidence (95/95) for the one-sided upper tolerance limit that depends on the number of 40 samples (Ni) in the critical power database. This formula is commonly used to determine a 41 95/95 one-sided upper tolerance limit for a normal distribution, which is appropriate for the 42 situation under consideration. The factor is generally attributed to D. B. Owen6 and was also 43 reported by M. G. Natrella7 as referenced in traveler TSTF-564, Revision 1. Example values of 44 are provided in Table 2 of the TSTF traveler. Table 1 of the TSTF traveler includes some 45 reference values of the MCPR95/95.

6 D. B. Owen, Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, Sandia Corporation, SCR-607, March 1963, ADAMS Accession No. ML14031A495.

7 M. G. Natrella, Experimental Statistics, National Bureau of Standards, National Bureau of Standards Handbook 91, August 1963.

1 2 As discussed by Piepel and Cuta8 for DNBR correlations, the acceptability of this approach is 3 predicated on a variety of assumptions, including the assumptions that the correlation data 4 comes from a common population and that the correlations population is distributed normally.

5 These assumptions are typically addressed generically when a critical power or critical heat flux 6 correlation is reviewed by the NRC staff, who may apply penalties to the correlation to account 7 for any issues identified. In its response to an RAI, the TSTF states that such penalties applied 8 during the NRCs review of the critical power correlation would be imposed on the mean or 9 standard deviation used in the calculating the MCPR95/95 (ADAMS Accession 10 No. ML18149A320). These penalties would also continue to be imposed in the determination of 11 the MCPR99.9%, along with any other penalties associated with the process of (or other inputs 12 used in) determining the MCPR99.9% (e.g., penalties applied to the MCPR99.9% SL for operation in 13 the Maximum Extended Load Limit Line Analysis Plus (MELLLA+) operating domain).

14 15 The NRC staff finds the definition of the MCPR95/95 will appropriately establish a 95/95 upper 16 tolerance limit on the critical power correlation and that any issues in the underlying correlation 17 will be addressed through penalties on the correlation mean and standard deviation, as 18 necessary. Therefore, the NRC staff concludes that the MCPR95/95 definition, as proposed, 19 establishes an acceptable fuel design limit and is acceptable.

20 21 3.3 Determination of Revised MCPR SL for Mixed Cores 22 23 The TSTF proposed that a core containing a variety of fuel types would evaluate the MCPR95/95 24 for all of the fresh and once-burnt fuel in the core and apply the most limiting (i.e., the largest) 25 value of MCPR95/95 for each of the applicable fuel types as the MCPR SL. As stated in 26 Section 3.1 of the TSTF traveler, this is because bundles that are twice-burnt or more at the 27 beginning of the cycle have significant MCPR margin relative to the fresh and once-burnt fuel.

28 In its response to an RAI (ADAMS Accession No. ML18149A320), the TSTF provided additional 29 justification for this assertion. The justification is that the MCPR for twice-burnt and greater fuel 30 is far enough from the MCPR for the limiting bundle that its probability of boiling transition is 31 very small compared to the limiting bundle and it can be neglected in determining the SL.

32 Results of a study provided in the RAI response indicate that this is the case even for fuel 33 operated on short (12-month) reload cycles. As discussed in the RAI, twice-burnt or greater fuel 34 bundles are included in the cycle-specific evaluation of the MCPR99.9% and the MCPR OL. If a 35 twice-burnt or greater fuel bundle is found to be limiting, it would be governed by the MCPR OL, 36 which will always be more restrictive than both the MCPR95/95 and the MCPR99.9%. The NRC 37 staff found this justification to be appropriate and determined that it is acceptable to determine 38 the MCPR95/95 SL for the core based on the most limiting value of the MCPR95/95 for the fresh 39 and once-burnt fuel in the core.

40 41 The NRC staff reviewed the information furnished by the TSTF and determined that the process 42 for establishing the revised MCPR SL for mixed cores ensures that the limiting fuel types in the 43 core will be evaluated and the limiting MCPR99.9% will be appropriately applied as the SL. The 44 NRC staff therefore found this process to be acceptable.

45 46 The size, mean, and standard deviation of the ECPR database may need to be provided by a 47 fuel vendor to determine the MCPR95/95 for a legacy fuel type. The value of depends on the 48 number of samples (Ni) in the critical power database. If the number of data points in the 8

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 database is not supplied by the vendor, the TSTF response to an RAI stated that a value of =

2 1.8 would be imposed on the MCPR95/95 determination, on the basis that any database used to 3 develop a critical power correlation will need at least 500 points to be acceptable.9 The limiting 4 value from either the new or legacy fuel would then be applied as the SL. The NRC staff finds 5 that there are potential circumstances where the number of data points used in determining the 6 correlations uncertainty may not correspond to a value of 1.8; for example, future correlations 7 may need fewer data points, or the subset of data used to determine a correlations uncertainty 8 may be smaller than the full correlation database. Therefore, the NRC staff determined that a 9 value of 1.8 for legacy fuel types where the number of data points N is not provided may not be 10 acceptable, and the used in determining the MCPR95/95 must be justified to be appropriate or 11 conservative for the fuel type and correlation in question. This determination does not affect the 12 overall acceptability of the process for determining the MCPR95/95 for a mix of fuel types as 13 discussed above. The NRC staff also notes that, as stated in Section 1.0 of this SE, this STS 14 change is only available to licensees through the CLIIP when using the fuel bundle types 15 specified in Table 1 of the traveler. Therefore, the use of legacy fuels, for which this 16 determination would be relevant, is outside the scope of a CLIIP application.

17 18 3.4 Relationship Between MCPR Safety and Operating Limits 19 20 In its response to an RAI, the TSTF discussed that MCPR99.9% is expected to always be greater 21 than the MCPR95/95 for two reasons. First, because the MCPR99.9% includes uncertainties not 22 factored into the MCPR95/95, and second, because the 99.9 percent probability basis for 23 determining the MCPR99.9% is more conservative than the 95 percent probability at a 95 percent 24 confidence level used in determining the MCPR95/95. The level of conservatism in the MCPR95/95 25 SL is appropriate because the lead fuel rod in the core (i.e., the limiting fuel rod with respect to 26 MCPR) is used to evaluate whether any fuel rods in the core are susceptible to boiling 27 transition, which is also discussed in the TSTF RAI response (i.e., RAI 2(a) (ADAMS Accession 28 No. ML18149A320)). This is consistent with evaluations performed for PWRs using a 95/95 29 upper tolerance limit on the correlation uncertainty as an SL.

30 31 The TSTF traveler proposed that the MCPR OL defined in LCO 3.2.2 would continue to be 32 evaluated using the MCPR99.9% as an input. The MCPR99.9% will continue to be evaluated in the 33 same way as it is currently, using the whole core. The TSTF traveler also changes TS 5.6.3, to 34 require the cycle-specific value of the MCPR99.9% to be included in the COLR. The methods 35 supporting the inclusion of the MCPR99.9% must also therefore be included in the list of COLR 36 references contained in TS 5.6.3.b.10 The changes to TS 5.6.3.b help to ensure that the 37 uncertainties being removed from the MCPR SL are still included as part of the MCPR OL and 38 will continue to appropriately inform plant operation.

39 40 The NRC staff therefore determined that the changes proposed by the TSTF will retain an 41 adequate level of conservatism in the MCPR SL in TS 2.1.1.2 while appropriately ensuring that 42 plant- and cycle-specific uncertainties will be retained in the MCPR OL. The NRC staff notes 43 that the MCPR95/95 represents a hard floor on the value of the MCPR99.9%, which should always 9

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, March 2011 (ADAMS Accession No. ML11102A076). This is more conservative than the for N =

500 data points, which would be 1.763.

10 The MCPR OL is already a COLR parameter and as such, the methodology to calculate it should already be included in TS 5.6.3.b. In current BWR methodologies for all major U.S. fuel suppliers, the MCPR SL (i.e., the MCPR99.9%) is calculated using the same methodology as the MCPR OL. Should this change, because the MCPR99.9% and the MCPR OL are both COLR parameters, both methodologies would need to be included in TS 5.6.3.b.

1 be higher since it accounts for numerous uncertainties that are not included in the MCPR95/95 (as 2 discussed in Section 3.1 of traveler TSTF-564, Revision 1, and the TSTFs response to RAI 7).

3 4 3.5 Implementation of the Revised MCPR SL in the TS 5

6 The value reported in TS 2.1.1.2 will be the value calculated using Equation 1 from traveler 7 TSTF-564, Revision 1, at a precision of two digits past the decimal point with the hundreds digit 8 rounded up. This is consistent with the current practice for PWR DNBR SLs and is acceptable 9 to the NRC staff. As previously discussed, the value of the MCPR OL provided in LCO 3.2.2 will 10 continue to be reported in the COLR. The COLR will be required to contain the cycle-specific 11 MCPR99.9% value and TS 5.6.3.b will continue to reference appropriate NRC-approved 12 methodologies for determination of the MCPR99.9% and the MCPR OL.

13 14 Traveler TSTF-564, Revision 1, added new language to the TS 2.1.1 Bases to provide the basis 15 for the redefined MCPR SL. In its response to an RAI, the TSTF revised the TS Bases to 16 specify the fuel type on which the SL is based. Though the traveler is intended to be applicable 17 to all types of fuel, the existing STS bases only discuss certain fuel vendors. As discussed in 18 RAI responses to RAIs 1, 10, and 12, the TSTF proposed changes to the STS bases for issues 19 directly related to the implementation of the revised MCPR SL.

20 21 The NRC staff reviewed the proposed TS and Bases changes and found that the TSTF 22 appropriately implemented the revised MCPR SL, as discussed in the proposed TSTF-564, 23 Revision 1 traveler.

24 25

4.0 CONCLUSION

26 27 The NRC staff reviewed traveler TSTF-564, Revision 1, which proposed changes to 28 NUREG-1433 and NUREG-1434. The NRC staff determined that the proposed definition of the 29 MCPR SL in TS 2.1.1.2 was acceptably modified and will be calculated in a manner consistent 30 with the new definition. Under the new definition, the MCPR SL will continue to protect the fuel 31 cladding against the uncontrolled release of radioactivity by preventing the onset of boiling 32 transition, thereby fulfilling the requirements of 10 CFR 50.36(c)(1) for SLs. The MCPR OL in 33 LCO 3.2.2 remains unchanged and will continue to meet the requirements of 34 10 CFR 50.36(c)(2) (and GDC 10 or the equivalent plant-specific design criterion) by ensuring 35 that no fuel damage results during normal operation and AOOs. The NRC staff determined that 36 the changes to TS 5.6.3 proposed in the traveler are acceptable; upon adoption of the revised 37 MCPR SL, the COLR will be required to contain the MCPR99.9%, supporting the determination of 38 the MCPR OL using current methodologies.

39 40 Principal Contributors: R. Anzalone, NRR/DSS 41 C. Tilton, NRR/DSS

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 33 10 percent rated core flow, MCPR shall be [1.07] for two recirculation loop operation 34 or [1.08] for single recirculation loop operation.] The MCPR SL ensures that 99.9 percent 35 of the fuel in the core is 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 15 single- and two-recirculation loop operation with a single limit since the revised SL is no 16 longer dependent 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 MCPR99.9% 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

Enclosure TSTF-564, Revision 2

BWROG-133, Rev. 0 TSTF-564, Rev. 2 Technical Specifications Task Force Improved Standard Technical Specifications Change Traveler Safety Limit MCPR NUREGs Affected: 1430 1431 1432 1433 1434 2194 Classification: 1) Technical Change Recommended for CLIIP?: Yes Correction or Improvement: Improvement NRC Fee Status: Not Exempt Changes Marked on ISTS Rev 4.0 See attached justification.

Revision History OG Revision 0 Revision Status: Closed Revision Proposed by: BWROG Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 30-May-17 Owners Group Comments Reviewed by the Owners Groups in October 2016.

A presubmittal teleconfernce was held on March 20, 2017. The traveler was revised to address NRC comments:

Owners Group Resolution: Approved Date: 18-Jul-17 TSTF Review Information TSTF Received Date: 02-Aug-17 Date Distributed for Review 02-Aug-17 TSTF Comments:

(No Comments)

TSTF Resolution: Approved Date: 28-Aug-17 NRC Review Information NRC Received Date: 28-Aug-17 NRC Comments:

Revised to reflect RAI responses.

Final Resolution: NRC Requests Changes: TSTF Will Revise 01-Aug-18 Copyright(C) 2018, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-133, Rev. 0 TSTF-564, Rev. 2 TSTF Revision 1 Revision Status: Active Revision Proposed by: BWROG Revision

Description:

Revised to reflect RAI response:

• The justification and Bases markup were revised and expanded to discuss application of the methodology to other fuel vendors.

• The changes to the TS 3.2.2 Applicability Bases were removed.

• Discussion of boiling transition was made more consistent in the justification and Bases.

• Reference to Westinghouse Optima3 fuel was removed.

• The SLMCPR value will be rounded up.

• The TS Bases states the fuel type on which the SLMCPR value is based.

Owners Group Review Information Date Originated by OG: 10-Apr-18 Owners Group Comments (No Comments)

Owners Group Resolution: Approved Date: 02-May-18 TSTF Review Information TSTF Received Date: 10-Apr-18 Date Distributed for Review 10-Apr-18 TSTF Comments:

(No Comments)

TSTF Resolution: Approved Date: 02-May-18 NRC Review Information NRC Received Date: 29-May-18 TSTF Revision 2 Revision Status: Active Revision Proposed by: BWROG Revision

Description:

Revised to reflect rounding up SL values. Updated GE methodology document in Reference 1. Changed GE14 SL from 1.05 to 1.06.

01-Aug-18 Copyright(C) 2018, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-133, Rev. 0 TSTF-564, Rev. 2 TSTF Revision 2 Revision Status: Active Owners Group Review Information Date Originated by OG: 01-Aug-18 Owners Group Comments (No Comments)

Owners Group Resolution: Date:

Affected Technical Specifications Bkgnd 2.1.1 Bases Reactor Core SLs S/A 2.1.1 Bases Reactor Core SLs SL 2.1.1.2 Safety Limit MCPR Bkgnd 3.2.2 Bases Minimum Critical Power Ratio (MCPR)

S/A 3.2.2 Bases Minimum Critical Power Ratio (MCPR)

LCO 3.2.2 Bases Minimum Critical Power Ratio (MCPR)

Appl. 3.2.2 Bases Minimum Critical Power Ratio (MCPR) 5.6.3 Core Operating Limits Report 01-Aug-18 Copyright(C) 2018, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-564, Rev. 2

1.

SUMMARY

DESCRIPTION The Technical Specification (TS) Safety Limit (SL) value and method of calculation for the Minimum Critical Power Ratio (MCPR) limit, SL 2.1.1.2, is revised for Boiling Water Reactor (BWR) plants. SL MCPR limits are provided for Global Nuclear Fuel and Westinghouse fuel designs. These and other fuel vendors may determine SL MCPR limits for other fuel designs using the described methodology. The revised calculation method is based on using the Critical Power Ratio (CPR) data statistics and is revised from ensuring that 99.9% of the rods would not be susceptible to boiling transition to ensuring that there is a 95% probability at a 95%

confidence level that no rods will be susceptible to transition boiling. The revised MCPR SL is consistent with the regulatory requirements while being cycle-independent, thereby minimizing the need for TS license amendment requests to revise this value for each operating cycle.

2. DETAILED DESCRIPTION 2.1. Current Design and Licensing Basis MCPR is defined in Section 1.1 of the BWR TS as:

The MCPR shall be the smallest critical power ratio (CPR) that exists in the core [for each class of fuel]. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations are developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Although fuel damage does not necessarily occur if a fuel rod experiences transition boiling, the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient metric for ensuring that fuel failures due to inadequate cooling do not occur.

The TS contain two limits on MCPR: a safety limit (herein referred to as the SLMCPR) and a Limiting Condition for Operation (LCO) operating limit (herein referred to as the OLMCPR).

Title 10 of the Code of Federal Regulations (10 CFR), Part 50, paragraph 50.36(c)(1) defines "safety limits" as 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. In the case of the MCPR safety limit, the physical barrier being protected is the fuel rod cladding. The current SLMCPR is calculated as the point at which 99.9% of the fuel rods are not susceptible to boiling transition (i.e., reduced heat transfer) during normal operation and anticipated operational occurrences, herein referred to as MCPR99.9%. The MCPR99.9% limit is calculated for each BWR on a cycle-by-cycle basis using approved methodologies.

Page 1

TSTF-564, Rev. 2 An LCO is defined in 10 CFR 50.36(c)(2) as the lowest functional capability or performance level of equipment required for safe operation of the facility. The OLMCPR LCO is required to be met to ensure that no fuel damage results during anticipated operational occurrences (AOOs).

To ensure that the measured MCPR does not exceed the SLMCPR during any AOO that occurs with moderate frequency, transients are analyzed to determine the largest reduction in critical power ratio. The limiting transient yields the largest change in CPR (CPR) during the event.

The largest CPR is combined with the MCPR99.9% value to determine the OLMCPR LCO limit.

Together, SLMCPR and OLMCPR ensure that no fuel damage occurs due to transition boiling during normal operation or AOOs.

2.2. Current Technical Specifications Requirements NUREG-1433 and NUREG-14341, Safety Limit 2.1.1.2 states:

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% rated core flow:

MCPR shall be [1.07] for two recirculation loop operation or [1.08] for single recirculation loop operation.

The values in brackets, [1.07] and [1.08], are plant-specific limits. The reactor steam dome pressure and core flow values are also plant-specific and differences do not affect the applicability of the proposed change.

NUREG-1433 and NUREG-1434, LCO 3.2.2, "Minimum Critical Power Ratio (MCPR)," states:

All MCPRs shall be greater than or equal to the MCPR operating limits specified in the

[Core Operating Limits Report] COLR.

NUREG-1433 and NUREG-1434, LCO 3.2.2 is applicable when thermal power is 25% of rated thermal power. Plant-specific TS may have a different Applicability, which does not affect the justification for the proposed change.

1 NUREG-1433 is based on the BWR/4 plant design, but is also applicable of the BWR/2, BWR/3, and, for some requirements, to the BWR/5 plant designs. NUREG-1434 is based on the BWR/6 plant design, and is applicable, for some requirements, to the BWR/5 plant design.

Page 2

TSTF-564, Rev. 2 NUREG-1433 and NUREG-1434, TS 5.6.3, "Core Operating Limits Report," states:

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:

[The individual specifications that address core operating limits must be referenced here.]

2.3. Reason for the Proposed Change The current SLMCPR (i.e., MCPR99.9%) is affected by the cycle-specific design, such as core power distribution, fuel type, and operating power-flow domain. These factors generally vary enough from cycle-to-cycle that changes to the SLMCPR TS values are common. The subsequent cycle core design is dependent on the core burnup of the previous cycle. As a result, the core design for the subsequent cycle is typically finalized late in the previous fuel cycle.

Consequently, license amendment requests to modify the SLMCPR typically request an accelerated NRC review (i.e., less than the typical period of one year) to support the scheduled start of the subsequent fuel cycle. A review of the NRC Agency-wide Documents Access and Management System (ADAMS) identified a number of approved license amendments to revise the SLMCPR in 2015 and 2016.

The Babcock & Wilcox, Westinghouse, Combustion Engineering, and Advanced Passive 1000 (AP1000) Standard Technical Specifications (NUREG-1430, NUREG-1431, NUREG-1432, and NUREG-2194) safety limits on fuel cladding are based on a Departure from Nucleate Boiling Ratio (DNBR) limit. The PWR DNBR limits are roughly analogous to the BWR SLMCPR, in that both protect fuel cladding integrity from inadequate cooling. The PWR DNBR safety limit corresponds to a 95% probability at a 95% confidence level that DNB will not occur, vice the BWR SLMCPR that is based on ensuring that 99.9% of the fuel rods will not be susceptible to boiling transition. Either approach is statistically valid, but this difference results in a PWR safety limit that is only dependent on the fuel type(s) in the reactor and the corresponding DNBR correlations. The PWR DNBR Safety Limits are not cycle dependent and are typically only revised when the type of fuel changes.

2.4. Description of the Proposed Change The proposed change revises the standard TS in NUREG-1433 and NUREG-1434 for all BWR plants. SL MCPR limits are provided for Global Nuclear Fuel and Westinghouse fuel designs.

These and other fuel vendors may determine SL MCPR limits for other fuel designs using the described methodology.

The proposed change substantially reduces the need for cycle-specific changes to the SLMCPR and eliminates the need for accelerated NRC review of those changes.

Page 3

TSTF-564, Rev. 2 The NUREG-1433 and NUREG-1434 Safety Limit 2.1.1.2 is revised to state:

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% rated core flow:

MCPR shall be [1.07] [for two recirculation loop operation or [1.08] for single recirculation loop operation].

The phrase "for two recirculation loop operation or [1.08] for single recirculation loop operation" are shown in brackets to retain compatibility for BWR plants that do not adopt the proposed change. The proposed SLMCPR methodology is not dependent on the number of recirculation loops in operation, so the distinction between a single loop and two loop operation is not needed.

Plants adopting the proposed change will revise their plant-specific SL to state:

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% rated core flow:

MCPR shall be [1.07].

The bracketed limit "[1.07]" will be replaced with a revised SLMCPR that ensures there is a 95%

probability at a 95% confidence level that no rods will be susceptible to transition boiling, and is referred to as SLMCPR95/95. The reactor steam dome pressure and core flow values are also plant-specific. Differences between the Standard Technical Specifications values and the plant-specific values do not affect the applicability of the proposed change.

The single SLMCPR95/95 is based on the fuel type in the reactor core.

Table 1: Proposed MCPR95/95 Values by Vendor and Fuel Bundle Type Vendor Fuel Type Proposed MCPR95/95 Global GE14 1.06 Nuclear Fuel Global GNF2 1.07 Nuclear Fuel Global GNF3 1.07 Nuclear Fuel Westinghouse Optima2 1.06 The derivation of these values is described in proprietary letters to the NRC from Global Nuclear Fuel and Westinghouse (References 1 and 2). Should these or different vendors desire to use this Page 4

TSTF-564, Rev. 2 approach for existing or new fuel types, the fuel vendor will describe to the NRC the derivation of the MCPR95/95 value for that fuel type. This description may be referenced by a licensee requesting a change to SLMCPR95/95. The TS Bases state the fuel type on which the SL is based.

For cores loaded with a mix of applicable fuel types, the SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR95/95 values for the fuel products that are fresh or once-burnt at the start of the cycle.

LCO 3.2.2 is not affected by the proposed change. However, licensees adopting the proposed change will include the MCPR99.9% value (i.e., the value equivalent to the current, cycle-dependent SLMCPR) in the COLR values for LCO 3.2.2.

TS 5.6.3, "Core Operating Limits Report," paragraph a, is revised to require the MCPR99.9%

value to be in the cycle-specific 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 (addition shown in italics):

[The individual specifications that address core operating limits must be referenced here.

The MCPR99.9% value used to calculate the LCO 3.2.2, "MCPR," limit shall be specified in the COLR.]

The proposed change is supported by changes to the TS Bases. The SL 2.1.1.2 Bases and TS 3.2.2 Bases are revised to reflect the proposed limits for Global Nuclear Fuel and Westinghouse fuel. In sections of the Bases applicable to all fuel types, the existing text and the proposed text are both presented in brackets, signifying that the licensee should choose the applicable description. A reviewer's note is added to the Bases to explain these options. The regulation at 10 CFR 50.36, states, "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications." A licensee may make changes to the TS Bases without prior NRC staff review and approval in accordance with the Technical Specifications Bases Control Program. The proposed TS Bases changes are consistent with the proposed TS changes and provide the purpose for each requirement in the specification consistent with the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 2, 1993 (58 FR 39132). Therefore, the Bases changes are provided for information and approval of the Bases is not requested.

A model application is included in the proposed change as Enclosure 1. The model may be used by licensees desiring to adopt the traveler following NRC approval.

3. TECHNICAL EVALUATION The proposed change revises the TS limit for the SLMCPR for the applicable plants and places the existing SLMCPR value (i.e., MCPR99.9%) in the COLR. The revised limit calculation method is based on using the Critical Power Ratio (CPR) data statistics and is revised from ensuring that 99.9% of the rods would not be susceptible to boiling transition to ensuring that there is a 95% probability at a 95% confidence level that no rods will be susceptible to transition Page 5

TSTF-564, Rev. 2 boiling. The revised SLMCPR, referred to as SLMCPR95/95, is based only on the CPR correlation uncertainty determined for the Global Nuclear Fuel or Westinghouse fuel type. Plant and cycle-specific uncertainties are not included in the SLMCPR95/95. These uncertainties are currently and will continue to be included in the OLMCPR LCO. Reactor coolant flow is one of the uncertainties removed from the SLMCPR calculation and retained in the OLMCPR.

Therefore, the SLMCPR for dual recirculation loop operation and single recirculation loop operation are replaced with a single SLMCPR95/95.

The LCO 3.2.2 limits (i.e., the OLMCPR values) are not changed and will be based on the existing SLMCPR, referred to as MCPR99.9%. The OLMCPR will continue to be determined based on the transient CPR components and the cycle-specific MCPR99.9% value that will be included in the COLR. Therefore, the margin to boiling transition remains unchanged.

3.1. Statistical Treatment of MCPR95/95 For each Global Nuclear Fuel and Westinghouse BWR fuel product (designated i), the MCPR95/95(i) is calculated using that products experimentally determined critical power statistics as follows:

MCPR95/95(i) = µi + i*i (Eq. 1)

Where,

µi is the mean Experimental Critical Power Ratio (ECPR),

i is the standard deviation of the ECPRs, 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.

i is a fuel product line, such as GE14, GNF2, GNF3, and OPTIMA2.

For each test data point "n", the Critical Power (CP) correlation is evaluated at the same conditions as the test to define the Experimental Critical Power Ratio (ECPRn) for that data point as:

( Measured Critical Power )n ECPR n (Eq. 1.1)

(Critical Power from Correlation)n From all appropriate N of these individual test data points the mean and standard deviation of the ECPR distribution is determined. The sample size N determines the value of .

The statistical parameter, i, is calculated using formulas attributed to Mary Gibbons Natrella (1963) as recommended by the National Institute of Standards and Technology (NIST) in their Page 6

TSTF-564, Rev. 2 Engineering Statistics Handbook (Reference 3). For a 95/95 probability/confidence level, the i values are shown in the table below as a function of database size (Ni).

Table 2: Statistical Parameter, i, at (95/95) for the One-Sided Upper Tolerance Limit Database Size, Ni i 500 1.7625 750 1.7401 1000 1.7270 1250 1.7181 1500 1.7115 2000 1.7024 Assuming a typical critical power database of 1000 data points with no bias (i.e., µi = 1.0), the following table illustrates representative MCPR95/95(i) values as a function of the database standard deviation.

Table 3: Representative MCPR95/95 Values for Ni=1000 Standard MCPR95/95 Deviation, i (%)

2.0 1.03 2.5 1.04 3.0 1.05 3.5 1.06 4.0 1.07 4.5 1.08 5.0 1.09 For cores loaded with a single fuel product, the SLMCPR95/95 is the MCPR95/95(i) value for that particular product line.

Page 7

TSTF-564, Rev. 2 For cores with a mix of fuel products, the corresponding SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR95/95(i) values for the product lines that are fresh or once-burnt at the start of the cycle. The MCPR95/95(i) values for product lines that are twice-burnt or more at the start of the cycle may be ignored, as these higher exposure bundles operate with considerable MCPR margin relative to the more limiting fresh and once-burnt bundles.

The SLMCPR95/95 will be reported in the TS to two digits past the decimal, rounding up. The SLMCPR95/95 also serves as the minimum value for the cycle-specific MCPR99.9%.

The revised method for calculation of the SL for Global Nuclear Fuel and Westinghouse fuel will continue to meet the regulatory definition of a safety limit and to reasonably protect the integrity of the fuel rod cladding against the uncontrolled release of radioactivity. The proposed change is also consistent with equivalent safety limits for other plant designs.

3.2. Cycle-Specific OLMCPR The current MCPR99.9% statistical limit calculation will continue to be performed using the approved methodology (e.g., References 4 through 7 or the plant-specific equivalents) and will be reported in the COLR. The OLMCPR limit in LCO 3.2.2 will continue to be determined based on the transient CPR component and the cycle-specific MCPR99.9% value. No changes to the method of determining the OLMCPR (i.e., the LCO 3.2.2 limit) are proposed, and the LCO limits and the MCPR99.9% value will be reported in the COLR.

4. REGULATORY EVALUATION Title 10 of the Code of Federal Regulations (10 CFR), Part 50, paragraph 50.36(c)(1)(i)(A) states:

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.

The purpose of the MCPR safety limit (SLMCPR) is to protect the physical barrier of the fuel cladding against the uncontrolled release of radioactivity. The SLMCPR is set such that no significant fuel damage is calculated to occur if the limit is met. Although it is recognized that the onset of transition boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. Therefore, the proposed change to the SLMCPR will continue to protect the fuel cladding physical barrier from uncontrolled release of radioactivity.

10 CFR 50, Appendix A, General Design Criterion 10 states that specified acceptable fuel design limits will not be exceeded during steady state operation, normal operational transients, and Anticipated Operational Occurrences (AOOs). Most plants have a plant-specific design criterion Page 8

TSTF-564, Rev. 2 similar to GDC 10. This design criterion will continue to be met. The OLMCPR, which is not affected by the proposed change, is established to ensure that no fuel damage results during normal operation, normal operational transients, and AOOs.

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

5. REFERENCES

1. Letter from Brian R. Moore, Global Nuclear Fuel, to U.S. NRC, "Revised Information Supporting TSTF-564 Safety Limit Minimum Critical Power Ratio," July 30, 2018, ADAMS Accession Nos. ML18212A018, ML18212A019, ML18212A020, ML18212A021.

2. Letter from James A. Gresham, Westinghouse Electric Company, to U.S. NRC, "Submittal of 'Calculation for Technical Specification SLM CPR Values Applying to Westinghouse Fuel in Support of TSTF-564'," May 16, 2017, ADAMS Accession No. ML17142A319.

3. NIST/SEMATECH e-Handbook of Statistical Methods, http://www.itl.nist.gov/div898/handbook/, April 2012.

4. GE Nuclear Energy, "General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," NEDO-10958-A, January 1977, ADAMS Accession No. ML102290144.

5. GE Nuclear Energy, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations," NEDC-32694P-A, August 1999, ADAMS Accession No. ML003740166.

6. GE Nuclear Energy, "Methodology and Uncertainties for Safety Limit MCPR Evaluations," NEDC-32601-P-A, August 1999, ADAMS Accession No. ML003740166.

7. ABB Combustion Engineering Nuclear Operations, "Reference Safety Report for Boiling Water Reactor Reload Fuel," CENPD-300-P-A, July 1996. ADAMS Accession No. ML110260388.

Page 9

TSTF-564, Rev. 2 Enclosure 1 Model Application

TSTF-564, Rev. 2

[DATE] 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 DOCKET NO.PLANT NAME 50-[xxx]

SUBJECT:

Application to Revise Technical Specifications to Adopt TSTF-564, "Safety Limit MCPR" Pursuant to 10 CFR 50.90, [LICENSEE] is submitting a request for an amendment to the Technical Specifications (TS) for [PLANT NAME, UNIT NOS.].

[LICENSEE] requests adoption of TSTF-564, "Safety Limit MCPR," Revision 2, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the [PLANT NAME, UNIT NOS] Technical Specifications (TS). The proposed amendment revises the Technical Specification (TS) safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL. provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked to show the proposed changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides existing TS Bases pages marked to show the proposed changes for information only.

No regulatory commitments are made in this submittal.

Approval of the proposed amendment is requested by [date]. Once approved, the amendment shall be implemented within [ ] days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated [STATE] Official.

[In accordance with 10 CFR 50.30(b), a license amendment request must be executed in a signed original under oath or affirmation. This can be accomplished by attaching a notarized affidavit confirming the signature authority of the signatory, or by including the following statement in the cover letter: "I declare under penalty of perjury that the foregoing is true and correct.

Executed on (date)." The alternative statement is pursuant to 28 USC 1746. It does not require notarization.]

If you should have any questions regarding this submittal, please contact [NAME, TELEPHONE NUMBER].

Page 1

TSTF-564, Rev. 2 Sincerely,

[Name, Title]

Attachments: 1. Description and Assessment

2. Proposed Technical Specification Changes (Mark-Up)

3. Revised Technical Specification Pages

4. Proposed Technical Specification Bases Changes (Mark-Up) for Information Only

{Attachments 2, 3, and 4 are not included in the model application and are to be provided by the licensee.}

cc: NRC Project Manager NRC Regional Office NRC Resident Inspector State Contact Page 2

TSTF-564, Rev. 2 ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

[LICENSEE] requests adoption of TSTF-564, "Safety Limit MCPR," Revision 2, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the [PLANT NAME, UNIT NOS] Technical Specifications (TS). The proposed amendment revises the Technical Specification (TS) safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation

[LICENSEE] has reviewed the safety evaluation for TSTF-564 provided to the Technical Specifications Task Force in a letter dated [DATE]. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-564. [As described herein,]

[LICENSEE] has concluded that the justifications presented in TSTF-564 and the safety evaluation prepared by the NRC staff are applicable to [PLANT, UNIT NOS.] and justify this amendment for the incorporation of the changes to the [PLANT] TS.

The [PLANT], Unit [Unit Numbers], reactor [is currently][will be] fueled with [TYPE] fuel bundles [describe multiple types of fuel bundles and which type limits the SLMCPR consistent with discussion in the traveler]. The proposed Safety Limit in [SL 2.1.1.2] is [1.07], consistent with [Table 1 of TSTF-564][ reference the fuel vendor or NRC documentation of the Safety Limit value for the limiting fuel type].

The MCPR value calculated as the point at which 99.9% of the fuel rods would not be susceptible to boiling transition (i.e., reduced heat transfer) during normal operation and anticipated operational occurrences is referred to as MCPR99.9%. Technical Specification 5.6.3, "Core Operating Limits Report (COLR)," is revised to require the MCPR99.9% value to be included in the cycle-specific COLR.

2.2 Variations

((LICENSEE] is not proposing any variations from the TS changes described in TSTF-564 or the applicable parts of the NRC staffs safety evaluation dated [DATE].] ((LICENSEE] is proposing the following variations from the TS changes described in TSTF-564 or the applicable parts of the NRC staffs safety evaluation: describe the variations]

[The [PLANT] TS utilize different [numbering][and][titles] than the Standard Technical Specifications on which TSTF-564 was based. Specifically, [describe differences between the plant-specific TS numbering and/or titles and the TSTF-564 numbering and titles.] These differences are administrative and do not affect the applicability of TSTF-564 to the [PLANT]

TS.]

Page 3

TSTF-564, Rev. 2

[The [PLANT] TS contain requirements that differ from the Standard Technical Specifications on which TSTF-564 was based, such as reactor steam dome pressure or core flow in SL 2.1.1.2, or Applicability in TS 3.2.2, but these differences do not affect the applicability of the TSTF-564 justification. [For differences other than reactor steam dome pressure, core flow, or applicability, describe the differences and why TSTF-564 is still applicable.]

[The traveler and Safety Evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). [PLANT] was not licensed to the 10 CFR 50, Appendix A, GDC. The [PLANT] equivalents of the referenced GDC are [reference including UFSAR location, if applicable]. [Discuss the equivalence of the referenced plant-specific requirements to the Appendix A GDC as related to the proposed change.] This difference does not alter the conclusion that the proposed change is applicable to

[PLANT].]

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis

[LICENSEE] requests adoption of TSTF-564, "Safety Limit MCPR," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the [PLANT NAME, UNIT NOS] Technical Specifications (TS). The proposed change revises the Technical Specifications (TS) safety limit on minimum critical power ratio (SLMCPR). The revised limit calculation method is based on using the Critical Power Ratio (CPR) data statistics and is revised from ensuring that 99.9% of the rods would not be susceptible to boiling transition to ensuring that there is a 95% probability at a 95% confidence level that no rods will be susceptible to transition boiling. A single SLMCPR value will be used instead of two values applicable when one or two recirculation loops are in operation. TS 5.6.3, "Core Operating Limits Report (COLR)," is revised to require the current SLMCPR value to be included in the COLR.

[LICENSEE] has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the Core Operating Limits Report (COLR). The SLMCPR is not an initiator of any accident previously evaluated. The revised safety limit values continue to ensure for all accidents previously evaluated that the fuel cladding will be protected from failure due to transition boiling. The proposed change does not affect plant operation or any procedural or administrative controls on plant operation that affect the functions of preventing or mitigating any accidents previously evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Page 4

TSTF-564, Rev. 2

2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?

Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. The proposed change will not affect the design function or operation of any structures, systems or components (SSCs). No new equipment will be installed. As a result, the proposed change will not create any credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. This will result in a change to a safety limit, but will not result in a significant reduction in the margin of safety provided by the safety limit. As discussed in the application, changing the SLMCPR methodology to one based on a 95%

probability with 95% confidence that no fuel rods experience transition boiling during an anticipated transient instead of the current limit based on ensuring that 99.9% of the fuel rods are not susceptible to boiling transition does not have a significant effect on plant response to any analyzed accident. The SLMCPR and the TS Limiting Condition for Operation (LCO) on MCPR continue to provide the same level of assurance as the current limits and do not reduce a margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, [LICENSEE] concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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

Page 5

TSTF-564, Rev. 2 4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

Page 6

TSTF-564, Rev. 2 Enclosure 2 Technical Specifications Proposed Changes

TSTF-564, Rev. 2 SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%

rated core flow:

THERMAL POWER shall be 25% RTP.

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10%

rated core flow:

MCPR shall be [1.07] [for two recirculation loop operation or [1.08]

for single recirculation loop operation.]

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.

General Electric BWR/4 STS 2.0-1 Rev. 4.0

TSTF-564, Rev. 2 MCPR 3.2.2 No Changes. Included for Reference 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.

APPLICABILITY: THERMAL POWER 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A.1 Restore MCPR(s) to within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limits. limits.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion POWER to < 25% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the Once within limits specified in the COLR. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 25% RTP AND

[ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter OR In accordance with the Surveillance Frequency Control Program ]

General Electric BWR/4 STS 3.2.2-1 Rev. 4.0

TSTF-564, Rev. 2 MCPR 3.2.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.2.2 Determine the MCPR limits. Once within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each completion of SR 3.1.4.1 AND Once within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each completion of SR 3.1.4.2 AND Once within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each completion of SR 3.1.4.4 General Electric BWR/4 STS 3.2.2-2 Rev. 4.0

TSTF-564, Rev. 2 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT

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:

[ The individual specifications that address core operating limits must be referenced here. The MCPR99.9% value used to calculate the LCO 3.2.2, "MCPR," limit shall be specified in the COLR.]

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:

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REVIEWERS NOTE----------------------------------------

Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the COLR will contain the complete identification for each of the Technical Specification referenced Topical Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details.

[ Identify the Topical Report(s) by number, title, date, and NRC staff approval document or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date. ]

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

[ The individual specifications that address RCS pressure and temperature limits must be referenced here. ]

General Electric BWR/4 STS 5.6-2 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2 for [both General Electric Company (GE) and Advanced Nuclear Fuel Corporation (ANF) fuel]. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

---

REVIEWER'S NOTE ------------------------------

In the Background and Applicable Safety Analysis sections, select the SLMCPR95/95 discussion or the 99.9% of the fuel rods discussion as the applicable SL 2.1.1.2 basis.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e.,

MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. [This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as SLMCPR95/95, which corresponds to a 95% probability at a 95% confidence level (the 95/95 MCPR criterion) that transition boiling will not occur.] [The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core are not susceptible to boiling transitiondo not experience transition boiling].

General Electric BWR/4 STS B 2.1.1-1 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

General Electric BWR/4 STS B 2.1.1-2 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. [The Tech Spec SL is set generically on a fuel ANALYSES product MCPR correlation basis as the MCPR which corresponds to a 95% probability at a 95% confidence level that transition boiling will not occur, referred to as SLMCPR95/95] [The reactor core SLs are established to preclude violation of the fuel design criterion that a MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be susceptible to boiling transitionexpected to experience the onset of transition boiling.]

The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR limit.

2.1.1.1a Fuel Cladding Integrity [General Electric Company (GE) Fuel]

GE critical power correlations are applicable for all critical power calculations at pressures 785 psig and core flows 10% of rated flow.

For operation at low pressures or low flows, another basis is used, as follows:

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi. Analyses (Ref. 2) show that with a bundle flow of 28 x 103 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be > 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.

With the design peaking factors, this corresponds to a THERMAL POWER > 50 % RTP. Thus, a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psig is conservative.

2.1.1.1b Fuel Cladding Integrity [Advanced Nuclear Fuel Corporation (ANF) Fuel]

The use of the XN-3 correlation is valid for critical power calculations at pressures > 580 psig and bundle mass fluxes > 0.25 x 106 lb/hr-ft2 (Ref. 3). For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a relatively high power and potentially can approach a critical heat flux condition. For the General Electric BWR/4 STS B 2.1.1-3 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 ANF 9x9 fuel design, the minimum bundle flow is > 30 x 103 lb/hr. For the ANF 8x8 fuel design, the minimum bundle flow is > 28 x 103 lb/hr. For all designs, the coolant minimum bundle flow and maximum flow area are General Electric BWR/4 STS B 2.1.1-4 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) such that the mass flux is always > 0.25 x 106 lb/hr-ft2. Full scale critical power tests taken at pressures down to 14.7 psia indicate that the fuel assembly critical power at 0.25 x 106 lb/hr-ft2 is approximately 3.35 MWt.

At 25% RTP, a bundle power of approximately 3.35 MWt corresponds to a bundle radial peaking factor of > 3.0, which is significantly higher than the expected peaking factor. Thus, a THERMAL POWER limit of 25% RTP for reactor pressures < 785 psig is conservative.

2.1.1.2a MCPR [GE and Westinghouse Fuel]

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. [The Technical Specification SL value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent. The value is based on the Critical Power Ratio (CPR) data statistics and a 95% probability with 95% confidence that rods are not susceptible to boiling transition, referred to as MCPR95/95.]

---

Reviewer's Note --------------------------------

The MCPR95/95 Values by Vendor and Fuel Product Type:

Vendor Fuel Type MCPR95/95 Global GE14 1.06 Nuclear Fuel Global GNF2 1.07 Nuclear Fuel Global GNF3 1.07 Nuclear Fuel Westinghouse Optima2 1.06

[The SL is based on [Fuel Type] fuel. For cores with a single fuel product line, the SLMCPR95/95 is the MCPR95/95 for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR values for the fuel product lines that are fresh or once-burnt at the start of the cycle.]

[However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which General Electric BWR/4 STS B 2.1.1-5 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.]

2.1.1.2b MCPR [ANF Fuel]

The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,

MCPR = 1.00) and the MCPR SL is based on a detailed statistical General Electric BWR/4 STS B 2.1.1-6 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty inherent in the XN-3 critical power correlation. Reference 3 describes the methodology used in determining the MCPR SL.

The XN-3 critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power, as evaluated by the correlation, is within a small percentage of the actual critical power being estimated. As long as the core pressure and flow are within the range of validity of the XN-3 correlation, the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. Still further conservatism is induced by the tendency of the XN-3 correlation to overpredict the number of rods in boiling transition.

These conservatisms and the inherent accuracy of the XN-3 correlation provide a reasonable degree of assurance that there would be no transition boiling in the core during sustained operation at the MCPR SL.

If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised. Significant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach. Much of the data indicate that BWR fuel can survive for an extended period of time in an environment of boiling transition.

2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

General Electric BWR/4 STS B 2.1.1-7 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 BASES SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs.

SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria,"

limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. NEDE-24011-P-A (latest approved revision).

3. XN-NF524(A), Revision 1, November 1983.

4. 10 CFR 100.

General Electric BWR/4 STS B 2.1.1-8 Rev. 4.0

TSTF-564, Rev. 2 MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2. The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs), and that 99.9% of the fuel rods are not susceptible to boiling transition if the limit is not violated.

Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref. 1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE ----------------------------------- REVIEWER'S NOTE -------------------------------

SAFETY Incorporate the MCPR95/95 discussion if applicable.

ANALYSES ---------------------------------------------------------------------------------------------.--

The analytical methods and assumptions used in evaluating the AOOs to establish the operating limit MCPR are presented in References 2, 3, 4, 5, 6, 7, and 8. To ensure that the MCPR Safety Limit (SL) is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (CPR). When the largest CPR is combined with added to the [SL] MCPR[99.9%] SL, the required operating limit MCPR is obtained.

[MCPR99.9% is determined to ensure more than 99.9% of the fuel rods in the core are not susceptible to boiling transition using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved Critical General Electric BWR/4 STS B 3.2.2-1 Rev. 4.0

TSTF-564, Rev. 2 MCPR B 3.2.2 Power correlations. Details of the MCPR99.9% calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties and the nominal values of the parameters used in the MCPR99.9%

statistical analysis.]

The MCPR operating limits are derived from [the MCPR99.9% value and]

the transient analysis, and are dependent on the operating core flow and power state (MCPRf and MCPRp, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Refs. 6, 7, and 8). Flow dependent MCPR limits are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Ref. 9) to analyze slow flow runout transients. The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.

General Electric BWR/4 STS B 3.2.2-2 Rev. 4.0

TSTF-564, Rev. 2 MCPR B 3.2.2 BASES APPLICABLE SAFETY ANALYSES (continued)

Power dependent MCPR limits (MCPRp) are determined by approved transient analysis modelsmainly by the one dimensional transient code (Ref. 10). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPRp operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The MCPR operating limits specified in the COLR [(MCPR99.9% value, MCPRf values, and MCPRp values)] are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the larger of the MCPRf and MCPRp limits[, which are based on the MCPR99.9% limit specified in the COLR.]

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.

Statistical analyses indicate that the nominal value of the initial MCPR expected at 25% RTP is > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels

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

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

Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is normally sufficient to restore the MCPR(s) to within its limits and is acceptable based on the General Electric BWR/4 STS B 3.2.2-3 Rev. 4.0

TSTF-564, Rev. 2 MCPR B 3.2.2 low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.

General Electric BWR/4 STS B 3.2.2-4 Rev. 4.0

TSTF-564, Rev. 2 MCPR B 3.2.2 BASES ACTIONS (continued)

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

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

[ The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation.

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

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REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

---

]

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

SR 3.2.2.2 determines the value of , which is a measure of the actual scram speed distribution compared with the assumed distribution. The MCPR operating limit is then determined based on an interpolation General Electric BWR/4 STS B 3.2.2-5 Rev. 4.0

TSTF-564, Rev. 2 MCPR B 3.2.2 BASES SURVEILLANCE REQUIREMENTS (continued) between the applicable limits for Option A (scram times of LCO 3.1.4, "Control Rod Scram Times") and Option B (realistic scram times) analyses. The parameter must be determined once within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each set of scram time tests required by SR 3.1.4.1, SR 3.1.4.2, and SR 3.1.4.4 because the effective scram speed distribution may change during the cycle or after maintenance that could affect scram times. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is acceptable due to the relatively minor changes in expected during the fuel cycle.

REFERENCES 1. NUREG-0562, June 1979.

2. NEDO-24011-P-A, "General Electric Standard Application for Reactor Fuel" (latest approved version).

3. FSAR, Chapter [4].

4. FSAR, Chapter [6].

5. FSAR, Chapter [15].

6. [Plant specific single loop operation].

7. [Plant specific load line limit analysis].

8. [Plant specific Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvements (ARTS) Program].

9. NEDO-30130-A, "Steady State Nuclear Methods," May 1985.

10. NEDO-24154, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," October 1978.

General Electric BWR/4 STS B 3.2.2-6 Rev. 4.0

TSTF-564, Rev. 2 SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%

rated core flow:

THERMAL POWER shall be 25% RTP.

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10%

rated core flow:

MCPR shall be [1.07] [for two recirculation loop operation or [1.08]

for single recirculation loop operation.]

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.

General Electric BWR/6 STS 2.0-1 Rev. 4.0

TSTF-564, Rev. 2 MCPR 3.2.2 No Changes. Included for Reference 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.

APPLICABILITY: THERMAL POWER 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A.1 Restore MCPR(s) to within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limits. limits.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion POWER to < 25% RTP.

Time not met.

General Electric BWR/6 STS 3.2.2-1 Rev. 4.0

TSTF-564, Rev. 2 MCPR 3.2.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the Once within limits specified in the COLR. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 25% RTP AND

[ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter OR In accordance with the Surveillance Frequency Control Program ]

General Electric BWR/6 STS 3.2.2-2 Rev. 4.0

TSTF-564, Rev. 2 Reporting Requirements 5.6 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:

[ The individual specifications that address core operating limits must be referenced here. The MCPR99.9% value used to calculate the LCO 3.2.2, "MCPR," limit shall be specified in the COLR.]

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:

---

REVIEWERS NOTE----------------------------------------

Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the COLR will contain the complete identification for each of the Technical Specification referenced Topical Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details.

[ Identify the Topical Report(s) by number, title, date, and NRC staff approval document or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date. ]

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

[ The individual specifications that address RCS pressure and temperature limits must be referenced here. ]

General Electric BWR/6 STS 5.6-2 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2 for [both General Electric Company (GE) and Advanced Nuclear Fuel Corporation (ANF) fuel]. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

---

REVIEWER'S NOTE ------------------------------

In the Background and Applicable Safety Analysis sections, select the SLMCPR95/95 discussion or the 99.9% of the fuel rods discussion as the applicable SL 2.1.1.2 basis.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e.,

MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. [This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as SLMCPR95/95, which corresponds to a 95% probability at a 95% confidence level (the 95/95 MCPR criterion) that transition boiling will not occur.] [The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core are not susceptible to boiling transitiondo not experience transition boiling.]

General Electric BWR/6 STS B 2.1.1-1 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

General Electric BWR/6 STS B 2.1.1-2 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. [The Tech Spec SL is set generically on a fuel ANALYSES product MCPR correlation basis as the MCPR which corresponds to a 95% probability at a 95% confidence level that transition boiling will not occur, referred to as SLMCPR95/95.] [The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core are not susceptible to boiling transitionwould not be expected to experience the onset of transition boiling.]

The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR limit.

2.1.1.1a Fuel Cladding Integrity [General Electric Company (GE) Fuel]

GE critical power correlations are applicable for all critical power calculations at pressures 785 psig and core flows 10% of rated flow.

For operation at low pressures or low flows, another basis is used, as follows:

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi. Analyses (Ref. 2) show that with a bundle flow of 28 x 103 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be > 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.

With the design peaking factors, this corresponds to a THERMAL POWER > 50% RTP. Thus, a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psig is conservative.

2.1.1.1b Fuel Cladding Integrity [Advanced Nuclear Fuel Corporation (ANF) Fuel]

The use of the XN-3 correlation is valid for critical power calculations at pressures > 580 psig and bundle mass fluxes > 0.25 x 106 lb/hr-ft2 (Ref. 3). For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a General Electric BWR/6 STS B 2.1.1-3 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 relatively high power and potentially can approach a critical heat flux condition. For the ANF 9x9 fuel design, the minimum bundle flow General Electric BWR/6 STS B 2.1.1-4 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) is > 30 x 103 lb/hr. For the ANF 8x8 fuel design, the minimum bundle flow is > 28 x 103 lb/hr. For all designs, the coolant minimum bundle flow and maximum flow area are such that the mass flux is always

> 0.25 x 106 lb/hr-ft2. Full scale critical power tests taken at pressures down to 14.7 psia indicate that the fuel assembly critical power at 0.25 x 106 lb/hr-ft2 is approximately 3.35 MWt. At 25% RTP, a bundle power of approximately 3.35 MWt corresponds to a bundle radial peaking factor of > 3.0, which is significantly higher than the expected peaking factor. Thus, a THERMAL POWER limit of 25% RTP for reactor pressures < 785 psig is conservative.

2.1.1.2a MCPR [GE and Westinghouse Fuel]

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. [The Technical Specification SL value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent. The value is based on the Critical Power Ratio (CPR) data statistics and a 95% probability with 95% confidence that rods are not susceptible to boiling transition, referred to as MCPR95/95.]

---

Reviewer's Note --------------------------------

The MCPR95/95 Values by Vendor and Fuel Product Type:

Vendor Fuel Type MCPR95/95 Global GE14 1.06 Nuclear Fuel Global GNF2 1.07 Nuclear Fuel Global GNF3 1.07 Nuclear Fuel Westinghouse Optima2 1.06

[The SL is based on [Fuel Type] fuel. For cores with a single fuel product line, the SLMCPR95/95 is the MCPR95/95 for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR values for the fuel product lines that are fresh or once-burnt at the start of the cycle.]

General Electric BWR/6 STS B 2.1.1-5 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1

[However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.]

General Electric BWR/6 STS B 2.1.1-6 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) 2.1.1.2b MCPR [ANF Fuel]

The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,

MCPR = 1.00) and the MCPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty inherent in the XN-3 critical power correlation. Reference 3 describes the methodology used in determining the MCPR SL.

The XN-3 critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power, as evaluated by the correlation, is within a small percentage of the actual critical power being estimated. As long as the core pressure and flow are within the range of validity of the XN-3 correlation, the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. Still further conservatism is induced by the tendency of the XN-3 correlation to overpredict the number of rods in boiling transition.

These conservatisms and the inherent accuracy of the XN-3 correlation provide a reasonable degree of assurance that there would be no transition boiling in the core during sustained operation at the MCPR SL.

If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised. Significant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach. Much of the data indicate that BWR fuel can survive for an extended period of time in an environment of boiling transition.

2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated General Electric BWR/6 STS B 2.1.1-7 Rev. 4.0

TSTF-564, Rev. 2 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

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

SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria,"

limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and the probability of an accident occurring during this period is minimal.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. NEDE-24011-P-A, (latest approved revision).

3. XN-NF524(A), Revision 1, November 1983.

4. 10 CFR 100.

General Electric BWR/6 STS B 2.1.1-8 Rev. 4.0

TSTF-564, Rev. 2 MCPR B 3.2.2 B 3.2 POWER DISTRIBUTIONS LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs), and that 99.9% of the fuel rods are not susceptible to boiling transition if the limit is not violated.

Although fuel damage does not necessarily occur if a fuel rod actually experiences boiling transition (Ref. 1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE ----------------------------------- REVIEWER'S NOTE -------------------------------

SAFETY Incorporate the MCPR95/95 discussion if applicable.

ANALYSES ------------------------------------------------------------------------------------------------

The analytical methods and assumptions used in evaluating the AOOs to establish the operating limit MCPR are presented in the FSAR, Chapters 4, 6, and 15, and References 2, 3, 4, and 5. To ensure that the MCPR Safety Limit (SL) is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (CPR). When the largest CPR is combined with added to the [SL]MCPR[99.9%] SL, the required operating limit MCPR is obtained.

[MCPR99.9% is determined to ensure more than 99.9% of the fuel rods in the core are not susceptible to boiling transition using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved Critical General Electric BWR/6 STS B 3.2.2-1 Rev. 4.0

TSTF-564, Rev. 2 MCPR B 3.2.2 Power correlations. Details of the MCPR99.9% calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties and the nominal values of the parameters used in the MCPR99.9%

statistical analysis.]

The MCPR operating limits are derived from [the MCPR99.9% value and]

the transient analysis, and are dependent on the operating core flow and power state (MCPRf and MCPRp, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Refs. 3, 4, and 5). Flow dependent MCPR limits are determined by steady state thermal hydraulic methods using the three dimensional BWR simulator code (Ref. 6) and the multichannel thermal hydraulic code (Ref. 7). MCPRf General Electric BWR/6 STS B 3.2.2-2 Rev. 4.0

TSTF-564, Rev. 2 MCPR B 3.2.2 BASES APPLICABLE SAFETY ANALYSES (continued) curves are provided based on the maximum credible flow runout transient for Loop Manual and Non Loop Manual operation. The result of a single failure or single operator error during Loop Manual operation is the runout of only one loop because both recirculation loops are under independent control. Non Loop Manual operational modes allow simultaneous runout of both loops because a single controller regulates core flow.

Power dependent MCPR limits (MCPRp) are determined by approved transient analysis modelsthe three dimensional BWR simulator code and the one dimensional transient code (Ref. 8). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, high and low flow MCPRp operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The MCPR operating limits specified in the COLR [(MCPR99.9% value, MCPRf values, and MCPRp values)] are the result of the Design Basis Accident (DBA) and transient analysis. The MCPR operating limits are determined by the larger of the MCPRf and MCPRp limits[, which are based on the MCPR99.9% limit specified in the COLR].

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a slow recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.

Statistical analyses documented in Reference 9 indicate that the nominal value of the initial MCPR expected at 25% RTP is > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies (Ref. 5) encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor (IRM) provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern.

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

General Electric BWR/6 STS B 3.2.2-3 Rev. 4.0

TSTF-564, Rev. 2 MCPR B 3.2.2 BASES ACTIONS A.1 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met.

Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is normally sufficient to restore the MCPR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.

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

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

[ The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation.

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

---

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

---

]

General Electric BWR/6 STS B 3.2.2-4 Rev. 4.0

TSTF-564, Rev. 2 MCPR B 3.2.2 BASES REFERENCES 1. NUREG-0562, June 1979.

2. [Plant specific current cycle safety analysis].

3. FSAR, [Appendix 15B].

4. FSAR, [Appendix 15C].

5. FSAR, [Appendix 15D].

6. XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors, Neutronics Methods for Design and Analysis," Volume 1 (as supplemented).

7. XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX Thermal Limits Methodology Summary Description," Volume 3, Revision 2, January 1987.

8. XN-NF-79-71(P), "Exxon Nuclear Plant Methodology for Boiling Water Reactors," Revision 2, November 1981.

9. "BWR/6 Generic Rod Withdrawal Error Analysis," General Electric Standard Safety Analysis Report, GESSAR-II, Appendix 15B.

General Electric BWR/6 STS B 3.2.2-5 Rev. 4.0