ML19276E475

Final Safety Evaluation for Framatome Inc. Topical Report ANP-10346, Revision 0, ATWS-I Analysis Methodology for BWRs Using RAMONA5-FA (Non-Prop Version)

ML19276E475
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Site:	PROJ0728, 99902041
Issue date:	10/30/2019
From:	Rowley J Licensing Processes Branch
To:	Peters G Framatome
Rowley J
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FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TOPICAL REPORT ANP-10346P, REVISION 0, ATWS-I ANALYSIS METHODOLOGY FOR BWRS USING RAMONA5-FA FRAMATOME, INC.

PROJECT NO. 728/DOCKET NO. 99902041

1.0 INTRODUCTION

By letter dated December 15, 2017 (Ref. 1), as supplemented by letter dated March 8, 2019 (Ref. 5), Framatome, Inc. (Framatome formerly known as AREVA, Inc.) submitted a topical report (TR) which presents a methodology for the evaluation of the Anticipated Transient Without Scram with Instability (ATWS-I) for boiling water reactors (BWRs) using a updated version of the RAMONA5-FA methodology. The TR is entitled, ATWS-I Analysis Methodology for BWRs Using RAMONA5-FA, and can be identified by its TR number, ANP-10346P (Ref. 2).

The RAMONA5-FA methodology was originally developed to predict the critical power response of a BWR core to regional oscillations. The methodology was approved by the U.S. Nuclear Regulatory Commission (NRC) for this purpose in TR EMF-3028P-A (Ref. 3). A number of the models from RAMONA5-FA were subsequently incorporated in the AISHA and SINANO methodologies described in ANP-3274P, which were approved for use to analyze the ATWS-I event for extended flow window (EFW) applications at Monticello Nuclear Generating Plant, Unit 1 (Monticello) (Ref. 4). Subsequently, Framatome updated the RAMONA5-FA methodology to incorporate enhancements to address important phenomena for the ATWS-I event on a generic (i.e., non-plant specific) basis. ANP-10346P provides information on the updated RAMONA5 FA methodology along with an ATWS-I specific phenomena identification and ranking table (PIRT), validation, and analysis procedure.

ANP-10346P provides: (1) a description of the models, correlations, and coupling routines within RAMONA5-FA relevant to ATWS-I modeling; (2) additional assessment of the revised RAMONA5-FA code to validate its predictions of the onset of instability and subsequent growth and mitigation of core power oscillations, as well as the thermal hydraulic response under oscillatory dryout/rewet conditions; (3) specific parameters and assumptions to be used during performance of ATWS-I analyses; (4) sensitivity studies or other technical justifications for generic analysis assumptions; and (5) sample ATWS-I problems. Since the NRC has previously approved multiple components of this methodology as part of the review of TR EMF-3028P-A and the Monticello EFW license amendment request, the primary focus of the NRC staff was on the aspects of this methodology that are novel approaches to ensure applicability on a generic basis, as well as the integration of multiple methodologies developed at different times into a single approach for generic ATWS-I analyses. However, emergent Enclosure

issues may arise which lead to questions about prior approved methodology components within the context of this methodology.

2.0 BACKGROUND

An ATWS-I event is defined as a scenario in which an anticipated operational occurrence (AOO) occurs followed by the failure of a reactor trip to occur, which results in thermal hydraulic conditions which can allow unmitigated, unstable power oscillations to grow. Typical AOOs for which this can occur for BWRs include the turbine trip event and the two recirculation pump trip (2RPT) event, both of which cause lower core flow coupled with decreasing feedwater temperature. If no scram occurs, then the decreasing feedwater temperature will cause the core power to increase. Eventually, the combination of lower core flow and increasing core power will cross the instability boundary, and power oscillations may occur. If the oscillations grow without mitigation, they may become large enough to cause loss of adequate cooling, an increase in cladding temperature, and subsequent fuel failure leading to loss of core coolability.

An acceptable means of ensuring that core coolability is maintained is to ensure that the PCT remains below 2200 F, which is consistent with similar core coolability requirements in Title 10 of the Code of Federal Regulations (10 CFR) 50.46, "Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants." Some further discussion of these specific events can be found in Section 3.2 of the TR.

The methodology described in ANP-10346P is intended to allow plant-specific ATWS-I analyses to be performed in order to support operation in the EFW domain on the power-flow map. The RAMONA5-FA code (Ref. 3) was originally developed to implement a long-term stability solution (LTSS), which ensures that the plant core monitoring system can detect and suppress potential power oscillations due to coupled neutronic-thermal hydraulic instabilities. An ATWS-I analysis methodology (Ref. 4) was subsequently approved for a plant-specific application at Monticello that utilized many of the constitutive models from RAMONA5-FA, but resolution of some issues required use of ['''''''''''''''''''''''' '''''''''''''''' ''''''''''''''''' ' '' ''''''''''''''''] that were implemented by use of the AISHA and SINANO methodologies. The updated RAMONA5-FA methodology described in ANP-10346P provides a generic ATWS-I analysis approach that can be applied at any plant utilizing the BWR/3 through BWR/6 designs. ANP-10346P also includes additional validation necessary to demonstrate the applicability of the RAMONA5-FA methodology, as updated, to the larger oscillations (relative to LTSS conditions) and dryout/rewetting conditions that may occur during an ATWS-I event. Since the methodologies documented in References 3 and 4 will be discussed repeatedly throughout this safety evaluation (SE), for ease of comprehension in this SE, these methodologies will henceforth be referred to as the RAMONA5-FA LTSS methodology (Ref. 3) and the Monticello ATWS-I methodology (Ref. 4).

The intended purpose of the ATWS-I analysis is to demonstrate that plant specific mitigation strategies are adequate to provide reasonable assurance that core coolability is maintained.

Such mitigation strategies typically include operator action to reduce the water level in the downcomer or activation of the Standby Liquid Control System (SLCS). The water level in the downcomer is reduced by terminating or reducing the flow of feedwater (FW) to the vessel. This reduction of cold FW injection causes the core inlet subcooling to decrease. Additionally, reduction of the water level in the downcomer below the level of the FW spargers causes the feedwater to fall through a steam environment and absorb additional heat, which also reduces the amount of core inlet subcooling. The reduced core inlet subcooling leads to reduction in the core power back below the instability boundary. The SLCS acts more directly to reduce core power by injection of soluble boron in the coolant.

Since the NRC review of ANP-10346P depends, in part, on the assumption that selected technical models have previously been reviewed and approved by the NRC for stability related calculations as part of the review of the RAMONA5-FA LTSS methodology and Monticello ATWS-I methodology, the SEs associated with the aforementioned documents were reviewed.

In the SE for the RAMONA5-FA LTSS methodology, several limitations were outlined associated with the ['''''''''''''''''''''''' '''' ''''''''''''''''''''''] neutronic methods. However, the methodology described in ANP-10346P for use of RAMONA5-FA to analyze the ATWS-I event utilizes the ADAPKIN module, which does not have any specific limitations. No additional limitations or conditions were imposed by the NRC on the use of RAMONA5-FA for analysis of instabilities and associated power oscillations. The SE for the Monticello ATWS-I methodology differs in that it only evaluates a plant-specific methodology and thus does not contain limitations and conditions, however, no specific concerns were identified for the models and correlations that were adopted as part of the methodology described in the TR.

3.0 REGULATORY EVALUATION

The regulation in 10 CFR 50.62 requires that the licensee/applicant provide an acceptable reduction of risk from ATWS events by inclusion of prescribed design features and demonstrating their adequacy in mitigation of the consequences of an ATWS event. Within the context of the review of ANP-10346P, the ATWS-I analyses are intended to demonstrate that the combination of automated plant functions and prescribed operator actions will be sufficient to preclude fuel failure.

The regulation in 10 CFR 50.46, Acceptance criteria for emergency core cooling systems

[(ECCS)] for light-water nuclear power reactors, is not directly applicable to the ATWS-I event because it is intended to address postulated loss-of-coolant accidents rather than ATWS events. However, this regulation does present a set of acceptance criteria for ensuring adequate cooling of fuel such that significant fuel failures do not occur.

General Design Criterion (GDC)-12, "Suppression of reactor power oscillations," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," addresses the ability of a plant to suppress power oscillations that may occur. The ATWS-I analyses show that this GDC is met by demonstrating that the prescribed operator actions are sufficient to suppress any power oscillations.

The Standard Review Plan (NUREG-0800, herein referred to as SRP) is the primary regulatory guidance document used by the NRC staff to support review of this TR. In particular, SRP Chapter 15.8, Anticipated Transients Without Scram (Ref. 7), establishes acceptance criteria for ATWS events. SRP 15.8 does include additional GDCs beyond those listed above, however, they define vessel, ECCS, and containment performance requirements. This is not a significant concern for ATWS-I events, therefore, these GDCs were not considered as part of this review.

ANP-10346P describes an application of an evaluation model to perform licensing analyses for an accident. As such, additional guidance for the evaluation may be found in SRP Chapter 15.0.2, Review of Transient and Accident Analysis Methods (Ref. 8). This chapter includes provisions for the review of submittals related to evaluation models intended for use in analyzing postulated accidents. This guidance is intended for use with design basis accidents, and as such, is not fully applicable to the ATWS-I event. However, the guidance provides a useful framework for the NRC staff to review this TR. In summary, the NRC staff used the review guidance in SRP Chapter 15.0.2 along with the applicable acceptance criteria in SRP

Chapter 15.8 in conducting its review of the TR. In accordance with SRP Chapter 15.0.2, the review covered the areas of: (1) documentation, (2) evaluation model, (3) accident scenario identification process, (4) code assessment, (5) uncertainty analysis, and (6) quality assurance plan. To the extent possible, the NRC staff leveraged the prior review and approval of the RAMONA5-FA LTSS methodology and the Monticello ATWS-I methodology.

4.0 TECHNICAL EVALUATION

ANP-10346P describes a methodology by which the RAMONA5-FA code can be used for analysis of the ATWS-I event. The NRC staff review of ANP-10346P was performed by following the key elements of the evaluation model development and assessment process (EMDAP) outlined in Regulatory Guide (RG) 1.203 (Ref. 12) and echoed in SRP Chapter 15.0.2 (Ref. 8). While this guidance was intended mainly to address design basis accidents, the general principles can be applied to ATWS-I analysis methodologies. In summary, the areas of review were as follows:

1. Accident scenario description and phenomena identification and ranking - Framatomes break-down of the ATWS-I event and its relevant phenomena, and characterization of the consequences. The NRC staff utilized other available approved PIRTs and relevant guidance to inform their assessment of whether all the relevant phenomena are appropriately addressed in the validation basis, acceptance criteria, and/or procedure used to confirm that the acceptance criteria are met.

2. Evaluation methodology - the proposed ATWS-I analysis methodology, including initial conditions, assumptions, and approach to ensuring that the acceptance criteria are met.

Since this methodology includes use of the evaluation model, by extension, this area includes the models and correlations within the RAMONA5-FA code.

3. Code assessment - the assessments performed by Framatome to validate the RAMONA5-FA performance for the thermal hydraulic and neutronics phenomena expected during ATWS-I events, particularly during unstable power oscillations and for the specific fuel designs currently used by Framatome customers.

4. Uncertainty analysis - This area is not formally required since the ATWS-I event is not a design basis event. However, the NRC staff did confirm that Framatome adequately addressed the parameters that have the most impact on the results of the analyses through conservative assumptions or sensitivity studies.

5. Documentation - The NRC reviewed Framatomes documentation of the various aspects of this analysis methodology, including ANP-10346P itself as well as various documents supporting the RAMONA5-FA code and calculational files or procedures that provide detail on the intended steps to be taken when performing ATWS-I analyses or qualifying the methodology for different plant configurations and fuel designs.

The documentation associated with ANP-10346P is captured by various calculational files, validation reports, technical references, code documentation, and ANP-10346P itself. The additional documentation reviewed by the NRC staff that were not formally submitted on the docket as References 2 or 5 are summarized in the audit report (Ref. 6). While this information was not necessary to make a safety finding, the NRC staff did confirm that the information was consistent with that presented in ANP-10346P and the request for additional information (RAI) responses. In order to resolve all key technical issues necessary to make a safety finding on

this TR, the documentation was found to include sufficient information for the NRC staff to understand the intended application and validation of the methodology described in the TR. As such, the NRC staff acceptance of the adequacy of the licensees discussion of each area implicitly includes acceptance of the licensee documentation associated with that area.

RG 1.203 also discusses a sixth key element of the EMDAP, quality assurance (QA) processes.

This aspect is not explicitly discussed in detail for this SE because the bulk of the QA processes are captured within the overall Framatome QA program, which is consistent with requirements in 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants. The NRC staff believes that reasonable assurance exists based on previous experience with Framatome, that the QA processes are adequate, and the documentation reviewed as part of the audit associated with ANP-10346P was consistent with a robust QA program.

Each of the first four aforementioned areas will be discussed and evaluated in the following subsections.

4.1 Accident Scenario Description and Acceptance Criteria As per the review guidance in Chapter 15.0.2 of the SRP, the accident scenario description and phenomena identification and ranking process is intended to ensure that the dominant physical phenomena influencing the outcome of the given accident scenario are correctly identified and ranked. Once an accident scenario has been described, then figures of merit (FoMs) can be determined for use in evaluating whether acceptance criteria are met. The subsequent phenomena identification and ranking process will determine the physical phenomena affecting the FoMs and rank them by their importance. By doing so, an applicant can demonstrate that reasonable assurance exists that they are accurately capturing and modeling the dominant physical phenomena necessary for evaluation of the accident scenario in question.

Section 4.0 of ANP-10346P provides an extensive description of the various characteristics of the large coupled neutronic/thermal-hydraulic oscillations that uniquely characterize the ATWS-I event. In addition, other potential characteristics of an ATWS-I event that are potentially important are discussed, including potential prompt criticality, the possibility of boiling within bypass flow channels, and the cyclical dryout/rewetting that may be experienced by fuel.

Framatomes understanding of the characteristics of the ATWS-I event were used to develop a PIRT which identifies specific physical processes and parameters that are expected to be relevant to the ATWS-I event.

The PIRT is intended to identify the dominant phenomena pertaining to ATWS-I licensing analyses. Because the RAMONA5-FA ATWS-I methodology is based on a preexisting approved methodology (the RAMONA5-FA LTSS methodology), Framatome used the ATWS-I PIRT to determine which equations and closure relations required development or enhancement in order to apply the methodology to ATWS-I. In addition to model development, Framatome also used the ATWS-I PIRT to define the types of validation and sensitivity studies that were needed to support the methodology.

An important step in the NRC staffs evaluation of the RAMONA5-FA ATWS-I methodology, therefore, was to determine whether the ATWS-I PIRT provided in ANP-10346P suitably encompassed all important phenomena for ATWS-I analyses, and whether the importance levels indicated in ANP-10346P were consistent with the NRC staffs current knowledge of the ATWS-I phenomena. In order to make this determination, the NRC staff reviewed PIRTs

developed in 2001 and 2011 under the guidance of the NRC (Ref. 9 and Section 5 of Ref. 17),

more recent NRC published studies of ATWS-I scenarios (Refs. 10 and 11), and other available sources of information from open literature or internal NRC experience based on reviewing ATWS-I methodologies.

An important basis for the PIRT is identification of appropriate FoMs that correlate with the acceptance criteria for the ATWS-I evaluation. The primary acceptance criterion is the PCT, since Framatome elected to use a 2200F upper limit on PCT to demonstrate that fuel/cladding damage sufficient to challenge core cooling will not occur. Secondary acceptance criteria are discussed in the Calculational Procedure section of this document, which are related to the timing of events in the ATWS-I accident progression (including any required mitigating actions).

When appropriate FoMs are identified, the phenomena expected to affect the FoMs can be identified as well as ranked in importance.

Framatome identified three FoMs, which are evaluated by the NRC staff below:

Oscillation inception, which is correlated with the decay ratio. Since the decay ratio describes the relative instability of a system, a higher decay ratio leads to earlier oscillation inception as well as a more rapid increase in oscillation magnitude. As such, this FoM directly affects the timing of failure to rewet, should it be predicted to occur.

This is consistent with the primary FoM for the PIRT developed by the NRC staff as part of Reference 17.

Limit cycle amplitude, which defines the worst possible oscillation that can occur for a given system and core configuration. The oscillations that arise during an ATWS-I will reach a maximum amplitude due to physical limitations on the severity of the density and power swings. Previous NRC experience (e.g., Ref. 10) indicates that the limiting amplitude is not well correlated with the decay ratio, therefore, to ensure that the worst-case power oscillations are captured, a separate FoM is necessary.

Post-dryout, which generally encompasses the dryout and rewetting behavior. This includes cyclical dryout and rewetting, as well as periods of extended dryout due to failure to rewet. This behavior directly affects the PCT, since loss of cooling due to dryout is the primary cause of any PCT increases during the ATWS-I event that are significant enough to challenge the 2200F limit.

Based on the NRC staffs knowledge of the ATWS-I event as correlated with the information presented in the TR, Framatome did an acceptable job of characterizing the event and the relevant phenomena. Framatome identified a key acceptance criterion, core coolability. Even though maintaining the PCT below 2200F is not a precondition for ensuring that core coolability is maintained during ATWS-I conditions, Framatome proposes use of this acceptance criterion as a proxy for core coolability. This proxy for core coolability is already considered in NRC regulatory requirements as an acceptable way to ensure that core coolability is maintained. In addition, Framatome identified that the timing of specific transitions would be important in providing reasonable assurance that the prescribed operator actions would occur in adequate time to mitigate the consequences of the event. This information was used with the event and phenomena characterization to develop a set of high importance phenomena that must be adequately captured by the ATWS-I analysis methodology. The ranking of phenomena as described in the PIRT documented in ANP-10346P is consistent with the NRC staffs understanding of the ATWS-I event.

The NRC staffs conclusions regarding the ATWS-I PIRT were used to guide other portions of the review, particularly the NRC staffs review of ANP-10346P, namely the Evaluation Model, Code Assessment, and Uncertainty Analysis sections of this SE. The following table indicates which section of this SE is associated with each of the high and medium ranked phenomena from the PIRT documented in ANP-10346P. The low ranked phenomena are also captured or otherwise dispositioned in the analysis methodology, but are not expected to have enough of an impact on the ATWS-I evaluation results to require a high level of fidelity or sensitivity studies.

While not explicitly mentioned in the below table, the integral benchmarks discussed in Sections 4.3.6 and 4.3.7 provide valuable validation of the ANP-10346P methodologys ability to conservatively predict the important phenomena affecting the FoMs associated with the ATWS-I event.

High Importance Phenomena NRC Evaluation (including relevant sections from this from Table 4-1 of ANP-10346P SE)

High Importance Phenomena NRC Evaluation (including relevant sections from this from Table 4-1 of ANP-10346P SE)

Medium Importance NRC Evaluation (including relevant sections from this Phenomena from Table 4-1 of SE)

ANP-10346P

As a result of the above discussion, the NRC staff has determined that Framatome appropriately characterized the ATWS-I scenario, identified the appropriate acceptance criteria, and constructed a PIRT that identified the most important phenomena and processes for the analysis methodology to capture. The NRC staff considerations in determining whether the analysis methodology is acceptable with respect to each phenomenon are discussed in the following sections, as described in the above table. In general, the oscillation inception FoM was addressed by examining how the given model or correlation affects the timing of oscillation onset. The limit cycle amplitude and post-dryout FoMs were generally considered through use of the PCT as a proxy, since a conservative application of these FoMs would be expected to increase the PCT. In several cases, the FoMs were not explicitly evaluated because the model or correlation was already established to accurately capture the phenomenon of interest.

4.2 Evaluation Methodology Chapter 15.0.2 of the SRP describes the review of the evaluation model as part of the transient and accident analysis methods. The associated acceptance criteria indicate that models must be present for all phenomena and components that have been determined to be important or necessary to simulate the accident under consideration. The chosen mathematical models and the numerical solution of those models must be able to predict the important physical phenomena reasonably well from both qualitative and quantitative points of view. Restated in terms of the review procedures provided in Section III of Chapter 15.0.2, it must be determined if the physical modeling described in the theory manual and contained in the mathematical models is adequate to calculate the physical phenomena influencing the accident scenario for which the code is used.

A number of models have been previously reviewed and approved by the NRC for similar purposes, so the scope of the NRC staff review was limited to that necessary to confirm the applicability of these models to the ATWS-I event. The models described in ANP-10346P are discussed in individual subsections below.

Major Assumptions Section 4.16 of the TR identifies major assumptions made in RAMONA5-FA ATWS-I relative to the previously approved RAMONA5-FA LTSS methodology. These major assumptions were identified and justified primarily through engineering judgment, based on extensive application experience with the approved RAMONA5-FA methodology which is largely similar to the current methodology.

The major assumptions identified in Section 4.16 of the TR are: (1) the three dimensional (3D) nodal adaptive neutron kinetics methodology is assumed adequate for ATWS-I and (2) new water property functions are used and ['' '''''''''''''''''''''''' '''''''' '''''''''''''''''''''' '''''''''']. The NRC staffs discussion and evaluation of these two assumptions is contained in Sections 4.2.1 and 4.2.3.11.5 of this SE, respectively.

4.2.1 Review of ANP-10346P Section 5.1 - Neutronics The ANP-10346P methodology uses an adaptive 3D neutron kinetics solution with ['''''''''' ''''''

'''''''''''''''''''''] to determine the time evolution of the 3D neutron flux distribution during anticipated transient events. This neutronic solution methodology is identical to that used in the RAMONA5-FA LTSS methodology and the Monticello ATWS-I methodology. This adaptive 3D neutron kinetics solution is a [''''''''''''''''' '''''''''''] methodology, which means that it solves for the

neutron flux level at each discretized axial level in each fuel assembly in the core, [''''' ''

'''''''' '''''' ''''''''''''''''' ' '' '''''''''''''' ''''''''''''''''' '''''''''''''']. The neutronic and thermal hydraulic solutions are coupled on a nodal level as well. Importantly, this means that the coupled neutronic/thermal-hydraulic methodology has sufficient fidelity to accurately resolve any anticipated axial and radial oscillation pattern, including core-wide and side-to-side radial mode behavior (including more complex modal interactions such as rotating modes) as well as single-channel instability. This was a key reason why the [''''''''''''''''' '' ''''''''''''' '''''''''''''''

''''''''''''' '''' '''''''''''''''' ' '''''''''' '''''''''''''''''''''''' '''' ''''''''].

During the large-amplitude oscillations that are characteristic of the ATWS-I event, up to and including limit cycles with dryout and failure to rewet, the neutronic solution becomes even more highly-peaked spatially and undergoes larger variations over time relative to LTSS applications with smaller oscillation amplitudes. However, based on the NRC staffs knowledge and experience with similar neutronics methodologies, this behavior is not expected to challenge the ability of the methodology to accurately represent the physical behavior under these conditions.

In fact, the [''' ''''''' ''''''''''' ''''''''''''''''''' ' '''''''''''''''''''''' '''''''''''''''''''' ' '' '''''' ''''''''''''''''''''

''''''' ' ''''''''''' '''''' ''''''''''']. Therefore, the NRC staff has concluded that the adaptive 3D neutron kinetics solution remains applicable and appropriate for this application.

Because the neutronic methodology in ANP-10346P did not change with respect to the RAMONA5-FA LTSS methodology, and this neutronic methodology remains suitable for ATWS-I applications, the NRC staff did not perform a detailed review of the neutronic methodology as a whole. However, the NRC staff did review the methodology for artificially applying noise to the neutronic solution during a transient calculation. The NRC staffs previous experience has indicated that the method of applying noise is important for correctly determining the timing of oscillation onset, which affects the ability of the methodology to predict whether operator actions occur sufficiently early to mitigate the potential public safety consequences of ATWS-I.

To assist in determining whether the implementation of artificial noise was acceptable for ATWS-I, the NRC staff asked for additional information in RAI-13. In the RAI response (Ref. 5),

Framatome discussed that the noise is applied [''''''''''''''''''''''''''' ' ''''''' ''''''''''''''' ''''''''''''''' ''

''''''''''''''''' ''''''''' '''''' '' '''''''''''''''''''''''' '''''''' '''''' ''''''''''''''''' '' '' '''''''''''''''''''] This information provides a high degree of confidence that both the in-phase and out-of-phase modes can be adequately and reliably excited in a timely fashion (i.e., shortly after one or both modes become unstable), [''''''''''''''''''''''''' ' '' ''''''' '''''' ''''' ''''''' ''' '''''''' '''''''''

'''''''''''''''''''' ' ''''' ''''''''''''''''' '' '].

The NRC staff finds the implementation of ['''''''''] in analysis of the ATWS-I event using RAMONA5-FA as described in ANP-10346P to be acceptable, with a limitation and condition to ensure that excessive tuning of ['' '''''''''] does not occur without justification. The use of the neutron kinetics solution implemented in RAMONA5-FA was also found to be acceptable, based on previous NRC approvals and the known ability of this methodology to capture neutron kinetics responses similar to those expected during an ATWS-I event.

4.2.2 Review of ANP-10346P Section 5.2 - Fuel Thermodynamics The following subsections discuss specific aspects of the fuel thermodynamics models that are relevant to accurately capture the temperature response in the fuel and subsequent heat transfer to the coolant during an ATWS-I event.

4.2.2.1 Review of ANP-10346P Section 5.2.1 - ATWS-I Fuel Pin Heat Conduction The methodology described in ANP-10346P determines the time-dependent axial and radial temperature distribution in the average rod within each fuel assembly, as well as in the hot rod (peak power rod) within each assembly. The average rod temperatures and heat generation rate are used [' '' ''' '''''''''''''''''''''''''' '''''''''''''' ''''''''''''''''' '''''''''''''' '' ''''''''''''''''''

''''''''' '' ''''''''''''''''''''''''' '''''''''''''''''''' ' ' '' ''' ''' '''''''''']

At each axial level in the assembly, a one-dimensional (1D) radial time-dependent transient temperature calculation is performed from the radial center of the fuel pin to the outer surface of the cladding, similar to the Monticello ATWS-I and RAMONA5-FA LTSS methodologies. This approach is on par with other state-of-the-art methods and provides a suitably accurate and realistic calculation for oscillatory ATWS-I conditions. Consistent with the previous methodologies, [''''' '''''' ''''''''''''''''''''' ' ''''''''''''''''''''], which is acceptable based on the fact that the model exhibits good agreement with experimental benchmarks, as discussed in Section 4.3 of this SE.

Unlike the ['''''''''''' '''''''''''] in the Monticello ATWS-I methodology, the ANP-10346P methodology solves the radial temperatures ['''''''''''''''' '''''''''' '''''''''''''''' '' ''''''''' '''''''''''''''''

''''''''''''''''''' ''''''''''' ''''' '''''''''''''''' '' '''''''''''' ''''''''''''''' ''''''' '']

Therefore, the NRC staff finds the ANP-10346P fuel rod conduction methodology to be acceptable based on its use of previously approved modeling approaches combined with state of art computational solution schemes appropriate for the intended application.

4.2.2.2 Review of ANP-10346P Section 5.2.2 - ATWS-I Heater Rod Conduction Model A separate heat conduction model is used for calculating time-dependent axial and radial temperature distribution in heater rods representative of the KATHY facility. The NRC staff determined that the only difference between this model and the one in the Monticello ATWS-I methodology was that the latter calculated the ['''''''''''''''''''''''' ''''''''''''''''''''''' '''''''''' ''' '''' '''''''''''

'' '''''''''' '' ''''''''''''''' ''''' '''''''''''''''''''' '''' ''''''''''' ''''''' ''' ''''''''''''''''''' ''''''''''''''' ']

Because the heater rod conduction model in the ANP-10346P methodology is more accurate than the previously-accepted model used in the Monticello ATWS-I methodology, and is used for the same scope and range of application - namely, to determine the heater rod temperature response during the KATHY ATWS-I experiments - the NRC staff concludes that the previous approval of the heater rod conduction model in the Monticello ATWS-I methodology is applicable to the ANP-10346P methodology, and no further review of the model was performed.

4.2.2.3 Review of ANP-10346P Section 5.2.3 - Heat Transfer Coefficient The ability of the fluid to transfer heat from the outer surface of the clad or heater rod is strongly dependent on the phase of the fluid (liquid, vapor, or both) and the ability of the liquid phase to contact the surface. The ANP-10346P methodology calculates a wetted heat transfer coefficient (HTC) under single-phase conditions using [''' ''''''''''''''''''''''''''''''] correlation, a wetted HTC under two-phase conditions using [ '''''''''''''' ''''''''''''''''''''] correlation, a dry HTC using [

''''''''''''''''''''' ''''''''''' ' ''''''''''''' '''''''''''''''''''''' '''''], and models for transitions between these regimes.

The [''''''''''''''''''''''''''''] single-phase liquid correlation and the [''''''''''''''' '''''''''''''''''''''''' '''''''''''']

correlation are the same as in the RAMONA5-FA LTSS methodology and the Monticello ATWS-I methodology. [''''''''''''''' '' ''''''''''''''''''''''''' '''''''''''''''''''''''' '''''' ' ''''' '''''''''' '

'''''''''''''''''' ' '''''' '''''''' '''' '''''''''''''''' '''''''''''''''''''''' ''''' ' ''''''''''''''''''' '' ''''''''''''] The NRC staff concludes that the single phase liquid and boiling heat transfer models are acceptable based on their previous validation and approved use in the RAMONA5-FA LTSS methodology, and that the regime transition and ['''' ''''''''' '''''''''''''] are acceptable because they are based on realistic physical principles and demonstrate good agreement with measured data in the benchmarks given in Section 6.0 which cover a wide range of conditions applicable to ATWS-I.

The dry HTC is determined using a correlation [''' ' '''' '''' ' ' '''''''''' '''''''''''''' '''''''''''

]

The NRC staff issued RAI-3 to address the NRC staffs concerns regarding the acceptability of using a ['''''''''''''' ''''''''''''''''''''' ''''''''''''''''''''' ''''''''' ''''' ''''''''''''''' ''''''''''''''''' ' ''''''''''' '''''

'''''''''''''''' ' '''''''''' ''' '' ''''']. In particular, RAI-3 requests more information to justify [''

''''''''''''''''''''' ''''' ' '''''''']. In the RAI response (Ref. 5), the licensee provided additional plots

[''''''''''''' '' ''''''''''''''''' '''''''''''''''''''''''''''''''' ''''''' '''' ''''''' '''' '''''''''' '''' ''''''''''' '''''''''''''' ''''''''

'''''''' '''''''' ''''' '''' ''''''''''''''''''' ' ''' '''''''''''']

However, the NRC staff observed the following:

(1) the magnitude of the power oscillations following failure to rewet was ['''''''''''''''''' '''''''

'' '''''''''''''''''' ' ''''' ''' ''''''''''''''' '''' '''''''''];

(2) the dry HTC correlation was [''''''''''''''''''' ' '''''''''''''''' ' ''' '''''''''''''' ''''''' '''''''''''

' '' '''''''''''''''''' '''''''''']; and (3) Framatome only used the [''''''''''''''''' '''''''' ' '' '''''''''''''''''''''''''''''''''''''''''''' ''''' ''''''''''

'''''''''''''' ''''''''''''''''' '''' '' '''''' '''''''' ' ''''''''''' '' ''' ''''''' ''''' ''' ''''''''''].

Additional data points were provided by Framatome from ['''''''' ''''''''''''''''''''' ''''''''''''''''''''''

''''' ' '] This is due partly to the fact that KATHY provides prototypical ATWS-I conditions, and the NRC staff does not expect the heat flux to go significantly beyond the KATHY data range without causing the fuel to exceed 2200 F.

The data provided by Framatome demonstrates a clear correlation between ['' '''' '''''' '''''

''''''''''''''''''''] that is reasonably bounding by incorporating the conservatisms discussed above, as items (2) and (3).

[''''' '''''''''''''''' '''''''''''''''''''''''' '''''''''''''''''' ' ''''''''''''' '''' '''''''''''''''' ''''''''''''''' ''''''''''''''''''''''''''' ''''

''''''''''''''''''''''' '' ''''' '''''''''''''' ''' ''''''''''''''' '''''''''' ''''''''''''''''' ''' '' ''''''' '''''''''''''''''' '' ]

4.2.2.4 Review of ANP-10346P Section 5.2.4 - Hot Fuel Pin Model As discussed in Section 4.2.2.1 of this SE, the ANP-10346P methodology provides a separate calculation for temperature and heat transfer in the hot fuel pin as opposed to the average fuel pin, which the NRC staff finds to be an acceptable approach to both calculate the maximum cladding temperature and provide realistic coolant temperatures and reactivity feedback for the neutronics solution.

['''''''''' ''' '''' ''''''''''''''' ''''''''''' ''' '''''' ''''''''''' '''''' '' ''''''''''''''''' '''''''''''''''''''''''''''''''''''''''' ''''''

''''''''' ' '''' ''''''''' '''''] The NRC staff finds this approach acceptable because it provides conservative [''' ''' '''''''' '''''''''''' '''''''''' '' '''''''], which is expected to increase the calculated PCT values during ATWS-I analyses.

4.2.2.5 Review of ANP-10346P Section 5.2.5 - Material Properties The ANP-10346P methodology uses fuel pellet and cladding thermophysical properties based on [''' '''''''''''''''' ''''''''''''''''' ''''''''''''''''''''''''' '''''''''''''''''''' ''']. The NRC staff finds this approach acceptable for use in the RAMONA5-FA ATWS-I calculations because these models account for all important fuel characteristics relevant to ATWS-I, including the ['''''''''''''''''''''''' ''''''''''''''''''''''''''

'''''''''''' ' '''''''''''''''''''''' ''''''''''''''' '''' '''''''''''' ''' '''' ''''''''].

Appendix A of Reference 5 includes an update to ANP-10346P that, among other changes, appends Appendix D, which presents modified fuel rod models that account for chromia doping of the UO2 fuel pellets. The fuel thermal conductivity model was adapted from the approved RODEX4 model in Reference 18. The [''''''''' '''' ''''''''''''] model was developed by benchmarking to the approved RODEX4 model in Reference 18. Because these models are based on previously reviewed and approved models for chromia doped fuel, the NRC staff finds these models acceptable for use in characterizing chromia doped fuel properties for ATWS-I analyses performed using the methodology as described in ANP-10346P.

4.2.2.6 Review of ANP-10346P Section 5.2.6 - Pellet Clad Gap Heat Transfer Coefficient The gas gap between the fuel pellet and cladding may introduce a large thermal resistance which affects both the amplitude and phase shift of fluctuations in heat flux at the cladding outer surface during a given oscillation period. In turn, the decay ratio and oscillation frequency of predicted ATWS-I oscillations may be significantly affected, which may impact whether the fuel remains protected within the time required for the ATWS-I mitigation actions to take effect. Due to burnup and history effects, the fuel-clad gap in twice- and even once-burned fuel will typically be closed at normal operating conditions, resulting in only a small thermal resistance; however, after the recirculation pump trip during the postulated turbine trip with bypass (TTWB) and 2RPT ATWS-I events, the gap will typically re-open and result in significant thermal resistance that must be accurately accounted for in the ATWS-I methodology.

['''' ''''''''''''''''''''''' ''''''''''''''''''''''''' ''''''' ' ''''''''''''' '''''''''''''''''''''''''''''''''' ''''' '''''''''''''''''''''' ''''''''''' ''''''''

'''''''''''''''' '' ''''''''''' '''''''''''''''' ' '''''''''' '''''' ''''''' '''''''' ''''''''''''''''''' '''''''''''''''''' ' '''''''''''''''']

['''' '''' '''''''''''''''''''' ' '''''''''''''''''''' ' '''''''''''''''' '''''''''''''''' '''''''''''' ''''' '''''''''' ''''

'''''''''''''''''''''''' ''''''''''' '''''''''''''''''''''''''''' '''''''''''''''''''''''' ''''''''''''''''''''']

In RAI-2, the NRC staff requested additional information on how the fitting parameters for the gap conductance model were determined from measured data, particularly when direct experimental validation for each parameter was not possible or not available. In particular, RAI-2 requested additional information to better understand the method of ['''''''''''''''' '' '''''''

'''''''''''''''''''' ''''''''''''''''''''''''''' ' ''''''''''''''''''''''''' '''' ''''' '' ''] Because these values cannot be directly measured but have a potentially significant effect on stability behavior, these values could have subsequently been adjusted [ '''''''''''''' ''' '''''''''''''''''''''' '''' '''''''''''''''''' ''''''''' '''''

'''''''''''''''''' '' '''''' ''''''''''''''''''''' ''''' ''''''''''''''''''] After examining the relevant models and experimental database, the NRC staff has concluded that the use of [''''''''' ''''''''''''''''' '''''''''

''''' ''''''''''''''''' '''' '''''''' '''''''''''' ''''''''''''''''''''' '''''' ''''' ''''''''''''''''''''''''''''' ' '''''''''''''] is acceptable because it provides a physically reasonable model of gap behavior as a function of fuel conditions (burnup, temperature, etc.) in relevant ATRIUM fuel types and, furthermore, the good agreement with [''''''''''''''''' ''''''''''''' '''''''''''''''''''''] provides strong evidence that these values remain applicable and acceptable under BWR stability conditions.

The NRC staffs review of the gap heat transfer coefficient model concluded that it includes the important physics required to calculate the intra-pin heat transfer behavior during a postulated ATWS-I event, including the initial transient, onset and growth of oscillations, dryout/rewet phase, and high-temperature failure-to-rewet phase of the event. The models for calculating

['''' '''''''''''''''''''''''' '''''''''''''''' ''''' '''' ''''''''''''''''''''''''' '' ''''''''''''''' '''''''''''' ''''''' '''''' ''''''

'''''''''''' ''''''''''''''''''''''''''' ''''' ''''''''''''''''''''''''' '''''''''].

['''' ''''''' '''''' ''''''' ''''' ''' ''''''''''''''''''' ''''' ''''''''' ' '''' ''''''''''''''''''' '' '''''''''''''''''

''''''''''''' '''''''''''''''''' ''''''''''' '''' ''''''''''''''''''' '''''' '''''''' ''''''''''''''''' '''''''''''''''''' '''''''''''''''' '''''''''']

Because of this, the NRC staff issued RAI-4 to request justification that the gap heat transfer coefficient model provides a reasonable and accurate representation of gap behavior during postulated ATWS-I events.

In the RAI response, Framatome indicated that the ['''' ''''' '''''''''' ' ''''''''''''''''''' '''' ''''

'''''''''''''''''''''''''''''''''''''''''' '''''''''''''''''''''''' '''''''''''''''' ' '''''''''''''''''''' '''''''''''''''''''' ''''], as discussed above.

['''''''''''''''''''''''''' '''''''''' '''' '''''''''''''' '''''''' ''''' ''''' '' ''''''''' ' ''' '''''''''''''''''''' ''''

'''''''''''''''''''''''''''' '''''''''''''''''' ''''''''''''''''''''''' ''''' '''''''''''''''''' '''''' '''''''''''''''''] While these data do not provide a direct separate-effects validation of the fuel and gap heat transfer models, the close overall agreement with these measured integral effects data, with little or no average bias in the errors, provided the NRC staff with confidence that the fuel and gap heat transfer was modeled in a reasonable and acceptable manner.

To provide the NRC staff with further understanding of the role of the gap model during ATWS-I events, the NRC staff issued RAI-11 which requests additional sensitivity results for one or more linear stability benchmark cases and a simulated ATWS-I event by adjusting the gap conductance values. In the RAI response, Framatome provided the sensitivity results for the linear and nonlinear cases by artificially adjusting the gap conductance [' ' ''''''' ' '' '

''''' ''''''''] Based on this [''''''''' ''''''''''' '' ''''''''' ''''''], the NRC staff concluded that the gap conductance model has ['''''' '''' ''''''''] on the stability predictions for regional mode cases.

For the single linear benchmark case in which the global mode was dominant [''''''''' '''''''' '

'''''''' ''''''], respectively, compared to the result with non-adjusted conductance. Framatome indicated that [''''''''''''' '''''''''''''''''' '' ''''''''''''''''''' ''''' '''''''''''''''''''''''' ''''''''''''''''' ''''''''''''''' '''''

'''''''''''''''''''''''' ''''''''' ''''''''''''''' ''''''''''''''' ''''''''' '''''''''' '''' ] In this particular case, the more significant sensitivities were such that the base calculation represented a reasonably conservative result. However, the interrelationship between the gap conductance model and the stability phenomena is expected to be at least somewhat sensitive to the specific scenarios being analyzed.

The ['' ' '''''''''''] adjustment in gap conductance also had a relatively mild effect for the ATWS-I sample problem, for which the sensitivity results were provided in RAI-11 as well. ['''

''''' ' '''''''''''''''''''''' ''''''''''''''''''''' ' '''' '' ''''''''''''' ''''''''''''''''''''''''''' ] Any impact due to uncertainties in the gap conductance model that may result in challenges to the regulatory limit would be expected to occur in situations where operator action is necessary to prevent the PCT from exceeding regulatory limits, and where relatively small margins exist between the licensing basis operator action time and the time at which operator action would be too late to stop the PCT from increasing beyond the regulatory limit. In such cases, a significant increase in the decay ratio may lead to an earlier failure to rewet and allow the PCT to increase for a longer time prior to mitigation. Once the margin in operator action time is appropriately justified, including any consideration of the gap conductance uncertainty, the sensitivities are not expected to change significantly from cycle to cycle. However, they may change when a new fuel design is introduced that changes the characteristics of the geometry or materials used in modeling the fuel rod, gap, and cladding. Consequently, a limitation and condition is necessary to verify that the uncertainties associated with the gap conductance model continue to be small enough to be readily accommodated by the available margins in operator action time.

Based on the similarity [''''' '' '''''''''''''''''' '''''''''''''''], inclusion of the important physics relevant to ATWS-I, close agreement of the RAMONA5-FA ATWS-I results to measured BWR stability data, and [''''''''''''' ''''''''''''''''''''''] of the stability results under most scenarios to variations in gap conductance, the NRC staff concludes that the fuel rod heat transfer model, including the gap conductance model, are acceptable for use in the ATWS-I analyses. In order to address specific scenarios where the gap conductance model may become important, a limitation and condition was imposed to require evaluation of the uncertainty in gap conductance for certain changes in fuel design, as described in the previous paragraph.

4.2.2.7 Review of ANP-10346P Section 5.2.7 - Radial Power Deposition Distributions in Fuel Pellets The ANP-10346P methodology determines the radial power distribution within fuel pellets using

['' ''''''''' ''''''''''''''' '''''''''''' '''''' ''' '''''''''''''''''' ''''''''''''''' '''''''''''''''''''''''' ''''''''' '''''

'''''''' '' '''''''' '''''''''''' '''''''''' ''' ''''''''''''''''''''''''''''''' '''''''' ']. The NRC staff has reviewed the methodology and determined that it provides the needed accuracy for calculating the radial power distribution in fuel pellets, including [''''''''''''''''''''''''''''''''''' ''''''''''' ''''''' ' '' ''''' ''''''''

''''''''''''''''' ''' '''''''''''''''' ''''' '''''''''''''''''''' ''''''''''''''''''''''' ''' ''''''''''''''''''' ' '''] Therefore, the NRC staff finds the radial power distribution methodology to be acceptable.

4.2.3 Review of ANP-10346P Section 5.3 - Thermal-Hydraulic Model The RAMONA5-FA ATWS-I methodology described in ANP-10346P utilizes a ['''''''''''''''''''''' ]

TH model comprised of ['''' ''''''' '''''''''''''''''''' '''''''' '''' '''''''''''' ''' ''''''''''''' ''''''''''''''''' ''''''''''

'''' ''''''''''' '''' '''' '''''''''''''''''''' ''''''''''''''''' '''' ''' '''''''''''''''''''''''' ''''''''''''']. A description and the NRC staff evaluation of the thermal-hydraulic model is given in the following subsections.

4.2.3.1 Review of ANP-10346P Section 5.3.1 - General Description of the System Considered The general system modeling in ANP-10346P consists of nine main components as shown in Figure 1 and is identical to the vessel methodology in the RAMONA5-FA LTSS methodology.

['''''' '''''''''''''''''''''' ''''''''''''''''' ''''''''''''''''''''''''' ' ''''''''''''''' ''''''''''' '''''' '' ''''''''''''''''''' ''''''''''''

''''''''''''''''''''''''''''' '''''''' ''''''''''' ''''''''''''''''' '''''''''''''''' '''''''''''''''' ' '''' ''''''''' '''''''''''''''''''].

Accurate modeling of the pressure losses and flow inertia in the vessel flow path is important for correctly determining flow rates and other core parameters during ATWS-I events; this is especially true for in-phase (core-wide) oscillations in which the total core flow rate experiences large time-dependent changes which become coupled to time-dependent flow rate changes in the surrounding components. Vessel flow inertia is particularly dependent on the recirculation pump model which is evaluated in Section 4.2.5.1 of this SE; however, the pressure losses and flow inertia in the remaining vessel components are relevant to ATWS-I analyses as well.

Figure 1 - Loop parts in the vessel hydraulics model (ANP-10346P Figure 5-3)

Because the ANP-10346P methodology contains the same vessel hydraulics treatment as the RAMONA5-FA LTSS methodology, which is approved for use in LTSS stability analyses including the analysis of in-phase oscillations where vessel pressure losses and flow inertia are important, these aspects of the vessel hydraulics model were not reviewed for ANP-10346P and the existing approval of these modeling aspects from the RAMONA5-FA LTSS methodology remains applicable. Furthermore, the RAMONA5-FA LTSS methodology is approved for 2RPT LTSS analyses, which is an identical event to ATWS-I 2RPT except that ATWS-I 2RPT assumes failure to scram. Although this failure to scram allows for larger-amplitude oscillations and therefore larger oscillations in flow rate and other thermal hydraulic parameters in the vessel components, these conditions do not impose additional physical modeling requirements on the vessel thermal hydraulic methodology, and this vessel methodology remains suitable for ATWS-I 2RPT.

The ATWS-I TTWB event is similar to ATWS-I 2RPT in that both involve a dual recirculation pump trip; however, the TTWB event requires modeling of turbine stop valve closure and turbine bypass valve opening, which impact the pressure response in the vessel including pressure wave propagation.

Additionally for the TTWB event, the decrease in feedwater temperature due to loss of feedwater heaters must be modeled, and the vessel model must be able to accurately model the mixing of the cold feedwater with the saturated liquid leaving the steam separator, and accurately transport this fluid through the downcomer and lower plenum to ensure proper timing

and magnitude of core inlet temperature decrease during the event. Accurate calculation of the time-dependent core inlet temperature is necessary to correctly predict the oscillation onset timing and magnitude. If the water level falls below the feedwater inlet to the vessel (feedwater spargers), significant heating of the subcooled liquid feedwater and condensation of the steam in the downcomer may occur, which may affect the core inlet temperature behavior as well.

Details on the steam line flow dynamics, recirculation pump model, jet pump model, steam separator model, and feedwater sparger condensation model are given Sections 5.4 through 5.5.4 of ANP-10346P and are evaluated later in this SE.

The core region consists of a number of parallel fuel assembly channels and [''''' ''''''''''''''''''] or bypass channel. The bypass channel accounts for the inter-channel flow (between channel boxes) as well as the flow through the internal water rods in each assembly. There are numerous leakage paths from the lower plenum to the bypass region. ['''''''''' '''''''''''''' '''''''

'''''' ''''''''''''''' '''''' ''''' '''' '''''''''''''''' '' ''''''''''''''''''''' '''''''''''''''''''' '''''''''''''''''' ' ''''''''''''' ]

The NRC staff issued RAI-1 to better understand the process for passing thermal hydraulic information from MICROBURN-B2 to RAMONA5-FA for ATWS-I analyses and ensure consistent solutions between the two codes during initial steady-state conditions. In the RAI response, Framatome provided a detailed description of the ['''''''''''''''''' ''''''' '''''' '''' '''''

''''' '''' ''''''''''''''''''''' ''''''''''''' '''''''''''''''''''' ''''' ''''''''''''''''' ''''''''''''''''''']. These inputs to the RAMONA5-FA ATWS-I methodology are the same as used in the approved RAMONA5-FA LTSS methodology. Both versions of the RAMONA5-FA code use these inputs to perform a thermal hydraulic calculation ['''''''''' '' '''''''''' ' '''' '''''''''''''''''''' ' ''''''''''''''''''''''''''''''''''' '''''

'''''''' '''' ''''' '''' '''''''''''''''''''''' ''''''''''' ''''''''''''''''''''''''' ''''''''''''' ''' ''''''''''' '''''''''''''''''' ' '''']

['''''''''''''' ' '''''''''''' ''''''''''''' ''''''' '''''''''''''' ' '''''''''''''''''''' ' '''''''''''''''''''''' ''''''''''''' '''

'''''''''''''''' ''''''''''''''''''''''''' ''''''''''' ''''''''''''' '' ''''''''''''''''''''''''''''''''''], the NRC staff concludes that the approach for determining the initial thermal hydraulic solution in the RAMONA5-FA ATWS-I methodology is acceptable.

Bypass flow, including water rod flow, constitutes a relatively small fraction of the total flow through the core, which limits its hydraulic impact during oscillations. However, direct gamma heating of the bypass flow may cause localized boiling in the bypass region, which may have a significant effect on the power level of neighboring fuel bundles due to neutronic feedback.

Bypass voiding is most likely when stagnant or reversed bypass flow is experienced, which

occurs at very low core flow rates due to the relatively large gravitational pressure head of the bypass liquid column.

However, including [''''''''''''' '''''''''''' '''''''''''''''''' ' ''' ''''''''' ''''''' ' '''''''''''''''''' '''''''''''''

''''''''''''''''' ''' ''''''' ''''' ''''''''''''''' ]. Therefore, the NRC staff concludes that modeling the bypass [' '''''''''' ''''''''''''' '''''''''''''''''''''' '' '''''''] is expected to give conservatively high PCT results and is therefore acceptable for the ATWS-I analyses.

Vessel Nodalization In RAI-9, the NRC staff requested additional information to ensure that the vessel nodalization provides sufficient fidelity for liquid and vapor transport in the vessel such that the system behavior including PCT is accurately predicted for ATWS-I events. In the response to RAI-9, Framatome specified the number of nodes used in each region of the vessel model under the base nodalization scheme, and these node numbers were [''''''''''''''''''' ' '''''''''' ' '''' '

'''''''''''''' ''' ''''''' ''''''' '''' '''' ''''''''' ' ''''' '''''''''] for the nodalization study in RAI-9.

Framatome clarified that the most limiting nodes in the model, [' '''''''' '''''''''''''''''''' '''''''' ''

'''''''''''''''' ''''''''' '' '''''''''''''''' '''''''''].

During in-phase oscillations, the coolant flow rate and void fraction in the primary circulation loop will oscillate along with the oscillations in the core. These time-varying thermal hydraulic quantities in the vessel will impact the recirculation loop momentum dynamics and may therefore impact the stability characteristics of the system. This impact is expected to be negligible for out-of-phase oscillations because the thermal hydraulic conditions outside the core remain essentially constant in this case.

In the vessel nodalization sensitivity cases provided in RAI-9, performed for the Brunswick sample problem, ['' '''''''''''''''''''''' ''''''''''' '''''' '''''''''''''' '''' ''''''''''''''' '''''''' ''''''''

''''''' ''''''''' '''''''''''''''' ''''''''''''' ], showed good agreement with the measured stability data using the base vessel nodalization. Furthermore, the NRC staff expects that other BWR plants will behave similarly with respect to vessel nodalization, with no significant differences that would be expected to materially change this finding. Therefore, the NRC staff concludes that the momentum-related effects associated with vessel nodalization are expected to be insignificant and the base vessel nodalization used in the TR is acceptable in this regard.

The vessel nodalization is expected to have an additional effect, with respect to the time-dependent core inlet subcooling. This is due to the effect of numerical diffusion on energy transport in the vessel liquid. Therefore, Framatome provided a plot of core inlet subcooling versus time during the Brunswick sample problem for each nodalization case. The time-dependent subcooling behavior was ['''''''''''''''''''' '''''''''''' '' ''' ''''''''''''''' ' '' '''''''' '''''''''

'''''''''''''''' '''''''''''''''''''' ''''''''''']

As a result of this small change in inlet subcooling, the [''''' ''''' ''''' ''''''''' '''''''''''''''''''''' ''''''''

''''''''' '''''''''] The relative insensitivity and [''''' '''''' ''''''''] in both the failure-to-rewet time and PCT, [''''' '''''' ''''''''''''''''' ''''' '''' ''''' ''''''''''''''''''''''''] will consistently produce results more or less conservative than other nodalizations, leads the NRC staff to conclude that the base vessel nodalization as specified in the RAI-9 response is reasonable and acceptable for use in the RAMONA5-FA ATWS-I methodology.

4.2.3.2 Review of ANP-10346P Section 5.3.2 - ['''''''''''''' '''''' '''''']

This section of ANP-10346P describes the spatial discretization scheme used by the RAMONA5-FA ATWS-I thermal hydraulic core channel solution. [''''' '''''''''''' ' ''''''''' '

''''''''''''''''''' '''''''''''''''''''' '''''' ''''''' '''''''''''' '''' '''''''''''' '''''''' ' ''''' ' '' '''' ]. This method is suitable for determining the flow behavior during normal conditions (i.e., upward flow through the bundle) as well as transient behavior such as periodic flow reversal expected during large-amplitude ATWS-I oscillations. Therefore, the NRC staff finds this spatial discretization scheme, which is the same as used in the RAMONA5-FA LTSS methodology, to be acceptable for ATWS-I applications.

4.2.3.3 Review of ANP-10346P Section 5.3.3 - Vapor Generation Rate The nodal vapor generation rate in ANP-10346P is calculated [' ' ' '''''''''''''''''''''' '''''''

]. This model is the same as in the Monticello ATWS-I methodology, and similar to the RAMONA5-FA LTSS model except for the modifications to allow for the [''''''''''''''''''''''''' '''''' '''''

'''''''''''''''''']; therefore, these modifications from the previously-approved methodology are appropriate. The NRC staff has reviewed the new model and determined that it acceptably models vapor generation during ATWS-I events.

4.2.3.4 Review of ANP-10346P Section 5.3.4 - Mass Conservation The ANP-10346P methodology solves separate liquid and vapor mass conservation equations, using [' '''' '''' '''''''''''''''''''''' ''''''''''' '' ''''''''''''''''''''''''''''''''''''' ''''''''''''''''' '' ''''''' ''''''''''''''

' ''''''' ''''''''' ''' '''''''''''''''''''''''' ']

Large-amplitude oscillations may exhibit any of the following flow scenarios: co-current upward flow (liquid and vapor both flowing upward), co-current downward flow (liquid and vapor both flowing downward), and counter-current flow (liquid and vapor flowing in opposite directions).

The NRC staffs review of the mass conservation model has determined that it properly accounts for all of these possible flow scenarios. This, in addition [ ''''''''''''''' '''

''''''''''''''''''''''''''''''''''''''''''''' '''''''''''''''''''''''''''''' '''''''''' ''''''''''''''''''''] and its ability to give realistic behavior at or near fully-voided conditions, has led the NRC staff to conclude that the mass conservation model in ANP-10346P is acceptable for ATWS-I analyses.

4.2.3.5 Review of ANP-10346P Section 5.3.5 - Energy Conservation The ANP-10346P methodology uses a [''''''''''''''''''''''''''''''''''''''''''''' '''''''''''''''''''''' ''''''''''''''' '''''''' '''''

''''''''''''''''' '''''' ''''' ''''''''' '''''''''' ''''''''''']

Since scalar quantities such as enthalpy are defined at the center of control volumes and vector (or directional) quantities such as mass flow rate and velocity are defined at the edges of control volumes, the energy balance which includes energy entering or leaving through each edge of the control volume uses different scalar cell indices depending on the direction of flow through each control volume edge. In principle, these directions can be different at the bottom and top edge of each volume. The NRC staff determined that the energy equation formulation properly accounts for all possible combinations of flow directions for the bottom and top edges by making suitable adjustments to the enthalpy index for the donor cell scheme in calculating the rate of energy flow in or out of the control volume for each phase. Thus, the energy balance is properly conserved for any flow situation for both the liquid and vapor phases.

The ANP-10346P methodology differs in the implementation of the overall core energy balance by including an [''''''''''''''''' ''''''' ''''''''' ''' ''''''''' ''''''' ''''''''' '''''''''''''''''' '''''''''''' ' '''''''''''

''''''''''' ''' ''''''''''''''''''' ' ''''] In RAI-10, the NRC staff requested additional information on the behavior of the ['''''''''''' ''''''''''''' ''''''] during postulated ATWS-I events. In the RAI response, Framatome provided a plot of ['' ''''''''''''''''''' ' ''' '''''''''' ''''''''''''' '''''''' ''' '''' ''''''''''''''

'''''''''''''''''''''' ''''' '''''''' ''''' ''''''''''''''''''''''''''' '''''''''' '''''''''' ''''''''' ' '''] In any case, the NRC staff expects that this effect would have only a small impact, at most, on the ATWS-I results.

Therefore, the NRC staff concludes that the implementation of the [''''''''''' ''''''''''' ''''''''] is acceptable.

4.2.3.6 Review of ANP-10346P Section 5.3.6 - [''''''''''' ''''''''''''''']

Section 5.3.6 of ANP-10346P discusses the approach for determining ['''''' ''''''''''''' ''''''''''

'''''''''''''''''''''''''''' '''''''''''''']

[' ' '''''''''''''''''''' '''''''''''''''''''' '' '''''''' '''''''''''''''''' '''''''''''''''' ''''''' ''''''''''''''''''''' ''''''''''' '

''''''' '' ''''''''''''''''''''' '''''' ' ''''''''''' ]

[

'''''''''''''''''' ''''''''''''' ' ''''''''''']

['''''''''' '''' '''''''''''''''' '' '''''''''''''''''''''''''' '''''''''' ''''''''''' ' '''''''''''''''''''' '' '''''''''' '''''''''' '''''''''

'''' ''''''''' ''''' ''''''''' ''''''''']

[''''''''''''''''' ''''''' ''''' ' ''''''''''''''''''' ''''''''''' '''''''' ''''''' '''''''' ''' '''''''''' '''''''' ' ''''' '''''''''''''''

''''''''' '''''''''''']

[''''''''''' ' '' '''''''' '''''''''''''''''''''' '' ''''''' '''''' '''' ''''''''''''''' ''' '''''''''''' '''''''''''''''''''''''' ''''''''''

''' ''''''''''''' ''''''''''''''''''''''' '' '' '' ''' '' ' ]

4.2.3.7 Review of ANP-10346P Section 5.3.7 - [''''''''''''' '''''''''''''''''''''' '''''''''''''''''''''''''']

The ANP-10346P methodology uses ['''''''''''''''''''''' ''' ''''''''' ''''''''''''''' ''''''''''''''''''''''' '''''''''''''

''''''''''''''''' ' '' ''''''''''''''''''''''''''''''' '''''''' '''''''''''''''''''''''], to account for flow inertia and acceleration terms and their effect on the time-dependent pressure drops. ['''' '''''' '''''''''''''''''' '''''''''''''''' '

' '''''' '' '''''''''' ''''' ''''''''] However, the NRC staff finds this implementation acceptable for the reasons given in Section 4.2.3.4 of this SE.

[''''' '''''''''''''' '''''''''''''''''''''' '''' '''''''''''''''''''''' ''''''''''''''''''''''''' '''''''''''''' ''''''''''''' '' ' ''''''''''''' ''''

''''''''''''''''''''''''''''''' '''''''' '''''''''''''' ''' '''''''''''''''' '''''''''' ']

[' ' '''''''''''''''''''' '' ''''''''''''''''''' ''''' '''''''''''''''''' ''''''''''''''' ''''''''''''''''''''' '''''''' ' ' ' ' '''''''''''

'''''''''''''''''''''''''''''''''' '''' ''''' ' '''''' ' ]

['''''' '''''''''''''''''' ' ''''''''''' ' ' '''''''''''' '''''''''''' ' '''' '''''''''' ''''''''''''''' '''''''''''''''''''' '''''''''''''''

'''''''''''''''''''' ''''''''''''''''''''''' ''''''''''''''''''' ' '' '''''''''''''''''''''''''''''''' ''''''''' '] With respect to momentum conservation, the basic phenomena and modeling requirements remain the same for large-amplitude oscillations characteristic of ATWS-I, and the ['''''''''''''' '''''''''''''''''''' ''''''''''''''''

''''''''''''''''''' ' ''''''''''''''''''] and acceptable for this use.

Special treatment is provided in ANP-10346P (the same as in the RAMONA5-FA LTSS methodology) to calculate the pressure response due to valve closures in the steam line, which is relevant for ATWS-I TTWB events. Details of this special treatment are given in Section 5.4 of ANP-10346P and evaluated in Section 4.2.4 of this SE. Since the pressure waves dissipate

rapidly once they reach the larger volumes of the vessel, this treatment is not necessary for the vessel and core regions. Therefore, the ['''''''''''' ''''''''''''''''''''' '''''''''''''''''] is acceptable for use in the vessel and core regions for the reasons stated above.

Acceptability of the ['''''''' '''''''''''''''''''' '''''''''''''''''' '''''''''''''''''']

In a BWR assembly, the liquid and vapor phases will in principle travel at different velocities; these velocities depend on a mass and momentum balance for each phase separately, as well as on mass and momentum exchange between the phases. [ ' '''''''''''''''' '''''''' ''''

'''''''''' '' '''''''' ''''''' '''''' '''' ''''''''''' ' ' ''' '''''']

In methodologies such as the one described in ANP-10346P, [''''''''' ''''''' ' '''''''' ''''''''''''''''''''''''

'''''''''''''''''''''' '''' '' ''' ''''''' ''''''''''''''''''''' '' '''''''''''' ''''''''''''' ''''''''''''''']

[''''' '''''''''''''''''''''''''''' ''''''''''''''' ' '' '''''''''''''''''''''''''''''''''''''''''''''''''''''''''''' '''''''''''''''''''''''' ''''''''''''''''

''''' '''' ''''''''''''''''' '''' ' '''''''''''''''''''''' '' '''''''''''''''''''''''''']

[''' '''''''''''''''' ' '''''''' ''''''' ' ''''''''''''''''''''''''''''''''''''''''' '''''''''''''''' '''''' '''''''''''''''''''''' ''''''''''''''''''''''''

''''''''''''''''''''''''''' ''''''''''''''''''''''''' ''''''' '''''''' '''''' ''''''''''''''''''' ''''' ''''' '']

['''''''''' ''''' '''''''''''''' ''''''''''' '''''' ''''''''''' ' ''' '''''''''''''''' ''''''''''' '''''''''''''''''''''' '''' '' '''''''''

'''''''''''''''''''''''''''''''' '''''''''''''' '''''''''''''''''''''''''''''''' '''''''''' '''''''''''''''''' '''']

[''''''''''''' ''''''''''''''''''''''' '''''''''''''''''''''''''''] Conclusion Because the ANP-10346P methodology uses [''''''''''''''''' '''' '''''''' '''''''''''''''''''''' ''''''''''''''''''''''''''

'''''''''''''''' ''' ''''''''''''''''''''''''''''''' '''''''''' '''''''''''''''''''''''''], and because this approach remains suitable for large-amplitude ATWS-I oscillations - including the ability to model reversed and counter-current flow and the demonstrated conservatism of the single-momentum-equation approach - the NRC staff finds the momentum conservation model in ANP-10346P to be acceptable.

4.2.3.8 Review of ANP-10346P Section 5.3.8 - Pressure Calculation Section 5.3.8 of ANP-10346P describes the methodology for calculating the time-dependent

[''''''''''' ''''''''''''''' ''''''' ' '''''''''''''''' '' '''''''''''' ''''''''''' '''''''''''''''''' '' ''''''''''' '''' ''''''' ''''''''''

''''''''''''''''''''''''''''''''''''''''''''''''''''''''''''' ''''' '' '''''''''' ''''''''''''' ''''' ''''''' ''''''''''' ' ''''''''''''''''''''''''''''''''']

['''''' ' '''''''''''''''' ' '' '''''''''''''''''' '''''''''''''''''' ''''''''''''' ''''''' ' '' ''''''''''''''''''''''''''''''' '''''''''

'''''''''''''''''''' '' ''' '''''''''''''''' '''''''''''''' ''''''''''''''' ' ''''''''''''''' ' '''''''''''''''''''] Therefore, the NRC staff finds the pressure calculation methodology to be acceptable.

4.2.3.9 Review of ANP-10346P Section 5.3.9 - Steam Dome Equations ANP-10346 describes an [''''''''''''''' '''''''''' ''''''''' '''''''''''''''''''''' ' '' ''''''''''''''''''''''''''''''' '''''''

''' ''''''''' ''''''' '' '''''''''' '''''''''''''''''' '''''''''''''''''''' ''''' '' '''''''''''''''''''''''' ''''''''''''''' '''''''''' ']

Therefore, the NRC staff finds the steam dome model acceptable.

4.2.3.10 Review of ANP-10346P Section 5.3.10 - Recirculation Flow ANP-10346 describes an ['''''''''''''''' '''''''''''''''''''''''' ''''' '''''''''''''''''''' ' '' '''''''''''''''''''''''''''''''' ''''''''

'''''' ''''' ''''''''' '''' '''''''' ''''''''' ' '' '''''''''''''''''''''''''''''''''' '''''''''''''''''''''' ''''']

Both the TTWB and 2RPT ATWS-I event scenarios involve a dual recirculation pump trip, which is also true for the LTSS analysis scenarios for which the RAMONA5-FA LTSS methodology has been previously approved. [''''''''''''''' ''''''' '' ''''''''''''''''' ''''' '''''''''''''''''''' '''''''''''''''' '''

''''''''''''''''''''''''' '''''' ''''''''''''''' '''''''''''''''''''' '''' ''''' ''''''''''''''']; therefore, the NRC staff finds the recirculation flow model acceptable for ATWS-I applications.

4.2.3.11 Review of ANP-10346P Section 5.3.11 - Constitutive Equations 4.2.3.11.1 Review of ANP-10346P Section 5.3.11.1 - Friction and Two-Phase Friction Multiplier ANP-10346 describes the same friction factor correlations as the RAMONA5-FA LTSS methodology and uses the [''''''''''''''''' ''''''''''''''''''' ''''''''''''''''' ''''''''''' ''''''''''''''''''] which is also available in the RAMONA5-FA LTSS methodology. The implementation in the ANP-10346P methodology, ['''''''''' ' ''''''''''''''''' ''' ' '''' '''''''''''''''''' ''''''''''''' '''''''''''''''''''''''' '''''''''' '''''' ''

'''''''''''''''''''''''''''' '''''''' '''''''''''''''''''''''''''] via additional accounting for reverse and counter-current flow. This change is necessary for properly treating the reversed and/or counter-current flow experienced during large-amplitude ATWS-I oscillations. The NRC staff has determined that the implementation of the friction and two-phase multipliers is acceptable and reasonable for this application.

4.2.3.11.2 Review of ANP-10346P Section 5.3.11.2 - Local Pressure Loss Models ANP-10346 describes essentially the same local pressure loss model as the RAMONA5-FA LTSS methodology, with the primary exception being that the ANP-10346P methodology accounts for the possibility of reversed flow. As in the previous section, the NRC staff has determined that this implementation, including treatment of reversed flow, is appropriate and acceptable for ATWS-I applications.

4.2.3.11.3 Review of ANP-10346P Section 5.3.11.3 - Abrupt Contraction/Expansion Pressure Change Model ANP-10346 describes a similar abrupt contraction/expansion reversible pressure change model as the RAMONA5-FA LTSS methodology, except for additional accounting for reverse and counter-current flow compared to the RAMONA5-FA LTSS methodology. This change is necessary for properly treating the reversed and/or counter-current flow experienced during large-amplitude ATWS-I oscillations. The NRC staff finds this treatment of reversed and counter-current flow to be acceptable and reasonable for this application.

4.2.3.11.4 Review of ANP-10346P Section 5.3.11.4 - [''''''''''''''''' ''''''' '''''''''''''' '''''''''''''''''']

['' ''''''''' ' ''''''''''''''''''' ''' ''''''' '''''' '''' ' '' '''''''''''''''' ''''''' '''''''''''''''''' '''''''''''''''' '

''''''''''''''''' '' '''''''''''' ''''' ''''''''''' ''''''''''''''''''] is acceptable for ATWS-I calculations.

4.2.3.11.5 Review of ANP-10346P Section 5.3.11.5 - Thermodynamic Steam-Water Properties The ANP-10346P methodology uses the IF97 steam tables for fluid properties as a function of enthalpy and pressure, the same as used in the Monticello ATWS-I methodology. This is an improvement over the RAMONA5-FA LTSS methodology, ['''''''' ''''''''''''' ''''''''''''''''' '''''''' '''''

'''''''''''''''''' ' ''' ''''''''''''''''''''''''' ''''''''''''''] This provides the best available representation of thermodynamic fluid properties at the full range of possible conditions during ATWS-I, and therefore the NRC staff finds this implementation acceptable.

4.2.3.11.6 Review of ANP-10346P Section 5.3.11.6 - [''''' ''''''''''''''''''''''']

In order to determine [''' '''''''''''''''''''''''' ''' ''''' '''' '''''''''''''''''''''' '' '''''''''''''''''''''''''''''

''''''''''''''''''''''' ''''''' ''' '''''' ' ''' '''''' '''''''''''''' ] As a result, the NRC staff finds the

[''''''''''''''''''''''''''''''' '''''''''''''''''''''] to be acceptable for the purpose of establishing the parameters of

[''' '''' '''' '''''''''''''''''''] outside the core.

4.2.3.12 Review of ANP-10346P Section 5.3.12 - Numerical Integration Techniques The RAMONA5-FA LTSS methodology utilizes [ ''''' ''''''''''' ''''' '''''''''''''''''''''''' '''''''''''''' ''''''

''''''''''' ''''''''''''''''''''''''']

[' '''''' '''''''''''''''' ''''''''''''''''' '' '''''''' ' ''''''''''''''' '''''''' ''''''''' ''''''''''''''''''''''''''''''' ''' '''''''''''''

'''''''''''''' '' ''''''''''''''''''''' ' '' ''''''''''''' ''''''' ''''''''''''''''' '' ''''''''''''' ''''''''''''''''''''''' '''''''''''''''' ''']

[''''' '''''''''''''''''''''''' '''''''''''''''''''''''''' '''''' ''''' ''''''''''' ''''''''''''' '''''''''''''''''' '' '' ''''''''' ''''

''''''''''''' '' '''''''''''' ' '''' '''''''''''' '' ''''''''' '''' '''''''' '''''''''''' '''' '''' ''''''''''']

['''' '''''''''''' ''''''''''' '''''''''''''''''''' ''''''''''''''''''''' ''''''''''''''''' ''''''''''''''''''''''''''''' '''''''''''''''''''''' '

'''' ' ''''''''''''' ''''' '''''''''''''''''' ' '''''''''''']. Furthermore, the benchmark results presented in ANP-10346P demonstrate the numerically stable and robust performance of the methodology up to and including large amplitude oscillations typical of ATWS-I analyses. Therefore, the NRC staff finds the numerical integration technique acceptable.

Core Axial Nodalization The NRC staffs experience has shown that the calculated decay ratio of thermal hydraulic oscillations in other codes may be significantly artificially dampened by numerical errors (numerical diffusion) in the underlying equations, and this effect may be strongly influenced by factors such as timestep size and spatial discretization scheme. ['''' '''''''''' '''''' '''''''''''''''''

'''' ' '''''''''']. In addition, the NRC staff has identified that the coarse nodalization associated with a 25 uniform axial nodalization scheme may lead to significant error in the oscillation decay ratio due to an insufficiently spatially resolved axial void profile, particularly near the bottom of the channel, and a resulting effect on neutronic feedback and oscillatory behavior.

The NRC staff issued RAI-8 to request justification that the axial nodalization scheme used in the ANP-10346P methodology provides sufficient numerical fidelity to accurately represent the stability behavior for ATWS-I. In the RAI response, Framatome provided both a discussion and numerical results to support this position. Framatome discussed three possible effects of numerical diffusion with regard to stability. The first effect is the kinematic diffusive spreading of solution variables over time as fluid moves along the channel. Framatome stated that this effect [' '' '''''''' ' ' ''''''' '''''''''' ''''''''''''''' ''''''''''''''''' '''''''''''''''' '' '''''''''''''''''''''''

''''''''''''''''''' ''''''''''''''''' '''''' '' ''''''' '''''''''''''''''' '''''''' ']. The NRC staff finds this reasoning to be logical and consistent with the theoretical formulation of the solution methodology presented in ANP-10346P.

['''' '''''''''' '''''''' ''''''''''''' ''''''''''''''''' ' ''''''''''''''''''''''' ' '' ''''''''''''''''' ''''''''''''''''''''' ''''

'''''''''] has only a small or minimal impact on the calculated stability behavior for the RAMONA5-FA ATWS-I methodology.

The third effect of diffusion discussed by Framatome is related to the momentum formulation itself, with respect to the effect on the momentum components of the density head and axial distribution of friction resulting from increased axial attenuation of density waves. Framatome stated that the improved axial resolution of void fraction gradients afforded by decreased node size, as proposed previously by the NRC staff and contractors, may be of more importance than the kinematic effect of numerical diffusion for which the Courant number plays a direct role.

['''''''''''''''''''''' ''''''' ''' ''''''''' ''''''' ''''''''''''' ''''' ''''''''''''' '''''''''''''''' '''''''' ''''''''''''' '''''''''''

'''''''''''''' ''''''''' '''''''''''''' ']

The discussion provided in the RAI-8 response supports a conclusion that the core axial nodalization scheme and associated numerical errors would be expected to be relatively small for ATWS-I applications. However, to confirm the effect that nodalization may have on the code results for ATWS-I applications, the NRC staff reviewed the results of the nodalization sensitivity study provided by Framatome in the RAI-8 response. In this study, Framatome increased the nodalization from ['' ' ''''''''''''] axial nodes in the core, which is the proposed value for the methodology, to ['' ' ''''''''''''] axial nodes in the core. The NRC staff reviewed Framatomes approach and determined that the nodalization increase was performed in a suitable manner, [' '''' '''''''''''''''''''' ''' '''''''''''''''''''''''''''''''' '''''''''''''''''''''''' ' '''' '''' '''''''''''''''''''''''

'' ' '' ''''''' '''''''''''''''''''''''''''''''''''''''' '''''''''''''' ' '' ''''''''''''''''''''''''''''''' '''''''']; this ensures a consistent approach for determining the neutronics and thermal hydraulics initial conditions to ensure that the conclusions of the RAMONA5-FA nodalization study are valid.

The finer nodalization resulted in increased decay ratios for all cases included in the nodalization study - namely, the KATHY stability tests, the linear reactor benchmarks, the Oskarshamn-2 nonlinear benchmark, and the Brunswick TTWB sample problem included in the nodalization study. For the KATHY linear stability tests and the linear reactor benchmarks, the decay ratio increased by amounts ranging from [''''' '''''' ' ''''' ' ''''''''''''''''''''''''''' '']

when comparing the finer nodalization results to the base nodalization results. [''''' ''''''''''''''''

'''''''' '''''''''''''' ''''' ''''''''''''' '''''''''''''''''''''''''' '''' ''''''''' ''''''' '''''''' ''''''''''''''''''''''''' ']. The effect of nodalization on frequency was [''''' ''''''' ''''''''''''''''''''''''''' ' ''''''] Hertz (Hz) change compared to the base nodalization case). For the Oskarshamn-2 nonlinear benchmark and the Brunswick sample TTWB problem, the decay ratio [''''''''''''''' ' '''''''''''''''''''''''' ''''''''' '' ''

'''''' '''''''''''''''''''''' ''''''''''''''''''''''''''' ' '' ''''''''''''''''''''''''''''''''' '''''''''''''''''''''''' '''''''''''''''' '''' '''''''''''''']

The larger growth rate also led to earlier failure to rewet by approximately 20 seconds in the Brunswick TTWB sample problem.

After failure-to-rewet, the time-dependent PCT values appeared to be ['''''''''''' ''''''''''

'''''''''''''''''''''''''''''''' ' ] degrees Celsius (C)) on average for the finer nodalization case compared to the base case. As a result, the maximum PCT throughout the event was

[''''''''''''''''''''''''] C higher with the finer nodalization for this problem. However, the NRC staff concludes that this difference is likely due to the finer nodalization case reaching failure to rewet earlier than the base case, allowing failure to rewet to extend to lower elevations on the hot rod before the oscillations are suppressed, compared to the base case. These lower elevations would likely correspond to higher average LHGR values. Even if failure-to-rewet did not extend lower in the finer nodalization case, the smaller node sizes mean that the limiting node in the base nodalization case is split into two nodes in the finer nodalization case, and the lower of these two finer nodes would have a slightly higher LHGR than the larger node overlapping this location in the base case. The NRC staff suspects a cause for the higher PCT [' ''''' '' ''''''

''''''''''''' ''' ''''''''''' ''''''''''''''''''''''''''''''' ''''''''''''''''''' '''''] Therefore, the NRC staff concludes that finer nodalization does not intrinsically cause higher PCT in the failure-to-rewet regime.

Therefore, no penalty or added conservatism is necessary to account for this apparent increase in PCT for finer nodalizations due to the modest nature of the increase in PCT, inherent conservatisms in the methodology, and the lack of evidence that finer nodalization would capture new phenomena which could have a significant impact on the PCT.

Although the finer nodalization resulted in greater instability - faster oscillation growth and earlier failure to rewet - in all linear and nonlinear analysis cases, the NRC staff determined that the base nodalization of [' '''''''''''''' ''''' ''''''''] in the core is acceptable because it gives the most consistent and non-biased overall agreement with the measured decay ratio values across the various stability benchmarks. Further rationale for the acceptability of this nodalization scheme was provided by Framatome - namely, ['''' '''''''' ''' '''''''''''''''''' '''''''''''''' ''''''''''''''''''''

'''''''''''''''']

The NRC staff concludes that the base nodalization of ['' ' ' ''''''''''''' '''''''] nodes in the core is acceptable for use in the RAMONA5-FA ATWS-I methodology. This determination is based primarily on the good agreement of the RAMONA5-FA ATWS-I methodology with measured stability data across the broad range of experimental conditions when using [' '

''''''''''''' '''''''] nodes in the core, compared to the ['''''''''''''''' '''''' ''''''''''''' '''' ' ''''''''''' '''''

''''''''''''''''' ' '' '''''''''''''''''''''' ''''''''' '''''''' ''''''' '' ' ''''''''''' ''''' ''''''''''''] As discussed previously in this section of the SE, the NRC staff also considered potential sources of error due to numeric diffusion and determined that they would not be significant for the RAMONA5-FA ATWS-I methodology.

4.2.4 Review of ANP-10346P Section 5.4 - Steam Line Flow Dynamics As discussed in Section 4.2.3.7 of this SE, the ANP-10346P methodology uses [' ''''''''''''''

''' ''''''''''' ''''' ]. However, a special model for the steam line is included in the ANP-10346P methodology (identically to the RAMONA5-FA LTSS methodology) to calculate the propagation of pressure waves only within the steam line. For ATWS-I, this is relevant for calculating the pressure response after turbine valve closures following a turbine trip, as well as the pressure response following safety relief valve (SRV) closure and re-opening which may occur during the oscillatory phase of the TTWB event.

[' ''''''''''''''''''' ''''''''''''''''''' ''''''''''' '''' ''''' ''''''' '''''''''''''' ''''' ' '''''' '''''''''' ''

'''''''''''''''''''' '''''''''''''''''''''' ' ] The NRC staff reviewed this model and found it to be a logical and acceptable method for determining pressure response in the steam line and vessel. This is primarily because, as discussed in Section 4.2.3.7 of this SE, [''' '''''''''''''''' ''''''''' '''''''''''''''''''''

''''''''''''''']. However, no detail is provided regarding how the steam line modeling and valve responses are verified to capture reasonable behavior during the ATWS-I event for specific plants. This is highly dependent on plant specific configurations, closure time, and setpoints, so a limitation and condition will ensure the resulting behavior (e.g., flow rates through the valves and pressure drop across the steam line(s)) from the steam line model is reasonably representative of expected plant-specific behavior during an ATWS-I event.

The NRC staff has determined that the steam line flow dynamics model provides an accurate and realistic approach for calculating pressure response during ATWS-I events including TTWB, and therefore the NRC staff finds this model acceptable with the condition that licensees must provide justification that their steam line modeling will appropriately capture expected variations in the pressure and flow boundary conditions for the ATWS-I event.

4.2.5 Review of ANP-10346P Section 5.5 - Special Models 4.2.5.1 Review of ANP-10346P Section 5.5.1 - Recirculation Pump Model The recirculation pump model determines the relationship between pump rotational speed, pump torque, pump flow rate, and pump head. These define the steady state operating characteristics of the recirculation pumps as well as their transient behavior. The primary relevance to stability analyses is in determining the recirculation pump coastdown behavior after a recirculation pump trip, as well as the recirculation pump inertia which has an important impact on the growth rate and limit cycle amplitude of global flow oscillations. Note that the effect on regional flow oscillations is much smaller, as the total core flow rate remains relatively constant in that case.

The recirculation pump model in ANP-10346P is identical to the model in the RAMONA5-FA LTSS methodology, which was approved for LTSS analyses including the case of in-phase oscillations. The NRC staff has reviewed these models for ANP-10346P and has concluded

that the models include all necessary physics and remain acceptable for instability events up to and including large-amplitude limit cycle oscillations.

4.2.5.2 Review of ANP-10346P Section 5.5.2 - Jet Pump Model Unlike the recirculation pump model, [' '''''''''' ''''''''''' '' ''''''''''''''' ' '' ''''''' '''''''''''' ''''''

'' ''''''' '''''''''''''''''' '''''' ' ''''''''''''''''''''''']. However, as with any pressure term in the primary loop, the calculated pressure head may affect the transient behavior during rapid pressure changes (such as immediately following a turbine trip) as well as affect the stability behavior particularly during global oscillations.

The jet pump model in ANP-10346P is identical to the model in the RAMONA5-FA LTSS methodology, which was approved for LTSS analyses, including the case of in-phase oscillations. The NRC staff has reviewed these models for ANP-10346P and has concluded that the models include all necessary physics and remain acceptable for instability events up to and including large-amplitude limit cycle oscillations.

4.2.5.3 Review of ANP-10346P Section 5.5.3 - Steam Separator Model The modeling of the steam separator - in particular, its flow inertia, as well as the flow rate of vapor leaving the circulation loops and entering the steam dome above the coolant level (known as carry-under) - may have a significant effect on the stability characteristics of the reactor system. Flow inertia has a particularly strong impact for core-wide (in-phase) oscillations.

The steam separator model in ANP-10346P is identical to the model in the RAMONA5-FA LTSS methodology. This model determines the steam separator flow inertia based on [ ''''''''''

''''''''''''' '''''''''''''' ''''''''' ''''''''''' ''''''' ''''''''''''''''' ''''' ''''' '''''''''''''' '''''''''' '' ''''''''''''' ''].

The flow conditions and behavior of the steam separator follow the same physical principles and exhibit the same general characteristics under ATWS-I conditions as under smaller amplitude LTSS oscillation conditions, and therefore the NRC staff concludes that the steam separator model, which was previously approved for the RAMONA5-FA LTSS methodology, is applicable and acceptable for ATWS-I applications in ANP-10346P.

4.2.5.4 Review of ANP-10346P Section 5.5.4 - Feedwater Sparger Condensation Model When the water level in the vessel downcomer is below the level of the feedwater inlet (feedwater spargers), significant heating of the subcooled feedwater liquid as well as condensation of the saturated steam may occur as the liquid flows downward through a steam environment. This can affect the core inlet temperature as well as the system pressure.

ANP-10346 describes the same model as the RAMONA5-FA LTSS methodology to model the condensation rate as a function of ['' ''''''''''' '''''''''''''''''''''''''' ''''''''' '''''''''''''''''''''' '''''' '''''

'''''''''''''''''' '''' ''''''' ''''''''''''''''' '' '''''].

Once the water level falls below the feedwater spargers, the nature of this condensation phenomenon is the same during ATWS-I as it is during other events currently approved for

analysis using the RAMONA5-FA LTSS methodology. Additionally, the model provides physically reasonable and realistic relationships with physical parameters. Therefore, the NRC staff finds the feedwater sparger condensation model acceptable for use in ANP-10346P for ATWS-I analyses.

4.2.5.5 Review of ANP-10346P Section 5.5.5 - Dryout and Rewetting Model The prediction of dryout and possible subsequent rewet of the hot rod is of primary importance to ATWS-I analyses due to the dramatic increase in PCT associated with sustained dryout.

Under sufficiently high cladding-to-coolant heat flux for a sufficient duration, all liquid in contact with the cladding surface evaporates, leaving only vapor in contact with the cladding surface (dryout conditions). Because vapor is much worse than liquid at conducting/convecting heat from the cladding surface, the temperature of the cladding (and also the fuel pellet) quickly increases after the onset of dryout. Due to the large, rapid changes in flow rate and thermodynamic quality of the coolant adjacent to the cladding surface during thermal hydraulic oscillations, there is a possibility that liquid will once again come into direct contact with the cladding surface (rewet), lowering the cladding temperature due to improved heat transfer.

However, rewetting of the cladding surface becomes more difficult as the cladding temperature (and therefore the evaporation capability) increases; this may lead to a runaway condition in which the liquid flow is no longer able to come into contact with the cladding surface for long enough to fully reverse the increase in cladding temperature. Under such a condition, the cladding temperature may ratchet up through multiple cycles of heatup and limited cooldown due to rewetting, or experience a continuous increase in temperature due to loss of rewetting ability. If the cladding temperature increase is not mitigated, very high cladding temperatures which might challenge the ATWS-I acceptance criteria may result. In this case, the cladding temperatures can only be brought down again by reducing the heat generation rate (power level) in the fuel; during ATWS-I, this is done either by increasing the average void fraction in the core (accomplished via water level reduction) or injection of soluble boron into the core (via the standby liquid control system).

Because of its strong impact on the PCT during ATWS-I events, the dryout and rewetting model in ANP-10346P was one of the primary focuses of the NRC staffs review. The ANP-10346P methodology provides the same fundamental approach as the Monticello ATWS-I methodology, but a fundamentally different approach than the RAMONA5-FA LTSS methodology, to determine dryout and rewet of the hot rod surface. The RAMONA5-FA LTSS methodology

[''''''''''''''''''' ''' '''''''''''''''' '''''''''''''' '''''''''''''''''''''''''''' '''''' ''''''''''' ''''''' ''''' '''''''''''''''''''

''''''''''''''''''''' '''''''''' '''''''''' '''''''''''''''''' ' '''''''''''''''''' ''''''''''''''''] The wetting or dryout status of the cladding surface is used to determine the heat transfer regime (nucleate boiling, transition boiling, or film boiling heat transfer regimes), and heat transfer coefficients are applied correspondingly.

However, based primarily on analysis of the KATHY dryout/rewet test data presented in Section 6.5 of ANP-10346P, Framatome concluded that a different modeling approach for determination of dryout and rewet behavior provided a better fit to the data under oscillatory conditions representative of ATWS-I. This model, similar to the one provided in the Monticello ATWS-I methodology, [''''''''''''''''''''''' ''' '''''''''''' ' '''''''''' '''''''' ''' '''''''''''''' '''''''''''' ''''''''''

'''''''''''''''''''' ' '' '''''''''''''''' ]

A thorough review of a similar dryout/rewet model was performed by the NRC staff in its review of the Monticello ATWS-I methodology. In that review, the NRC staff concluded that the model was acceptable for the plant-specific application for which the methodology was submitted. For the ANP-10346P review, the NRC staff reviewed the dryout/rewet model with particular focus on determining the acceptability of differences in the model relative to the Monticello ATWS-I methodology, as well as the applicability and acceptability of the model for generic application.

The NRC staffs review determined that the ANP-10346P dryout/rewet model is largely similar to the dryout/rewet model used in the Monticello ATWS-I methodology, including [''' '''''''''''''''''

'''''''''' ''''''''''''''']. Although the NRC staffs review and approval of the Monticello ATWS-I methodology was performed on a plant-specific basis, in its evaluation of the Monticello ATWS-I methodology dryout/rewet model the NRC staff did not note any particular limitations or shortcomings of the model which may specifically limit its use for other plants or operating conditions. For ANP-10346P, the NRC staff further examined the model and determined that the experimental benchmarking, as discussed in Section 4.3.5 of this SE, covered a sufficiently broad range of representative ATWS-I conditions such that the model may be acceptably applied to the current fleet of BWRs on a generic basis.

However, some differences were noted between the Monticello ATWS-I methodology and ANP-10346P dryout/rewet models, and these are evaluated in additional detail in the remainder of this section. These differences include the ['''''' '' ''''''''''''''''' ''''''''''''''''''''''' '''''''''''''' '

'''''' '''''''''''''''''' ''''''' '''' '' '''''''''''''' ' ''''''' ''' '''''''' ''''''''''' ''''''''' '''''' ''''''].

[''''' '''''' ''' '''''''''''''''''' '''''''''''''''''''' '''''''''''''' ''''''''' '''' ''''''''''''' '''''''''' ''' ''''''''''''''''''''

'''''''' '''''''] The NRC staff finds the [''''''''''''''' '''''''' ''''' '''' ''''''''''''''''''''''''' ' '' '''''''''''''''''''

''' ] to be conceptually reasonable on a physical basis, and the strong - and perhaps slightly improved - agreement of the dryout/rewet model with the cyclic dryout/rewet behavior observed in the KATHY experiments leads the NRC staff to find this revised model to be acceptable.

The NRC staff examined the addition of a [''''''''' ''''''''''''''''''''''' ''''''] to the ANP-10346P model, relative to the Monticello ATWS-I methodology, based on consideration of the flow conditions under which the dryout/rewet model was derived. [''''' ''''''''''''''''''''''''' '''''''''''''''''''''' '''''''' '''''''''''

''''''''' ''''''''' ''' '''''''''' '' '''''''''''' ''''' '' ' ''''''''''' ''''''''''''''''''' ''''''''''''''''''] In RAI-5, the NRC staff asked for further information to justify the applicability [ ''' ''''''''''''''''''''''' '''''''''''

''''''''''''''''''''''''' ''''''' ''''''''''''''' '''' ''''''''''''''''''''' '''''''''' '''' '' ' '''''''''].

In the RAI response, Framatome indicated that ['' '''''''' '''' '''''''''''''''''''''' ' ''' ''''''''''''''''''''''

'''''''''''''''''''' '''' ''''''''''''' ''''' ''''''''''''''' '''''''''''''' ''''''''''''''''' ]

Specifically, Framatome applied the [''''''''''''''''''''''''''''''''''' ''''''''''''''' ' ''''''' '''''''''''''''''''' '''''''] to the base critical power reduced order model (CPROM). As depicted in Figures 8 through 14 of the RAI-5 response, the [''''''' ''''''''''''''''''''' ' ''''''''''''''''''''' ''''''' '''''''''' ' '''''''''''' ' '' ''''

'''' ''''''''''''''''''''''''''' '''''''' ''''''''''''''''''' '''' '' ''''''''''''''''''' '''''''''''''''' ''''''''' '' ''], the NRC staff concludes that the dryout/rewet model with ['' '''''' ''''''''''''''''''] provides an acceptable representation of dryout behavior for the full range of quality and void fraction conditions expected during ATWS-I.

ANP-10346 describes a process for fitting the CPROM correlation to steady-state CPR data which differs from the [''''''''''''''''''''] process used for the Monticello ATWS-I methodology. This new process was determined by the NRC staff to be acceptable based on the evaluation provided in Section 4.2.9 of this SE.

In RAI-2a, the NRC staff requested additional information on how the fitting parameters for the dryout/rewet model were determined from measured data, particularly when direct experimental validation for each parameter was not possible or not available. In the RAI response, Framatome described the fitting process in detail, including a combination of [''''''''''''''''''''''''''''''

'''' ''''''''''''''''''''''''''''' ''''''' ''' ''''''''''''''''''''''']. The NRC staff reviewed the RAI response in detail and determined that all parameters were fitted to the data in a consistent, logical, and well-defined fashion to provide the most accurate and acceptable prediction of both steady state and transient behavior applicable to ATWS-I.

The NRC staff has reviewed the dryout/rewetting model in detail, including evaluating the generic applicability of the model as well as evaluating the differences relative to the similar model in the Monticello ATWS-I methodology. The NRC staff has concluded that the model is acceptable because it is based on realistic physical principles, exhibits close agreement to measured CPR data under steady state conditions, and agrees closely with measured data under transient conditions for a wide range of operating conditions which reasonably encompass the expected range of conditions expected to occur during postulated ATWS-I events in the current fleet of BWRs.

4.2.6 Review of ANP-10346P Section 5.6 - Plant Control and Protection Systems The NRC staff reviewed the plant control and protection systems methodology provided in Section 5.6 of ANP-10346P, including the implementation of:

Pressure control system consisting of turbine control, bypass valve and safety and relief valve (SRV),

Plant protection systems (PPS) including recirculation pump trips, Feedwater control system, for water level control The pressure control system model is required to accurately model the system pressure response following a turbine trip and possible cycling of the SRVs during a TTWB event.

Modeling of the recirculation pump trip function of the PPS is relevant for both the TTWB and 2RPT ATWS-I events, to determine realistic timing of the recirculation pump trip and associated core flow rate reduction. Modeling of the feedwater control system is relevant for both the TTWB and 2RPT ATWS-I events to allow the water level to automatically adjust to the setpoint value by adjusting the feedwater flow rate, both during the initial event progression as well as after the operator action to reduce the water level setpoint.

These models are essentially the same as in the RAMONA5-FA LTSS methodology, with the primary exception that ANP-10346P added a manual operator actions model to reduce the water level during an ATWS-I event by allowing the user to specify the start time of operator actions as well as a setpoint to which the water level will be reduced. The model assumes the feedwater pumps trip at the specified start time and calculates the coast down behavior of the feedwater pumps, after which the feedwater controller maintains the water level at the new lower level based on the user-defined setpoint.

The NRC staff has reviewed these models and has concluded that these models are acceptable because they realistically and adequately represent the plant control and protection systems behavior during postulated ATWS-I events.

4.2.7 Review of ANP-10346P Section 5.7 - Numerical Time Integration The numerical scheme used for time integration (time marching) in the neutron kinetics, thermal hydraulics, and fuel rod thermodynamics equations has a significant effect on the numerical robustness and accuracy of the solution. In particular, the large temporal and spatial gradients associated with rapidly-changing conditions within the core and vessel during ATWS-I oscillations increase the potential for large numerical errors in the solution, particularly during sharp changes in the solution such as the expected changes due to the dryout and rewet phenomena. Such errors may lead to results which depart significantly from reality and prevent accurate determination of the system response with respect to the ATWS acceptance criteria.

In the ANP-10346P methodology, the neutron kinetics, fuel thermodynamics, and vessel hydraulics equations are solved separately, with ['' ''''''''' ''''''''''''''' '''' ''''''' ] integrate each of these three calculation domains. Enforcing ['' '''''''' '''''''''''''''' ''''] in each domain ensures numerical consistency, improves numerical robustness, and avoids loss of accuracy in the overall coupled solution.

As described in Sections 4.2.1 and 4.2.2 of this SE, the neutron kinetics and fuel thermodynamics equations are integrated [''''''''''''''] in time. The most significant impact of this is [ '''''''''''''''''' '' '''''''''''''''' '''''' ''''''''''''''''''''''' ''''''''''''''''''''' ''''' '''''''''''' '''''''''''''''''' ''''''''''''''''''

''''''''' ''''' ' '' '''''''''''' ''''''''''''''''' '].

[' '''''''''''''''' '''''''''''' ''''''''''''' ''''' '''' ''' ''''''''''' ''''''''''''''''''' ''''''' '' '''''''''' ''''''''''''''''''''''

''''''''''''''' '''' '' ''''''' '''''''''''''''''' ''''''' ' '''''''''' ''''''''''''' ''''''''''''''']. Discussion and sensitivity studies regarding vessel and core nodalization were requested in RAI-9 and RAI-8, respectively, and these are discussed in Sections 4.2.3.1 and 4.2.3.12 of this SE.

The NRC staff issued RAI-12 to request additional details to determine the acceptability of the timestep control scheme and the values used for the benchmarks and sample problem in ANP-10346P. In the RAI response, Framatome described the timestep control parameters specified in the input file. The NRC staff reviewed these parameters and determined that they provide adequate capability to control the numerical timestep size to ensure robustness and accuracy of the solution. Importantly, the NRC staff determined from the RAI response that the same timestep control parameters were used for all benchmark cases and the sample problem in ANP-10346P. This ensures consistency and validity of the methodology across all benchmarks. ['''''''''''''''''''' ''''''''''''''' ''''' ''''''''''''''''' '''''''' ''''''''''' ''' '''''''''''''''''' ' ''''''''''''

'' '''''''''''''''' ' ''' ''''' '' '''''' ' ''''''''''''''''''''''''''''''''' ''''''''''''''']. The NRC staff accepts this disposition and has placed a limitation and condition on the methodology to enforce this approach for selecting timestep control parameter values.

Results of a timestep sensitivity study were provided in the RAI-12 response, by varying the values of the timestep control parameters for the Brunswick sample problem. These sensitivity cases, which modeled timestep size differences that varied by up to [' '''''''' '] throughout the event, exhibited only minor differences in oscillation growth rate and failure-to-rewet times (no more than 10 seconds across the sensitivity cases). The failure-to-rewet time varied in an unpredictable fashion with no clear trend as a function of the timestep control parameter values.

Because of this lack of a clear trend, and because the base timestep control values

demonstrated good agreement with measured data across all benchmarks documented in ANP-10346P, the NRC staff finds the base timestep control values to be acceptable for use in RAMONA5-FA ATWS-I applications.

4.2.8 Review of ANP-10346P Section 8.0 - Calculation Procedure A portion of the ATWS-I analysis methodology described in ANP-10346P is not captured within the RAMONA5-FA code or various assessments of the performance of its constituent models and correlations. In order to perform an ATWS-I analysis, an analyst must follow prescribed steps in order to ensure that the plant-specific analyses are performed in a manner consistent with the assumptions within the code and the intent of the methodology in demonstrating regulatory compliance. A discussion of the key guidance provided in ANP-10346P for performance of ATWS-I analyses is provided in the following subsections.

Statepoint Definition Section 8.0 defines the procedure to be used for plant-specific ATWS-I analyses using the ANP-10346P methodology. The procedure defines the characteristics of the statepoint to be analyzed. This includes [' ''' ' '''' '''' '' ''' '''''''''''' ''''''' ''' ''''''''' '' '''' ''''' ''''''

''''''''''''''''''' ] The NRC staff has reviewed the proposed statepoint definition and determined that it is specified in an acceptable manner for ATWS-I and is consistent with the approach for MELLLA+TM applications that has previously been reviewed and approved by the NRC staff (e.g., Ref. 14).

Overview of the Analysis Procedure Defined in ANP-10346P Details of the analysis procedure and their acceptability are described in this section. Note that ATWS-I plant-specific calculations are performed for the first cycle planned for utilization of the EFW operating domain, and subsequently, new calculations are performed only when a new fuel is introduced or another change that could impact the ATWS-I analysis is implemented that requires a new license amendment. Therefore, the calculation procedure must ensure that the

ATWS-I results bound cycle specific variations, such that the analyses remain applicable for all EFW cycles that would be allowed by implementation of the proposed license amendment.

The calculation procedure relies on the evaluation of ['''' ''''''''''''''''''''''''' '''''''''''''' '''' ''''''

'''''''''''''''''''' '''''''''''''''' ''''''' ' ' ''''''''''''' '' ''' ''' ''''' ' ''''''''''' ' ''''''''''''']:

[''''' ' ''''''''''' '''''''''''''''''''''''''''' ''''''''''''''' '''''' ''''''''' '' '''''''''''''''' '''''''''' ''''''''''''''''''' ''''

''''''''''''''' ''' '''''''''''''''''''''']

['''' '''''''''''''''' ' ''' ''''''''''''''''''''' ' ''''''''''''''''''''' '''' '''''''' '''''' ' ''''' ''' ''''' '''''''

''''' ''''' '''' ''''' '''''''' '''' '''''''''''''' ''''''''''''''''' ''''''''''']

['''''' '' ''''''''''''''' ' ''''''''''''''''''''' ' ''' '''''''''''''''' '''''''''''''''' '''''''''''''''' '' '''''' ''''''''''''''''''

''''''''' '''''''''''' ' '''''''''' ''''''''''' ''''''''''''''''' '' ''''''''''''''''' ' ''''''''''''''''''' ''''' '''']

[ '''' ''''''''''''' '''''''''''''''''''' ''''''' ''' ' '''''''''''''''''''' ''''' ''''''''''''''''' '''''''''''''''''''' '''''' ''

''''''''''''''' '''''''''''''''''''''''' '''''''''''''''''''' '''''''''''''''''' '''''''''' ''''''' '''''''' ' '''''''''''''''' '''''''''' ''']

Evaluation of Step 3 After review, the NRC staff identified concerns with the calculation procedure, beginning with the first paragraph described under Step 3. ['''' '''''' '''''' ''''''''' '''' '' ''''''''''' ''''' '''''''''''''

''''''''''''' '''''''' ''''' ''' ''''''' ''''''''''' ''''''''' '''' '''''''''''''''''']

[' '''''''''' ''''''''''''' ' ''''''' '''''''''''''''' ''''''''''''' '''''''' '''''''''' ''' '''''''''''''' '''''''''' '''''' '''''''''''

'''''''''''''''' '''''']

[''''''''''''''''''''' '' ''''''''''''''' ''''' ''''''''''' ''''''''' '''''''''''''''''''''' '''''' ' ''''''''''''' ' '''''''''''''''' ''''''''''

''''''''''''' ''''''''''' ''' ''''''' ''''''''' ''''''''''''''''''''''''''''''' ]

As a result of these concerns, the NRC staff has determined that the procedure defined in the first paragraph of Step 3 of ANP-10346P does not give sufficient assurance that a PCT of 2200°F will not be exceeded for all operating cycles and exposure points.

For reasons discussed below, the NRC staff has concluded that the procedure described in Steps 3.a through 3.c, as well as in the response to RAI-15 (with the additional requirement to address core designs that deviate significantly from the reference equilibrium core design used in the ATWS-I analyses, such as transition cores), provide an acceptable means of obtaining reasonable assurance that a PCT of 2200°F will not be exceeded for any cycle and exposure point during EFW operation. Alternatively, plant-specific applications may choose to augment the analyses and/or discussions provided in the first paragraph of Step 3 to provide this reasonable assurance, in lieu of performing Steps 3.a through 3.c. The NRC staff will review these justifications on a plant-specific basis.

Evaluation of Step 3.a through Step 3.c, and the RAI-15 and RAI-16 Responses In RAI-15, the NRC staff requested details of a specific process that may be used to determine that the margin described in Step 3.a is sufficient to ensure applicability to all EFW cycles. In RAI-16, the NRC staff requested additional information to ensure that the modeling assumptions remain appropriate when considering their effect on the time of oscillation onset.

The process provided by Framatome in the RAI-15 response involves ['''''''''''''''''''' '' '''''''''''''''

'''''' '''' '' '''''''''''''''''''' ''''''''''''''' '''''''''''''' '''''''''' '' '''' '''''''''' ' ' ''''''''' '''''' '''' ''' ']

The response to RAI-15 clarifies that sufficiency, in the context of sufficient margin to failure to rewet, refers to satisfying either of the two acceptance criteria listed above; in other words, the margin is sufficient if failure to rewet is avoided or if failure to rewet occurs and the PCT remains below 2200°F. In the proposed methodology, if the margin is sufficient based on these criteria, no further TTWB analyses are required, as discussed in Step 3.b. If the margin is insufficient, Step 3.c must be followed.

The NRC staff concludes that the definition, in the RAI-15 response, [ ''' ''' '''''''''''''''''''

'' ''''''''''''' ''''''''''''''''''' '''''''''''''''' '''''''' '' ''''''''' ''''''''''''''''' ''''''''' '''''''''''''''' ''' '''''''''''''''' ''''''']

Based on the NRC staffs concerns, as discussed above, that cycle-specific variations may potentially lead to large changes in PCT even if failure to rewet has occurred, the NRC staff has concluded that all transition cycles (i.e., up to two transition cycles) occurring during EFW operation must be addressed, either by explicit analysis as in Step 3.a or by appropriate justification, to provide adequate assurance that the most limiting cycle is analyzed. As such, a limitation and condition is being imposed to ensure that this is the case.

The RAI-15 response applies a ['''' '''''''''''''''' ''''''''''''''''''''''''' '''''''''''''''''''''''''' '''''''''

''''''''''' '''''''''''''''''''''''' ' '''''''' ' ''' '''''''''''''''''' ''' '], based on the NRC staffs experience this is expected to provide sufficient added conservatism to compensate for the possibility of the most limiting point not occurring precisely at one of the [''' ''''''''''''''''''' ''''''''''] analyzed.

In the RAI-16 response, Framatome discussed that the process described in RAI-15 employs conservative assumptions with respect to the time of oscillation onset. [''''' '''''''''''''''''

''''' '''''''''''''' '''''''''' '' ''''''''''''''''''''''' ''' '''''''] Framatome also discussed that the assumed FW temperature reduction rate in all cases must conservatively bound the expected plant-specific behavior. Furthermore, the definition of exposure points and transition cycles to be analyzed, as well as ['' ''''''''''''''''''' ' ' '''''''''''''''''' ''''''''''''''''''''''''' ''''''''''''''''''''''''''] discussed above, provides additional assurance that the results remain bounding when considering cycle-specific and exposure-specific variations in oscillation onset time. The NRC staff finds this approach acceptable consistent with the discussions above, provided that the applicable limitations and conditions are met.

The NRC staff finds Step 3.c acceptable because the NRC staff will review the approach used for Step 3.c on a plant-specific basis.

Selection of the Limiting Event Framatome proposed to perform Steps 3 through 3.c for the TTWB event, and - as discussed in Step 4 - 2RPT analyses are only required if TTWB did not experience failure to rewet. The rationale for this, as understood by the NRC staff, is as follows. The feedwater heaters remain active throughout the 2RPT event but not in the TTWB event, resulting in higher feedwater temperatures and therefore lower average power, less severe oscillations, and lower PCT values than in the TTWB event.

However, because the time-critical operator action is defined with respect to the time of identification of an ATWS, and because the 2RPT event involves manual initiation of scram by the operator if the RPV level does not increase to the high level turbine trip setpoint (compared

to automatic scram initiation in the TTWB event), plant-specific analyses must justify the time required for operators to initiation manual scram following the 2RPT and must add this time to the time-critical operator action time following identification of ATWS. This effective delay in operator actions relative to the initiating event allows additional time for the feedwater temperature to decrease during the 2RPT event, such that the feedwater temperature at the time of operator actions may potentially be less than that for the TTWB event. Furthermore, for plants with steam-driven feedwater pumps, the TTWB event may include a trip of the FW pumps prior to operator action, which may mitigate the consequences of the ATWS-I event (due to reduction in water level before action is taken). Consequently, the TTWB event may not be limiting for all plants.

Additionally, the TTWB event, due to turbine valve closure, leads to a higher system pressure than the 2RPT event. Based on the NRC staffs experience, competing effects may exist which lead to an unclear relationship between system pressure and stability behavior, which is difficult to ascertain a priori.

Because of these two concerns, the NRC staff concludes that plant-specific applications must justify the operator action time and feedwater temperature assumptions for both the TTWB and 2RPT events, and must perform analyses for both events using these assumptions to determine which event is limiting with respect to PCT. However, the NRC staff acknowledges that the limiting event is expected to be primarily a function of the operator action and feedwater temperature assumptions used in the analyses, which remain the same regardless of operating cycle or exposure conditions. Therefore, there is reasonable confidence that the limiting event

- 2RPT or TTWB - will remain the same across all cycles and exposure points for a given plant-specific application.

As a result, the NRC staff requires that both TTWB and 2RPT analyses be performed initially for Step 3; and once the most limiting event is determined, only that event must be considered for the additional justifications for Step 3 or the additional analyses in Steps 3a through 3c.

Plant-specific applications must justify the operator action and feedwater temperature assumptions for both events. The NRC staff also notes that due to the additional work necessary to justify cycle-independent application of the ATWS-I analyses, the value of providing the results of the non-limiting event to the NRC for review will be limited.

Boron Injection In addition to taking action to reduce water level, the reactor operators must initiate SLCS boron injection within the time-critical action interval following identification of an ATWS. The effect of SLCS injection is to deliver borated water to the core which provides sufficient negative reactivity to shut the core down. This shutdown is capable of terminating any oscillations as well as limiting the impact of high core power on the containment heat load. However, the operator actions to reduce water level are expected to mitigate the oscillations well before the borated water reaches the core. ['''''''''' ' '''' ''''''''''''''''''''''''' ''' '''''''''''''''''''''''''''' '''''''''''''

'''''' '''''''''''''' ''''''''''''''].

[ '' '''''''''''' ''''''''''''''' '''' '''''''''''''''' ''''''''' '''''''''''''' ' ''''''''''''' ' ''''''''''''''' ''''''''''''''''''''

''''''''''' ''''''''''' ']

Calculation Procedure Conclusions Based on the evaluations given above, the NRC staff finds the calculation procedure to be acceptable for its intended purpose, with four limitations and conditions as discussed in Section 5.0 of this SE.

4.2.9 Review of ANP-10346P Appendix A - Steady State Dryout Correlation CPROM Dryout of a fuel rod, unless it is quickly followed by rewet, leads to very large increases in cladding temperature. Therefore, accurate calculation of the timing and location of dryout in the fuel bundles is of high importance in determining the PCT and has a strong effect on whether the ATWS acceptance criteria are met during postulated ATWS-I scenarios. The approach used to develop the models for dryout (and rewet) in ANP-10346P is twofold. ['''''''

''' ''''''''''''''''' '' ''''''' ''''' '''''''''''''''' ''''''''''''''''''''' ] dryout/rewet model described in Section 5.5.5 of ANP-10346P and evaluated in Section 4.2.5.5 of this SE.

The CPROM correlation was previously presented in the Monticello ATWS-I methodology and was reviewed and accepted by the NRC staff for plant-specific application at Monticello.

However, the correlation was developed from ['' '''''''' ''''''''''''''''''''''' '''''''''''''''''' ' '' ''''''

''''''''''' '''''''' '''''''''''''''''''''' '''' ''''''''''' ' '''''''' '''''''''' '''''''''''''''' '''''''''''''''''']:

['''''' ''''''' ' '''''''''''''''''''' '' '''''''''''''' '''''''' '' ''''']

[''''''''''''''''' '' '''' ' '']

['''''''''''''''''''''''' ' ''''''''''''''' ''' '''''''''''''''' '''''''''''' ' '''' '''''''']

[''''''''''' ' '']

[''''''' ''''''''''' ''''''''''' ''''''''''''''''''''''' '''''''''' ''''' '''''''''''']

['''' '''''' ''''''''''''''''''' ''''''' ''''''''''''' ' '' '''''''''''''''''''' '''''''''''''''' '''''''''''''''''''''''''''' '''' '''''''''''''''''''

'''''''''''''''' '' ''''''''''' ''''''''''''''' '' '''''' '''''''' ''''''' '''''''''''''' ''''''''''''''' ' '' ''''''''''''''''''''' '' ]

With the exception of stagnant or reversed flow - which may occur during ATWS-I, especially near the inlet of the bundle - the range of operating conditions shown above encompasses the expected ATWS-I conditions for the current BWR fleet and is therefore suitable for generic use in BWR ATWS-I analyses. [''''''''''''''''' '' ''''''''''''''' ''''''''''''''''''' ''''''' ' '''' ''''''''''''''''' '''''''''''

''''''''''''''''''''] For these reasons, the NRC staff finds the use of the CPROM correlation within the dryout/rewet model in ANP-10346P to be generically acceptable for ATWS-I analyses.

[ '''''''''''''''' '' '''' ''''''''''''''' ''''''''''''''' '''''''''' '''' '''''''''''''''' ''''''''''''''' '''''' ''''''

''''''''''''''''''''''''''' ' ''''''''''''''' '' '''''''''''''''''''' ''''''''''''' ], to be acceptable.

Because of the wide range of operating conditions over which the ANP-10346P CPROM

['''''''''''' ''''''' '''''''''''''''''''] is validated, and because the modifications relative to the previously reviewed the Monticello ATWS-I CPROM [''''''''''' '''''' ''''''''''''''''''''] are reasonable and result in comparable or improved accuracy relative to the measured data, the NRC staff finds the CPROM steady state correlation in ANP-10346P to be acceptable for use as the underpinning critical power correlation for use in the ANP-10346P [''''''''''''''''' '''''''''''''''''''''''] model on a generic basis for limiting ATWS-I analyses.

4.3 Code Assessment Following the review guidance provided in Chapter 15.0.2 of the SRP, the next area of review for transient and accident analysis methods focuses on assessment of the code. The associated acceptance criteria indicate that all models need to be assessed over the entire range of conditions encountered in the transient or accident scenarios. The review procedures provided in Section III of Chapter 15.0.2 of the SRP also indicate that the assessment of these models is commensurate with their importance and required fidelity. This assessment is generally performed via comparison of predicted results against both separate effects tests and integral effects tests.

Separate effects tests are generally used to demonstrate the adequacy of individual models and the closure relationships contained therein. Complementary to these types of tests are integral tests, which are generally used to demonstrate physical and code model interactions that are determined to be important for the full-size plant. The NRC staff evaluation of the individual elements of the code assessment suite provided in ANP-10346P are presented in the following subsections.

4.3.1 Review of TR Section 6.1 - Test Suite and Acceptance Criteria The following acceptance criteria were proposed by Framatome for validation against measured data:

Calculated void fraction [''''''''' '' '''''] of measured Calculated pressure drop [''''''''' ' '] percent of measured Calculated decay ratio ['''''''''' '' ''] of measured, with the exception that higher decay ratios above this range are considered acceptable Calculated oscillation frequency ['''''''' ' ''' '''] of measured

['''''''''''''''''''''''''''' ''''''' ''''''''''''''' '''''''''''''''' '''''' '''''''''''''''''''''' ''''''''' '''''''' '''' '''''''''''''''

'''''''''''''''''' '' '''''''''''''''''''''''''' '''''''''''''''''' ''''''''']

Acceptance criteria for nonlinear benchmarks given on a case-specific basis These acceptance criteria ranges for void fraction, pressure drop, decay ratio, and oscillation frequency correspond to the uncertainties determined in the prediction of these parameters in STAIF (Ref. 15) and the RAMONA5-FA LTSS methodology. As discussed in Section 4.4, ATWS analyses are not required to explicitly account for modeling uncertainties, as is required for design-basis event analyses such as those for which STAIF and the RAMONA5-FA LTSS methodology are used. However, the NRC staff finds the approach of defining validation acceptance criteria for ANP-10346P based on relevant uncertainty bounds for the previously-approved stability analysis methods STAIF and RAMONA5-FA to be acceptable because this demonstrates that the ANP-10346P methodology has similar or not significantly greater modeling uncertainty than the previously approved stability methodologies.

The NRC staff finds the added stipulation that calculated decay ratios are allowed to be more than 0.2 higher than measured to be acceptable for this application because higher decay ratios are conservative and [''''' ' ''''''' '''''''] in calculated-versus-measured decay ratio was seen in the benchmarking to KATHY stability tests (Section 4.3.4 of this SE), indicating good predictive capability of the ANP-10346P methodology across a wide range of conditions.

The suitability of the acceptance criteria for the pin-dependent CPROM term and for the nonlinear benchmarks is discussed as part of their own separate subsections later in this SE.

4.3.2 Review of TR Section 6.2 - Benchmarking to Void Fraction Tests As described in Sections 4.2.3.4 and 4.2.3.11.6 of this SE, the ANP-10346P methodology uses the [''''''''''''''''''''''''''''''''''''' ''''''''''''''''''''''] to determine the relationship between quality and void fraction in the fuel bundles. The ['''''''''''''''''''''''''''''''''''''' ''''''''''''''''''''] was approved for use in the RAMONA5-FA LTSS methodology.

Section 6.2 of ANP-10346P states that the following steady-state void fraction data sets were used to validate this correlation:

FRIGG (314 test points)

ATRIUM-10 KATHY (['' '''' '''''''''''])

ATRIUM 10XM KATHY ([' '''' ''''''''''])

These data include a wide range of pressure, inlet subcooling, and mass flow rate conditions.

The ATRIUM-10 and ATRIUM 10XM data include a maximum void fractions of [''''' ''''''''''''']

and ['''''' '''''''''''], respectively. These data are the same as were used to validate a different void fraction correlation in the Monticello ATWS-I methodology. However, all data collected for pressures outside the range of ['''' '' '''''''' ''] were discarded for the current application.

The NRC staff finds this acceptable because this pressure range encompasses the expected pressure during postulated limiting ATWS-I events. Additionally, [''''''''''''''''''''''' '''''' '''' ''''

''''' '''''''' ''''''' '''''''''''''''' ' ''''''''''''''''''''''''' '''''''''''''' '' '''''''''''''''''''' '''''''''''' '''''''''''''''''''''''''''].

However, the remaining data used for validation - especially the ATRIUM-10XM data, which covers the broadest range of conditions - provides very good coverage of the range of expected operating conditions during ATWS-I. The calculated void fraction demonstrates good agreement with the measured data, and includes benchmarking directly relevant to the specific geometric configuration of current fuel types (ATRIUM-10 and ATRIUM 10XM).

The [''''''''''''''''''''''''''''''''' ''''''''''''''''''''' '''''''''''''''''' ' '''''' ''''''''''''''''''''' '' '''''''''' '''''''''''' '''''''''''''''

'''''''' '''''''''''''''''' '' '''' ''''' ''''''''' '''''' ''''''''' ''''''''''''''''''''' ''''''''''''''''''''] As discussed in Section 4.2.3.7 of this SE, the dynamic transient behavior - especially under rapidly-changing conditions such as large amplitude oscillations - may cause significant departure of the relative phase velocity behavior from the behavior under steady-state conditions, and this behavior is not directly validated by the steady-state void fraction tests.

However, the linear and nonlinear stability tests discussed in later sections provide an integral validation of the overall code behavior, including the void fraction correlation. The calculated stability behavior (e.g., oscillation growth rate) is highly sensitive to the void fraction correlation due to its impact on the pressure drop response and density reactivity feedback; therefore, the close agreement of the ANP-10346P methodology with measured linear and nonlinear stability data under a wide range of conditions provides additional assurance that the void fraction correlation does not impose any significant nonconservative error trend or bias in the calculated results under expected oscillatory conditions during ATWS-I.

4.3.3 Review of TR Section 6.3 - Benchmarking to KATHY Pressure Drop Tests Results from the RAMONA5-FA ATWS-I code for pressure drop were validated against the following steady state pressure drop measurements:

KATHY ATRIUM-10 KATHY ATRIUM 10XM Benchmarking against pressure drop data allows for validation of the total pressure drop calculated in the ANP-10346P methodology under steady-state conditions. For single-phase flow, the total pressure drop depends directly on the single-phase friction factor (as well as the liquid density thermophysical correlation, which has low uncertainty). For two-phase flow, the total pressure drop depends primarily on the single-phase friction factor, two-phase friction

multiplier, and void-quality correlation (which determines the density and velocity of the fluid).

Because the void-quality correlation was directly validated by the steady-state void fraction benchmarks and shown to give good agreement, the pressure drop tests are particularly useful in validating the single-phase friction factor and two-phase multiplier.

The measurements include a broad range of pressure, inlet temperature, and mass flow rate. A total of ['''' '''''''''''''''' '''''''' '''' '''' '''''''''''''''''''''''' ''''''' ''''''''] were included in the validation, covering single-phase and two-phase conditions. For ATRIUM-10, the mean relative error for the single-phase (two-phase) data points was [''''' ''''''''''''' ''''''' '''''''''''''' '''' ''''''''''''''''''

''''''''''''''''' '' ''''''''''''''' ''''' '''''''''']. For ATRIUM 10XM, the single-phase (two-phase) mean relative error was [''' '''''''''''' '''' '''''''''''''' '''' ''''''''''''''''' ''''''''''''''' ''' ''''''''''''

'''''''''''''''''''''''']. Both the single-phase and two-phase tests (for both fuel types) demonstrate close agreement between calculated and measured total pressure drop with no observable trends, and the NRC staff concludes that the ANP-10346P methodology is well-validated for calculating pressure drop over a wide range of operating conditions. As with the ['''''''''''''''''''''

'''''''''''''''''''] discussed above, the calculation of pressure drop is an important parameter to correctly determine thermal hydraulic stability, and the close agreement with measured linear and nonlinear stability data discussed below provides added assurance that the single-phase friction factor and two-phase multiplier and their implementation for oscillatory transient conditions is accurate and acceptable for ATWS-I applications.

4.3.4 Review of TR Section 6.4 - Benchmarking to KATHY Stability Tests Results from the RAMONA5-FA ATWS-I code for single-assembly thermal hydraulic stability were compared to measured stability data in KATHY for the following fuel designs:

ATRIUM-10 (['] test points)

ATRIUM 10XM (['] test points)

These tests included stable conditions (decay ratio less than one) as well as unstable conditions (decay ratio greater than one). For the stable test points, the decay ratio and resonance frequency were determined from analysis of noise in the output signals using well-established numerical techniques. For the unstable points, the decay ratio and resonance frequency were determined from analysis of the coherent oscillation signals above noise level in the output data.

The benchmarking results show acceptable agreement between measured and calculated decay ratios, with the majority of the calculated decay ratios [''''''' '' ''' '''''''''''''''''' ''''''''].

Most of the points that have [''''''''' '''''' '''''' ''] are near or above the stability boundary, and the RAMONA5-FA ATWS-I code [''''''''''''''''''''' '' '''''''' ''''' ''''''''''''''''' '''''''''''''''''''' '''''' '

'''''''' '''''''' '''''''''' ''''''''''''''' ''''''''''''''''''''''''']. The mean error in calculated versus measured frequency is [''''''''''''''''''''''''' ''''' ''''''' '''], with very few points exhibiting error larger than

['''''''''''''']. The comparison of calculated to measured decay ratio and frequency satisfies the acceptance criteria discussed in Section 4.3.1 of this SE and demonstrates the ability of the RAMONA5-FA ATWS-I methodology to accurately predict the channel thermal hydraulic stability behavior of ATRIUM fuel types. These tests provide an integral validation of the thermal hydraulic phenomena important for stability, including fluid mass, momentum, and energy transport as well as constitutive relations such as the void-quality, friction factor, and wall heat transfer coefficients in the subcooled and two-phase nucleate boiling regimes.

4.3.5 Review of TR Section 6.5 - Benchmarking to KATHY Dryout/rewet Tests The RAMONA5-FA ATWS-I code was benchmarked against KATHY dryout/rewet stability tests for the following fuel types:

ATRIUM-9 ([] test points)

ATRIUM 10XM (['] test points)

These experiments were reviewed and evaluated in the SE for the Monticello ATWS-I methodology and were found to provide a realistic representation of the thermal hydraulic behavior, including the impact of neutronic feedback, during large amplitude oscillations characteristic of ATWS-I conditions up to and including failure to rewet. In particular, the inclusion of realistically simulated neutronic feedback allows significant inlet flow reversal to occur and promotes the occurrence of dryout at low elevations as expected for ATWS-I events.

The NRC staff examined Figures 6-6 through 6-20 in ANP-10346P and has concluded that the RAMONA5-FA ATWS-I methodology provides a reasonable and realistic agreement with the qualitative behavior of the KATHY dryout/rewet tests, including the onset and growth of oscillations, cyclic dryout and rewet, and eventual failure to rewet. Furthermore, for each test case, ['''''''''''''''''' ''''''' ' ''''''''''''' ' '' ''''''''' ''''''''''''' ' ''''''''''' '''''''' ''''''' '''''''''''''''''

''''''''''' ' ''' ''''''''' ''''''''''''''' ' '''''''' '''''''' ''''''''''' '' '''''''''''''''''''''' '''''''''''''''''''''''''''''] As discussed in Section 4.2.5.5 of this SE, [' ''''''''''''''''''''' ''''''''' ''''''''''''''''''' '''''''' '' ''''''''''''''''

'''''''''' ''''''''''''''''' '''''''''''''''''' '''''''''''''''''''''''''' '''''''''''''']

[''''''''''' ''''''''''''''''''''''''''' ''''''' ''''' '''''' ' '''''''''''''' '''' ''''''''' '''' ''' ''''' '''''''''''''' ''''''''''''''''''''

''''''''''''''''''' ''''''' ''''']

The KATHY dryout/rewet experiments serve as an extension of the model validation described in the previous section and additionally provide validation of [''''''''''''''''''''''''' ''''''' ''''''''''''''' '

'' ''''''' '''' ''''''''''''''''''' '''''''''' '''' '' ''''''' '''''''''''''' '''''''''''''''''''' '''''''''''].

4.3.6 Review of TR Section 6.6 - Benchmarking to Linear Reactor Stability Benchmarks Section 6.6 of ANP-10346P describes the benchmarking performed with the RAMONA5-FA ATWS-I code for linear reactor stability data for the following BWR plant events:

[''''''''''''''''''''''''''''''' ''''' ' '''''''' ' '''''''''''''' '''''''''''''''' ''''']

['''''''''''''''''''''''''''''''''''''' '''' '''''' ' ''''''''''''' ''''''''''''''''' ''''''''''''''''''']

[''''''''''''''' ''''''' '''''''''''' '''''''''''''''''' '''']

['''''''''''''''' ''''''''''''''''''' ''''''''''' '''''''' ' ''''''''' ''''''''''''''''' ''''''']

These linear reactor stability benchmarks were also included in benchmarking suites for the approved RAMONA5-FA LTSS and STAIF methodologies. These events involved measured oscillations with a decay ratio of approximately ['' ''''''''''''' ''''''''''''' '''''' ''''' '''''''' '''''' '

'''''''' ]. Therefore, similarly to the KATHY stability tests, these benchmarks provide an integral validation of the fluid mass, momentum, and energy transport as well as constitutive relations in terms of their impact on system stability. Specifically, these benchmarks are used to validate the prediction of the timing of stability onset and the oscillation growth rate, but not the prediction of dryout or rewet that could potentially occur during the later stages of postulated ATWS-I events.

However, compared to the KATHY single-assembly stability tests, the linear reactor stability benchmarks also provide validation of the effect of neutronic feedback - including the effect on wall-to-fluid heat transfer as a function of space and time - on the overall system oscillation characteristics.

Another key difference relative to the KATHY stability tests is that these benchmarks involve mixed cores with multiple fuel types other than ATRIUM 10 and ATRIUM 10XM. The NRC staff issued RAI-6 to obtain additional information regarding the linear reactor stability benchmarks, including information on operating conditions and fuel types for each benchmark case. In the response to RAI-6, Framatome provided the requested list of fuel types and conditions present in each of the benchmarked cores. Framatome indicated that the majority of fuel-specific inputs were available from the past benchmarking for the STAIF and RAMONA5-FA codes. In some cases, some input for the [''''' '''' ''' '''''''''''''''' '''''''''''''' '''''' ' '''''''''''''''''''''''''''''''''''''' ''''

''''''' ''''''''''''' '''''''''' ' ''''''''' ''' ''''''''''''' ]. The NRC staff reviewed the information that was inferred and determined that this inference was performed in an acceptable manner and that any differences between the inferred and actual values for these fuel types would be expected to have a minor impact on the results. For these benchmarks, all neutronic and thermal

hydraulic data were taken directly from the benchmarking suites for the approved STAIF and RAMONA5-FA codes, with no additional neutronic or thermal hydraulic data required.

Also in the response to RAI-6, Framatome provided a table of operating conditions and power distribution information for each of the linear reactor benchmarks. The benchmarks consist of stability tests performed at off-rated conditions during reactor startup ([''' '''''''''''''''''''''''''''''''' '''''

'''''''''''''''' ''''''' '''''''' '''''''''']) or a stability event from off-rated conditions ([''' '''''''''''''''''''

'''''''''''''''''''' ''''''''''' ''''''''''''''''''' ''''''' '''''''' ''''''''']). Although these operating conditions may not strictly correspond to the conditions that would occur during a 2RPT or TTWB ATWS event at each plant, the tests encompass a reasonably wide range of power, flow rate, inlet subcooling, outlet quality, and axial power shapes (including highly bottom-peaked axial power profiles) similar to those expected during the initial oscillation growth phase of ATWS-I events. The oscillation decay ratios and frequencies calculated by the RAMONA5-FA ATWS-I code were within the acceptance criteria listed in Section 4.3.1 of this SE for all [''''] linear benchmark cases, which included both regional and core wide oscillation modes, and the results showed no discernible trend with respect to operating conditions, fuel types, or other plant-specific differences that exist among the four benchmark cases. Therefore, the NRC staff concludes that the linear stability benchmarks provide good additional assurance, beyond the single-assembly KATHY stability benchmarking, that the code is able to predict the onset and initial growth rate of oscillations that would occur during postulated ATWS-I events on a generic basis.

4.3.7 Review of TR Section 6.7 - Benchmarking to Non-Linear Reactor Benchmarks Section 6.7 of ANP-10346P describes the benchmarking performed with the RAMONA5-FA ATWS-I code for the following two nonlinear stability events:

o Oskarshamn turbine trip with non-linear oscillation o BWR A feedwater temperature transient with non-linear oscillation Unlike the linear reactor benchmarks, the nonlinear reactor benchmarks provide direct validation of the system response during events leading up to reactor instability, as well as validation of the onset timing and growth of oscillations up to relatively large amplitude (approximately

['''''''''''''''''''''''''''''' ''''''''''] in both benchmarks) before oscillation suppression via scram.

Limited details were given in ANP-10346P for the boundary conditions and assumptions used in the RAMONA5-FA ATWS-I methodology for these benchmarks. To assist the NRC staff in determining whether the events were analyzed in an acceptable manner, the NRC staff requested further details in RAI-7 on the fuel types, fuel-specific data, boundary conditions, and other assumptions used for these two cases. In the RAI response, Framatome indicated that

['' '''''''''' ''''''' ''''' '''''''''''''''''''''' '''''' ''''''''' ' ''''''''''''''''''''''' ''''' '''' '' '''''''''''''''''''''

'''''''''''''''' ''''' ''''''''''''' '''''''''''''''' '''''' ' '''' ''''''' '''''' ''''''''''''''']. However, as was the case for the linear reactor benchmarks, some input data for ['' ''''' '''' '''' ''''''''''''' ''''''''''''']

were not directly available for all fuel in the core, and in these cases the additional input data were inferred by comparisons to similar ['''' '''' ''''''''''']. The NRC staff finds this acceptable for the same reasons as stated in the previous section.

Additionally, in the response to RAI-7, Framatome provided further information on the boundary conditions used in both models. For the Oskarshamn-2 benchmark, most initial conditions and

boundary conditions were taken from the OECD/NRC Oskarshamn-2 BWR Stability Benchmark specifications (Ref. 16). Notably, this included the feedwater temperature time-dependent behavior provided in the benchmark specifications, in which the feedwater temperature was assumed to decrease earlier than the measured value to account for the heat-conduction-related time delay between the actual and measured feedwater temperature during the event. Feedwater flow rate was decreased from the measured data to ensure reasonable water level calculation in the RAMONA5-FA ATWS-I code, and two sets of runs were made for pump speed: one run using the measured pump speed versus time and a second run using a modified pump speed to more closely match the measured core flow rate.

The NRC staff reviewed these assumptions and found them to constitute a reasonable and acceptable representation of the Oskarshamn-2 instability event.

['''' ''''''''''' '''''''' '''''''''''''''' ''''' '''' ''''''''''''''''''''''' '''''''''''''''' ''' ''''''''''''''''''''' ''''''''''''''''''''''''''''

'''''''''''''''' ''''''''' '''''' ' ''''''''''''''''''''''' ''''''''''''''' ''''''''''''''''''''''' '''''' '' ''''''''''''''''''''''''''''''' ''''''']

For the BWR-A benchmark, [ '''''' '''' ''''''''''''''''' '''''''''''''''''''' '''''''' ''''''' '''''' '''''''''''''''''

''''''''''''''''''''''''''''' '''''''''''' ' ''''' ''''''' '' ''' '''''''''''''''''''''''''''''''' '''''''''''''''''' '''''''']

Results for the two benchmarks are shown in Figures 6-24 through 6-30 of ANP-10346P. In the Oskarshamn-2 case, the calculated core-average power matches the measured core-average power closely up to the point of instability, and the calculated oscillation onset time and oscillation frequency appear to match the measured values closely while the calculated oscillation growth rate appears to be noticeably larger than measured, which is conservative.

The two different assumptions used for pump speed affected the oscillation growth rate, but this growth rate was higher than measured in both cases.

['''''''''''''''' '' '''''''''''' ''''''''''''''''' ''''''''' ''''''''''''' ''''''''''''''''' '''' ''''''''''''''''''''''''' ''''''''''''' ' ''

''''' '''' '''''''''''''''''' '''''''''''''' '''''''''' '' '''''''''''''''''''''' ]

Therefore, the NRC staff concludes that both nonlinear reactor benchmark cases demonstrate the accurate or possibly conservative prediction of oscillation onset time and growth rate during measured BWR instability events with relatively large oscillation amplitude.

4.4 Uncertainty Analysis As opposed to analyses for design basis events, which should explicitly account for modeling uncertainties to ensure that the safety criteria are met, ATWS analyses may use best-estimate or reasonably bounding modeling approaches to demonstrate acceptable consequences to the public under limiting ATWS events. No explicit requirement or guidance is given for analyzing uncertainties in the calculated results for these events. The rationale for this is that ATWS events, which are beyond design basis events, have very low probabilities of occurrence compared to design basis events.

Although ATWS analyses may use a best-estimate approach, some understanding of the impact of variations in specific parameters or modeling assumptions in relation to satisfying the ATWS acceptance criteria is important, in order to properly evaluate the models and assist in determining acceptable input and modeling requirements for the given application. Framatome provided a PIRT in Section 4.15 of the TR to assist in this process. As discussed in Section 4.1 of this SE, Framatome ranked various phenomena by their importance for three figures of merit

- oscillation inception, limit cycle amplitude, and post-dryout - which affect the ATWS-I event progression in different ways and contribute to the overall PCT in a given calculation. The NRC staff performed its review of the RAMONA5-FA ATWS-I methodology partly based on this PIRT, as a means of focusing the review preferentially on phenomena and corresponding models with higher importance. For example, the NRC staff issued RAI-4 to obtain more information on the validation of the gap model and RAI-11 to request sensitivity studies by adjusting the gap conductance values, in part because the ['''''''''''''' ''''' '''''''''''''''''''' ''''''''''''''''' ''' ''' ''''

''''''''''''''''' '''''''''''''''' '' ''''''''''''''''''''''' '''' '] were dispositioned as parameters of high or medium importance for the oscillation inception and post-dryout figures of merit. Additionally, although the gap model was applicable to the linear and nonlinear core benchmarks, it was not applicable to ['' '''''''''''''' '''''''''''' ''''''' '''' '' ''''' ''''''''''''''''''''' ''''''''' ''''' ' '''''''' '''''''''''''''''''''''].

Therefore, numerical sensitivity analyses were particularly useful for this model to understand the models impact on stability calculations. The NRC staffs evaluation of these RAIs is presented in Section 4.2.2.6 of this SE.

Several other highly-ranked phenomena such as total core power, total core flow, feedwater temperature, core size, and core design are determined or justified uniquely for each plant-specific application and therefore were not subjected to sensitivity studies.

Additional highly-ranked phenomena such as ['''''''''''''''''' '''''''''''''''' ''''' '''''''''''''''' '' ''''

'''''''''''''''''''' ''''' ''''''''''''''''''''''''''' '''''''] are related to processes involving fluid transport, heat transfer, and neutronic coupling, which impact the stability behavior of the system. The models used to determine ['''''''''''''' '''''''' ''''' ''''''''''''' ''''' '''''' '''''' '''''''''' ''''''''''''''' '''''''''''''

''''''' '''' '''' ]. Their behavior under transient conditions - specifically, their impact on stability - was validated in an indirect manner through their impact on the KATHY and full-core stability benchmarking, which provide an integral validation of the stability-related dynamic processes as discussed in Sections 4.3.4-4.3.7 of this SE. Due to the extensive benchmarking of the stability predictions of the code under a wide range of operating conditions, which provides confidence that the relevant phenomena are calculated accurately, additional sensitivity calculations were not requested by the NRC staff for these models.

However, in RAI-8 and RAI-9, the NRC staff did request justification that the core and vessel nodalization were sufficient to provide reasonable and accurate prediction of PCT during

ATWS-I events. The discretization used in the numerical solution of the models impacts the transport of mass, momentum, and energy in the system - in particular, by impacting numerical diffusion - and therefore it may have high importance on the calculated stability behavior.

Discussion and evaluation of the response to RAI-8 and RAI-9 is given in Sections 4.2.3.12 and 4.2.3.1 of this SE, respectively.

Additional sensitivity studies were provided by Framatome in ANP-10346P, including sensitivities on ['''''''''''''''''' ''''''''''''''''''''''''' ''''''''''''''''''''''''''''' '''''''''''''''' '''''''''' '''''''' ''''''' '''''''''''''''''''

''''''''''' ''''' '''''''' ''''''''''''''''''''''' '''' '''''''''''''''''''' ' ''''''''''''''''' ''''''''''''' ''''' ''''''''' '''''''''''''''''''''']

['''' ''''''''''''''''' '''''''''''''' ' ''''''''''''''''''' '''''''''''''''''''''''''' '''' ''''''''''''''' ''''''''' ''''' '''''''' ''''''''''''''''''

'''''''''''''''' '''''''' ''''''''''' ''''''''' '''''''''''''' '''''''''] The NRC staffs experience shows that the rod node associated with initial failure-to-rewet is not necessarily the hottest (highest peaking factor) node in the core, and additional, higher-power nodes which fail to rewet later in the event may cause large, rapid increases in PCT, potentially on the order of hundreds of degrees.

In addition to possible PCT increases associated with changes to the limiting PCT node location, the PCT at a given limiting node location is also expected to increase with increasing core inlet subcooling because this increases the core average power level. This behavior was observed in the time-dependent results for the sample problem provided in ANP-10346P as well as the additional sensitivity results provided in the responses to RAI-8 through RAI-12. In these results, the PCT following failure to rewet appears to increase and decrease in tandem with the core inlet subcooling.

['' '''''''''''''''' '''''''''''''''''' ''''''' '''''' ''''''''''''''''' ' '''''''' '' ''''''''''''' ''''''''''''''''' '''''''''''''''''''''''

'' '''''''' ''''''''''''''' ' ''''''' ''''''''''']

The determination of core inlet subcooling is not straightforward and depends on the codes ability to accurately model the mixing of injected feedwater from the vessel feedwater spargers into the vessel downcomer liquid - or vapor, if the water level is low enough. Additionally, the code must accurately model the mass and energy transport of this fluid through the vessel downcomer and lower plenum in order to properly determine the core inlet subcooling as a function of time. This calculation further depends on the codes ability to accurately model the feedwater flow rate, which responds to changes in the steam flow rate exiting the vessel and is characterized by a time delay based on the balance-of-plant dynamics. As a result of these dynamic effects, the core inlet - and, relatedly, the PCT - may continue to increase for a significant length of time ([' '' '''''''' ''''''' '''''''''''''' ' '' ''''''''''''''''''''' ''''''''''''' ''''''''''''''])

after operator actions are performed, such that the timing and magnitude of peak PCT is determined by competing dynamic effects associated with feedwater flow rate, water level, and the mass and energy transport of fluid through the vessel.

Based on the sensitivity results provided in ANP-10346P as well as in the responses to RAI-8 through RAI-12, the discretization assumptions - including vessel nodalization, core nodalization, and timestep size - as well as core modeling assumptions such as gap conductance appear to have a ['''''''''''''''''' '' '''''''''''''''''''' '''''''''] effect on the core inlet temperature response for the Brunswick sample problem. These results highlight the importance of accurately modeling the vessel and recirculation loop as well as the balance of plant dynamics on a plant-specific basis, as these are expected to be the primary determinants of the core inlet temperature response during ATWS-I events. Modeling assumptions which impact the core response, such as core nodalization and gap conductance, may have some effect on core inlet temperature, but the more important effect of these parameters appears to be in impacting the stability behavior of the core itself and the timing of oscillation growth with respect to operator actions. The NRC staff expects that this conclusion likely also holds for cycle-specific changes because such changes would primarily impact the core behavior - via, for example, changes in the radial and axial power distribution - while the recirculation loop and balance of plant dynamics would typically not change between cycles. As discussed in Section 4.2.8 of this SE, any changes to the plant-specific configuration other than cycle-specific changes to the fuel in the core will require an evaluation to ensure that the ATWS-I analyses remain bounding, or reanalysis of the ATWS-I event with the updated plant configuration. In addition, ATWS-I analyses must be justified to reasonably bound the behavior for future cycles when considering possible changes in oscillation onset timing and mode behavior which are expected to be caused primarily by differences in core fuel loading and operational changes even when the ex-core plant configuration remains unchanged.

Even though the NRC guidance for beyond design basis accidents such as the ATWS-I event does not require uncertainties to be accounted for within the analysis conclusions, Framatome provided some sensitivity analyses to demonstrate the relative sensitivity of the ATWS-I results to specific parameters that were not explicitly evaluated through code validation. As discussed above, most of the sensitivities were relatively modest, except for ['' '''''''''''''' ''''''''' '''''''' '''''

''''''''''''''''' ''''''''''''''''''''' '''''''''''''''' '''']. The NRC staff found that the guidance provided in ANP-10346P, as supplemented in the RAI responses and augmented by the limitations and conditions listed in Section 5.0 of this SE, was adequate to ensure that the important sensitivities are adequately addressed for each application of this methodology.

5.0 LIMITATIONS AND CONDITIONS As discussed previously in this report, the following conditions and limitations have been applied to NRC approval for use of the RAMONA5-FA methodology to analyze the ATWS-I event as described in ANP-10346P. Table 8-1 of ANP-10346P contains additional limitations and conditions imposed on the applicability of the RAMONA5-FA ATWS-I methodology by Framatome; these limitations and conditions are considered to be part of the proposed methodology and are not included below.

1. The gap conductance sensitivity shall be repeated or otherwise justified for transitions to new fuel designs.

2. If the acceptance criteria for the first paragraph in Step 3 of Section 8.0 of the TR are met, additional justification must still be provided to demonstrate adequate margin in operator action timing for variations in neutron kinetics response from specific core designs. This justification may be provided by following Steps 3.a through 3.c, as amended by the response to RAI 15, or providing an alternative justification on a plant-specific basis.

3. Plant-specific evaluations that are intended to be bounding of all core designs must be confirmed to provide reasonable assurance that neutron kinetics characteristics such as possible differences in dominant oscillation modes or the potential for multiple oscillation modes to be active simultaneously are bounded by the analysis of record.

4. Due to the unique neutron kinetics characteristics associated with transition cycles, all transition cycles must be dispositioned in a manner consistent with Limitations and Conditions #2 and #3.

5. The ATWS-I analysis must be performed for both the TTWB and 2RPT events during the initial implementation of this methodology, to confirm which event is limiting.

Subsequent evaluations may only consider the event determined to be limiting, except when changes are made to the plant design or operation that may affect stability behavior during ATWS, such as: turbine bypass capability, fraction of steam-driven feedwater pumps, and changes expected to significantly increase core inlet subcooling during ATWS events.

6. The steam line and valve modeling options shall be confirmed to accurately capture the expected plant-specific system performance during ATWS-I events.

7. Plant-specific applications must justify that the selected settings and modeling options are appropriate, including core and vessel nodalization, time step control parameters, and noise parameters. In particular, the modeling should be reasonably consistent with both the characteristics of the plant in question and the validation basis for the RAMONA5-FA ATWS-I methodology as discussed in this SE.

6.0 CONCLUSION

S In ANP-10346P, Framatome presented a proposed methodology to analyze the ATWS-I event using the RAMONA5-FA code. The following conclusions are provided here in summary as they apply to BWR/3-6 submittals.

ANP-10346P presents a description of the ATWS-I event, the relevant phenomena, the applicable Figures of Merit (FoMs), and a ranking of the phenomena for any applicable FoMs.

This information was reviewed and compared to similar information available to the NRC staff (e.g., Ref. 9) and confirmed to be consistent with previous approvals of ATWS-I or other stability related methodologies.

The application of the RAMONA5-FA code for the purpose of analyzing ATWS-I events involved the incorporation of several new models in the RAMONA5-FA code relative to what the NRC staff has previously reviewed and approved for LTSS analyses. Many of these models had been reviewed and approved by the NRC staff as part of a plant-specific ATWS-I methodology adopted at Monticello. The NRC reviewed the previously approved RAMONA5-FA models, the

previously approved models from the Monticello ATWS-I application, and new models developed specifically for the purpose of the ANP-10346P methodology. The NRC staff confirmed that the previously approved models and new models are applicable to analysis of the ATWS-I event.

ANP-10346P also presents a procedure for analysis of the ATWS-I event, which [''''''''''''''''

'''''''''''''''' ''''''''''''' ''''''''''''''''' ' ''''''''''''''''' '''''''''''' '''''''''''''''''''''' ''''''''''' ''' '''']. Since the intent of the proposed ATWS-I analysis methodology is to perform a single evaluation upon initial implementation at a specific plant without subsequent confirmatory analyses on a cycle-specific basis, the NRC staff carefully considered how different characteristics of future cycles might affect the results of a cycle-independent evaluation. In addition to changes in fuel assembly designs (including transition core designs), the NRC staff considered whether cycle or plant configuration changes might affect the limiting PCT or the margin to operator action timing.

As a result of sensitivities of the coupled neutronic/thermal-hydraulic feedback to cycle-specific variations in the core neutronic or plant system response, conditions (1) through (5) were found to be necessary to ensure that a plant-specific analysis will bound future core designs and changes to relevant plant system parameters. Additionally, conditions (6) and (7) were identified to ensure that the plant-specific models used for analysis of the ATWS-I event are consistent with the underlying validation and assessment of the methodology as described in ANP-10346P and the RAI responses.

In order to demonstrate the capability of the RAMONA5-FA code to analyze the ATWS-I event, assessments were made against separate effects tests and integral benchmarks. Separate effects tests helped validate the RAMONA5-FA code for prediction of parameters important to the ATWS-I event, such as void fraction, pressure drop, single channel stability characteristics, and dryout/rewetting response during large amplitude oscillations. The integral benchmarks provided confidence in the RAMONA5-FA codes ability to model full scale stability events. In some cases, sensitivity studies were used to demonstrate that the RAMONA5-FA ATWS-I methodology was either conservative or insensitive to variations in specific parameters. This provided assurance that relevant uncertainties in the ATWS-I analysis methodology and model parameters would not change the conclusions of an ATWS-I evaluation done in accordance with ANP-10346P. Based on a general review of the tests, benchmarks, and sensitivity studies, the NRC staff determined that the methodology was appropriately confirmed to yield acceptable predictions for all parameters and phenomena important to the ATWS-I event, provided that the limitations and conditions are met.

In summary, the NRC staff finds that the assessment of the RAMONA5-FA code, as described in ANP-10346P and responses to NRC staff RAIs, adequately demonstrates that RAMONA5-FA is suitable to analyze the ATWS-I event by demonstrating acceptable performance in each of the highly ranked phenomena. In addition, the NRC staff finds that the procedure described in ANP-10346P for performance of the ATWS-I analyses provides appropriate guidance to perform ATWS-I analyses that will bound cycle-specific variations at a given plant, subject to limitations and conditions. As such, NRC staff approval of the ANP-10346P methodology for analysis of the ATWS-I event is contingent on adherence to the conditions and limitations set forth in Section 5.0.

7.0 REFERENCES

1. Letter from Gary Peters, Director, Licensing & Regulatory Affairs, AREVA, Inc., to USNRC Document Control Desk, Request for Review and Approval of ANP-10346P, Revision 0,

ATWS-I Analysis Methodology for BWRs Using RAMONA5-FA, dated December 15, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17355A231).

2. AREVA NP Inc. Topical Report ANP-10346P, Revision 0, ATWS-I Analysis Methodology for BWRs Using RAMONA5-FA, December 2017 (ADAMS Accession No. ML17355A233 (Non-Proprietary)/ML17355A235 (Proprietary)).

3. AREVA NP Inc. Licensing Topical Report EMF 3028P-A, Volume 2, Revision 4, RAMONA5-FA: A Computer Program for BWR Transient Analysis in the Time Domain:

Theory Manual (ADAMS Accession No. ML131550602 (Proprietary)).

4. AREVA NP Inc. Report ANP-3274P, Revision 2, Analytical Methods for Monticello ATWS-I, July 2016 (ADAMS Accession No. ML16221A275 (Non-Proprietary)/ML16221A278 (Proprietary); Approved via License Amendment in ADAMS Accession No. ML17054C394).

5. Letter from Gary Peters, Director, Licensing & Regulatory Affairs, Framatome, Inc., to USNRC Document Control Desk, Response to Request for Additional Information Regarding ANP-10346P, Revision 0, ATWS-I Analysis Methodology for BWRs Using RAMONA5-FA, dated March 8, 2019 (ADAMS Accession No. ML19071A274).

6. Letter from Jonathan Rowley, Project Manager, Licensing Processes Branch, Division of Policy and Rulemaking, USNRC, to Gary Peters, Director, Licensing & Regulatory Affairs, Framatome, Inc., Audit Report for the June 11-13, 2018, Audits in Support of the Review of ANP-10346P, Revision 0, ATWS-I Analysis Methodology for BWRs Using RAMONA5-FA (EPID: L-2017-TOP-0067), April 25, 2019 (ADAMS Accession No. ML18249A225).

7. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: Light Water Reactor Edition, Chapter 15.8, Anticipated Transients Without Scram, Revision 2 (ADAMS Accession No. ML070570008).

8. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: Light Water Reactor Edition, Chapter 15.0.2, Review of Transient and Accident Analysis Methods, Revision 0 (ADAMS Accession No. ML070820123).

9. NUREG/CR-6743, Phenomenon Identification and Ranking Tables (PIRTs) for Power Oscillations Without Scram in Boiling Water Reactors Containing High Burnup Fuel, USNRC, September 2001 (ADAMS Accession No. ML012850300).

10. NUREG/CR-7179, BWR Anticipated Transients Without Scram in the MELLLA+ Expanded Operating Domain, Part 1: Model Development and Events Leading to Instability, USNRC, June 2015 (ADAMS Accession No. ML15169B064).

11. NUREG/CR-7180, BWR Anticipated Transients Without Scram in the MELLLA+ Expanded Operating Domain, Part 2: Sensitivity Studies for Events Leading to Instability, USNRC, June 2015 (ADAMS Accession No. ML15169A168).

12. Regulatory Guide 1.203, Transient and Accident Analysis Methods, December 2005 (ADAMS Accession No. ML053500170).

13. AREVA NP Inc. Licensing Topical Report BAW-10247PA, Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, February 2008 (ADAMS Accession No. ML081340208 (Non-Proprietary)/ML081340383 and ML081340385 (Proprietary)).

14. GE-Hitachi Nuclear Energy Licensing Topical Report NEDC-33006P-A, Revision 3, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, June 2009 (ADAMS Accession No. ML091800530).

15. Siemens Power Corporation Licensing Topical Report EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, August 2000 (ADAMS Accession No. ML090750216 (Proprietary)).

16. Kozlowski, T., Gajev, I., Lefvert, T., Christer, N., Roshan, S., et al., BWR Stability Event Benchmark based on Oskarshamn-2 1999 Feedwater Transient, Royal Institute of Technology Stockholm, Sweden, OECD Nuclear Energy Agency, 2013.

17. Yarsky, P., Applicability of TRACE/PARCS to MELLLA+ BWR ATWS Analyses -

Revision 1, USNRC Office of Nuclear Regulatory Research, November 18, 2011 (ADAMS Accession No. ML113350073).

18. Framatome Inc. Licensing Topical Report ANP-10340P-A, Revision 0, Incorporation of Chromia-Doped Fuel Properties in AREVA Approved Methods, May 2018 (ADAMS Accession No. ML18171A119 (Non-Proprietary)/ML18171A120 (Proprietary)).

Attachment:

Resolution of Comments Principal Contributors: Scott Krepel, NRR/DSS/SNPB Aaron Wysocki, Oak Ridge National Laboratory Date: October 30, 2019.