ML25248A330
| ML25248A330 | |
| Person / Time | |
|---|---|
| Site: | Westinghouse |
| Issue date: | 11/25/2025 |
| From: | Lenning K Licensing Processes Branch |
| To: | Ewing J Westinghouse |
| References | |
| EPID L-2024-NTR-0005 WCAP-18850-P/NP | |
| Download: ML25248A330 (0) | |
Text
Enclosure 1 U. S. NUCLEAR REGULATORY COMMISSION FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR WESTINGHOUSE ELECTRIC COMPANY TOPICAL REPORT WCAP-18850-P/NP, REVISION 0, ADAPTATION OF THE FULL SPECTRUM LOCA (FSLOCA) EVALUATION METHODOLOGY TO PERFORM ANALYSIS OF CLADDING RUPTURE FOR HIGH BURNUP FUEL DOCKET NO. 99902038 EPID: L-2024-NTR-0005
1.0 INTRODUCTION
By letter dated February 29, 2024, Westinghouse Electric Company (Westinghouse) submitted to the U.S. Nuclear Regulatory Commission (NRC) Topical Report (TR) WCAP-18850-P/NP, Revision 0, Adaptation of the FULL SPECTRUM LOCA' (FSLOCA') Evaluation Methodology to Perform Analysis of Cladding Rupture for High Burnup Fuel (WCAP-18850-P/NP, Revision 0) (Ref. 1). WCAP-18850-P/NP, Revision 0, proposes to update the WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) (FSLOCA EM) (Ref. 2) methodology to perform cladding rupture evaluations to demonstrate that cladding of fuel rods susceptible to fine fragmentation will not rupture during a postulated loss-of-coolant accident (LOCA). This TR also extends the applicability of the FSLOCA EM to high burnup and high enrichments, as described in limitations and conditions (L&Cs) 3 and 11.
As outlined in WCAP-18850-P/NP, Revision 0, the proposed cladding rupture methodology would affect analyses and evaluations that support the demonstration of reactor safety during a LOCA in several areas. The TR discusses the following:
Phenomena Identification Ranking Table (PIRT) Review Fuel Rod Model Updates Kinetics and Decay Heat Model Updates Fuel Rod Cladding Rupture Calculational Methodology Sensitivity Analyses By letters dated January 31, 2025 (Ref. 3), and March 31, 2025 (Ref. 4), Westinghouse submitted substantial information in the form of responses to the NRC requests for additional information (RAIs).
The NRC staff's safety evaluation (SE) assesses the information contained in WCAP-18850P/NP, Revision 0, as well as the supplementary submittals discussed above.
The phenomena of fuel fragmentation, relocation, and dispersal (FFRD) are presumed to be credible as fuel rods are operated into higher burnup ranges (beginning at approximately 55 gigawatt days per metric ton of uranium (GWd/MTU) nodal exposure) with sufficiently high local power in those rods. The NRC staff assessed whether the increased initial U-235 loading (leading to higher local power at higher burnups) and extended licensed burnup limits (leading to increased fuel pellet cracking) would result in FFRD of fuel to an extent that could adversely impact the mitigation of a postulated LOCA. Because Westinghouses approach is intended to prevent the rupture of susceptible fuel rods, the NRC staff focused its review on Westinghouses modeling approaches for phenomena occurring prior to the point of cladding rupture and fuel dispersal.
FFRD phenomena can occur as result of postulated reactor coolant system (RCS) pipe breaks.
As the ceramic fuel pellet is irradiated the sintered uranium dioxide (UO2) matrix begins to form cracks in its structure. This process continues creating smaller and smaller fragments of the pellet as irradiation continues over its lifetime. Under normal operations the fragments are held together and contained by the fuel rod cladding, especially once pellet to cladding contact is attained. Pellet fragments are also prevented from relocating axially in any significant way because of the other pellets in the pellet stack. The major concern at lower burnups is pellet-clad interaction (PCI) cladding failures that can occur as result of local stresses created by rapid local power changes.
During a postulated LOCA, the RCS is rapidly depressurized resulting in a high differential pressure across the fuel rod cladding. The high rod internal pressure is from the initial rod fill gas, the fission gases that are released from the ceramic fuel pellet matrix during a LOCA event, and the thermal expansion of the gas as it heats up. At higher burnups, higher local power histories, and higher initial local power at the time of the events, these effects are maximized. If the fuel rod cladding is at a sufficiently high temperature and the differential pressure exceeds the mechanical limitations of the cladding material and structure, the cladding will begin to deform outward in a plastic manner creating a bulge or local ballooning in the cladding. During this process, the ceramic pellet structure is no longer contained and may allow the finely fragmented pellet pieces to relocate and pack into the ballooned area. If sufficient differential pressure across the cladding is achieved during the event that cladding strain limits are exceeded, then the cladding may rupture, expelling the finely fragmented pellet pieces into the RCS.
Given the difficulty of assessing the impacts of dispersal and the potential risk of dispersal to plant personnel and the public, Westinghouse has decided the best way to evaluate this phenomenon is to prevent dispersal as the evaluation acceptance criterion for their fuel product lines.
This methodology builds upon decades of proven neutronics, thermal-hydraulic, thermal-mechanical, non-LOCA, and LOCA transient methods reviewed and approved by the NRC staff to demonstrate with reasonable assurance susceptible fuel rods will not rupture during a range of postulated RCS breaks. The NRC staff reviewed the changes necessary to Westinghouses existing methods as well as the new research and information provided to justify the use of the WCAP-18850-P/NP, Revision 0, EM for FFRD at higher enrichments and fuel burnup. This review is documented in this SE with L&Cs for the EM as applied to certain plant and fuel designs.
2.0 REGULATORY EVALUATION
Section 1.3 of WCAP-18850-P/NP, Revision 0, identifies significant regulatory guidance that Westinghouse deemed pertinent to its methodology for cladding rupture calculations. Key applicable regulatory requirements are also noted in this and other sections of WCAP-18850-P/NP, Revision 0.
Applicable regulatory requirements and guidance identified by the NRC staff are listed in the following subsections of this SE, organized by topical area.
2.1 Fuel Design Regulatory requirements and guidance for the design of nuclear fuel are intended to ensure acceptable behavior under normal operation, anticipated transients, and postulated accidents:
General Design Criterion (GDC) 10, in Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs).
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition (SRP), SRP Section 4.2, Fuel System Design (Ref. 5), provides review guidance to the NRC staff concerning the establishment of specified acceptable fuel design limits to assure:
o the fuel system is not damaged as a result of normal operation and AOOs o fuel system damage is never so severe as to prevent control rod insertion when it is required o the number of fuel rod failures is not underestimated for postulated accidents o core coolability is maintained 2.2 Loss-of-Coolant Accident Regulatory requirements and guidance addressing the LOCA are relevant to the present review because the methodology discussed in WCAP-18850-P/NP, Revision 0, describes an evaluation model for demonstrating satisfactory emergency core cooling system (ECCS) performance and cladding integrity with respect to fuel in the high burnup range up to [ ]
WCAP-18850-P/NP, Revision 0, proposes [
] As described in Section 7.3 of WCAP-18850-P/NP, Revision 0, this method does not propose to directly satisfy the requirements of 10 CFR 50.46, although it provides a technical basis to help address concerns about the impact of fuel dispersal on core coolability.
GDC 35 establishes minimum requirements for water-cooled nuclear power plants with respect to emergency core cooling. In particular, GDC 35 requires abundant core cooling capable of transferring heat from the reactor core following any loss of reactor coolant, such that fuel and cladding damage that could interfere with continued effective core cooling is prevented and cladding metal-water reaction is limited to negligible amounts.
GDC 35 further incorporates design requirements for ECCSs, addressing redundancy, leak detection and isolation, and functionality both with and without offsite power.
Appendix K to 10 CFR Part 50 consists of two parts, the first of which specifies required and acceptable features of LOCA evaluation models, and the second of which specifies documentation required for LOCA evaluation models. Appendix K incorporates requirements for modeling significant physical phenomena throughout all phases of the LOCA event, including relevant heat sources, fuel rod performance, and thermal-hydraulic behavior.
In addition to these regulatory requirements, the NRC staffs review further considered significant regulatory guidance for the LOCA including the following:
SRP Section 15.6.5, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary (Ref. 6), provides guidance to support the NRC staff's review of LOCA analyses.
Regulatory Guide (RG) 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance (Ref. 7), provides guidance concerning realistic modeling of the LOCA with explicit accounting for relevant uncertainties.
SRP Section 15.0.2, Review of Transient and Accident Analysis Methods (Ref. 8),
provides guidance to support the NRC staff's review of analytical evaluation models used to perform safety analyses for nuclear reactors.
RG 1.203, Transient and Accident Analysis Methods (Ref. 9), provides regulatory guidance to licensees and applicants concerning the development and assessment of evaluation models used to perform safety analyses for nuclear reactors.
SRP Chapter 15.0, Introduction - Transient and Accident Analyses, which provides guidance to support the NRC staff's review of analytical evaluation models used to perform safety analyses for nuclear reactors, as well as specific guidance for evaluating different types of AOOs, as categorized by their impact on the reactor response.
3.0 TECHNICAL EVALUATION
3.1 Phenomena Identification and Ranking (WCAP-18850-P/NP, Revision 0, Section 2)
Westinghouse performed a review of the PIRT discussed in the FSLOCA EM and an industry PIRT for phenomena that are related to cladding rupture in Section 2 of WCAP-18850-P/NP, Revision 0. The goal was to identify phenomena where current code models or modeling approaches should be reviewed for adequacy relative to cladding rupture calculations for high burnup and assign new rankings as appropriate.
Since the approval of the FSLOCA EM, Westinghouse has been working on the extension of the FSLOCA EM to higher burnup and higher initial fuel rod enrichments to support industry needs.
That extension includes addressing new phenomena associated with high burnup fuel rods response to a postulated LOCA, such as the potential for FFRD. While the licensed FSLOCA EM does account for fuel fragmentation and relocation, it does not account for fuel dispersal.
Westinghouse, therefore, has proposed the [
] As such, these PIRTs do not include any changes that may be required to account for fuel rod geometrical changes beyond rupture or tracking fuel material outside the fuel rod and their accompanying impacts.
3.1.1 FULL SPECTRUM Loss-of-Coolant Accident Methodology Phenomena Identification Ranking Table Review 3.1.1.1 Fuel Rod Stored Energy Westinghouse identified that stored energy requires additional considerations for cladding rupture calculations. These considerations are discussed further in Sections 3.2.1 and 3.2.4 of this SE.
Clad Oxidation Westinghouse maintains the high importance PIRT ranking of cladding oxidation for cladding rupture calculations. Westinghouse asserts that transient cladding oxidation [
] Therefore, the NRC staff determined that [ ] is acceptable for evaluation of transient cladding oxidation for cladding rupture calculations.
Decay Heat Westinghouse identified that its WCOBRA/TRAC-TF2 decay heat models need to be updated to be applicable up to the proposed enrichment and burnup limits. This evaluation is discussed further in Section 3.3 of this SE.
Clad Deformation Westinghouse identified that the cladding deformation models need to be updated for use in cladding rupture calculations. This evaluation is discussed further in Section 3.2.2 of this SE.
3.1.1.2 Core Critical Heat Flux Westinghouse states that [
] for cladding rupture calculations and high burnup evaluations. Westinghouse further states that [
] The NRC staff
determined that [ ] is acceptable because the expanded applicability in terms of burnup and enrichment will not affect the predictive capability of Westinghouse CHF methods for the FSLOCA EM.
Post-Critical Heat Flux Heat Transfer/Steam Cooling Westinghouse states that the cladding rupture evaluation and increased applicability to high burnup and enrichment will not affect post-CHF heat transfer and steam cooling because the
[
] The NRC staff determined that the disposition of post-CHF heat transfer and steam cooling is acceptable.
Rewet/Tmin Westinghouse states that the cladding rupture method and increased applicability for high burnup and enrichment require no additional considerations with respect to rewet and Tmin for similar purposes as discussed above for CHF. The NRC staff determined that the disposition of rewet and Tmin is acceptable.
Heat Transfer to Covered Core Westinghouse states that heat transfer to a covered core requires no additional considerations for cladding rupture calculations because cladding rupture will not occur when the core is covered. As seen in the sensitivity studies, [
] Therefore, the NRC staff determined that the disposition of the heat transfer to a covered core PIRT ranking is acceptable.
Radiation Heat Transfer Westinghouse states that the cladding rupture analysis method does not affect the PIRT ranking of radiation heat transfer. Radiation heat transfer refers to the surface-to-surface transfer of thermal energy and is affected most by material properties and surface temperature, neither of which are affected by the proposed cladding rupture methodology. Therefore, the NRC staff determined that the disposition of the radiation heat transfer PIRT ranking is acceptable.
3-D Flow/Core Natural Circulation Westinghouse states that the 3-D flow and core natural circulation parameters are accounted for using the current core nodalization scheme and no additional considerations are needed for the cladding rupture method. The NRC staff determined that the disposition of the 3-D flow and core natural circulation PIRT ranking is acceptable because the cladding rupture method is not expected to significantly alter the core-wide flow characteristics.
Void Generation/Void Distribution Westinghouse states that the PIRT ranking of void generation and distribution for the cladding rupture method and increased burnup and enrichment is unaffected. The NRC staff determined that the disposition of the void generation and distribution PIRT ranking is acceptable because
the cladding rupture method is not expected to significantly alter the core-wide flow characteristics.
Entrainment/De-entrainment Westinghouse states that the PIRT ranking of entrainment and de-entrainment is unaffected by the analysis of cladding rupture. The NRC staff determined the disposition of the entrainment and de-entrainment PIRT ranking is acceptable because the cladding rupture method is not expected to significantly alter the coolant flow characteristics important to this phenomenon.
Flow Reversal/Stagnation Westinghouse states that the PIRT ranking of flow reversal and stagnation is unaffected by the analysis of cladding rupture. The FSLOCA EM treatment of these phenomena is dependent on
[ ] in the cladding rupture method. Therefore, the NRC staff determined that the disposition of the flow reversal and stagnation PIRT ranking is acceptable because the current FSLOCA EM approach is unaffected.
Flow Resistance Westinghouse states that the PIRT ranking of flow resistance is unaffected by the analysis of cladding rupture. The NRC staff determined that the disposition of the flow resistance PIRT ranking is acceptable because the current FSLOCA EM approach remains applicable and is not impacted by changes proposed in the cladding rupture method.
Water Storage in Barrel/Baffle Region Westinghouse states that the PIRT ranking of water storage in the barrel and baffle regions is unaffected by the analysis of cladding rupture. The NRC staff determined the disposition of the water storage in the barrel and baffle region is acceptable because the cladding rupture method is not expected to affect core-wide flow characteristics.
3.1.2 Industry Phenomena Identification Ranking Table Review 3.1.2.1 Plant Transient Phenomena Gas Pressure and Rod Free Volume Gas pressure and rod free volume are identified as important parameters for cladding rupture calculations. The rod internal pressure is one of primary drivers leading to the ballooning and bursting of high burnup rods. This phenomenon is discussed further in Section 3.2.6 of this SE.
Cladding Temperature Westinghouse states that parameters important to cladding temperature are already addressed in the FSLOCA EM. This phenomenon remains highly important in the cladding rupture calculations and [ ] Westinghouse asserts that no additional considerations are needed to evaluate the cladding temperature. The NRC determined that the disposition of cladding temperature is acceptable because the range of temperatures and
parameters important to cladding temperature are not treated differently in the cladding rupture method.
Burst Criteria The method described in WCAP-18850-P/NP, Revision 0, proposes to preclude cladding burst for fuel rods susceptible to fine fragmentation. Therefore, the burst criteria must be precisely defined such that [ ] can be accurately calculated. The WCAP-18850-P/NP, Revision 0, burst criteria are described in Section 3.2.3 of this SE.
Location of Burst Westinghouse states that the burst location is unimportant as WCAP-18850-P/NP, Revision 0, provides a method for evaluating if a rod will burst so as to preclude rod burst during a postulated LOCA. If a rod bursts at all it is considered a failed outcome. The NRC staff determined that the disposition of burst location is acceptable because the method intends to preclude rod burst; therefore, burst location is irrelevant.
Time-Dependent Gap-Size Heat Transfer Westinghouse maintains the importance of time-dependent gap-size heat transfer in the cladding rupture methodology. This phenomenon is discussed in detail in Sections 3.2.1 and 3.2.2 of this SE.
3.1.2.2 Transient Fuel Rod Phenomena Heat Resistance in the Gap Heat resistance in the gap was addressed by the discussion related to gap conductance in Section 3.2.1 of this SE.
Heat Resistance in the Oxide Westinghouse notes that the cladding oxide layer can increase the amount of stored energy in the fuel rod. The effects of cladding oxidation are discussed in Section 3.2.4 of this SE.
Cladding Oxidation Magnitude The cladding oxidation magnitude was dispositioned above as part of the FSLOCA PIRT review.
It was determined that [ ] for LOCA evaluations at high burnups.
Size of Burst Opening Westinghouse states that the size of burst opening is unimportant as WCAP-18850-P/NP, Revision 0, provides a method for evaluating if a rod will burst so as to preclude rod burst during a postulated LOCA. If a rod bursts at all it is considered a failed outcome. The NRC staff determined that the disposition of the size of burst opening is acceptable because the method intends to prevent rod burst; therefore, burst opening size is irrelevant.
Time of Burst Westinghouse states that the time of burst is unimportant as WCAP-18850-P/NP, Revision 0, provides a method for evaluating if a rod will burst so as to preclude rod burst during a postulated LOCA. If a rod bursts at all it is considered a failed outcome. The NRC staff determined the disposition of the time of burst is acceptable because the method intends to preclude rod burst; therefore, time of burst is irrelevant.
Conclusions The NRC staff reviewed WCAP-18850-P/NP, Revision 0, Section 2.0, Phenomena Identification and Ranking, and determined that Westinghouse has adequately identified all the important phenomena needed to establish an acceptable EM for preventing fuel dispersal by precluding cladding rupture. The NRC staff also determined that Westinghouse has identified all necessary changes to its existing approved methods and, where necessary, created new methods to model existing fuel product lines when operated in certain plant designs as outlined in Section 4.4 of this SE.
The NRC staff concluded that the information and justifications provided by Westinghouse in WCAP-18850-P/NP, Revision 0, Section 2.0, are acceptable.
3.2 WCOBRA/TRAC-TF2 Fuel Rod Model Updates (WCAP-18850-P/NP, Revision 0, Section 3)
In Section 3 of WCAP-18850-P/NP, Revision 0, Westinghouse presented an assessment of the fuel rod models identified in the associated PIRTs (Section 2 of WCAP-18850-P/NP, Revision 0).
Specifically, the models were assessed in terms of their applicability to fuel analysis for, as appropriate, a rod or assembly average burnup up to [ ] Westinghouse stated that the WCOBRA/TRAC-TF2 fuel rod models from Section 8.3 of TR WCAP-16996-P-A (Ref. 2) were maintained without modification where appropriate. The models from Section 8.3 of WCAP-16996-P-A were approved for a burnup up to a rod average of [ ] Any proposed changes in the WCOBRA/TRAC-TF2 fuel rod models were described in Section 3 of WCAP-18850-P/NP, Revision 0.
3.2.1 Pellet-Cladding Gap Conductance Model The pellet-cladding gap conductance model used in this methodology is the WCOBRA/TRAC-TF2 model that was used in the FSLOCA EM.
In Section 3.1 of WCAP-18850-P/NP, Revision 0, Westinghouse stated that the pellet-to-cladding gap conductance model is discussed in Section 8.3.2 of WCAP-16996-P-A (Ref. 2), as well as the response to Part 3 of RAI 37 of that review. Because the modeling approaches in WCOBRA/TRAC-TF2 and PAD5 are not identical, Westinghouse stated that it is not possible to ensure that the fuel rod initial stored energy and gap conductance between the two codes are both maintained identically. Westinghouse indicated that it found maintaining a prototypical prediction of stored energy to be paramount for correctly predicting the LOCA response.
Because of this, Westinghouse [
] This methodology was approved by the NRC staff for rod average burnups up to [ ] as part of the WCAP-18446-P-A, Revision 0, review (Ref. 10).
To justify the methodology for burnups up to [ ] Westinghouse performed a comparison of the gap conductance predicted by WCOBRA/TRAC-TF2 as a function of gap width at a burnup of [ ] (see response to RAI-1) and the gap conductance predicted by PAD5 as a function of gap width at [ ] For these similar burnups, the Westinghouse predictions show that the WCOBRA/TRAC-TF2 gap conductance is around
[ ] for a gap width of [ ] and around [ ] for a gap width of [ ] In comparison, the PAD5-predicted gap conductance is around
[ ] for a gap width of [ ] and around [ ] for a gap width of [ ] The information provided by Westinghouse shows that the WCOBRA/TRAC-TF2 gap conductance is lower than the PAD5 gap conductance, which is conservative.
In RAI 1, the NRC staff also questioned the use of PAD5 above its currently approved burnup limit, specifically up to [ ] In response to RAI 1, Westinghouse provided data to show that the PAD5 predictions follow well-behaved trends up to [ ] which supports its use up to this burnup level. In addition, Westinghouse re-stated the L&C that an NRC-approved fuel performance code will need to be used for actual applications using the proposed methodology. The NRC staff determined that this is acceptable.
The NRC staff performed Fuel Analysis under Steady-state and Transients (FAST) calculations to confirm the PAD5 predictions at [ ] and performed FAST calculations of gap conductance at [ ] as shown below in Figure 1. The FAST calculations of gap conductance were in reasonable agreement with the PAD5 predictions at [ ] and at
[ ] Overall, the FAST gap conductance followed the same trend as the PAD5 gap conductance, with the PAD5 results being generally more conservative than the FAST results. In all cases, the WCOBRA/TRAC-TF2 results were conservative relative to the FAST calculations.
The NRC staff's review of Westinghouses modeling of gap conductance for the WCAP-18850-P/NP, Revision 0, methodology also relies significantly upon the NRC staffs previous review of Westinghouses discussion of the modeling of gap conductance in the FSLOCA EM. In particular, the NRC staff's review of WCAP-16996-P-A (Ref. 2) found that Westinghouse had presented adequate evidence in its response to FSLOCA RAI 37 to demonstrate that, after [
] are not significant. The NRC staffs conclusion further relies upon the additional evidence provided during the review of WCAP-18446-P-A on incremental burnup, including Westinghouses response to the incremental burnup RAI 20, which demonstrated gap heat transfer coefficients of similar magnitude for PAD5 and WCOBRA/TRAC-TF2.
The NRC staff concluded Westinghouses modeling of the pellet-to-cladding gap heat conductance to be acceptable in support of calculations with the WCAP-18850-P/NP, Revision 0, methodology.
Figure 1: Comparisons of gap conductance versus gap width between PAD5, FAST, and WCOBRA/TRAC-TF2 at high burnups. The top figure shows FAST results for a flat power history at [ ] and the bottom figure shows results for a power history similar to a fuel design limit power history Cladding Deformation In Section 3.2 of WCAP-18850-P/NP, Revision 0, Westinghouse stated that the modeling of cladding deformation in WCOBRA/TRAC-TF2 is described in Section 8.4.1 of WCAP-16996-P-A, and that its modeling of cladding deformation [
]
[
]
The models proposed by Westinghouse were previously approved for [
] cladding materials for burnups up to [ ] (Ref. 10). In WCAP-18850-P/NP, Revision 0, Westinghouse proposes to extend the burnup range to
[ ] The NRC staff reviewed the data presented and determined that the hydrogen contents and oxide thicknesses used in testing conservatively cover the range of hydrogen contents and oxide thicknesses expected in AXIOM cladding for burnups up to [ ]
Because of this, the NRC staff has determined that the new proposed burnup range is acceptable.
In WCAP-18850-P/NP, Revision 0, Westinghouse proposes to model the cladding creep response of [ ] on a [ ] basis instead of a [ ] basis, to account for local phenomena of importance in LOCA and fuel rod rupture modeling. The NRC staff determined that the [ ] basis provides a more detailed picture of the fuel rod behavior when compared to the [ ] basis, which is important for the more accurate determination of fuel rod behavior. In addition, the NRC staff previously determined that the treatment of the [
] is conservative. Of these bases, the NRC staff determined that the more refined modeling approach is acceptable.
Section 3.2 of WCAP-18850-P/NP, Revision 0, briefly discusses the fuel rod nodalization used in the FSLOCA EM. In response to RAI 13, Westinghouse clarified that the [
] for the higher burnup range. This is, in part, due to
[
] Further, Westinghouse states that the [
] RAI 13 also discusses the heating behavior of the [ ] when located adjacent to a higher power feed assembly. The feed assembly may be at a higher temperature and provide some additional heating to the [ ] via radiation heat transfer and neutron and gamma heating. The effects of neutron and gamma heating are discussed further in Section 3.3 of this SE. The WCOBRA/TRAC-TF2 code [
] The NRC staff concluded that the described conservatisms bound the estimated radiation heat transfer from feed assemblies, such that the proposed methodology will overestimate the true PCT for the analyzed transient.
3.2.2 Cladding Rupture 3.2.2.1 Effect of Hydrogen Uptake in the Cladding Westinghouse stated that for burnups up to [ ] the cladding hydrogen content will remain below [ ] for AXIOM cladding. Westinghouse also stated that [
] in the context of cladding burst testing. The NRC staff agreed with these statements as documented in WCAP-18446-P-A, Revision 0, and in WCAP-18546-P-A,
Revision 0 (Refs. 10 and 13). As a result, Westinghouse performed LOCA burst testing on AXIOM test specimens [ ] and presented the results of this testing in Figures 3.3-1 through 3.3-4 of WCAP-18850-P/NP, Revision 0. The NRC staff reviewed the test results and confirmed Westinghouses statement that for [
] of AXIOM cladding. Therefore, the NRC staff determined that including both [
] to develop the rupture model is acceptable for [
] which conservatively covers the expected [ ] for burnups up to
[ ]
3.2.2.2 Cladding Rupture Models In Section 3.3.2 of WCAP-18850-P/NP, Revision 0, Westinghouse proposes a cladding rupture model based on [
] Westinghouse first presented data [
] In RAI 14, the NRC staff inquired about additional data that was mentioned but not shown in any figures. The response to RAI 14 provided the additional data, as well as a discussion of this data. In addition, Westinghouse discussed proposed updates to WCAP-18850-P/NP, Revision 0, to include the additional data. The NRC staff reviewed the expanded unirradiated and irradiated data sets as well as the burst criteria to confirm that Westinghouses assertions were correct. The NRC staff did note that the additional data slightly reduced the amount of data that is bounded by the proposed model, from 97 percent to 95 percent, but found that 95 percent is an acceptably high confidence level that the model bounds the phenomena. The NRC staff determined that the
[ ] burst criterion curve is conservative for the [ ] rod burst data, which is acceptable.
Westinghouse studied the difference between [
] The NRC staff evaluated the additional information provided and concluded that the temperature at the burst location was indeed [
] Regarding this temperature difference, Westinghouse stated in response to RAI 2 that [
] Assuming that the actual burst temperature is constant for a given rod internal pressure (with all other parameters being identical), the NRC staff agrees with this statement. Thus, the NRC staff determined that Westinghouses application of [ ]
is acceptable.
Using this approach, Westinghouse showed that the proposed rupture model is conservative relative to the data, see Figure 2-9 from Reference 4 below.
The light blue diamonds are shown as [
] is conservative relative to what Westinghouse believes are the [
] for the burst criterion data. The x axis labeled Pressure (psi) is the differential pressure across the cladding. The NRC staff concluded that [
] remained conservative and acceptable, and that the proposed rupture model is thus conservative and acceptable.
3.2.3 Fuel Rod Initialization In Section 3.4 of WCAP-18850-P/NP, Revision 0, Westinghouse stated that the same approach that was approved in Reference 2, which was determined to be [
] Based on these prior determinations and approvals and based on the fact that the PAD5 code is approved for use with AXIOM cladding up to a burnup of [ ] the NRC staff determined that fuel rod initialization for WCOBRA/TRAC-TF2 based on PAD5 is acceptable up to a rod average burnup of [ ]
Westinghouse stated that for burnups up to [
] Based on these observations, Westinghouse proposes that [
] The NRC staff evaluated Westinghouses proposal based on engineering judgement and determined that this fuel rod initialization procedure is acceptable for burnups up to [ ] but not higher without further information. This burnup limit is specified in L&C 3 of this SE.
3.2.4 Susceptibility to Fine Fragmentation In Section 3.5 of WCAP-18850-P/NP, Revision 0, Westinghouse presented fine fragmentation data from the [ ] LOCA tests.
Westinghouse initially made the argument that there was limited data that showed fine fuel fragmentation below a [
] The NRC staff examined the data presented, compared Westinghouses proposal with the thresholds defined by the NRCs Office of Nuclear Regulatory Research in Research Information Letter (RIL) 2021-13 (Ref. 14), and found that the RIL proposed a more conservative threshold than Westinghouse. Based on this finding, the NRC staff asked RAI 3 to request further justification for the proposed threshold based on [
] In its response to RAI 3, Westinghouse presented information to support [
] This new proposed threshold is comparable to the thresholds defined by the NRC in RIL 2021-13, which the NRC staff determined to be acceptable.
3.2.5 Loss-of-Coolant Accident Transient Fission Gas Release 3.2.5.1 Test Data Review and Model Development In Section 3.6.1 of WCAP-18850-P/NP, Revision 0, Westinghouse stated that the available database for transient fission gas release (FGR) [
]
Westinghouse also cited several independent sources that show that [
] The NRC staff evaluated the information and arguments presented by Westinghouse and concluded that the database used to develop the transient FGR model is conservative relative to the expected actual LOCA transient FGR and is thus an acceptable database.
The model proposed by Westinghouse is a function of [
] The NRC staff determined that this approach is conservative and should bound most locations along the fuel rod. The fact that there is a small possibility that some portions of the fuel rod would not be bounded by the proposed approach is acceptable because Westinghouse
[
]
Westinghouse proposed a [
] to bound the data used in the model development. The NRC staff reviewed the data and the model along with its uncertainty bands and confirmed that [
] This is conservative and thus acceptable. In addition, Westinghouse compared their model to available data from [
]
[
] In view of the data and model comparisons presented by Westinghouse, the staff determined that the proposed model was acceptable because it bounded all available data (after some acceptable adjustments, as described above).
The NRC staff asked RAI 4 to clarify how the model is used [
] Westinghouse responded that [
] Based on Westinghouses response to RAI 4, and the clarification edits proposed for in WCAP-18850-P/NP, Revision 0, the NRC staff determined that the proposed use of the model, equation 3-1A (below) and validity bounds are adequate and acceptable.
[
]
3.2.5.2 Onset of Gas Release In Section 3.6.2 of WCAP-18850-P/NP, Revision 0, Westinghouse referenced scientific literature to state that [
] The NRC staff reviewed the information presented and determined that the proposed [ ] threshold for transient FGR is conservative and acceptable.
Westinghouse also referenced scientific literature to state that [
] The NRC staff reviewed the information provided and determined based on engineering judgement that the proposed [ ] threshold for transient FGR is very low and likely to be exceeded rapidly in nodes where significant [ ] is predicted to occur.
These are also the nodes where the [ ] threshold is likely to be exceeded, and thus transient FGR would not be prevented by the small threshold proposed. As a result, the NRC staff determined that the [ ] threshold is acceptable.
Finally, Westinghouse stated that any transient FGR would be assumed to [
] Although there is little data on the [ ] FGR, the NRC staff determined that assuming that all predicted FGR will occur [
] is conservative and acceptable.
Westinghouse stated that FGR is applied to [
] The [ ] are the basis for cladding rupture determinations, which is the primary objective of this TR. In addition, the effect of [
] will be the most limiting in the [
] Finally, the [
] Because of these reasons, the NRC staff determined that the proposed approach is sufficiently comprehensive to capture fuel rod ruptures in an accurate and acceptable manner.
3.2.5.3 Axial Gas Communication In Section 3.6.3 of WCAP-18850-P/NP, Revision 0, Westinghouse stated that [
] Westinghouse also stated that it performed hand calculations to confirm this engineering prediction. Based on its assessments, Westinghouse proposed to assume that [
] From the response to RAI 12, the NRC staff obtained more details about Westinghouses gas communication analyses. In addition, the NRC staff performed independent confirmatory hand calculations based on fission gas predictions from the FAST code. As shown in Figure 2, the NRC staffs analyses were in good agreement with Westinghouses results presented in the response to RAI 12, and the relative ranking of gas communication assumptions agreed. Based on these observations, the NRC staff determined that Westinghouses proposal to [
] is conservative and acceptable.
Figure 2: Comparison of NRC and Westinghouse (WEC) predictions of [
]
3.2.6 Pre-Burst Axial Fuel Relocation 3.2.6.1 Packing Fraction Assessment with Burst In Section 3.7.1 of WCAP-18850-P/NP, Revision 0, Westinghouse presented its packing fraction model for axial fuel relocation. Westinghouse initially stated that [
] The NRC staff confirmed that this sampling range covered the available data at low burnup. Westinghouse then added [
] The NRC staff evaluated the model and data and asked RAI 5 to inquire about the
[ ] Based on this RAI, Westinghouse
reanalyzed the packing fraction data in more detail and showed that the data follows [
] In addition, Westinghouse showed in the response to RAI 5 that [
] To ensure reasonable conservatism, Westinghouse also [
] The NRC staff verified that the updated model correctly represents and bounds all available data with adequate conservatism. On this basis, the NRC staff concludes that Westinghouses approach is acceptable.
3.2.6.2 Conditions for Relocation In Section 3.7.2 of WCAP-18850-P/NP, Revision 0, Westinghouse stated that [
] The bases for these conditions are that [
] The NRC staff determined that these arguments to be acceptable based on engineering judgement as well as experimental evidence from the testing that NRC was a part of at Halden and at Studsvik. Finally, based on a review of available data, Westinghouse stated that the [
] On this basis, the NRC staff concluded that the conditions chosen by Westinghouse for relocation are acceptable.
3.2.6.3 Pre-Burst Relocation Test Data Review In Section 3.7.3 of WCAP-18850-P/NP, Revision 0, Westinghouse reviewed available pre-burst relocation test data. Westinghouse described the test characteristics and results from [
] Westinghouse pointed out that some characteristics of these tests were
[ ] and stated that [ ] tests had [
] The NRC staff did not agree that the [
] used by Westinghouse was an adequate representation of this data and asked RAI 5 to obtain further justification.
In response to RAI 5, Westinghouse revised the model so that the pre-burst packing fraction is
[
]
such that it is between [
] The NRC staff verified that the updated model correctly represents and bounds all available data with the adequate conservatism, which the NRC staff determined to be acceptable.
3.2.6.4 Pre-Burst Fuel Relocation Model In Section 3.7.3 of WCAP-18850-P/NP, Revision 0, Westinghouse summarized its pre-burst fuel relocation model. The NRC staff reviewed each item in this model and determined the following:
[
]
[
]
[
]
[
]
[
]
3.2.7 Properties of Nuclear Fuel Rod Materials 3.2.7.1 Uranium Dioxide Thermal Conductivity In Section 3.8.2 of WCAP-18850-P/NP, Revision 0, Westinghouse stated that the [
] The NRC staff verified that both of these models account for thermal conductivity degradation as a function of burnup and confirmed that both models [ ] which is the burnup range of applicability of this TR. On these bases, the NRC staff determined that the UO2 thermal conductivity model proposed for use in this methodology is acceptable for burnups up to
[ ]
3.2.7.2 Uranium Dioxide Density In Section 3.8.2 of WCAP-18850-P/NP, Revision 0, Westinghouse stated that its UO2 density model is described in Reference 2, which was approved by the NRC. This model scales with the fraction of the theoretical density of UO2. The NRC staff determined that this model continues to be physically valid and acceptable for both UO2 and Advanced Doped Pellet Technology (ADOPT') fuel pellets based on the fact that it can be analytically derived from simple physics theory.
Conclusions The NRC staff reviewed the changes and updates to WCOBRA/TRAC-TF2 fuel rod models and determined that Westinghouse has made adequate changes to its methodology or justified the existing methodology for pellet-to-cladding gap conductance, cladding deformation, cladding rupture, fuel rod initialization, susceptibility to fine fragmentation, LOCA FGR, pre-burst axial fuel relocation, and the properties of nuclear fuel rod materials for evaluating FFRD phenomenon up to the point of burst. No updates or changes were made to evaluate dispersal.
The NRC staff concluded that the changes, updates, and justifications for fuel rod models described by Westinghouse in Section 3.0 of WCAP-18850-P/NP, Revision 0, are acceptable.
3.3 WCOBRA/TRAC-TF2 Kinetics and Decay Heat Model Updates (WCAP-18850-P/NP, Revision 0, Section 4)
The WCOBRA/TRAC-TF2 kinetics and decay heat model, which is part of the FSLOCA EM, calculates various heat sources which are important in determining the PCT. As described in Section 9 of WCAP-16996-P-A (Ref. 2), the primary heat sources during a LOCA are fission product decay heat, fission heat, actinide decay heat, and cladding chemical reaction.
Westinghouse determined that updates to this model are necessary for modeling fuel in the high burnup and high initial enrichment domains for the purpose of cladding rupture calculations as this model is important in determining the [ ] The methodologies used to update the WCOBRA/TRAC-TF2 model have been previously approved by the NRC in WCAP-18446-P-A and WCAP-18773-P-A, Higher Enrichment for Westinghouse and Combustion Engineering Fuel Designs (Refs. 15 and 16). The modifications to this method, as described in WCAP-18850-P/NP, Revision 0, apply solely to the analysis of fuel for the purposes of [ ]
calculations and do not affect the approved FSLOCA EM described in WCAP-16996-P-A (Ref. 2). The proposed updates are discussed in the following sections.
3.3.1 Nuclear Physics Data Westinghouse proposed updated nuclear physics data for WCOBRA/TRAC-TF2 generated by Westinghouses neutronics code, PARAGON2, described in WCAP-18443-P-A, Revision 0 (Ref. 15). Section 4.1 of WCAP-18443-P-A, Revision 0, establishes a range of applicability in which PARAGON2 may be applied. The relevant parameters described in this section are the maximum permissible initial fuel enrichment and average peak pellet burnup. PARAGON2 has been approved for a maximum permissible initial enrichment of 10 wt% U-235 and a maximum average peak pellet burnup of [ ] GWd/MTU. WCAP-18850-P/NP, Revision 0, proposes a maximum initial enrichment of [ ] wt% U-235 and a maximum rod average burnup of
[ ] GWd/MTU. When considering a conversion of rod average to pellet-average burnup, the proposed burnup and enrichment ranges for WCAP-18850-P/NP, Revision 0, are within the range of applicability of PARAGON2. Thus, use of PARAGON2 to supply nuclear physics data for WCOBRA/TRAC-TF2 for cladding rupture calculations is acceptable.
Increases in the maximum permissible fuel burnup and initial enrichment are known to have various effects on several nuclear physics parameters. The updated parameters proposed by Westinghouse include fission fractions, effective delayed neutron fraction, prompt neutron lifetime, prompt and delayed energy release, delayed neutron precursor decay constants by group, and U-238 capture-to-fission ratio. These parameters are shown in Figures 4.1-1 through 4.1-14 of WCAP-18850-P/NP, Revision 0, as function of burnup and initial fuel enrichment. The figures presented in this section contain calculated data between [ ] GWd/MTU rod
average burnup and initial fuel enrichments of [ ] wt% U-235. Therefore, the proposed data spans the entire requested approval range of WCAP-18850-P/NP, Revision 0.
The NRC staff has compared Figures 4.1-1 through 4.1-14 to those previously approved in the TRs WCAP-18446-P-A and WCAP-18773-P-A within their respective approval ranges. As expected, the calculated data matches as the same methodology is used in all three TRs. Given that this methodology has been previously approved twice, the NRC staff focused on the combinations of enrichment and burnup that are unique to WCAP-18850-P/NP, Revision 0. The figures demonstrate that PARAGON2, as the provider of nuclear physics data for WCOBRA/TRAC-TF2, captures expected trends associated with the analyzed parameters with no noticeable discontinuities or adverse trending that would suggest that the model is inadequate at modeling fuel rods in the requested burnup and enrichment ranges.
The NRC staff has determined that the updated nuclear physics data presented in Section 4.1 of WCAP-18850-P/NP, Revision 0, provides a reasonable estimation of the various neutronics parameters relevant in calculations performed by WCOBRA/TRAC-TF2 for rod average burnups up to [ ] GWd/MTU and initial fuel enrichments up to [ ] wt% U-235 and is therefore acceptable.
3.3.2 Neutron Capture Correction The decay heat model used in the WCOBRA/TRAC-TF2 methodology is based on the ANSI/ANS-5.1 1979 decay heat standard. This standard includes a neutron capture correction factor, which is determined by Equation 11 in the standard. Equation 11 has the following limitations:
(1) The shutdown time, t, is less than 10,000 seconds.
(2) The total operating time, T, is less than 4 years.
(3) The fissions per initial fissile atom,, is less than 3.
Westinghouse has indicated the condition (2) is violated for certain nuclear designs, especially those employing higher enrichments or burnups. Westinghouse has provided justification demonstrating that the calculated neutron capture correction factor for such cycles under the increased enrichment and burnup conditions requested in this TR is acceptable. Westinghouse has performed calculations using [
] The result of this analysis demonstrated that [
] In general, the [
] Westinghouse has provided results of the comparisons between the [ ] in Figures 4.2-1 through 4.2-4. The NRC staff performed some confirmatory calculations comparing [
] and found that [
] adequately bounds [
] up to 10,000 seconds. The [ ] also remove the applicability of condition (2). Therefore, Westinghouses approach to modeling the neutron capture correction factor with [ ] is acceptable.
Westinghouse has demonstrated applicability of condition (3) in Figure 4.2-5 of the TR. The figure contains [
] The figure contains adequate data up to fuel enrichments of [ ] wt% U-235 and a rod average burnup of [ ] GWd/MTU. Up to these enrichment and burnup limits, the [ ] thus condition (3) is satisfied.
The NRC staff concluded that Westinghouses justification for the use of [
] to calculate the neutron capture correction factor is acceptable because the applicability conditions described in the decay heat standard are acceptably dispositioned. This approach will ensure a conservative calculation of neutron capture correction factor for shutdown times up to 10,000 seconds for nuclear designs with initial enrichments up to
[ ] wt% U-235 and rod average burnups of up to [ ] GWd/MTU.
3.3.3 Normalized Fission Interaction Frequency The normalized fission interaction frequency (NFIF) is described in Section 9.3 of WCAP-16996-P-A (Ref. 2). Westinghouse defined the NFIF as the product of neutron velocity, prompt energy release per fission, and macroscopic fission cross section. In WCAP-16996-P-A, Westinghouse determined that, for the conditions relevant to simulating a LOCA, the NFIF can be modeled as [ ]
Data presented in Figure 4.3-1 of WCAP-18850-P/NP, Revision 0, demonstrates that [
] As a result, an updated model inclusive of higher enrichments is required to demonstrate acceptability of the NFIF up to enrichments of [ ] wt% U-235. The updated model preserves the original form of the equation apart from [
] The only other change to the NFIF is updated coefficients, which are documented in Table 4.3-1 of the TR. Westinghouse then compares the updated NFIF model to a NFIF calculated by PARAGON2 for various enrichments in Figures 4.3-2 through 4.3-9 of the TR. The data demonstrates that the updated NFIF model reasonably predicts the NFIF for [
]
The NRC staff determined that the updates to the NFIF model are acceptable because [
] in the calculation of the NFIF, and the updated coefficients are determined using the Westinghouses approved neutronics code, PARAGON2. PARAGON2 is approved for initial enrichments up to 10 wt%
U-235 and rod average burnups of up to [ ] Therefore, the updated coefficients are determined using an acceptable method and will ensure an acceptable calculation of the NFIF.
3.3.4 Gamma Energy Redistribution The gamma energy redistribution model and generalized energy deposition model (GEDM) is described in Section 9.6 of WCAP-16996-P-A (Ref. 2). This model calculates the heat deposition of several types of radiation, most notably penetrating radiation such as neutrons and gamma photons. Unliked charged particles, these types of penetrating radiation will not deposit their energy within the fuel pin they originated from but will instead radiate outward depositing
energy in the surrounding coolant and fuel pins. Accounting for the heat deposition from surrounding fuel pins will ensure an accurate estimation of key figures of merit.
Similar to other models described in this section, the GEDM is updated to use data generated by PARAGON2. This new data supersedes the data presented in WCAP-16996-P-A (Ref. 2).
The results of the GEDM recalculation are presented in Tables 4.4-1 and 4.4-2 of the TR. During an audit, the NRC staff reviewed the results which support Westinghouses conclusion that
[
] The NRC staff confirmed that the results adequately support this conclusion. Furthermore, the geometric model used in this analysis is the same as what was presented in WCAP-16996-P-A (Ref. 2). Westinghouse stated that no changes were necessary to support the increase in maximum initial enrichment and burnup and to accommodate the use of PARAGON2 for the geometric model. The NRC staff determined this to be an acceptable approach because the geometric model is [ ] which is conservative.
The NRC staff noted in RAI 7 that the description of the energy deposition model in Section 9.6 of WCAP-16996-P-A (Ref. 2) was not consistent with discussions held during the audit regarding the treatment of energy deposition in the [
] in the proposed WCAP-18850-P/NP, Revision 0, methodology. The inconsistency suggests that WCOBRA/TRAC-TF2 [
] as described during audit discussions.
Westinghouse provides clarification related to the energy redistribution model in response to RAI 7. Westinghouse indicates that WCOBRA/TRAC-TF2 predicts [
] The NRC staff inquired about the adequacy of the transfer matrices developed by Westinghouse. Westinghouse stated that the transfer matrix result sensitivity is unaffected by [
] The NRC staff examined Table 9-13 of WCAP-16996-P-A (Ref. 2) to confirm Westinghouses conclusion related to transfer matrix size and noted a similarity in results between the [ ] The NRC staff determined that the described gamma energy redistribution model and transfer matrix size remains appropriate for use in the WCAP-18850-P/NP, Revision 0, methodology because [
] Further, Westinghouses comparison evaluation of transfer matrix size indicates that a [ ] transfer matrix is adequate in modeling gamma energy redistribution.
Westinghouse also clarifies how the surrounding assembly powers are determined. It is described that the surrounding rod powers are [
] The NRC staff determined that the methodology for determining surrounding rod power, described in WCAP-16696-P-A, Revision 1, remains acceptable to use in the WCAP-18850-P/NP, Revision 0, methodology because the surrounding rod power is modeled conservatively high, which would reduce the heat transfer from the [ ]
Westinghouse discussed, in more detail, the neutron and gamma energy redistribution in response to RAI 7 Part D. With regard to neutron energy redistribution, Westinghouse uses a conservative approach which involves [
] With regard to gamma energy redistribution, Westinghouse references a demonstration analysis in the original submittal of WCAP-16696-P.
In comparing [
]
Westinghouse concludes that the gamma energy redistribution is adequately accounted for using the [ ] The NRC staff concluded that the proposed treatment for both neutron and gamma energy redistribution is acceptable because both approaches conservatively increase power of adjacent assemblies more than would typically be expected.
Conclusions The NRC staff reviewed the changes and updates to WCOBRA/TRAC-TF2 kinetics and decay heat model and determined that Westinghouse has made adequate changes to its methodology or justified the existing methodology for nuclear physics data, neutron capture correction, NFIF, and gamma energy redistribution for evaluating FFRD phenomenon up to the point of burst. No updates or changes were made to evaluate dispersal.
The NRC staff concluded that the changes, updates, and justifications for the kinetics and decay heat model provided by Westinghouse in Section 4.0 of the TR WCAP-18850-P/NP, Revision 0, are acceptable.
3.4 Fuel Rod Cladding Rupture Calculation Method (WCAP-18850-P/NP, Revision 0, Section 5) 3.4.1 Treatment of Regions The FSLOCA EM identifies two break regions of interest, Region I and II, representing small break LOCA (SBLOCA) and large break LOCA (LBLOCA), respectively. The intermediate range of breaks have been shown to be non-limiting with respect to the 10 CFR 50.46 ECCS acceptance criteria within WCAP-16996-P-A (Ref. 2). WCAP-18850-P/NP, Revision 0, proposes new acceptance criteria, and the limiting behavior discussed in WCAP-16996-P-A is not necessarily maintained. Analyzing the entire break spectrum also provides additional assurance of adequate protection against fuel dispersal.
The WCAP-18850-P/NP, Revision 0, cladding rupture methodology identifies several differences in the FSLOCA EM treatment between Regions I and II. The proposed treatment for Region IB is best described as a mix between Region I and II treatments with a few unique characteristics, which are discussed throughout Section 3.4 of this SE.
Westinghouse describes the break sizes that define each region and analyses required.
Region I analyses require a [
] is performed to determine the
[ ] Region I breaks consider the smallest break sizes up to a [ ]
where Region IB starts. The NRC staff determined that the proposed treatment for Region I cladding rupture calculations is acceptable because the [
] and it is consistent with the treatment described in WCAP-16996-P-A (Ref. 2).
Region IB is defined as break sizes greater than Region I breaks and smaller than Region II breaks (1 ft2 or ~13.5 inches). The transition from Region I to Region IB is defined by a change
[ ] Figure 5.1-9 provides a qualitative definition of this transition where the [
] with particular sensitivity in different plant classes (2-, 3-, and 4-loop plants). Westinghouses response to RAI 15A provides a description of the phenomenological demarcation between Region I and IB.
Region IB analyses have been shown to [
] Westinghouse has provided Region IB break spectrum analyses for Westinghouse 2-, 3-, and 4-loop plants to justify that the [
] in Figures 5.1-5 through 5.1-8 of the TR WCAP-18850-P/NP, Revision 0. The NRC staff observed from the figures that the [
] Region IB sensitivity to offsite power conditions is discussed in Sections 3.4.2.4 and 3.5.1.1 of this SE where the NRC staff found that the [
]
Region II is defined as all break sizes greater than 1 ft2 up to a double-ended guillotine break.
The proposed treatment for cladding rupture calculations [
] The NRC staff determined that the proposed treatment for Region II breaks is acceptable because [
]
Westinghouse states that Region II cladding rupture analyses [
] Westinghouse concludes that [
] As demonstrated in the figures provided in Section 5.1.1 of WCAP-18850-P/NP, Revision 0, [
] Increasing the PCT will [ ] therefore,
[ ] Westinghouse provides more clarification in response to RAI 15C attributing the [ ] Region II behavior to
[ ] The NRC staff determined that Region II cladding rupture analyses [
]
Some figures, such as Figure 5.1-9 from Reference 1 (below), appear to suggest that [
]
Figure 5.1-9: Illustration of Regions for the Cladding Rupture Calculations The NRC staff issued RAI 10 to obtain a comprehensive description of the modeling practices for Region IB breaks, similar to the definitions of Region I and II breaks contained in WCAP-16996-P-A (Ref. 2). In response to RAI 10, Westinghouse provided detailed descriptions of important characteristics related to Region IB breaks. These are listed below along with the relevant sections of this SE where they are discussed in detail.
Break Type and Size (3.4.1)
[ ] (3.4.2.3)
Offsite Power Availability (3.4.2.4 and 3.5.1.1)
[ ] (3.4.3.1)
Treatment of [ ] Fuel Assemblies (3.4.3.2)
Control Rod Insertion (3.4.3.3)
Counter-Current Flow Limitation (3.4.3.4)
Steam Generator Tube Plugging (3.4.3.5 and 3.5.1.3)
Containment Pressure (3.4.3.6)
Safety Injection Flow (3.4.3.7)
Operator Action (3.4.3.8)
The NRC staff concluded that the additional information related to Region IB modeling practices is acceptably comprehensive. Table 10-1 (below) from Reference 3 illustrates the modeling differences for each region.
Table 10-1: Summary of Modeling Differences Between Regions I, IB, and II in the FSLOCA EM and the Cladding Rupture Methodology
Figure 5.1-6 of WCAP-18850-P/NP, Revision 0, shows wide variation in the PCT across a small variation in break size. The NRC staff inquired in Part D of RAI 15 whether the behavior is a result of physical or numerical phenomena. Westinghouse attributes this behavior to [
] The NRC staff concluded that the characterization of the described Region IB break spectrum analysis behavior is acceptable. A description of the phenomena causing the variations was described and no code or method deficiencies were identified.
3.4.2 Uncertainty Contributors 3.4.2.1 Peaking Factor Uncertainty Westinghouse maintains the same treatment for peaking factor uncertainties as described in WCAP-16996-P-A (Ref. 2) and as updated by WCAP-18446-P-A (Ref. 10) for higher burnups.
The rod bow uncertainties described in WCAP-16996-P-A [ ]
Due to the increase in maximum rod burnup, Westinghouse updates the rod bow uncertainties in Table 5.2-1 of WCAP-18850-P/NP, Revision 0, consistent with the methodology approved in WCAP-18446-P-A. Therefore, the NRC staff determined that the proposed treatment for the peaking factor and rod bow uncertainties in the cladding rupture methodology is acceptable.
3.4.2.2 Decay Heat Uncertainty The FSLOCA EM [ ]
consistent with L&C 4 of WCAP-16996-P-A (Ref. 2). This treatment is maintained for WCAP-18850-P/NP, Revision 0, cladding rupture evaluations and is described in L&C 2 of this SE. The NRC staff determined that the proposed treatment is acceptable because it is conservative.
3.4.2.3 [ ]
The uncertainty parameters [
] The treatment for each of these parameters differs depending on the evaluated region.
Westinghouse describes the Region-specific treatment in Section 5.2.3 of WCAP-18850-P/NP, Revision 0. Therein, Westinghouse states that the [
] In response to RAI 9A, Westinghouse provides a more detailed description of the experimental tests and data used to justify the revised treatment for these uncertainty parameters. The NRC staff reviewed the referenced test data and found good agreement between the test data and the conclusions drawn by Westinghouse to support the proposed methodology changes. The sections below will go into more detail about each uncertainty parameter and the proposed change to its treatment in each Region and the proposed treatment for the newly defined Region IB.
The NRC staff issued RAI 9 to acquire more information related to the experimental tests and data, as well as various calculations performed by Westinghouse, used to support the revisions described below. The response also contains a discussion of the code changes to
WCOBRA/TRAC-TF2. The NRC staffs findings related to this supplemental information are included in the discussions below.
[ ]
[ ] Westinghouse proposes to
[
] Westinghouse updated WCOBRA/TRAC-TF2 during the review of WCAP-16996-P-A (Ref. 2) to fix the deficiencies and [ ]
However, Westinghouse did not credit the impacts of these code changes in the review of WCAP-16996-P-A. Instead, Westinghouse sought to conclude the complex FSLOCA review based on its perception [
]
Section 29.1.5 of WCAP-16996-P-A defines the [ ] Figures 5.2-3 and 5.2-4 of WCAP-18850-P/NP, Revision 0, show the performance of WCOBRA/TRAC-TF2 with modifications to address [ ] The NRC staff reviewed the figures and determined that the predicted behavior is much more consistent with the measured data after modification. [
] Therefore, the NRC staff determined that the proposed modifications to WCOBRA/TRAC-TF2, with respect to the [ ] are acceptable because the apparent deficiencies identified in the WCAP-16996-P-A have been resolved. Westinghouse's proposal to [
]
Westinghouse proposes to [ ] for Region IB evaluations.
The NRC staff concluded that the proposed treatment in Region IB is acceptable because Westinghouse has demonstrated that the [
]
[ ]
[ ] Westinghouse proposes to maintain the same treatment for both Region I and II for cladding rupture calculations. [
]
This treatment has been shown to be acceptable in WCAP-16996-P-A and Westinghouse determined that no changes were necessary to incorporate cladding rupture calculations.
To determine an acceptable treatment of the [ ] for Region IB evaluations, Westinghouse performed sensitivity studies [
] The NRC staff reviewed Figure 5.2-5 and determined that [
] Westinghouse notes that
[ ]
Therefore, Westinghouse proposes to [
] The NRC staff determined that the proposed treatment is acceptable because it is conservative.
[ ]
[
] Westinghouse proposes to modify the treatment of [ ] for Region I evaluations to be consistent with Region II evaluations. That being, to [
] Westinghouse justifies the proposed modification in two parts, discussed in response to RAI 9. Firstly, the impact on PCT as a result of [
] The NRC staff determined that the revised treatment of [ ] for Region I analyses is acceptable because the impact of
[
]
Westinghouse proposes to use a combination of the treatments of [ ] in Region I and II evaluations for Region IB evaluations. This treatment involves [
] Section 29.1.7 of WCAP-16996-P-A defines the [
] and the proposed treatment is consistent with this definition. Based on the discussion related to [
] The NRC staff concluded that the proposed treatment related to [
] is acceptable because the [
]
3.4.2.4 Offsite Power Availability Section 5.2.4 of WCAP-18850-P/NP, Revision 0, discusses Westinghouses treatment of offsite power availability in break spectrum analyses to determine the limiting break size and power availability configurations to evaluate in a detailed uncertainty analysis. There are two offsite power configurations considered, LOOP and offsite power available (OPA). A notable difference between the two configurations is the availability of the reactor coolant pumps (RCPs). During a LOOP, the RCPs immediately trip and during OPA the RCPs trip after operator action, which is assumed to be performed within a pre-determined, bounding time window. In WCAP-18850-P/NP, Revision 0, the bounding RCP trip time is determined to be [ ] Justification and discussion of the bounding RCP trip time is discussed in Section 3.4.2.5 of this SE.
] The NRC staff determined that this approach is acceptable because the limiting power availability configuration and break size will be identified using the proposed approach. Because Region IB is a newly defined region in the FSLOCA EM, Westinghouse
submitted sensitivity studies for the proposed treatment in Section 6.1.1 of WCAP-18850-P/NP, Revision 0, which is discussed in Section 3.5.1.1 of this SE. The proposed treatment is dependent on plant class and is as follows:
2-Loop pressurized water reactors (PWRs) will consider [
]
3-Loop PWRs will consider [
]
4-Loop PWRs will consider [
]
The NRC staff concludes that the proposed offsite power availability configurations are acceptable in determining the limiting configurations for Region IB analyses based on the sensitivity studies discussed in Section 3.5.1.1 of this SE. The proposed treatments will adequately identify the bounding configurations to ensure plant safety during a postulated LOCA and prevent cladding rupture of high burnup rods.
3.4.2.5 Reactor Coolant Pump Trip Timing The NRC staff issued RAI 8 to obtain information related to the assumed RCP trip timing of
[ ] questioning whether an earlier trip timing was more likely and could potentially be more limiting. In response to RAI 8B, Westinghouse describes the effect tripping the RCP has on reactor coolant. Notably, the RCP [
] This behavior is seen in Figures 8-2 through 8-4 of Reference 4. The [
] As the break size increases [
]
Figures 8-5 through 8-7 provide additional details describing the relationship between [
] The NRC staff determined this understanding of the relationship between these parameters to be key when understanding the effect RCP trip timing has on the overall transient.
Westinghouse compares RCP trip timings of [ ] The break spectrum results are shown in Figure 8-8. The NRC staff observed [
]
[
]
As break size increases, [
] This is attributed to [
]
The above conclusions were drawn from 4-loop analyses. Westinghouse provided a similar break spectrum analysis for 3-loop plants in Figure 8-22 for RCP trip timings of [
] The break spectrum analyses demonstrate similar results, ultimately concluding that
[
]
The NRC staff have concluded that Westinghouses treatment of RCP trip timing is acceptable because it produces a limiting PCT. In general, the NRC observed that running the RCP [
] The end results are a mixture of deeper core uncovery, delayed and longer cladding heatup, and a higher PCT.
3.4.3 Miscellaneous Considerations 3.4.3.1 [ ]
The [ ] is discussed in Section 30 of WCAP-16996-P-A (Ref. 2) and Sections 4.7.6 and 4.7.7 of the NRC staff final SE for WCAP-16996-P-A. The FSLOCA methodology uses a [
]
The FSLOCA methodology has slightly different treatments for the [ ] requirements for Region I and Region II breaks. Region I break size analyses require [
] Region II break size analyses require [
] to demonstrate compliance with the regulatory limits. This is described in more detail in Section 30 and L&Cs 11 and 15 of WCAP-16996-P-A.
Westinghouse in WCAP-18850-P/NP, Revision 0, proposes to [
] The NRC staff determined that the [
] is acceptable, as this is consistent with the [
]
Westinghouse in WCAP-18850-P/NP, Revision 0, also proposes to allow a [
] The technical basis for limiting the [
] Given the potentially more limiting conditions in the high burnup regime for cladding rupture calculations, it may be necessary to [
] to satisfy regulatory requirements. Further [
] This approach was previously accepted for Region II break sizes. The NRC staff determined that this approach is acceptable for Region I breaks because it results in a more accurate determination of the [ ] and reduces potential over-conservatism.
WCAP-18850-P/NP, Revision 0, defines a new break size region, known as Region IB. This region was not previously defined in WCAP-16996-P-A. Given the change in methodology to treat [ ] the NRC staff determined that it is acceptable to treat [ ]
The NRC staff determined that a [
] is acceptable because [ ] which provides reasonable assurance of adequate protection.
Limitation and Condition 11 of WCAP-16996-P-A Limitation and Condition 11 of WCAP-16996-P-A states, in part:
[
] is given in Table 30-1 (see page 30-16) in WCAP-16996-P, Revision 1. Note that the [ ]
Per the discussion above, Westinghouse revised the [ ] requirement of [ ] to [ ]
per proposed L&C 7, discussed below.
Limitation and Condition 15 of WCAP-16996-P-A Limitation and Condition 15 of WCAP-16996-P-A states, in part:
The [ ] to provide the required 95/95 probability, confidence statement that addresses the three major criteria of PCT, MLO, and CWO. This condition should be consistent with limitation number 11 in the table for
[ ] for each set.
To maintain consistency with L&C 11 of WCAP-16996-P-A and proposed L&C 7 of WCAP-18850-P/NP, Revision 0, the [ ]
and is addressed in proposed L&C 7 of WCAP-18850-P/NP, Revision 0.
Proposed Limitation and Condition 7 of WCAP-18850-P/NP, Revision 0 Proposed Limitation and Condition 7 of WCAP-18850-P/NP, Revision 0, states, in part:
The [ ] to provide the required 95/95 probability, confidence statement that addressed the margin to rupture.
Based on the discussions above related to [ ] the NRC staff determined that the proposed L&C 7 of WCAP-18850-P/NP, Revision 0, is acceptable.
3.4.3.2 Treatment of [ ] Fuel Assemblies The FSLOCA EM allows for [
] The NRC staff determined that this treatment is acceptable because of the similar characteristics between Region I and IB analyses. Further, [
] Feed assemblies can typically be omitted from cladding rupture calculations because they generally have not reached the fine fragmentation burnup threshold and therefore are not within the [ ] considered in the TR WCAP-18850-P/NP, Revision 0.
3.4.3.3 Control Rod Insertion Westinghouse proposes to [
] The NRC staff determined that the proposed treatment for control rod insertion is acceptable because [
]
3.4.3.4 Counter-Current Flow Limitation Westinghouse proposes to [
] The various phenomena important to LOCA evaluations are similar between Region I and Region IB, such that [
] Therefore, the NRC staff determined that the proposed CCFL treatment is acceptable.
3.4.3.5 Steam Generator Tube Plugging Westinghouse proposes to model Region IB breaks with [
]
Westinghouse clarifies that [ ] SGTP is limiting for Region IB breaks because
[ ] leading to a higher overall PCT compared to [ ]
SGTP cases. The justification and sensitivity studies are in Section 6.1.3 of WCAP-18850-P/NP, Revision 0, RAI 11, and Section 3.5.1.3 of this SE. As discussed further in Section 3.5.1.3 of this
SE, the NRC staff determined that modeling a [ ] SGTP for Region IB breaks is acceptable.
3.4.3.6 Containment Pressure In response to RAI 10, Westinghouse provided additional information related to the treatment of containment pressure for cladding rupture analyses. Westinghouse states that [
] The NRC staff determined that the proposed approach for modeling containment pressure is acceptable because the methods are conservative.
3.4.3.7 Safety Injection Flow In response to RAI 10, Westinghouse provided additional information related to the treatment of safety injection (SI) flow for cladding rupture analyses. Westinghouse states that [
] The NRC staff determined that the treatment of SI flow for cladding rupture analyses is acceptable because [
]
3.4.3.8 Operator Action The FSLOCA EM [
] Westinghouse proposes [
] The NRC staff determined that the proposed treatment related to operator action for Region IB is acceptable because [
]
Conclusions The NRC staff reviewed the Fuel Rod Cladding Rupture Calculation Methodology in Section 5.0 of WCAP-18850-P/NP, Revision 0, and determined that Westinghouse has made an adequate treatment of regions, uncertainty contributors, and other miscellaneous considerations for evaluating FFRD phenomenon up to the point of burst. No new methods were made to evaluate dispersal.
The NRC staff concluded that the changes, updates, new methods, and justifications for the fuel rod cladding rupture calculation methodology described by Westinghouse in Section 5 of TR WCAP-18850-P/NP, Revision 0, are acceptable.
3.5 Sensitivity Studies and Demonstration Analysis (WCAP-18850-P/NP, Revision 0, Section 6) 3.5.1 Intermediate Break LOCA Sensitivity Studies Westinghouse has performed numerous sensitivity studies to justify the methodology proposed in WCAP-18850-P/NP, Revision 0, and to show the impact of key parameters on intermediate break LOCA (IBLOCA) transient progression. Region IB is a newly defined region in the WCAP-18850-P/NP, Revision 0, methodology and is treated differently compared to Region I and Region II analyses. The sensitivity studies assist in understanding the phenomenology of the IBLOCA transient and the basis for the modeling assumptions described in Section 5 of WCAP-18850-P/NP, Revision 0. Westinghouses evaluation and associated NRC conclusions related to each sensitivity study are discussed in Section 3.5.1.1 through 3.5.1.9 of this SE.
Some of these sensitivity studies are included for information only to aid in phenomenological understanding of the IBLOCA transient and are not used to justify modeling decisions in the WCAP-18850-P/NP, Revision 0, methodology. As such, those sensitivity studies do not require NRC approval, and this will be noted in their respective sections. Section 3.5.2 of this SE discusses Westinghouses demonstration analysis of an IBLOCA transient utilizing the proposed method presented in WCAP-18850-P/NP, Revision 0.
The NRC staff questioned in RAI 11 the level of variability in the sensitivity studies presented in Section 6 of WCAP-18850-P/NP, Revision 0. The NRC staff was particularly concerned with whether or not the studies presented were the product of a single simulation or representative of a population of simulations. Each of sensitivity studies detailed in Sections 3.5.1.1 through 3.5.1.9, below, received different treatment in terms of the number of simulations performed in order to draw conclusions from the results of that study. Each section below will describe the treatment used to generate a representative sensitivity study from which Westinghouse could draw reasonable conclusions. While not discussed in detail in this section, the NRC staff determined that the simulation treatments, described in response to RAI 11A, are acceptable for drawing reasonable conclusions because the variance in each of the sensitivity studies is reduced compared to only using a single simulation as a representative result, thus providing a reasonable level of confidence in the results.
The NRC staff issued RAI 15B to understand the dominant parameters for maximizing the Region IB PCT. Westinghouse states that the dominant parameters include [
] Some of these parameters have competing effects depending on break size and are explored in some of the sensitivity studies below. Westinghouse identifies [
] These parameters are discussed in more detail in Section 3.2 of this SE.
3.5.1.1 Offsite Power Availability Westinghouse provided a set of sensitivity studies related to the offsite power availability for Westinghouse built 2-, 3-, and 4-loop plants in Section 6.1.1 of the TR WCAP-18850-P/NP, Revision 0. The goal of this study is to determine the [
] for each plant class. The results of this sensitivity study are used to justify the offsite power availability assumed in the WCAP-18850-P/NP, Revision 0, methodology and therefore requires NRC approval. As clarified in response to RAI 11A, each of these studies are selected from the break spectrum analyses performed in Section 5.2.4 of the TR and represent the limiting configuration for each respective plant class and offsite power availability. Each break
[
]
4-Loop Pressurized Water Reactor Study The 4-loop Region IB break spectrum analysis is documented in Figure 5.2-8 of the TR. The analysis concludes that the limiting power availability configurations are the [
] Region IB break size for LOOP and [ ] Region IB break size for OPA. These two configurations are [
] in implementations of WCAP-18850-P/NP, Revision 0, for 4-loop Region IB evaluations. The NRC staff determined that the selections of limiting break size for each offsite power configuration are acceptable because the break spectrum analysis results clearly demonstrate the limiting behaviors for each configuration.
The results of this sensitivity study are presented graphically in Figures 6.1.1-1 through 6.1.1-11 of the TR. The TR describes the evolution of the transient and the relevant phenomena.
Figure 6.1.1-10 compares the hot rod PCT for the LOOP and OPA cases. In this figure the LOOP and OPA PCTs are approximately [ ] respectively. The NRC staff determined that the approximate [
] due to phenomenological characteristics and not statistical variation.
Therefore, the NRC staff concluded that the decision to [ ] cases for 4-loop Region IB evaluations is acceptable because this [ ] calculated PCT and [ ]
3-Loop Pressurized Water Reactor Study The 3-loop Region IB break spectrum analysis is documented in Figure 5.2-7 of the TR. The analysis concludes that the limiting power availability configurations are the [ ] Region IB break sizes for LOOP and [ ] Region IB break size for OPA. Westinghouse considers the [ ] break size as being representative for larger Region IB break sizes
[ ] The results presented in Figure 5.2-7 may be considered representative due to [ ] in the larger Region IB break sizes.
These two configurations are [
] in implementations of WCAP-18850-P/NP, Revision 0, for 3-loop Region IB evaluations. The NRC staff determined that the selections of limiting break size for each offsite power configuration are acceptable because the break spectrum analysis results clearly demonstrate the limiting behaviors for each configuration.
The results of this sensitivity study are presented graphically in Figures 6.1.1-12 through 6.1.1-26 of the TR. The TR describes the evolution of the transient and the relevant phenomena. Figure 6.1.1-17 compares the hot rod PCT for [ ] break LOOP and OPA cases. In this figure the LOOP and OPA PCTs are approximately [ ]
respectively. Figure 6.1.1-25 compares the hot rod PCT for the [ ] break LOOP and OPA cases. In this figure the LOOP and OPA PCTs are approximately [ ]
respectively. The NRC staff determined that the approximate [
] Therefore, the NRC staff concluded that the decision to [
] cases for 3-loop Region IB evaluations is acceptable because this [
] calculated PCT and [ ]
2-Loop Pressurized Water Reactor Study The 2-loop Region IB break spectrum analysis is documented by Westinghouse in Figure 5.2-6 of the TR WCAP-18850-P/NP, Revision 0. The analysis concludes that the limiting power availability configurations are the [ ] Region IB break size for both LOOP and OPA. Westinghouse compares these two configurations with two additional LOOP and OPA configurations at a [ ] break size, which are more representative of the transient behavior throughout the Region IB spectrum. The shift in transient behavior occurs at the [ ] break size for both LOOP and OPA and results in [
] These configurations are compared to [
] in implementations of WCAP-18850-P/NP, Revision 0, for 2-loop Region IB evaluations. The NRC staff determined that the selections of limiting break size for each offsite power configuration are acceptable because the break spectrum analysis results clearly demonstrate the limiting behaviors for each configuration.
The results of this sensitivity study are presented graphically in Figures 6.1.1-27 through 6.1.1-41 of the TR. The TR describes the evolution of the transient and the relevant phenomena. Figures 6.1.1-34 and 6.1.1-35 compare the hot rod PCT for the LOOP and OPA cases for the [ ] breaks, respectively. The LOOP and OPA PCTs for the
[ ] break are approximately [ ] respectively. The LOOP and OPA PCTs for the [ ] break are approximately [ ] respectively. The NRC staff determined that the approximate [ ] difference in calculated PCTs does [
] As described in Section 5.2.4 of the TR, the WCAP-18850 -P/NP, Revision 0, methodology requires a [
] The NRC staff determined that the decision to perform [
] is acceptable because this treatment will result in a limiting calculated PCT and [ ]
Conclusions The results of the above sensitivity studies conclude which offsite power configurations will be evaluated for plant-specific Region IB analyses. In summary, 4-loop Region IB analyses consider [ ] cases, 3-loop Region IB analyses consider [ ] cases, and 2-loop Region IB analyses consider [ ] cases. An uncertainty analysis for [
] will be performed. The decision to [ ] for the 3-and 4-loop [ ] cases is justified by significant margins in the maximum PCT comparisons.
The NRC staff concluded that this sensitivity study adequately justifies the treatment of offsite power availability on a generic basis for Westinghouse 2-, 3-, and 4-loop plants as the [
] will be selected for the uncertainty analysis.
3.5.1.2 Upper Head Temperature Westinghouse performed a sensitivity study of the upper head temperature during a Region IB transient for Westinghouse-designed 4-and 3-loop plants. The sensitivity study results presented in Section 6.1.2 is the result with a [ ] Therefore, this sensitivity study is reasonably representative of plant behavior on a generic basis.
Westinghouse submitted the upper head temperature sensitivity study for information only.
These results are not used to justify any analysis treatment described in Section 5 of the TR WCAP-18850-P/NP, Revision 0. Therefore, NRC approval is not necessary. This SE will only provide a high-level summary and interpretation of the results.
Upper head temperature is an important LOCA parameter as it is used to characterize the flow and fluid properties in the upper head of the vessel. This affects the draining of entrained liquid into the core.
4-Loop Pressurized Water Reactor Study Each sensitivity includes [
] The results of this analysis are presented graphically in Figures 6.1.2-1 through 6.1.2-4 of the TR WCAP-18850-P/NP, Revision 0. In general, the results demonstrate an [
]
3-Loop Pressurized Water Reactor Study Each sensitivity study includes [
] The results of this analysis are presented graphically in Figures 6.1.2-5 through 6.1.2-9 of the TR WCAP-18850-P/NP, Revision 0. This analysis yields similar results and conclusions to the 4-loop sensitivity analysis. There is slightly different behavior between the two plant classes, but the ultimate outcome is same, where [ ]
Conclusions The NRC staff reviewed the upper head temperature sensitivity studies to understand the effect on cladding heatup. The NRC staff observed [ ] between upper head initial temperature and PCT.
3.5.1.3 Steam Generator Tube Plugging Westinghouse has provided sensitivity studies related to Region IB SGTP for Westinghouse built 2-, 3-, and 4-loop plants in Section 6.1.3 of the TR WCAP-18850-P/NP, Revision 0. The goal of this study is to justify the [ ] of SGTP described in Section 5.3.5 of the TR, which requires NRC approval. As clarified in response to RAI 11A, the sensitivity studies considered in this section were [ ]
Westinghouse describes several phenomena affected by SGTP and the associated impact on PCT in Section 5.3.5 of the TR and in response to RAI 11D. Most importantly, cladding heatup and PCT are highly dependent on the [
] Small changes in this time difference may have a significant effect on the resulting PCT. Differences in SGTP result in changes to various phenomena important in evaluation of a LOCA response. Some of these phenomena include fluid flow through the steam
generator, steam venting capability, initial RCS inventory, break flow, and RCS pressure. As discussed in WCAP-16996-P-A, [
] This conclusion is discussed below for each plant class.
The SGTP sensitivity studies are presented graphically in Figures 6.1.3-1 through 6.1.3-30 of the TR. These figures are modified in response to RAI 11 and presented in Figures 11-1 through 11-27 of Reference 4 (Figures 11-1 through 11-24 correspond to 6.1.3-1 through 6.1.3-24 and Figures 11-25 through 11-27 correspond to Figures 6.1.3-28 through 6.1.3-30, with Figures 6.1.3-25 through 6.1.3-27 being unmodified). The modifications to these figures include
[
]
4-Loop Pressurized Water Reactor Study Section 6.1.3.1 of the TR discusses the 4-loop sensitivity study for SGTP. The results of this study are presented graphically in Figures 11-1 through 11-8 of Reference 4. The results demonstrate a [ ] for all the parameters presented in the figures. Figure 11-8 presents the PCTs of these cases which demonstrate that [ ] Westinghouse further submitted a cumulative distribution function (CDF) plot of the 4-loop PCTs in Figure 11-25. This figure shows that [ ]
The NRC staff concluded that [ ] for Westinghouse built 4-loop PWR Region IB LOCA evaluations. This [ ] and
[ ]
3-Loop Pressurized Water Reactor Study Section 6.1.3.2 of the TR WCAP-18850-P/NP, Revision 0, discusses the 3-loop sensitivity study for SGTP. The results of this study are presented graphically in Figures 11-9 through 11-16 of Westinghouses response to RAI 11. The results are consistent with relative behavior observed in the 4-loop SGTP sensitivity study, that being [
] The primary difference between the 4-loop and 3-loop results is that there is [
] The CDF presented in Figure 11-26 confirms this observation as [
]
The NRC staff concluded that because [
] Westinghouses proposal to model [ ] is acceptable for Westinghouse built 3-loop Region IB LOCA evaluations when determining [ ]
2-Loop Pressurized Water Reactor Study Section 6.1.3.3 of the TR discussed the 2-loop sensitivity study for SGTP. The results of this study are presented graphically in Figures 11-17 through 11-24 of Westinghouses response to
RAI 11. The results demonstrate that [
] This behavior is attributed to [ ]
The NRC staff concluded that because [
] Westinghouses proposal to model [ ] is acceptable for Westinghouse built 2-loop Region IB LOCA evaluations when determining [ ]
Conclusions The NRC staff evaluated the proposed treatment for SGTP for Region IB LOCA analyses. The data provided by Westinghouse demonstrates that [ ] SGTP is acceptable for cases that would produce the 95/95 analysis result. Westinghouse acknowledges that [ ] SGTP is not strictly limiting across all cases, most notably for low PCT 3-loop cases. Furthermore, the NRC staff determined that [
] Therefore, the NRC staff concluded that the proposed treatment for SGTP in WCAP-18850-P/NP, Revision 0, is acceptable.
3.5.1.4 Upper Core Plate Flow Area Westinghouse performed a sensitivity study of the upper core plate flow area for a Region IB LOCA for Westinghouse-designed 4-and 3-loop plants. The sensitivity study results presented in Section 6.1.4 of WCAP-18850-P/NP, Revision 0, is the result with a [
] Therefore, this sensitivity study is reasonably representative of plant behavior on a generic basis. Westinghouse submitted the upper core plate flow area sensitivity study for information only. These results are not used to justify any analysis treatment described in Section 5 of the TR. Therefore, NRC approval is not necessary. This SE will only provide a high-level summary and interpretation of the results.
The upper core plate flow area is an important parameter in LOCA evaluations as it influences the availability of liquid in the upper regions of the reactor vessel to reach the core. The upper core plate flow area varies from reactor to reactor.
4-Loop Pressurized Water Reactor Study The sensitivity study includes two different initializations of the upper core plate flow area where the flow area was modeled with a [
] The results of this analysis are presented graphically in Figures 6.1.4-1 through 6.1.4-3 of the TR. The results demonstrate that a reduced flow area results in reduced flow from the upper plenum to the core and a slightly earlier cladding heatup.
Figure 6.1.4-3 appears to show [
]
3-Loop Pressurized Water Reactor Study The sensitivity study includes two different initializations of the upper core plate flow area where the flow area was modeled with a [
] The results of this analysis are presented graphically in Figures 6.1.4-4 through 6.1.4-6 of the TR WCAP-18850-P/NP, Revision 0. The results demonstrate that a reduced flow area results in reduced flow from the upper plenum to the core
and an increase in cladding heatup. [
] The important phenomena influencing these results is that the nominal flow area case allows for more water to drain into the core, providing slightly more cooling and reducing the PCT.
Conclusions The NRC staff reviewed the upper core plate flow area sensitivity studies to understand the effect on cladding heatup. The NRC staff observed that [
]
3.5.1.5 Main Steam Safety Valve Setpoint Westinghouse performed a sensitivity study of the main steam safety valve (MSSV) setpoint for a Region IB LOCA for Westinghouse-designed 4-and 3-loop plants. The sensitivity study results presented in Section 6.1.5 is the result with a [ ] Therefore, the sensitivity study is reasonably representative of plant behavior on a generic basis.
Westinghouse submitted the MSSV setpoint sensitivity study for information only. These results are not used to justify any analysis treatment described in Section 5 of the TR WCAP-18850-P/NP, Revision 0. Therefore, NRC approval is not necessary. This SE will only provide a high-level summary and interpretation of the results.
The MSSV setpoint is an important parameter in LOCA evaluations as it influences the primary-to-secondary side heat transfer.
4-Loop Pressurized Water Reactor Study The sensitivity study includes three different initializations of the MSSV setpoint where the setpoint is modeled at [ ]
The results of this analysis are presented graphically in Figures 6.1.5-1 through 6.1.5-3 of WCAP-18850-P/NP, Revision 0. The results demonstrate that a [
]
3-Loop Pressurized Water Reactor Study The sensitivity study includes three different initializations of the MSSV setpoint where the setpoint is modeled at [
] The results of this analysis are presented graphically in Figures 6.1.5-4 of the TR WCAP-18850-P/NP, Revision 0. The results demonstrate [
]
Conclusions The NRC staff reviewed the MSSV setpoint sensitivity studies to understand the effect on cladding heatup. The NRC staff observed that [
]
3.5.1.6 Accumulator Cover Pressure Westinghouse performed a sensitivity study of the accumulator cover pressure for a Region IB LOCA for Westinghouse-designed 4-and 3-loop plants. The sensitivity study results presented in Section 6.1.6 of the TR WCAP-18850-P/NP, Revision 0, is the result with a [
] Therefore, the sensitivity study is reasonably representative of plant behavior on a generic basis. Westinghouse submitted the accumulator cover pressure sensitivity study for information only. These results are not used to justify any analysis treatment described in Section 5 of the TR. Therefore, NRC approval is not required. This SE will only provide a high-level summary and interpretation of the results.
The accumulator cover pressure is an important parameter in LOCA evaluations as it influences the timing, rate, and duration of accumulator injection. Accumulator injection is typically the key inflection point in LOCA analyses that concludes cladding heatup. The PCT usually occurs a few seconds after accumulator injection before quickly cooling.
4-Loop Pressurized Water Reactor Study The sensitivity study includes three different initializations of the accumulator cover pressure where the pressure was set to [ ] This covers a wide range of pressures that are expected in Westinghouse 4-loop plants. The results of this analysis are presented graphically in Figures 6.1.6-1 through 6.1.6-4 of WCAP-18850-P/NP, Revision 0. The results demonstrate that a [
] This is the expected behavior.
3-Loop Pressurized Water Reactor Study The sensitivity study includes three different initializations of the accumulator cover pressure where the pressure was set to [ ] This covers a wide range of pressures that are expected in Westinghouse 3-loop plants. The results of this analysis are presented graphically in Figures 6.1.6-5 through 6.1.6-8 of WCAP-18850-P/NP, Revision 0. The results demonstrate that a [
] This is the expected behavior.
Conclusions The NRC staff reviewed the accumulator cover pressure sensitivity studies to understand the effect on cladding heatup. The NRC staff observed that [
]
3.5.1.7 Accumulator Water Volume Westinghouse performed a sensitivity study of the accumulator water volume for a Region IB LOCA for Westinghouse-designed 4-and 3-loop plants. The sensitivity study results presented in Section 6.1.7 of WCAP-18850-P/NP, Revision 0, is the result with a [
] Therefore, the sensitivity study is reasonably representative of plant behavior on a generic basis. Westinghouse submitted the accumulator water volume study for information only. These results are not used to justify any analysis treatment described in Section 5 of the TR. Therefore, NRC approval is not required. This SE will only provide a high-level summary and interpretation of the results.
Accumulator water volume is an important parameter in LOCA evaluations as it influences accumulator injection rate and duration, ultimately affecting the duration of cladding heatup.
4-Loop Pressurized Water Reactor Study The sensitivity study includes three different initializations of the accumulator water volume where the volume was set to [
] The evaluated range of water volumes encompass the expected range in Westinghouse 4-loop plants. The results of this analysis are presented graphically in Figures 6.1.7-1 through 6.1.7-4 and 6.1.7-9 of WCAP-18850-P/NP, Revision 0. The results demonstrate that a [
] The NRC staff observed [ ] accumulator water volume in 4-loop plants.
3-Loop Pressurized Water Reactor Study The sensitivity study includes three different initializations of the accumulator water volume where the volume was set to [
] The evaluated range of water volumes encompass the expected range in Westinghouse 3-loop plants. The results of this analysis are presented graphically in Figures 6.1.7-5 through 6.1.7-8 and 6.1.7-10 of WCAP-18850-P/NP, Revision 0. The results demonstrate a similar behavior as described for the 4-loop plants above. A notable difference between the 3-and 4-loop studies is that 3-loop plants appear [
] Westinghouse performed an additional sensitivity study with [
] The NRC staff did not review the additional sensitivity studies, but the conclusions submitted by Westinghouse appear reasonable on the basis of engineering judgement. Note that this sensitivity study is not used to justify any method treatments.
Conclusions The NRC staff reviewed the accumulator water volume sensitivity studies to understand the effect on cladding heatup. The NRC staff observed that [
]
3.5.1.8 Accumulator Temperature Westinghouse performed a sensitivity study of the accumulator water temperature for a Region IB LOCA for Westinghouse-designed 4-and 3-loop plants. The sensitivity study results presented in Section 6.1.8 of WCAP-18850-P/NP, Revision 0, is the result with a [
] Therefore, the sensitivity study is reasonably representative of plant behavior on a generic basis. Westinghouse submitted the accumulator temperature study for information only. Therefore, NRC approval is not required. This SE will only provide a high-level summary and interpretation of the results.
Accumulator water temperature is an important parameter in LOCA evaluations as it primarily influences the collapsed liquid level in the downcomer and lower plenum. This may influence the duration of cladding heatup.
4-Loop Pressurized Water Reactor Study The sensitivity study includes three different initializations of the accumulator temperature which are [
] The results of this analysis are presented graphically in Figures 6.1.8-1 through 6.1.8-6 of WCAP-18850-P/NP, Revision 0. The NRC staff observed
[
]
3-Loop Pressurized Water Reactor Study The sensitivity study includes three different initializations of the accumulator temperature which are [
] The results of this analysis are presented graphically in Figures 6.1.8-7 through 6.1.8-12 of WCAP-18850-P/NP, Revision 0. The NRC staff observed similar behavior as described in the 4-loop study. One notable exception is the [
] is seen. This results in larger differences in PCT, but not necessarily more limiting behavior. The PCT typically [
] In any event, [ ] behavior may be experienced during LOCA events, and such behavior does not necessarily represent reduced predictive capability of the code.
Conclusions The NRC staff reviewed the accumulator temperature sensitivity studies to understand the effect on cladding heatup. The NRC staff observed [
]
3.5.1.9 Accumulator Line Resistance Westinghouse performed a sensitivity study of the accumulator line resistance for a Region IB LOCA for Westinghouse-designed 4-and 3-loop plants. The sensitivity study results presented in Section 6.1.9 of WCAP-18850-P/NP, Revision 0, is the result with a [
] Therefore, the sensitivity is reasonably representative of plant behavior on a generic basis. Westinghouse submitted the accumulator line resistance study for information only.
Therefore, NRC approval is not required. This SE will only provide a high-level summary and interpretation of the results.
Accumulator line resistance is an important parameter in LOCA evaluations as it primarily influences the accumulator injection rate and duration, which may affect cladding heatup duration.
4-Loop Pressurized Water Reactor Study The sensitivity study includes three different initializations of the accumulator line resistance which is the [ ] The evaluated range of line resistances is reasonably representative of expected values in operating plants. The results of this analysis are presented graphically in Figures 6.1.9-1 through 6.1.9-4 of WCAP-18850-P/NP, Revision 0. The NRC staff observed [
]
3-Loop Pressurized Water Reactor Study The sensitivity study includes three different initializations of the accumulator line resistance which is the [ ] The evaluated range of line resistances is reasonably representative of expected values in operating plants. The results of this analysis are presented graphically in Figures 6.1.9-5 through 6.1.9-8 of WCAP-18850-P/NP, Revision 0. The NRC staff observed [
]
Conclusions The NRC staff reviewed the accumulator line resistance sensitivity studies to understand the effect on cladding heatup. The NRC staff observed [
]
3.5.2 Demonstration Analysis Westinghouse has performed a full demonstration analysis for a Region IB break. This analysis is consistent with what might be included in an implementation license amendment request (LAR) and is provided for information only to exercise the method. Therefore, no NRC approval is required, but this SE will provide a high-level summary and interpretation of the results.
3.5.2.1 Demonstration Analysis Inputs Tables 6.2.1-1 through 6.2.1-3 provide the input parameters considered in this demonstration analysis. The analysis assumes a break size of [ ] in a Westinghouse 4-loop plant.
[
]
3.5.2.2 Discussion of Results Westinghouse identified in Table 6.2.2-1 that [
]
Westinghouse provides more detail for [ ] in Figures 6.2.2-1 through 6.2.2-7. These figures demonstrate the expected behavior of a Region IB transient.
3.5.2.3 Conclusions from Demonstration Analysis The NRC staff reviewed the demonstration analysis described in Section 6.2 of WCAP-18850-P/NP, Revision 0. The NRC staff observed an example of an uncertainty analysis evaluation that would be included in a typical implementation LAR for this TR. Acceptable performance is demonstrated because [
]
Conclusions The NRC staff reviewed the information, evaluations, and justifications related to the sensitivity studies and demonstration analysis and determined that Westinghouse made adequate changes to its methodology or justified the existing methodology for Region IB LOCA sensitivity studies for evaluating FFRD phenomenon up to the point of burst. Because the approach is intended to prevent cladding rupture, no information was provided by Westinghouse to evaluate dispersal.
The demonstration analysis provided by Westinghouse shows that the EM can reasonably predict, in a conservative and reproducible manner, the onset of fuel rod burst during various LOCA scenarios. The NRC staff determined that when this EM is applied with plant and fuel product specific inputs (as outlined in Section 4.0 of this SE), the health and safety of the public are protected with reasonable assurance by demonstrating that fuel rod burst is prevented during LOCA events for rods susceptible to fine fragmentation.
The NRC staff concluded that the information, demonstrations, and justifications for the sensitivity studies and demonstration analysis provided by Westinghouse in Section 6 of the TR WCAP-18850-P/NP, Revision 0, are acceptable.
4.0 LIMITATIONS AND CONDITIONS 4.1 Review of Limitations and Conditions on the FULL SPECTRUM Loss-of-Coolant Accident Methodology
Westinghouse performed an assessment of the 15 L&Cs in WCAP-16996-P-A, Revision 1 (Ref. 2), to determine their applicability to the proposed WCAP-18850-P/NP, Revision 0, cladding rupture methodology.
L&C 1 is related to the FSLOCA EM not being approved to demonstrate compliance with 10 CFR 50.46(b)(5). Westinghouse states that WCAP-18850-P/NP, Revision 0, is only applicable for short-term LOCA response evaluations. Therefore, the long-term cooling requirements of 10 CFR 50.46(b)(5) are not applicable to WCAP-18850-P/NP, Revision 0. The NRC staff determined that the disposition of L&C 1 is acceptable.
L&C 2 is related to the applicability of the FSLOCA EM to Westinghouse-designed 3-and 4-loop plants. WCAP-18850-P/NP, Revision 0, is proposed to be applicable to Westinghouse-designed 2-, 3-, and 4-loop plants. WCAP-16996-P-A, Revision 1, and WCAP-18850-P/NP, Revision 0, may not be applied to 2-loop plants until approval of WCAP-16996-P/NP, Revision 1, Supplement 1, Revision 0, Extension of FULL SPECTRUM' LOCA (FSLOCA') Evaluation Methodology to 2-loop Westinghouse Pressurized Water Reactors (PWRs) with Information to Satisfy Limitations and Conditions Specific to 2-loop Plant Types, which is currently under the NRC review. This is documented in L&C 1 of Section 4.4 of this SE. The NRC staff determined that Westinghouses assessment of L&C 2 is acceptable.
L&C 3 is related to Region II containment pressure calculations. WCAP-18850-P/NP, Revision 0, proposes to maintain this condition for the cladding rupture methodology. No changes to the treatment of containment pressure calculations were determined to be necessary related to the cladding rupture method. Therefore, the NRC staff determined that Westinghouses assessment of L&C 3 is acceptable.
L&C 4 is related to the decay heat uncertainty multiplier. Westinghouse proposed an updated version of this L&C, which is documented by the NRC staff as L&C 2 in Section 4.4 of this SE.
The proposed revision removes the reporting requirements of decay heat uncertainty multipliers.
The NRC staff determined that this revision is acceptable due to the extensive reporting history of these multipliers, also noting that the treatment in the method will remain the same.
L&C 5 is related to the applicable burnup range of WCAP-16996-P-A, Revision 1 (Ref. 2).
WCAP-18850-P/NP, Revision 0, proposes a new burnup limit, therefore, this L&C is no longer applicable. L&C 3 in Section 4.4 of this SE is an updated version of L&C 5, documenting the new burnup limit approved in this TR. The NRC staff determined that Westinghouses assessment of L&C 5 is acceptable.
L&C 6 is related to the fuel performance data used in FSLOCA EM analyses. Westinghouse maintains the applicability of this L&C, requiring an approved version of a fuel performance code with the appropriate applicability range for high burnup and enrichment. L&C 4 in Section 4.4 of this SE is updated version of L&C 6. The NRC staff determined that Westinghouses assessment of L&C 6 is acceptable.
L&C 7 is related to the [ ] The TR WCAP-18850-P/NP, Revision 0, methodology proposes to [ ] and is discussed in Section 3.4.2.3 of this SE. The NRC staff determined that Westinghouses assessment of L&C 7 is acceptable, and it does not need to be carried forward to WCAP-18850-P/NP, Revision 0.
L&C 8 is related to the [ ] WCAP-18850-P/NP, Revision 0, proposes to maintain the treatment of the [ ] but proposes a different treatment of the [ ] consistent with the discussion in Section 3.4.2.3 of this SE. An updated version of this L&C is provided as L&C 5 in Section 4.4 of this SE. The NRC staff determined that Westinghouses assessment of L&C 8 is acceptable.
L&C 9 is related to [
] for non-Westinghouse 3-loop PWRs. The requirements of this L&C related to [ ] are not applicable to WCAP-18850-P/NP, Revision 0, based on the discussion in Section 3.4.2.3 of this SE.
Westinghouse has already docketed the requested [ ] for Westinghouse 2-and 4-loop PWRs (Refs. 17 and 18). In WCAP-18850-P/NP, Revision 0, Westinghouse acknowledged that a similar sensitivity study will need to be submitted prior to implementing the methodology for Combustion Engineering (CE)-designed PWRs. The requirement to complete any outstanding sensitivity studies is captured in L&C 1 of this SE.
Therefore, this L&C 9 from the NRC staff's SE on WCAP-16996-P-A is not applicable to WCAP-18850-P/NP, Revision 0. The NRC staff determined that Westinghouses assessment of L&C 9 is acceptable.
L&C 10 is related to sensitivity studies that need to be submitted to the NRC to confirm that the phenomenology and limiting behavior is accurately evaluated for non-Westinghouse 3-loop LOCA analyses. Westinghouse has already submitted to the NRC the requested sensitivity studies demonstrating acceptable predictive capability for Westinghouse 2-and 4-loop LOCA analyses. In WCAP-18850-P/NP, Revision 0, Westinghouse acknowledged that a similar sensitivity study will need to be submitted prior to implementing the methodology for Combustion Engineering (CE)-designed PWRs. Applicability to Westinghouse 2-loop LOCA analyses still needs to be confirmed per L&C 1 in Section 4.4 of this SE. Westinghouse proposes not to include this L&C in WCAP-18850-P/NP, Revision 0. The NRC staff determined that Westinghouses assessment of L&C 10 is acceptable.
L&C 11 is related to analysis inputs and documentation. Westinghouse included an updated version documented as L&C 6 in its response to RAI 6. A modified version of the L&C proposed by Westinghouse is included in Section 4.4 of this SE, which is consistent with the discussion in Section 3.4.3.1 of this SE. Item 2 of L&C 11 was originally determined to not be applicable to the TR WCAP-18850-P/NP, Revision 0, but later revised per Westinghouses response to RAI 6.
The NRC staff determined that Westinghouses assessment of L&C 11 is acceptable.
L&C 12 is related to plant-specific dynamic pressure loss from the steam generator secondary side to the MSSVs. Westinghouse states that this limitation is applicable to WCAP-18850 and proposes no revisions. The NRC staff determined that Westinghouses assessment of L&C 12 is acceptable.
L&C 13 is related to the [
] Westinghouse states that this limitation is applicable to the TR WCAP-18850-P/NP, Revision 0, and proposes no revisions. The NRC staff determined that Westinghouses assessment of L&C 13 is acceptable.
L&C 14 is related to compliance with the 10 CFR 50.46 oxidation criterion. WCAP-18850-P/NP, Revision 0, does not propose to address compliance with 10 CFR 50.46. Therefore, L&C 14 is
not applicable to WCAP-18850-P/NP, Revision 0. The NRC staff determined that Westinghouses assessment of L&C 14 is acceptable.
L&C 15 is related to Region II offsite power configurations and the [
] The portion of L&C related to analyzed offsite power configurations is applicable to the TR WCAP-18850-P/NP, Revision 0. The [
] consistent with the discussion in 3.4.3.1 of this SE. A revised version of this L&C is included as L&C 7 in Section 4.4 of this SE.
The NRC staff determined that Westinghouse has acceptably addressed all of the L&Cs documented in WCAP-16996-P-A, Revision 1 (Ref. 2). All L&Cs determined to be applicable to WCAP-18850-P/NP, Revision 0, have been included in Section 4.4 of this SE, either as written in WCAP-16996-P-A or revised, consistent with the NRC staffs safety findings.
4.2 Limitations and Conditions for WCAP-18850-P/NP, Revision 0 Based upon its review of the TR WCAP-18850-P/NP, Revision 0, the NRC staff determined that it is necessary to impose certain L&Cs upon the proposed methodology to ensure acceptable implementation.
The L&Cs listed below include 11 limitations proposed by Westinghouse in Section 7.2 of WCAP-18850-P/NP, Revision 0. The NRC staff has adopted these L&Cs in this SE as L&Cs 1-11, albeit in some cases in modified form; as such, the list of L&Cs in this SE supersedes the self-identified limitations Westinghouse included in Section 7.2 of WCAP-18850-P/NP, Revision 0.
4.2.1 Details Associated with the Fuel Rod Average Burnup Limitation WCAP-18850-P/NP, Revision 0, references numerous models with varying ranges of applicability, particularly with respect to fuel rod burnup. Westinghouse provided a brief summary of the ranges of applicability which are summarized below:
The fuel pellet thermal conductivity model is applicable up to a rod average burnup of
[ ]
The kinetics and decay heat model is applicable to a fuel rod average burnup of
[ ]
The neutron capture correct model is applicable up to a fuel rod average burnup of
[ ]
Other fuel rod models discussed in the TR WCAP-18850-P/NP, Revision 0, are applicable up to a fuel rod average burnup of [ ]
The above burnup limits do not supersede the maximum fuel rod average burnup of
[ ] as the upper limit found to be acceptable for use of the methodology described within this TR.
4.3 Compliance with 10 CFR 50.46/10 CFR 50.46c Acceptance Criteria WCAP-18850-P/NP, Revision 0, does not provide a method capable of directly demonstrating compliance with the requirements of 10 CFR 50.46. WCAP-18850-P/NP, Revision 0, provides a method capable of demonstrating that cladding rupture will not occur, which may be used in
combination with other analyses to confirm that the core coolability criterion in 10 CFR 50.46 is met. In general, the cladding rupture acceptance criterion proposed in WCAP-18850-P/NP, Revision 0, can be more restrictive than the 10 CFR 50.46 ECCS acceptance criteria for cladding temperature, oxidation, and hydrogen generation.
While the TR WCAP-18850-P/NP, Revision 0, references the acceptance criteria from the discontinued 10 CFR 50.46 proposed rulemaking, these criteria are not currently part of the regulations and were not considered by the NRC staff.
4.4 Limitations and Conditions
- 1. The TR WCAP-18850-P/NP, Revision 0, is applicable to Westinghouse-designed 2-loop PWRs equipped with upper plenum injection, 3-loop PWRs with cold-side injection, and 4-loop PWRs with cold-side injection as well as CE-designed PWRs. The methodology for the LOCA cladding rupture calculations can only be applied to the Westinghouse 2-loop PWR and CE PWR designs once the FSLOCA EM is approved for these designs and any plant-design-specific sensitivity studies have been acceptably completed. Any applicable differences in the approved methodology for these plant designs must be addressed in the cladding rupture calculations. Furthermore, any deviations from the method described in WCAP-18850-P/NP, Revision 0, shall be described and justified.
- 2. The following conditions apply to decay heat modeling and sampling in the cladding rupture calculations for all regions:
a) the decay heat uncertainty will be [
] and b) the cladding rupture calculations cannot be performed for transient times longer than 10,000 seconds following shutdown unless the decay heat model is shown to be acceptable for the analyzed core conditions. The latter limitation is [
]
- 3. The maximum fuel rod length-average burnup and fuel assembly average burnup permitted with WCAP-18850-P/NP, Revision 0, is [ ] Details behind the various burnup-related limitations associated with this TR are provided in Section 7.2.1 of WCAP-18850-P/NP, Revision 0.
- 4. The fuel performance data utilized to initialize the fuel rods for the cladding rupture calculations shall be obtained from a fuel performance code which includes the effect of thermal conductivity degradation and is NRC-approved through the fuel rod average burnups and initial fuel enrichments that are analyzed. The [
] and the generation of all the fuel performance data shall adhere to the NRC-approved methodology or acceptably justified.
- 5. The [
]
- 6. For each cladding rupture analysis performed:
a) The [ ] analysis seed(s), and the analysis inputs shall be declared and documented prior to performing the cladding rupture calculations. The [ ]
and the analysis seed(s) will not be changed once they have been declared and documented.
b) Should a plant-specific application of the cladding rupture methodology deviate from the originally declared analysis inputs for the intended purpose of demonstrating that cladding rupture does not occur with high probability, all modifications shall be discussed in the analysis submittal to the NRC. The calculated preliminary analysis result for each such case shall be summarized for information only in the analysis submittal to the NRC. Because these preliminary analyses and results are not intended to demonstrate that no cladding rupture occurs during the evaluated limiting LOCA, formal 10 CFR Part 50, Appendix B, verification and archival documentation of these preliminary analyses are not required. All final calculations used to demonstrate the final evaluation are subject to formal 10 CFR Part 50, Appendix B, verification and archival documentation rules.
c) Plant operating ranges which are sampled for the cladding rupture calculations shall be provided in the analysis submittal associated with the cladding rupture calculations.
- 7. Plant-specific applications of the cladding rupture methodology for Region II shall include two complete sets of sampled statistical evaluations:
a) a complete set with OPA, and b) a second complete set with a LOOP. For each set, the calculated statistical results at the 95/95 probability, confidence level should result in margin to cladding rupture.
The [ ] to provide the required 95/95 probability, confidence statement that addresses the margin to rupture.
- 8. L&Cs 3, 12, and 13 WCAP-16996-P-A, Revision 1, shall be satisfied for the application of this cladding rupture methodology.
- 9. The TR WCAP-18850-P/NP, Revision 0, is applicable to standard UO2 or ADOPT fuel with AXIOM cladding. However, prior to applying the methodology to new fuel designs approved subsequent to issuance of this SE, Westinghouse shall confirm the applicability of the methodology to the new fuel design or propose any modifications necessary to acceptably model the new fuel design.
- 10. The TR WCAP-18850-P/NP, Revision 0, is applicable to un-poisoned fuel, fuel with integral fuel burnable absorber (IFBA), and fuel with Gadolinia. This limitation does not preclude the use of wet annular burnable absorbers or other discrete burnable absorbers during the lifetime of an assembly.
- 11. A maximum of [ ] is permitted with WCAP-18850-P/NP, Revision 0. This limitation results from the maximum enrichment that is supported by the WCOBRA/TRAC-TF2 kinetics and decay heat module.
5.0 CONCLUSION
S The NRC staff has reviewed WCAP-18850-P/NP, Revision 0, which describes Westinghouses proposed methodology for cladding rupture analyses using the FSLOCA EM for Westinghouse-designed 2-, 3-, and 4-loop PWRs and CE PWRs. The proposed method provides a means to evaluate whether cladding rupture occurs for high-burnup fuel rods susceptible to fine fragmentation. An analysis result demonstrating that no cladding rupture occurs during the evaluated limiting LOCA for fuel susceptible to fine fragmentation is considered an acceptable result which precludes fuel dispersal into the coolant.
The NRC staff concluded that the proposed method in the TR WCAP-18850-P/NP, Revision 0, is acceptable for use in plant-specific applications and licensing. Westinghouse has adequately demonstrated acceptable predictive capability of evaluating the susceptibility of high burnup fuel rod cladding to rupture during a postulated LOCA provided that the L&Cs in Section 4.4 of this SE are fully met.
6.0 REFERENCES
- 1. Westinghouse, WCAP-18850-P/NP, Revision 0, Adaptation of the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology to Perform Analysis of Cladding Rupture for High Burnup Fuel, dated February 2024 (Agencywide Documents Access and Management System Accession No. ML24060A160).
- 2. Letter from J. A. Gresham, Westinghouse, to NRC DCD, Submittal of WCAP-16996-P-A/WCAP-16996-NP-A, Volumes I, II, III and Appendices, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) (TAC No. ME5244) (Proprietary/Non-Proprietary), LTR-NRC 66, October 2, 2017 (ML17277A131).
- 3. Westinghouse, Set 1 of Requests for Additional Information on Westinghouse Topical Report WCAP-18850-P/NP, Revision 0, Adaptation of FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology to Perform Cladding Rupture Calculations for High Burnup Fuel, dated January 2025 (ML25034A075).
- 4. Westinghouse, Set 2 of Requests for Additional Information on Westinghouse Topical Report WCAP-18850-P/NP, Revision 0, Adaptation of FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology to Perform Cladding Rupture Calculations for High Burnup Fuel, dated March 2025 (ML25090A297).
- 5. NRC, Section 4.2, Revision 3, Fuel System Design, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, dated March 2007 (ML070740002).
- 6. NRC, Section 15.6.5, Revision 3, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, dated March 2007 (ML070550016).
- 7. NRC, RG 1.157, Revision 0, Best-Estimate Calculations of Emergency Core Cooling System Performance, dated May 1989 (ML003739584).
- 8. NRC, Section 15.0.2, Revision 0, Review of Transient and Accident Analysis Method, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, dated March 2007 (ML070820123).
- 9. NRC, RG 1.203, Revision 0, Transients and Accident Analysis Methods, dated December 2005 (ML053500170).
- 10. Westinghouse, WCAP-18446-P-A, Revision 0, Incremental Extension of Burnup Limit for Westinghouse and Combustion Engineering Fuel Designs, dated August 2024 (ML24227A590).
- 11. [
]
- 12. [
]
- 13. Westinghouse, WCAP-18546-P-A, Revision 0, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel, dated March 2023 (ML23089A063).
- 14. NRC, RIL 2021-13, Interpretation of Research on Fuel Fragmentation, Relocation, and Dispersal at High Burnup, dated December 2021 (ML21313A145).
- 15. Westinghouse, Submittal of WCAP-18443-P-A, Revision 0, and WCAP-18443-NP-A, Revision 0, Qualification of the Two-Dimensional Transport Code PARAGON2, dated August 2021 (ML21214A335).
- 16. Westinghouse, WCAP-18773-P-A, Revision 0, Higher Enrichment for Westinghouse and Combustion Engineering Fuel Designs, dated September 2025 (ML25274A115).
Principal Contributors:
Jeremy Dean (Lead Reviewer)
Patrick Raynaud Brandon Wise John Lehning Jack Vande Polder Date: November 25, 2025 NRC Resolution to Westinghouse proprietary markings and voluntary comments Westinghouse submitted on September 4, 2025, the proprietary markup and voluntary comments (Package ML25248A040 (Non-publicly available/Proprietary)) following the NRC issuance on July 28, 2025 (ML25168A301 (Non-publicly available/Proprietary)), of the proprietary draft SE for proprietary review. The NRC staff reviewed and incorporated Westinghouses proprietary markings requests (including Westinghouses requests for removal of some proprietary markings) and voluntary editorial comments into the final SE.