ML18208A146
| ML18208A146 | |
| Person / Time | |
|---|---|
| Site: | Technical Specifications Task Force |
| Issue date: | 10/03/2018 |
| From: | NRC/NRR/DSS |
| To: | |
| Honcharik M, NRR/DSS, 301-415-1774 | |
| Shared Package | |
| ML18207A380 | List: |
| References | |
| CAC MG0161, EPID L-2017-PMP-0007, TSTF-564, Rev 1 | |
| Download: ML18208A146 (9) | |
Text
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 specification (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 The U.S. Nuclear Regulatory Commission (NRC) staff transmitted requests for additional 35 information (RAIs) to the TSTF by letter dated April 12, 2018 (ADAMS Accession 36 No. ML18095A229). Responses to these RAIs were transmitted from the TSTF by letter dated 37 May 29, 2018 (ADAMS Accession No. ML18149A320).
38 39 1.1 Background on Boiling Transition 40 41 During steady-state operation in a BWR, most of the coolant in the core is in a flow regime 42 known as annular flow. In this flow regime, a thin liquid film is pushed up the surface of the fuel 43 rod cladding by the bulk coolant flow, which is mostly water vapor with some liquid water 44 droplets. This provides effective heat removal from the cladding surface; however, under 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 certain conditions, the annular film may dissipate, which reduces the heat transfer and results in 2 an increase in fuel cladding surface temperature. This phenomenon is known as boiling 3 transition or dryout. The elevated surface temperatures resulting from dryout may cause fuel 4 cladding damage or failure.
5 6 1.2 Background on Critical Power Correlations 7
8 For a given set of reactor operating conditions (pressure, flow, etc.), dryout will occur on a fuel 9 assembly at a certain power, known as the critical power. Because the phenomena associated 10 with boiling transition are complex and difficult to model purely mechanistically, 11 thermal-hydraulic test campaigns are undertaken using electrically heated prototypical fuel 12 bundles to establish a comprehensive database of critical power measurements for each BWR 13 fuel product. These data are then used to develop a critical power correlation that can be used 14 to predict the critical power for assemblies in operating reactors. This prediction is usually 15 expressed as the ratio of the actual assembly power to the critical power predicted using the 16 correlation, known as the CPR.
17 18 One measure of the correlations predictive capability is based on its validation relative to the 19 test data. For each point j in a correlations test database, the experimental critical power ratio 20 (ECPR) is defined as the ratio of the measured critical power to the calculated critical power,2 21 or:
22 Measured Critical Power 23 ECPR =
Calculated Critical Power 24 25 For ECPR values less than or equal to 1, the calculated critical power is greater than the 26 measured critical power and the prediction is considered to be non-conservative. Because the 27 measured critical power includes random variations due to various uncertainties, evaluating the 28 ECPR for all of the points in the dataset (or, ideally, a subset of points that were not used in the 29 correlations development) results in a probability distribution. This ECPR distribution allows the 30 predictive uncertainty of the correlation to be determined. This uncertainty can then be used to 31 establish a limit above which there can be assumed that boiling transition will not occur (with a 32 certain probability and confidence level).
33 34 1.3 Background on Thermal-Hydraulic Safety Limits 35 36 To protect against boiling transition, BWRs have implemented an SL on the CPR, known as the 37 minimum critical power ratio (MCPR) SL. As discussed in NUREG-1433 and NUREG-1434 for 38 General Electric BWR designs, the current basis of the MCPR SL is to prevent 99.9 percent of 39 the fuel in the core from being susceptible to boiling transition. This limit is typically developed 40 by considering various cycle-specific power distributions and uncertainties, and is highly 41 dependent on the cycle-specific radial power distribution in the core. As such, the limit may 42 need to be updated as frequently as every cycle.
43 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