ML24311A072

FOIA-2024-000042 - Technical Evaluation Report

ML24311A072
Person / Time
Issue date:	05/31/2023
From:	NRC/OCIO
To:	Elliott G Framatome
References
FOIA-2024-000042
Download: ML24311A072 (1)

Text

Pacific Northwest NATIONAL LABORATORY PNNL-34230 Technical Evaluation Report of Topical Report ANP-10339P Revision 0 ARITA-ARTEMIS/RELAP Integrated Transient Analysis Methodology May 2023 OFFICIAL USE ONLY:

Kenneth J Geelhood David W Engel Travis J Zipperer Bruce E Schmitt David J Richmond May b empt from public release under the Freedom of Info USC 552),

mption number(s) and category:

May 8, 2023 Date Prepared for the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Under Contract DE-AC05-76RL01830 lnteragency Agreement: 31310019N0001 Task Order Number: 31310020F0051 61-1-ICliett USE 6NLY

DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor Battelle Memorial Institute, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof, or Battelle Memorial Institute. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

PACIFIC NORTHWEST NATIONAL LABORATORY operated by BATTELLE for the UNITED STATES DEPARTMENT OF ENERGY under Contract DE-AC05-76R.L0l830 OFFIOl,lcl 1:J&E ONLY

PNNL-34230 Technical Evaluation Report of Topical Report ANP-10339P Revision 0 ARITA-ARTEMIS/RELAP Integrated Transient Analysis Methodology May 2023 Kenneth J Geelhood David W Engel Travis J Zipperer Bruce E Schmitt David J Richmond Prepared for the U.S. Department of Energy under Contract DE-AC05-76RL01830 Pacific Northwest National Laboratory Richland, Washington 99354

PNNL-34230 Summary This technical evaluation report (TER) provides documentation of the Pacific Northwest National Laboratory's (PNNL) review of the Licensing Topical Report (L TR), ANP-10339P Revision 0, ARITA - ARTEMIS/RELAP Integrated Transient Analysis Methodology. This report documents the review of the models in ARITA, a technical assessment of the requested application of ARITA, and the methodology that has been proposed to calculate various figures of merit in a safety analysis for comparison with acceptance criteria. This report documents the acceptability of each of these items and proposes limitations for consideration in a safety evaluation report produced by NRC staff.

PNNL-34230 Summary ii

PNNL-34230 Acronyms and Abbreviations AFW ANS AO AOO ARC ARO ASME BE BOC BPVC BWR CE CET CFR CHF CMA Cp eves DNB DNBR DPC OTC ECCS EM EOC EOP ESFAS FAST FCM FGR FMA FoM FRM FRR FW (G)

Acronyms and Abbreviations Auxiliary Feedwater American Nuclear Society Axial Offset Anticipated Operational Occurrence Automatic Control Rod All Rods Out American Society of Mechanical Engineers Best Estimate Beginning of Cycle Boiler and Pressure Vessel Code Boiling Water Reactor Combustion Engineering Component Effects Tests Code of Federal Regulations Critical Heat Flux Conservative Methodology Assumptions heat capacity Chemical and Volume Control System Departure from Nucleate Boiling Departure from Nucleate Boiling Ratio Doppler Power Coefficient Doppler Temperature Coefficient Emergency Core Cooling System Evaluation Model End of Cycle Emergency Operating Plan Engineered Safety Features Actuation System Fuel Analysis Steady State and Transient (NRC fuel performance code)

Fuel Centerline Melt Fission Gas Release Foundation Methodology Assessments Figure of Merit Fuel Rod Module Fuel Rod Response Feedwater parameters with generic approval iii

GCN HCF HEM HUM IET ITC k

HZP LAR LBB LCO LOCA LOOP LP LSCM LTR LWR MDNBR MFIV MFW MSIV MSL MSLB MSS MSSV MTC NI NRC OTDT PA PC pcm PCS PDF PIRT PNNL PORV ppm Acronyms and Abbreviations Global Core Neutronics Hot Channel Factor Homogeneous Equilibrium Model Human Control System Integral Effects Test Isothermal Temperature Coefficient thermal conductivity Hot Zero Power License Amendment Request Leak Before Break Limiting Conditions for Operation Loss-of-Coolant Accident Loss of Offsite Power Low Pressure Loss of Subcooled Margin Licensing T apical Report Light Water Reactor Margin to Departure to Nucleate Boiling Ratio Main Feedwater Isolation Valve Main Feedwater Main Steam Isolation Valve Main Steam Line Main Steam Line Break Main Steam System Main Steam Safety Valve Moderator Temperature Coefficient Nuclear Instrument Nuclear Regulatory Commission Overtemperature Delta-Temperature Postulated Accident Power Coefficient per cent mille (one one-thousandth of a percent)

Primary Coolant System Probability Density Function Phenomenon Identification and Ranking Table Pacific Northwest National Laboratory Power Operated Relief Valve parts per million PNNL-34230 iv

(PS)

PSV PTO PWR PZRO QA RAI RCP RCS REA RPS RG RTD SAFDL SOBS SEC SET SG SGO SGTR SRP TBD TCS TCV TER TG TGV TH THAP TIC THM TM TS TSV UB us UTL OD Acronyms and Abbreviations parameters with plant specific approval Pressurizer Safety Valve Plant Transient Data Pressurized Water Reactor Pressurizer Overfill Quality Assurance Request for Additional Information Reactor Coolant Pump Reactor Coolant System Rod Ejection Accident Reactor Protection System Regulatory Guide Resistance Temperature Detector Specified Acceptable Fuel Design Limit Steam Dump and Bypass System Secondary Coolant System Separate Effects Test Steam Generator Steam Generator Overfill Steam Generator Tube Rupture Standard Review Plan To Be Determined Transient Cladding Strain Turbine Control Valve Technical Evaluation Report Turbine Generator Turbine Governor Valve Thermal Hydraulic Thermal Hydraulic Additional Parameters Time in Cycle Thermal Hydraulic Module Thermal Margin Technical Specifications Turbine Stop Valve Upper Bound United States Upper Tolerance Limit Zero Dimension PNNL-34230 V

PNNL-34230

~ ett Effective Delayed Neutron Fraction cr standard deviation

µ mean Acronyms and Abbreviations vi

PNNL-34230 Contents Summary................................................................................................................................... iii Acronyms and Abbreviations...................................................................................................... iii Contents................................................................................................................................... vii 1.0 Introduction..................................................................................................................... 1 2.0 Regulatory Evaluation..................................................................................................... 3 3.0 Technical Evaluation....................................................................................................... 5 3.1 Executive Summary............................................................................................. 5 3.2 Background.......................................................................................................... 5 3.2.1 Description of Physical Phenomena...................................................... 5 3.2.2 Relevant History.................................................................................... 7 3.3 Documentation................................................................................................... 10 3.3.1 Necessary Documentation................................................................... 10 3.4 Code Application................................................................................................ 11 3.5 Evaluation models.............................................................................................. 12 3.5.1 Individual Codes.................................................................................. 12 3.5.2 EM Implementation.............................................................................. 13 3.5.3 Assessment of Evaluation Models....................................................... 14 3.6 Accident Scenario Identification Process........................................................... 16 3.6.1 Target Scenarios................................................................................. 16 3.6.2 Phenomena Identification and Ranking............................................... 18

3. 7 Code Integral Assessment................................................................................. 19

3. 7.1 Range of Code Assessment................................................................ 19 3.7.2 S-RELAPS Assessment....................................................................... 20 3.7.3 ARTEMIS Nodal Assessment.............................................................. 21 3.7.4 ARTEMIS THM Assessment............................................................... 21 3.7.5 GALILEO Assessment......................................................................... 22 3.7.6 Coupled EM Assessment.................................................................... 23 3.8 Uncertainty Analysis.......................................................................................... 24 3.8.1 Treatment of Uncertainties for Highly Ranked Phenomena................. 25 3.8.2 Non-parametric Uncertainty Analysis Process..................................... 41 3.8.3 Uncertainty Parameters for Non-Parametric Uncertainty Analysis.......45 3.8.4 Application of Non-Parametric Statistics for Licensing Analyses.......... 97 4.0 Conclusions................................................................................................................. 100 5.0 References.................................................................................................................. 1 02 Contents vii

Figures Figure 1.

Figure 2.

Figure 3.

Tables Table 1.

Table 2.

Table 3.

Table 4.

Table 5.

Table 6.

Table 7.

Table 8.

Table 9.

Table 10.

Table 11.

Table 12.

Table 13.

Table 14.

Contents PNNL-34230 GALILEO nominal, 95/95 upper bound, and 95/95 lower bound ramp test hoop strain predictions for M5 and Zircaloy-4 assessment database................... 8 GALILEO nominal, 95/95 upper bound, and 95/95 lower bound centerline temperature predictions for assessment database............................................... 9 Data flow of system and core coupled calculations (Figure 4-1 of L TR)............. 13 Figures of merit and acceptance criteria for ARITA.............................................. 6 Transients performed with ARITA and EM used for each................................... 16 Treatment of PIRT processes and phenomena within ARITA............................ 26 Wilks Method Required Number of Observations for Each Order Statistic for at least 95/95 Tolerance...............................................................................41 Evaluation of PCS Uncertainty Parameters........................................................46 Evaluation of SEC Uncertainty Parameters........................................................ 66 Evaluation of TG Uncertainty Parameters.......................................................... 75 Evaluation of HUM Uncertainty Parameter......................................................... 76 Evaluation of GCN Uncertainty Parameters....................................................... 76 Evaluation of LCN Uncertainty Parameters........................................................ 79 Evaluation of TH Uncertainty Parameters.......................................................... 81 Evaluation of THAP Uncertainty Parameters...................................................... 82 Evaluation of FRR Uncertainty Parameters........................................................ 83 Event Initiators for Transients performed with ARITA......................................... 89 viii

PNNL-34230 1.0 Introduction Framatome has submitted to the U.S. Nuclear Regulatory Commission (NRC) topical report ANP-10339P Revision 0[1], entitled "ARITA-ARTEMIS/RELAP Integrated Transient Analysis Methodology" for review and approval. This topical report describes the ARITA transient analysis methodology consisting of three distinct evaluation models (EMs). The ARITA methodology is a statistical methodology for the analysis of pressurized water reactor (PWR) non-LO CA events in Chapter 15 of the SRP [2]. The methodology is capable of evaluating criteria defined for non-LOCA events in Chapter 15, including SAFDLs (DNB, FCM, and/or TCS), and selected non-SAFDL criteria such as the primary and secondary system pressure.

Consistent with the SRP Chapter 4 and Chapter 15, the TCS SAFDL is only evaluated for anticipated operational occurrences (AOOs). This methodology is not applied for evaluation of the Control Rod Ejection event.

In this methodology, S-RELAP5 is used for the system thermal-hydraulic analysis, ARTEMIS is used for core analysis, and GALILEO is used for thermal-mechanical analysis. The three different EMs allow running S-RELAP5 and ARTEMIS, either independently or in a coupled fashion, depending on the event being considered. In most events where the figure of merit is a SAFDL, S-RELAP5 and ARTEMIS are run in a coupled fashion. ARTEMIS standalone is used to verify SAFDLs for static events. S-RELAP5 is run as a standalone code for events where the figure of merit is a non-SAFDL criterion.

The ARITA methodology uses a statistical approach to ensure coverage of the potential conditions for each event. A conservative population of the potential event outcome domain (results) is defined, and a Monte Carlo approach is used to sam le that domain. The statistical approach involves simulation of model parameters, Cb)<4>

geometric uncertainties using the S-RELAP5, ARTEMIS, and co es.

non-parametric approach is used to estimate the value of the figure of merit (results) for each criterion at 95% probability with 95% confidence on an individual and a group basis.

Pacific Northwest National Laboratory (PNNL) has acted as a consultant to the NRC in this review. Two rounds of requests for information were produced by NRC and PNNL [4,5) and were transmitted to Framatome based on PNNL's review of the subject topical report.

Framatome provided a response to these round RAl's.[6,7,8,9, 10) These RAls are referred to as, RAI-##, when referenced in this report. These RAls and responses are discussed at appropriate points throughout this report.

The separate codes, S-RELAP5, ARTEMIS, and GALILEO have already been reviewed and approved. As such, this review focuses less on the acceptability of these codes except as this methodology may apply these codes in areas they have not previously been applied to. This review, as document in this TER, address the following items:

Ability of identified EM's to produce acceptably conservative values of relevant figures of merit and the acceptability of identified acceptance criteria (Section 3.2.1)

Ability of identified EM to acceptably model each transient identified in Chapter 15 of the SRP (Section 3.6.1)

The overall assessment of the codes used in the methodology (Section 3.7)

Introduction OFFICl~L USE ONLY 1

PNNL-34230 The uncertainty analysis, including:

Introduction o

The treatment of uncertainties for highly ranked parameters (Section 3.8.1) o The acceptability of the non-parametric uncertainty analysis process (Section 3.8.2) o The acceptability of each uncertainty parameter considered in the process (Section 3.8.3) o The application of non-parametric statistics for licensing analysis (Section 3.8.4)

OFFICIAL YSE ONLY 2

PNNL-34230 2.0 Regulatory Evaluation As discussed below, licensees must evaluate the consequences of various transients and accidents that could occur at its nuclear power plant. A transient, or Anticipated Operational Occurrence (AOO), is an event which is expected to occur one or more times during the life of the nuclear power plant. Examples of transients include tripping of the turbine generator, isolation of the main condenser, and loss of offsite power (LOOP). An accident, or a Postulated Accident (PA), is an event which is not expected to occur during the life of the nuclear power plant, but must still be examined because of its potential to release significant amounts of radiation to the public. Examples of accidents include a major pipe rupture on a primary loop, a major pipe rupture on the secondary loop, and ejection of a control rod assembly.

Licensees use a variety of methods to evaluate the transients and accidents that could occur at its nuclear power plant. The PNNL staff has reviewed these methods to ensure that they provide a realistic or conservative result and that they adhere to the requirements of Title 10 of the Code of Federal Regulations (10 CFR). Regulations which are applicable to transient and accident analysis methods are found in 10 CFR 50.34, "Contents of Applications; Technical Information," which provides the requirements for the Final Safety Analysis Report required for each plant including the analysis of transients and accidents; and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," which provides the requirements for a LOCA analysis; and 10 CFR 50 Appendix K, "ECCS Evaluation Models," which provides further requirements for a LOCA analysis.

Additionally, because the results of the transient and accident analysis methods are important to the safety of nuclear power plants, these methods must be maintained under a quality assurance program which meets the criteria set forth in 10 CFR 50 Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants."

To assure the quality and uniformity of NRC staff reviews, the NRC created the Standard Review Plan (SRP) [11] to guide the staff in performing their reviews. Regulatory guidance for this review is provided in Chapter 15 of the SRP, "Transient and Accident Analysis," and Section 15.0.2 of the SRP, "Review of Transient and Accident Analysis Methods". Similar guidance is also set forth for the industry in Regulatory Guide (RG) 1.203, "Transient and Accident Analysis Methods," [12].

SRP Section 15.0.2 describes six areas of PNNL staff review for transient and accident analysis methods. These areas are listed below and the section of this TER that documents PNNL's review in each area is noted:

Documentation (Section 3.3)

Evaluation Model (Section 3.5)

Accident Scenario Identification Process (Section 3.6)

Code Assessment (Section 3. 7)

Uncertainty Analysis (Section 3.8)

Regulatory Evaluation OFFIGIAL USE ONLY 3

PNNL-34230 QA Program (PNNL's review did not address the QA program).

The PNNL staff's review of ARITA was based on the SRP guidance in the first five of these six areas, evaluation of the technical merit of the submittal, and any other applicable regulations associated with the review of topical reports. PNNL also provided a review of the requested code application, which is documented in Section 3.4.

Regulatory Evaluation OFFIOl~L USE ONLY 4

PNNL-34230 3.0 Technical Evaluation This section documents the technical evaluation that was performed for ARITA.

3.1 Executive Summary The ARITA LTR has been reviewed and the following conclusions have been made.

The figures of merit (FoM) and acceptance criteria are identified for this submittal (Section 3.2.1 ).

The necessary documentation to review the L TR has been provided (Section 3.3.1)

Reactor applicability of code has been provided (Section 3.4)

Transient and accident scenarios for which the code and methodology will be used to perform safety analyses have been provided (Section 3.6.1)

The four main phenomenological evaluation models in ARITA; thermal hydraulics (S-RELAP5), neutronics (ARTIMIS Nodal and ARTIMIS THM), systems modeling (S-RELAP5), and thermal-mechanical fuel performance (GALILEO) have been reviewed and have been found to be acceptable (Sections 3.5.1 & 3.7)

The identification of target accident scenarios and phenomena identification and ranking has been reviewed and is acceptable (Section 3.6.2)

The ARITA code assessment relative to applicable data has been reviewed and found to be acceptable (Section 3.7)

The ARITA uncertainty analysis, which includes treatment of uncertainties for highly ranked phenomena, the non-parametric uncertainty analysis process, the uncertainty parameters for the non-parametric uncertainty analysis, and the application of the non-parametric statistics for licensing analyses, have been reviewed and have been found to be acceptable within the scope of application that was requested (Section 3.8)

Section 4.0 reiterates these conclusions and states applicability and limitations of the reviewed codes and methods. Section 4.0 also provides recommended limitations and conditions that should be applied to the approval of ARITA.

3.2 Background

The ARITA methodology is part of a larger suite of codes and methods that address the analyses required to demonstrate compliance with NRG requirements and regulations for normal operation, AOOs, and PAs. Specifically, ARITA demonstrates compliance with NRG requirements for AOOs and some postulated accidents. This section will provide a description of the phenomena that ARITA models as well as background history on each code used within this methodology.

3.2.1 Description of Physical Phenomena ARITA will be used to calculate a number of figures of merit for both SAFDLs and Non-SAFDLs during specific transient events from SRP 15. These events are described in greater detail in Section 3.6. The figures of merit that ARITA will be used to calculate are shown in Table 1. The original ARITA submittal [1] was not explicitly clear on the acceptance criteria for each figure of Technical Evaluation Ol=l=ICI AL USE O~ILY 5

PNNL-34230 merit, but instead referred back to other approved topical reports where these acceptance criteria were given for other codes and methods.

PNNL recommends that acceptance criteria be associated and approved separately for each code and method so that acceptance criteria and conservatisms within these criteria are appropriate for that code and method and any conservatisms within the code and method. NRC requested in RAl-1 that Framatome explicitly state the acceptance criteria for SAFDLs and non -

SAFDLs that will be used with ARITA so the staff can judge if these acceptance criteria will provide assurance that SRP limits will not be exceeded when analyses are performed using the ARITA methodology. Framatome responded by providing tables of acceptance criteria for each figure of merit. These are shown in Table 1. This RAI also removed "Loss of Natural Circulation" from the figures of merit that had been shown in the original L TR.

PNNL reviewed these acceptance criteria and found them to be acceptable. The DNB limit is similar to that applied in other thermal hydraulic methods. The FCM and TCS limit are similar to those applied in other thermal mechanical codes such as GALILEO and COPERNIC. The RCS and Secondary pressure limits are acceptable from a structural point of view. The SGO, PZRO, and LSCM all use a 95/95 lower bound and are acceptable.

It is noted that the TCS is not analyzed for postulated accidents and only for AOO. NRC requested in RAl-20 that Framatome provide justification for not considering the TCS limit to determine fuel failure for radiological consequences in postulated accidents. Framatome did not provided a particularly satisfactory response to this RAI, but SRP-4.2 [3] lists cladding strain as "fuel system damage" to be evaluated for normal operation and AOO's. The possibility of exceeding any of these damage criteria does not necessarily indicate fuel cladding failure.

These limits are listed later in SRP-4.2 under "Fuel Rod Failure". Historically, analysis of transient cladding strain has not been used for postulated accidents to indicate fuel rod failure, and additionally, cladding is not expected to realistically fail at the limits of TCS shown in Table

1.

Table 1. Figures of merit and acceptance criteria far ARITA Figure of Merit SAFDLs Departure from Nucleate Boiling (DNB)

Fuel Centerline Melt (FCM)

Transient Cladding Strain (TCS)

Non SAFDLs Technical Evaluation Acceptance Criteria Minimum DNBR > 95/95 DNBR limit

• Shall not exceed for AOO

• Assume fuel failure for PA P. 4-34 Gives GALILEO melting temperature for U02 and U02-Gd20 3

• Shall not exceed for AOO

• Assume fuel failure for PA The strain criterion is defined as a transient-induced, uniform tangential deformation, elastic plus inelastic; steady-state creepdown and irradiation growth are excluded M5: 1 % up to burnup of 62 GWd/MTU Zircaloy-4: 1 % P>

><-t~),.,..--.....,,.,.___,,.,....,......,,.........,,..=,-=-=-------'

Other approved limits will roll into ARITA

• Shall not exceed for AOO

• Not analyzed for PA OFFICIAL YSE ONLY 6

PNNL-34230 Peak RCS pressure Peak Secondary Pressure Based on ASME BPVC Section Ill

• 110% of design pressure for AOO

• 120% of design pressure for PA Based on ASME BPVC Section Ill

• 110% of design pressure for AOO

• 120% of design pressure for PA Steam Generator Overfill (SGO)

Margin to steam generator overfill is defined as the difference between the total steam generator secondary side internal free volume and steam generator secondary side collapsed liquid volume. The acceptance criterion used to demonstrate that the steam generator does not fill is the lower bound 95/95 steam generator overfill margin must be greater than zero cubic feet.

Pressurizer Overfill (PZRO)

Margin to pressurizer overfill is defined as the difference between the total pressurizer internal free volume to the height of the safety relief valve inlet piping penetrations and pressurizer collapsed liquid volume. The acceptance criterion used to demonstrate that the pressurizer does not fill is the lower bound 95/95 pressurizer overfill margin must be greater than zero cubic feet.

Loss of Subcooled Margin (LSCM)

Subcooled margin is defined as the difference between local saturation and fluid temperatures along the active RCS flow 3.2.2 Relevant History path. The acceptance criterion used to demonstrate that subcooled margin is maintained is the lower bound 95/95 subcooled margin at the limiting location along the active flow path must be greater than zero degrees.

The ARITA methodology uses three previously approved codes to simulate various transients and postulated accidents. These codes are: GALILEO, S-RELAP5, and ARTEMIS. A brief overview of each of these codes is given below.

GALILEO GALILEO calculates the fuel thermal/mechanical response of a single fuel rod under normal operation and AOO. It also calculates rod internal pressure, cladding oxide thickness, and cladding hydrogen content. The main inputs of GALILEO are fuel rod component geometry, rod power history, axial power profile, and coolant conditions. The main outputs of GALILEO are through life predictions of temperature in every fuel rod component and stress and strain in the fuel and cladding. Additional outputs are rod internal pressure and gas composition, cladding oxide thickness and hydrogen content.

GALILEO has been approved (b)(4) roved for PWR a lications for UO2 and UO2-Gd2O3 fuel. It is (b)(4) e approva o met o o ogy or pe orm,ng ca cu at,ons o fuel centerline melt or cladding transient strain, the code has been assessed to perform adequate predictions of fuel centerline temperature and cladding hoop strain, and appropriate model and manufacturing uncertainties have been identified that will result in a 95/95 upper bound prediction of both fuel centerline temperature Technical Evaluation OFFICIAL YSE ONLY 7

PNNL-34230 and cladding hoop strain. Figure 1 and Figure 2 show the measured and predicted cladding hoop strain and centerline temperature comparisons respectively. These figures also show the upper and lower bound predictions. ARITA will use the same uncertainty parameters in its prediction of TCS and FCM.

rb)(4)

Figure 1. GALILEO nominal, 95/95 upper bound, and 95/95 lower bound ramp test hoop strain predictions for MS and Zircaloy-4 assessment database Technical Evaluation OFFICIAL YSE ONLY 8

PNNL-34230 (6)(4)

Figure 2. GALILEO nominal, 95/95 upper bound, and 95/95 lower bound centerline temperature predictions for assessment database S-RELAPS S-RELAPS is a thermal hydraulic system code used for LOCA and Non-LOCA analysis. S-RELAPS code solves a two-fluid, six equation model plus one continuity equation of non-condensable gas, and a boron tracking equation for flow of a two-phase steam-water mixture, which can contain a non-condensable in the vapor phase and a soluble in the liquid phase. S-RELAPS also solves for reactor kinetics and simulates control and trip systems. The main inputs of S-RELAPS are PCS and SEC geometries and thermal properties, core neutronic properties, control system logic and setpoints, and core power distribution and the neutronic solution. The main outputs of S-RELAPS are water pressure and temperature throughout the primary and secondary coolant systems. For non-static events where the FoM is a SAFDL criterion, S-RELAPS and ARTEMIS are run in a coupled fashion. The state parameters from S-RELAPS provided to ARTEMIS include: core outlet pressure, boron concentration, inlet flow, control rod positions, and inlet temperature distribution for time. ARTEMIS provides S-RELAPS the core power distribution and neutronic solution. Where the FoM is non-SAFDL criterion S-RELAP5 is run as a standalone code.

S-RELAPS has previously been approved for LOCA applications but not explicitly for non-LOCA events such as those that are addressed with ARITA. NRC and PNNL asked in RAl-39 and RAl-40 to describe how the model options would be selected and to justify the application of S-RELAPS to these events. Framatome provided clarification and adequate justification such that PNNL concludes that the use of S-RELAPS is acceptable for the events that ARITA is requesting approval for.

Technical Evaluation OFFICIAL YSE ONLY 9

PNNL-34230 ARTEMIS ARTEMIS is a steady-state and transient 30 reactor core simulator with pin-power reconstruction and microscopic depletion. Previously, ARTEMIS was approved for use in PWR's as part of the ARCADIA reactor analysis system. Within ARITA, two ARTEMIS calculations are included, a core nodal model and a detailed core model.

ARTEMIS nodal calculation The ARTEMIS nodal calculation is a multi-group 30 core nodal simulator for steady state and transient applications that generates 30 power and burnup distributions as a function of core conditions (i.e., calculates the core response). This code contains a coupled neutronic thermal hydraulic and fuel rod module. The main inputs of the ARTEMIS nodal calculation are core inlet flow temperature, core inlet flow, core outlet pressure and optional inputs of boron concentration, control rod positions, and SCRAM signal. The main outputs of the ARTEMIS nodal calculation are nodal power produced, nodal fuel surface power, nodal moderator temperature, excore signals, and direct moderator heating fraction.

ARTEMIS detailed calculation The ARTEMIS detailed calculation is a full core ICb)(4)

I model that computes the detailed solution for evaluations of DNB using the ARTEMIS Thermal Hydraulic Module and Fuel Rod Module. The main inputs of the ARTEMIS detailed calculation are pin powers from the dehomogenization module in ARTEMIS nodal calculation, coolant pressure, inlet temperature and inlet flow. The main output of the ARTEMIS detailed calculation is the fuel rod powers at each transient time step. Coupled calculations use S-RELAP5 to obtain the system thermal-hydraulic response and transfer core boundary conditions to ARTEMIS.

ARTEMIS returns core power distribution and neutronic solutions to S-RELAP5. ARTEMIS is run as a standalone code to verify SAFDLs for static events.

3.3 Documentation 3.3.1 Necessary Documentation Framatome provided the topical report describing the ARTEMIS/RELAP Integrated Transient Analysis (ARITA) code and to define a methodology to analyze the non-Loss-of-Coolant Accident (non-LOCA) events defined in Chapter 15 of the Standard Review Plan (SRP) using a statistical approach (with the exception of the Control Rod Ejection event). This document is ANP-10339P Revision 0.

ARITA relies on a number of other topical report of previously approved codes and methods.

The following is a list of other documents that were necessary to conduct the review of the ARITA code and methodology.

ANP-10297P-A_Rev0 _ Supplement1 P-A: The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results ANP-10297P _Rev0: The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results ANP-10338P-A_Rev0: AREA' - ARCADIA Rod Ejection Accident Technical Evaluation OFFIOl~L liGE OHL¥ 10

PNNL-34230 ANP-10341P-A_Rev0: The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations EMF-10311 P-A_Rev1: COBRA-FLX: A Core Thermal-Hydraulic Analysis Code EMF-2103P-A R3: Realistic Large Break LOCA Methodology for Pressurized Water Reactors EMF-2328(P): PWR Small Break LOCA Evaluation Model, S-RELAP5 Based EMF-2328PA_Rev0_Supplement1 PA: PWR Small Break LOCA Evaluation Model, S-RELAPS Based ANP-10323PA_Revision 1 GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors EMF-1961 (P)(A) Revision 0 Statistical SetpoinUTransient Methodology for Combustion Engineering Type Reactors EMF-92-081 (P)(A) Revision 1 Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors 3.4 Code Application Framatome has requested approval of the ARITA methodology presented in the LTR for application to US licensed PWR reactors including Westinghouse 2, 3, and 4-loop plants and CE plants.

ARITA will be applied to all the PWR SRP 15 events except large break LOCA, small break LOCA, and Rod Ejection Accident. The original L TR stated that ARITA would be used for station blackout, but RAl-34 clarified that ARITA would not be used for this event. Table 2 in Section 3.6.1 lists each event ARITA will be applied to as well as the specific EM that will be used to model each event.

Reactor conditions that Framatome requested application of ARITA for include:

RCS flow rate: Cb)C4)

ICbX4)

IAn exception to this is for CHF analyses, which are limited to the flow as defined by the CHF correlationlCbX4)

ICb)(4) las defined in t~he_ A_R __ T __ E_M_l....,..S_m_o_d_e_l _ ----------~

RCS fluid temperature: !CbX4) t

)(4)

(4)

RCS system pressure: l(b)(4)

~

l(b)(4)

I For DNB calculations, the pressure is limited by the pressure range of t e CHF correlation.

Secondary System Pressure: )(4) l(bX4)

I Technical Evaluation OFFICIAL YSE ONLY 11

PNNL-34230 Fuel Rod Power: ICb)C4)

F )(4)

(4)

(6)(4) on the Reactor Protection System: The RPS must be capable of being modeled using S-RELAPS control variables.

Other requested applications are for the analysis of a mixed core (either several fuel assemblies of different designs or fuel assemblies from multiple vendors), and analysis of reconstituted fuel assemblies (assemblies where a failed fuel rod is removed and replaced with inert fuel assemblies).

3.5 Evaluation models The ARITA methodology consists of three evaluation models (EMs) that contain more or less modeling fidelity as required by the event being considered. The treatment of events within the ARITA methodology fall into the following three categories:

1. Evaluation of AOOs and PAs against the SAFDLs, where the RCS and main steam system (MSS) conditions are important to event consequences.

2. Evaluation of AOOs and PAs against criteria that do not include SAFDLs, including when the SAFDL consequences of the event are dispositioned as being bounded by another event, such that only non-SAFDL criteria are explicitly analyzed.

3. Evaluation of AOOs and PAs against SAFDLs, where the RCS and MSS conditions are not important to event consequences.

3.5.1 Individual Codes The codes used within the ARITA methodology have been described in Section 3.2.2 and are:

ARTEMIS - Multi-group 3D core nodal simulator for steady state and transient applications. At the nodal level a coupled neutronic, Thermal-Hydraulic Module (THM) and Fuel Rod Module (FRM) solution is performed at each specified core condition. The thermal-hydraulic solution is an open-channel calculation performed by the ARTEMIS THM, COBRA-FLX.

S-RELAPS - RELAPS-based thermal-hydraulic system code used for performing LOCA and non-LOCA analyses.

GALILEO - Best-estimate fuel rod performance code which models the thermal and mechanical behavior of individual PWR fuel rods during normal operation and transient conditions.

Technical Evaluation 12

PNNL-34230 3.5.2 EM Implementation NRC and PNNL requested in RAl-2 and RAl-9 for Framatome to explicitly list which of the three EMs will be used for each of the target scenarios. Framatome provided this information, and it is reflected in Table 2. Each EM is briefly described in the following sections. Section 3.5.3 will discuss PNNL's review of the acceptability of the use of each EM for the stated target scenarios.

3.5.2.1 EM1: Coupled System TH and Neutronics EM1 is the most detailed EM. In the coupled EM, ARTEMIS and S-RELAPS are run as a coupled system to solve time-de endent multi-h sics roblems. GALILEO also la s a role with this EM (b)C4)

(b)C4) igure 3 shows a graphical illustration of how these codes are coupled.

EM1 is applied to the target scenarios where it is necessary to evaluate AOOs and PAs against the SAFDLs, where the RCS and main steam system (MSS) conditions are important to event consequences.

(bX4)

Figure 3. Data flow of system and core coupled calculations (Figure 4-1 of L TR)

Technical Evaluation OFFICIAL YSE ONLY 13

PNNL-34230 3.5.2.2 EM2: 00 System TH EM2 is essentially S-RELAP5 run in standalone mode with inputs from other models. In EM2, S-RELAP5 is not coupled to ARTEMIS. Rather reactor power is determined using a point kinetics model based on reactivity coefficients from ARTEMIS and decay heat is determined using the f1>X4)

IThe fuel rod heat structure in S-RELAP5 is changed from a set of heat structures to a simplified model, intended to produce a reasonable rediction of heat conduction during rapid fuel rod heat-up and cooling transients Cb)<4)

(b)(4)

EM2 is applied to the target scenarios where it is necessary to evaluate AOOs and PAs against criteria that do not include SAFDLs, including when the SAFDL consequences of the event are dispositioned as being bounded by another event, such that only non-SAFDL criteria are explicitly analyzed. Examples of this include over-pressure, pressurizer overfill, and steam generator overfill.

3.5.2.3 EM3: Static Core EM3 is essentially ARTEMIS run in standalone mode with system parameters specified as boundary conditions.

EM3 is applied to the target scenarios where it is necessary to evaluate AOOs and PAs against SAFDLs, where the RCS and MSS conditions are not important to event consequences. This EM is used to analyze events that do not require a system thermal-hydraulic solution. These events include static misaligned control rod events and misloaded fuel assembly.

3.5.3 Assessment of Evaluation Models PNNL agrees that for the events where evaluation of AOOs and PAs is performed against the SAFDLs EM1 is the most appropriate. GALILEO calculates both TCS and FCM and has been previously approved to be suitable to perform these calculations. Likewise, the ARTEMIS detailed calculation calculates the DNB ratio using previously approved assembly specific critical heat flux (CHF) correlations. The S-RELAP5 and the ARTEMIS nodal calculations provide realistic boundary conditions for the two previous codes to improve on previous methodologies that relied on limiting and conservative transient power and coolant conditions that would cover a larger range of possible AOO and PA events.

PNNL agrees for those events that are only related to the cooling system performance, EM2 is acceptable and for those events that are only related to the core physics performance, EM3 is acceptable.

For events where a SAFDL and a non-SAFDL are simultaneous challenged, Framatome clarified in RAl-14 that )(4)

PNNL has reviewed the target scenarios in Table 2 and determined that the stated EM is appropriate to model each scenario and will provide realistic values for output parameters of interest.

Technical Evaluation OFFIGIAL YSE ONLY 14

PNNL-34230 Because EM2 and EM3 rely primarily on running previously approved codes in standalone mode, the only assessment that must be performed is that the limits of these code approvals coincide with the application for these target scenarios, that existing assessment cases have been run in a similar manner to ARITA, and that inputs from other codes are taken appropriately. This assessment is primarily made in Sections 3.7.2 and 3.7.3 for S-RELAP5 and ARTEMIS, respectively.

For EM 1 PNNL performed a more detailed review on how the separate codes are coupled which is described below.

PNNL and NRC requested additional detail on the coupling procedure that is used within ARITA to combine the various models (RAl-25 and RAl-26). Specifically, more detail was provided on the time step management in RAl-25, and of importance was that previously the time step manager in ARTEMIS had not been approved for SRP Chapter 15 events such as the ARITA tar et scenarios. In the response to RAl-25, Framatome provided CbX4)

Cb>(4) figures of merit. More detail on the thermal hydraulic coupling of S-RELAP5 and IS was thermal-hydraulic coupling was provided in RAl-26. PNNL reviewed this information and concluded that the thermal-hydraulic coupling between S-RELAP5 and ARTEMIS is acceptable.

PNNL and NRC asked in RAl-30 to discuss the nodalization that will be used within ARITA for Westinghouse and CE plants. Framatome committed to applying the sample nodalizations from the L TR and if any deviations were used then these could be described and justified in the specific plant submittal. Regarding Westinghouse 2-loop and 3-loop plants, the sample 4-loop plant will be modified to reduce the number of primary coolant loops. Other specific nodalization changes will be justified in the plant submittal.

PNNL and NRC asked Framatome in RAl-35 to justify the use of different versions of the ICbX4) f0X4)

IEM2 (OD System TH) uses the ICb)(4)

~ ith EM1 (Coupled EM) uses t~h-e-~

._I ____

_. Framatome explained that the EM useslCb>(4)

!version because this is the version in the previously approved S-RELAP5 topical report and was the version used in the assessment of this code. Framatome provided demonstration that EM2 OD EM im lements the same modeling approach and assumptions for )(4)

~..;..._ __

__.......-=,---,-----:-----:--:-----:---:-----:-:---:--..,.....,...-...,..,.......

Cb)(4) conservative! hi h rediction (b)(4)

Framatome also provided detail about how the CbX4) is implemented in EM1 (Coupled EM). PNNL reviewed this approach and concludes that this will also result in a conservatively high estimate!CbX4) j PNNL and NRC asked Framatome in RAl-22 and RAl-24 to clarify how actual equipment behavior will be modeled including safety and non-safety grade equipment. Framatome clarified that timing of events will be taken from the plant licensing basis and !Cb)~4)

I ICb)(4)

~ ill not be credited as a benefit for the calculation of figure of merit. Additionally, any deviation from uncertainties stated in Section 3.8.3 will be justified and approved during the plant LAR.

Finally, PNNL and NRC asked Framatome to provide additional technical details regarding the modeling of turbulent mixing (RAl-27, RAl-28, RAl-29) and the modeling the steam generator and pressurizer (RAl-31, RAl-32, RAl-33, RAl-42, RAl-44, RAl-45, and RAl-46). Framatome provided adequate responses to these requests, and as a result of these questions several additional uncertainty parameters were added to those in Section 3.8.3. The applicability of Technical Evaluation OFFICIAL Uii O~ILY 15

PNNL-34230 these additional uncertainties were reviewed in conjunction with those uncertainties listed in the original L TR and are discussed in Section 3.8.3.

Overall, PNNL's technical review of the three EMs concludes that the target scenarios identified in Table 2 can be adequately modeled using the EM identified for each scenario. Additionally, PNNL concludes that each EM is acceptable to realistically model the target scenario that it will be applied to.

3.6 Accident Scenario Identification Process This section discusses the events that ARITA will be used to assess as well as the PIRT that was performed to identify the important phenomena that should be addressed by ARITA.

3.6.1 Target Scenarios Table 2 lists the transients from SRP 15 [2] and the EM version that Framatome will use for each event. This information was not clear in the original submittal, so NRC asked in RAl-2 for Framatome to provide a table of every Chapter 15 event and identify which EM variant would be used to analyze each on. Framatome provided this information in RAl-9. Table 2 shows Framatome's response to what scenarios will be addressed by this methodology and what EM variant will be used for each event.

Table 2. Transients performed with ARITA and EM used for each SRP#

Event Type of Event ARITA EM Used 15.1.1 Decrease in Feedwater AOO (b)(4)

Temperature 15.1.2 Increase in Feedwater Flow AOO 15.1.3 Increase in Steam Flow AOO 15.1.4 Inadvertent Opening of a AOO Steam Generator Relief or Safety Valve 15.1.5 Main Steam Line Break PA 15.2.1 Loss of External Load AOO 15.2.2 Turbine Trip AOO 15.2.3 Loss of Condenser AOO Vacuum 15.2.4 Closure of Main Steam AOO Isolation Valve 15.2.5 Steam Pressure Regulator AOO Failure (BWR) 15.2.6 Loss of Nonemergency AC AOO Power to the Station Auxiliaries 15.2.7 Loss of Normal Feedwater AOO Flow 15.2.8 Feedwater System Pipe PA Breaks Inside and Outside Containment Technical Evaluation OFFIGIAL YSE QNbY 16

PNNL-34230 15.3.1 Loss of Forced Reactor AOO

)(4)

Coolant Flow Including Trip of Pump Motor 15.3.2 Flow Controller AOO Malfunctions 15.3.3 Reactor Coolant Pump PA Rotor Seizure 15.3.4 Reactor Coolant Pump PA Shaft Break 15.4.1 Uncontrolled Control Rod AOO Assembly Withdrawal from a Subcritical or Low Power Startup Condition 15.4.2 Uncontrolled Control Rod AOO Assembly Withdrawal at Power 15.4.3 Control Rod Misoperation AOO/PA (Single Rod W/D) 15.4.3 Control Rod Misoperation AOO/PA (Dropped Rod/Bank) 15.4.3 Control Rod Misoperation AOO/PA (Misaligned Rod}

15.4.4 Startup of an Inactive Loop AOO or Recirculation Loop at an Incorrect Temperature 15.4.5 Flow Controller Malfunction AOO Causing an Increase in BWR Core Flow Rate 15.4.6 Inadvertent Decrease in AOO Boron Concentration in the Reactor Coolant System (PWR) 15.4.7 Inadvertent Loading and AOO Operation of a Fuel Assembly in an Improper Position 15.4.8 Spectrum of Rod Ejection PA Accidents (PWR) 15.4.9 Spectrum of Rod Drop PA Accidents (BWR) 15.5.1 Inadvertent Operation of AOO ECCS 15.5.2 Chemical and Volume AOO Control System Malfunction that Increases Reactor Coolant Inventory 15.6.1 Inadvertent Opening of a AOO PWR Pressurizer Pressure Relief Valve Technical Evaluation 17 OFFIGIAL USE ONLY

PNNL-34230 15.6.2 Radiological PA (b)(4)

Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment 15.6.3 Radiological PA Consequences of Steam Generator Tube Failure 15.6.4 Radiological PA Consequences of Main Steam Line Failure Outside Containment (BWR) 15.6.5 Loss-of-Coolant Accidents PA Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary 3.6.2 Phenomena Identification and Ranking The ARITA methodology uses a Phenomena Identification and Ranking Table (PIRT) process to establish important parameters and conditions for event analyses. The phenomena encountered during events ARITA will be used to analyze (Table 2) are identified and ranked relative to the importance of the phenomena with respect to predicting margins to event acceptance criteria (Table 1 ). The process is performed within the range of applicability as defined in Section 3.4.

Phenomena are categorized by 1) plant systems, and 2) the core. The two categories are subdivided as shown below.

Plant Systems o

RCS o

Secondary coolant system Core o

Thermal-hydraulics o

Neutronics o

Fuel rod thermal and mechanical performance NRC did not feel that the L TR contained adequate information to review and approve this PIRT and requested that Framatome provide supplemental information pertaining to the process by which the six event classification phenomena identification and ranking tables (PIRTs) were developed, including all of the phenomena considered for each of the PIRTs, their associated rankings, and the justification for the rankings.[13]

PNNL reviewed this supplemental information and found it to be adequate. All of the high and medium-ranked phenomena are addressed through the ARITA methodology (RAl-23 and RAl-

9) and the approach used for each is shown in Section 3.8.1 along with a summary PNNL's review of each a roach. The PIRTICbX4)

I Cb)(4) uncertainty parameters (Section 3.8.3) that have been selected to a

ress t e 19 y ran e p enomena (Section 3.8.1) identified are used the application of ARITA. Nevertheless, Framatome has stated in the response to RAl-37 that no further changes to the PIRT will be made without a revision to the ARITA L TR. Because of this, the uncertainty Technical Evaluation 18

PNNL-34230 parameters will also not change except for those that are approved to be determined on a plant-specific basis as discussed in Section 3.8.3.

PNNL and NRC had concerns that the ARITA LTR stated that parameters associated with henomena and rocesses havin a medium or hi h rankin are Cb)(4)

(RAl-43, RAl-48, and

...,.....,r-r-...,.....,-_,...,.-::-a-=-se-::-::r--=-o-::-n-=:-,1c:::-s---=c:--::o--=-n-::-ce-=-r::::n,-,.....-::-ra=--:m=ar-::oc::m-=-e:--::-:re:-:-v,.,.,,s:-:e:-::ri=-=-e--::p:--::r-=-oc-=-e::-:s:-::sC""'l'":::'o-::-r -::-co-=-nc:--:!sidering highly ranked phenomena and add several uncertainty parameters. These are included in the tables in Section 3.8.1 and 3.8.3).

PNNL and NRC asked Framatome to clarify in RAl-36 if the uncertainty treatment will apply on an event-specific basis (e.g., SRP Cha ter 15.1.1 or on an event cate o level e.. SRP Cha ter 15.1. Framatome clarified 1>)(-t)

IS.

Overall, PNNL concludes that the PIRT that is described in the L TR is acceptable for the development of the ARITA methodology but notes that the methodology as described in the L TR and RAI responses should not be altered without further NRC review and approval.

3. 7 Code Integral Assessment This section describes the assessment of the ARITA code. This section will assess the applicability of ARITA to model the target scenarios identified in Section 3.6.1. Although all the individual codes that are used within ARITA are approved, each code has limits of its approval, so this section will also ensure that the requested use of these codes within ARITA does not challenge the range of applicability of these codes.

3.7.1 Range of Code Assessment The assessment of ARITA is in two general areas. These are individual code assessment and assessment of the coupled EM.

PNNL and NRC expressed concern in RAl-39 that many of the assessments were performed for situations that are different than those that ARITA will be applied to (e.g., LOCA). In response to this RAI, Framatome provided discussion of each individual code assessment to clarify the applicability of the assessment for the target scenarios requested. This will be discussed in more detail in the following sections on individual code assessment.

The total assessment of the code, as documented in the L TR and various Responses to RAls, adequately demonstrates that ARITA is suitable to model all US licensed Westinghouse and CE PWR reactors and AOO and PA conditions listed in Table 2. The following sections briefly summarize each the assessment of the individual codes and the coupled EM.

Categories of tests used in the assessment of the phenomena for each of the codes are listed below:

Separate Effects Tests (SET) - used to develop and assess empirical correlations and other closure models.

Component Effects Tests (CET) - used to assess individual components.

Technical Evaluation QFFICIAb YSE ONLY 19

PNNL-34230 Integral Effects Tests (IET) - used to assess system responses and global code capability.

Plant Transient (or Steady State) Data (PTO - test at operating plants; used to demonstrate code performance based on a full-scale plant.

Foundation Methodology Assessments (FMA) - used to demonstrate the fundamental approach and solution scheme adopted in each code.

Conservative Methodology Assumptions (CMA) - conservative data based on external testing and/or engineering analyses as applicable.

3. 7.2 S-RELAP5 Assessment The main use of S-RELAP5 within ARITA is to provide coolant temperature, pressure, and mass flow within the primary and secondary coolant system during the events of interest. In order to make these predictions, S-RELAP5 needs to have adequate models for the following phenomena.

Primary Coolant System Reactor Vessel Mixing Pressurizer Phenomena Flashing of Coolant in the Reactor Vessel Upper Head Choking/Critical Flow Secondary Coolant System Primary/Secondary Heat Transfer Flashing of MFW Downstream of Isolation Valve Choking/Critical Flow Flashing of Tube Rupture Effluent Boiler Region Mixture Level With regard to the applicability of the original S-RELAP5 assessment to the ARITA target scenarios (RAl-39), Framatome provided detailed discussion of the applicability of the S-RELAP5 assessment in RAl-40 and a detailed response of the differences in the nodalization and modeling practices in RAl-90. In this response, Framatome acknowledged that the validation of S-RELAP5 for non-LOCA SETs and IETs is not as extensive as for LOCA, but that some of the LOCA benchmarks contain elements that can be used to validate non-LOCA processes. Additionally, Framatome provided rationale for including specific LOCA benchmarks tests, and discussion regarding modeling options and nodalization that has been found to be acceptable. PNNL has reviewed the provided S-RELAP5 benchmarks and the discussion on applicability to non-LOCA events and finds that S-RELAP5 is reasonably well benchmarked.

NRC asked a question about discrepancy between measured and predicted values for the f1i)(4) l(RAl-89). Framatome provided clarification regarding the purpose of that particular benchmark and provided updated figures to better show the com arison. Similar! in RAl-91 NRC asked about a discre anc betweenlCb)C4)

I CbX4)

Again, Framatome prov, e can IcatIon as to t e purpose o t PNNL has reviewed the assessment provided for S-RELAP5 and concludes that S-RELAP5 nodal will provide reasonable values for the phenomena listed above during the target scenarios.

Technical Evaluation OFFICIAL YSE ONLY 20

PNNL-34230

3. 7.3 ARTEMIS Nodal Assessment The main use of the ARTEMIS nodal model within ARITA is to provide fuel power, power distribution, and burnup during various events. In order to make these predictions, ARTEMIS needs to have adequate models for the following phenomena.

Global core neutronics Delayed Neutron Fraction Mean Generation Time Doppler Reactivity Moderator Reactivity Scram Worth Shutdown Margin Rod Worths Boron Concentration Boron Worth Power Asymmetries Excore Flux Decay Heat Xenon Reactivity Multi-Dimensional Kinetics Local power distributions Allowed Initial Power Distribution Event-Specific Power Distributions Energy Deposition Factor Rod and Assembly Bow Penalties With regard to the applicability of the original ARTEMIS assessment to the ARITA target scenarios (RAl-39), Framatome confirmed that the ARTEMIS models for the benchmarks presented in the L TR were developed in a manner that is consistent with the recommended ARITA modeling scheme and that the methods used to determine the ARTEMIS axial and radial mesh consistent with those used in the original validation of ARTEMIS.

PNNL has reviewed the assessment provided for ARTEMIS and concludes that ARTEMIS nodal will provide reasonable values for the phenomena listed above during the target scenarios.

3. 7.4 ARTEMIS THM Assessment The main addition to ARTEMIS nodal model within the ARTEMIS THM model is the addition of COBRA-FLX. Within ARTEMIS THM, COBRA-FLX is used to evaluate the DNBR response to the target scenarios. In order to make these predictions, ARTEMIS THM needs to have adequate models within COBRA-FLX for the following phenomena.

Core Thermal-Hydraulics Non-uniform Core Inlet Flow Distribution Non-uniform Flow Resistance in the Core Fuel Assembly Bow Fuel Rod Bow Crossflow Resistance Diversion Crossflow Grid-Induced Flow Mixing Technical Evaluation OFFICIAL YSE ONLY 21

PNNL-34230 Turbulent Energy Exchange Two-Phase Flow/ Voiding Spacer Grid Flow Resistance Bare Rod Friction DNB / CHF Correlation DNB Propagation Additional Phenomena and Processes Manufacturing Tolerances (Local Power Distribution and DNB Effects)

Cladding Surface Heat Transfer Radial and Axial Nodalization Impact of TH Parameters on Nodal Model Results With regard to the applicability of the original COBRA-FLX assessment to the ARITA target scenarios within ARTEMIS THM (RAl-39), Framatome confirmed that the COBRA-FLX validations were updated to be consistent with the ARITA model requirements and stated that following this update the maximum difference in predicted axial mass velocity and enthalpy was l(b)(4)

!between the updated cases and the original cases. PNNL concludes that this difference is negligible.

With regard to the axial nodalization, PNNL and NRC asked in RAl-52 for more information on the sensitivity study that was performed. Adequate detail had been provided for the radial nodalization study but not for the axial nodalization study. Framatome provided this information, and it was found to be acceptable.

PNNL has reviewed the assessment provided for COBRA-FLX and concludes that ARTEMIS THM will provide reasonable values for the phenomena listed above during the target scenarios.

It should be noted that there may be applications where the COBRA-FLX assessed range as described in the previously approved topical report are more restrictive than ARITA globally.

This may also change case to case, based on fuel type and plant type. An example would be when applying correlations with specific assessed limits. Any approved range for ARITA or other coupled codes does not override these limits.

3.7.5 GALILEO Assessment The main use of GALI LEO within ARITA is to provide predictions of fuel temperature and predictions of cladding hoop strain during power ramps.

In order to make these predictions, GALILEO needs to have adequate models for the following phenomena.

Normal Operation Power History and Fluence History Fuel Pellet Radial Power Profile Fuel Rod Axial Power Distribution Transient Power Conditions Transient Coolant Conditions Fission Gas Release from Pellets Heat Conduction through Pellet Pellet-Cladding Gap Conductance Pellet-Cladding Gap Closure Cladding Corrosion, Crud Deposition, and Cladding Hydriding Burnup of Fuel Rod Technical Evaluation OFFIOl~L USE ONLY 22

PNNL-34230 During the review and approval of the GALILEO L TR [14] it was determined that GALIELO contained appropriate models for each of these phenomena, that GALILEO could provide a best-estimate prediction, of fuel temperature and cladding hoop strain, and that with appropriate model uncertainties described in Table 13, GALILEO could provide 95/95 UB predictions of these two outputs. See also Figure 1 and Figure 2.

PNNL has reviewed the assessment provided for GALILEO and concludes that GALILEO will provide reasonable values for the phenomena listed above during the target scenarios.

3.7.6 Coupled EM Assessment The coupled EMs utilize existing models that have previously been assessed and allow for data exchange between models during the assessment of the transients. There are no new physical models that have been added for the S-RELAP5 and ARTEMIS or for the use of ARTEMIS THM or GALILEO within ARITA. Therefore, the assessment of the coupled EM will focus on:

1. Frequency of data exchange between S-RELAP5 and ARTEMIS

2. Data exchange between the geometric models of the codes

3. Overall assessment of the coupled EM r

...,._::_: ________________________ ~I PNNL has reviewed this approach (see also Section 3.5.3) and finds it to be acceptable.

~

I l(bX4)

I Calculated parameters at the end of the steady-state portion are compared to those from the individual standalone models. Additionally, a null transient is performed to verify that the coupled model remains stable without steady state control. PNNL and NRC asked in RAl-56 for more information on this null transient. Framatome clarified that the null transient is a symmetric event that maintains a steady-state condition without the use of artificial controllers. Due to expected symmetry and stableness of this event, it is a good indicator of issues in calculational stability and geometric data exchange. If the geometric data exchange is unbalanced in the radial direction, this would manifest in an asymmetry in the coupled null transient and due to the symmetry in the event, even small asymmetries are noticeable. Therefore, if the model is able to maintain the balance and stability of the coupled null transient, there is confidence that the model is stable and the geometric data exchange is adequate. PNNL agrees that this approach is reasonable and that the geometric data exchange within ARITA is acceptable.

A number of benchmark calculations were provided to demonstrate the adequacy of the ARITA EMs. The benchmark transients include asymmetry, and the results provide confidence of adequate geometric data exchange. Additionally, limited data comparisons are provided to available data from a loss of external load test occurring at an operating plant. These benchmarks and data provide some confidence in the coupled EM performance, but the majority of the confidence in the acceptability in the coupled EM performance comes from the strong assessment of the individual models (S-RELAP5, ARTEMIS, COBRA-FLX, and GALILEO) and in the assessment of the techniques used to couple these models.

Technical Evaluation QFFICIAb blSE ONLY 23

PNNL-34230 PNNL has reviewed the assessment provided for the coupled EM and concludes that ARITA will provide reasonable values for the relevant parameters within the target scenarios based on the assessments listed above as well as the assessment of the individual models within ARITA.

3.8 Uncertainty Analysis In order to provide high confidence that certain limits will not be exceeded during identified SRP chapter 15 AOOs and PAs, the ARITA L TR describes a non-parametric uncertainty analysis that will be performed. As clarified in RAl-10 the ARITA methodology is intended to support the following statistical statement:

Fo~(b)(4)

I if this event occurred from within the licensed operating space, the limiting MDNBR (FCM, TCS, peak pressure, etc.) margin to the established limit is W(X, Y, Z,... NJ with at least 95% probability at 95% confidence.

The following sections will describe the ARITA uncertainty analysis and provide PNNL's review of this methodology. Section 3.8.1 describes how the high and medium ranked phenomena identified in the PIRT (Section 3.6.2) are addressed in the uncertainty analysis. Section 3.8.2 describes the non-parametric uncertainty analysis process that will be used to determine an upper bound output for a given event. Section 3.8.3 describes the individual parameters that will be perturbed within this uncertainty analysis. Section 3.8.4 describes how the non-parametric statistics will be applied within licensing analyses.

Through the review of ARITA, NRC and PNNL staff asked a number of RAls and participated in numerous audits with Framatome staff. As a result of these RAls and audits a number of significant changes were made by Framatome to the ARITA methodology. These changes include:

Normal distributions will not bef:bX4)

I Rather, when physically

...._ ________ __,_~

appropriate, normal distributionsl(b)(4)

I For parameters with known unner and/or low limits, ).'h)(4)

I

~~--.-----------------~

(b)(4)

I For oarametersl(b)(4)

I Kb)(4)

(b)(4) lare described in Section 3.8.3. (RAl-7)

~

For events that challen (b)(4)

(b)(4)

(RAl-8)

For a given event when certaint:t,)(4)

(b)(4)

(b)(4)

[RAl-11)

Technical Evaluation OFFIOIAL USE o~,bY I

I 24

PNNL-34230 For instances when the ARITA methodolo indicates a SAFDL will be exceeded durin the core desi n hase for a lant CbX4)

)(4)

RAl-12)

(b)(4) 3.8.1 Treatment of Uncertainties for Highly Ranked Phenomena Framatome provided a table for each highly ranked process or phenomena identified in the PIRT in the response to RAl-9. This table describes how each process or phenomena will be treated within the ARITA methodology. Many phenomena are treated through parameters that are either biased to a conservative direction or sampled within the non-parametric uncertainty analysis (Section 3.8.2). The individual parameters are reviewed individually in Section 3.8.3.

Table 3 shows the conclusions of PNNL's review of the treatment of each highly ranked phenomenon. Categories of tests used in the assessment of the phenomena for each process or phenomenon are listed below:

Separate Effects Tests (SET) - used to develop and assess empirical correlations and other closure models.

Component Effects Tests (CET) - used to assess individual components.

Integral Effects Tests (IET) - used to assess system responses and global code capability.

Plant Transient (or Steady State) Data (PTO) - test at operating plants; used to demonstrate code performance based on a full-scale plant.

Foundation Methodology Assessments (FMA) - used to demonstrate the fundamental approach and solution scheme adopted in each code.

Conservative Methodology Assumptions (CMA) - conservative data based on external testing and/or engineering analyses as applicable As discussed in Section 3.6.2, Framatome was requested to justify not explicitly modeling uncertainties in every medium and highly ranked phenomena (RAl-43). Framatome provided detailed review to identify important phenomena where uncertainties are treated in a nominal or best-estimate manner based upon the perceived accuracy of the ARITA methodology and for each of the identified phenomena, provided either a basis for neglecting the uncertainty or a description of the uncertainty treatment. As a result of this review, Framatome included some additional uncertainty parameters and made some changes to their treatment of highly ranked phenomena. The treatment of each item in Table 3 and the associated uncertainty parameters in Table 5 through Table 13 have been updated to include these additions and PNNL's technical review was performed on the as-modified phenomena treatment and uncertainty parameters.

Technical Evaluation 25

Table 3. Treatment of PIRT processes and phenomena within ARITA Process/Phenomenon PCS-1. Stored Energy (metal, fuel and coolant)

PCS-2. Nonuniform Power Distribution PCS-3. Direct Moderator Energy Deposition PCS-4. Decay Heat PCS-5. Kinetics /

Reactivity Feedback Technical Evaluation Associated Test Type Parameters FMA Cb)C4)

CMA CMA FMA FMA CMA CMA FMA Evaluation The parameters identified are sufficient to ade uatel model stored energy in the rimar coolants stem. The abilit of Cb)C4) upper o erance ImI wI to result in a limited e 5.

The l'-lnr>"!"":m=o::re::r11"""'1,~-C'.CAJ'.'l'-r::ro=-e=-:s=-n=-o=-=e::-v:-::ar.u-:--::a:re::.-,,.,.,.'1"'\\T""".c=-=r:rr1 e=-r~,a=-=-a=-n::rn=-:e:-::re::-1:ore uses a point kinetics model to simulate core power generation. Input for the S-RELAP neutronic model are taken from ARTEMIS. This is sufficient within the EM2 model.

P,X4) lis sufficient to address hie direct moderator energy deposition w1th1n the EM1 model. The EM2 model (S-RELAP) uses a plant specific value. This is sufficient within the EM2 model.

The decay heat addressed in Cb)C4) s sufficient to address the decay heat. PNNL finds the use of )(4) acceptable within EM2 and th

>C4) acceptable within EM1 The EM2 model (S-RELAP) does not evaluate SAFDL criteria and therefore uses a point kinetics model to simulate core power generation.

Input for the S-RELAP kinetics model are taken from ARTEMIS. Reactivity parameters are selected that yield the most liming non-SAFDL event consequences. This is sufficient within the EM2 model. Cb)(4)

)(4) are sufficient to address the kinetics and reactivity feedback within the EM1 model.

Qf:f:ICIJ\\L Uili 0NLY PNNL-34230 26

Process/Phenomenon Test Type PCS-6. Boron Injection CMA

/ Dilution FMA PCS-7. ECCS Operation PCS-8. Control Rod Motion CMA CMA PCS-9. Fuel Rod Heat FMA Transfer PCS-10. Cladding FMA Surface Heat Transfer Technical Evaluation Associated Parameters (6)(4)

Evaluation The parameters identified are sufficient to adequately model boron in*ection and dilution in the primary coolant system. The ability of )(4)

b)(4) result in a upper o erance 1m1 wI e 5.

The parameters identified are sufficient to adequately model ECCS operation in the primary coolant system. The ability of )(4) upper to erance ImIt wI The arameters identified in con*unction with CbX4)

b)(4) are sufficient to a equate y mo e store energy In t e primary coo ant system. The ability of (6)(--t)

(6)(4) resu in a assesse in able 5.

)(4)

)(4) are sufficient o a ress ue ro Framatome states f:b)(4)

PNNL agrees that the individual model qualification of S-RELAP5 and COBRA-FLX are acceptable for the OD EM and the coupled EM, respectively.

OFFIGl,tcl USE ONLY PNNL-34230 27

Process/Phenomenon PCS-11. Component Flow Resistance &

RCP Performance Test Type CMA FMA PCS-12. RCP & RCS/ CMA Core Flow Coast Down PCS-13. In Reactor SET Vessel Coolant Mixing PTO CMA PCS-14. lncore CMA Coolant Mixing PCS-15. Bypass of CMA Coolant Flow PCS-16. Excore IET Nuclear Instrumentation Power Measurement PCS-17. Coolant Loop CMA

/ Reactor Vessel Temperature Measurement Technical Evaluation Associated Parameters (6)(4)

Evaluation The parameters identified are sufficient to adequately model component flow resistance & RCP Performance in the rimar coolants stem. The ability of the )(4)

(6)(4) resu m a ImI e e

assessed in Table 5.

The parameter identified is sufficient to adequately model core flow coast down in the primary coolant system. The ability of the IC6X4)

I IC6)(4)

Iresult in a limited 95/95 upper tolerance 1JmIt will be assessed in I able 5.

The parameter identified is sufficient to adequately model in-reactor vessel coolant mixing in the primary coolant system. The ability of the magnitude of the (6)(4) result in a limited 95/95 upper tolerance limit will be assessed in Table 5.

(6)(4)

The parameter identified is sufficient to adequately model b ass of coolant flow in the primary coolant system. The ability of the )(4)

~

Ire~

upper tolerance hm,t will be assessed m I able 5.

The parameter identified in conjunction with )(4) o)(4) is sufficient o a equa e y mo e excore nuc ear ms rumen a,on power measurement in the primary coolant system. The ability of the IC6X4)

I f0X4)

!result in a limited 95/95 upper tolerance limit will be assessed in Table 5.

The parameters identified are sufficient to adequately model the coolant loop and reactor vessel temperature measurements in the primary coolant system. The ability of thelC6)(4)

I 1Cb)(4)

I result in a limited 95/95 upper tolerance limit will be assessed in Table 5.

OFFIGIAL ~SE ONLY PNNL-34230 28

Process/Phenomenon PCS-18. Coolant Flow Rate Measurement Test Type CMA PCS-19. Pressurizer &

CMA RCS Pressure Measurement PCS-20. Pressurizer Level Measurement PCS-21. Pressurizer Phenomena PCS-22. Surge Line Flow Resistance /

Pressure Drop PCS-23. Natural Circulation Flow PCS-24. Makeup

&Letdown Flow Technical Evaluation CMA IET IET IET CMA Associated Parameters (6)(4)

Evaluation The parameters identified are sufficient to adequately model coolant flow rate measurement in the rima coolant s stem. The abilit of the rb)(4) resu erance ImI wI The parameters identified are sufficient to adequately the pressurizer and RCS pressure measurements in the primary coolant system. The ability of the Cb)(4) fbX4Jlresu t rn a ImIte 5 upper to erance ImIt wI The parameters identified are sufficient to adequately model pressurizer level measurement in the rima coolants stem. The abili of the

)(4) result in a limited 95/95 upper tolerance limit will be assessed in Table 5.

PNNL agrees that S-RELAP5 correct! calculates the ressurizer henomena. The abili of the

)(-t)

(b)(4) res~u~ 1-n-a~-~----~~---

wi e assesse In a e.

The parameter identified is sufficient to adequately model surge line flow resistance and ressure dro in the rima coolants stem. The abili of the Cb)(4) result in a limited 95/95 upper tolerance limit will be assessed in Table 5.

PNNL agrees that S-RELAP5 correctly calculates the natural circulation flow. The ability of the (6)(4)

~..,.,....,.-....,,........,.,...---,--,=-=-:=:-----,-.,..------,,,....-,.,..-...,.,,...,,--~

!CbX4)

I result in a limited 95/95 upper tolerance limit will be assessed in Table 5 The parameters identified in conjunction with the boron injection and dilution and the ECCS operation addressed in PCS-6 and PCS-7 are sufficient to adequately model the makeup and letdown flow in the primary coolants stem. The abili of the!Cb)(4)

I (b)(4) esult in a limited 95/95 upper tolerance limit WI OFFIOl,lcl USE OHLY PNNL-34230 29

Process/Phenomenon PCS-25. RCP Bearing

& Seal Flow PCS-26. Pressurizer Relief Valve Flow PCS-27. RCS Heat Structure Latent Heat Storage & Release PCS-28. Flashing of Coolant in the Reactor Vessel Upper Head PCS-29. Heat Addition from RCP Operation PCS-30. Choking /

Critical Flow PCS-31. RCS metal thermal expansion SEC-1. Stored Energy Technical Evaluation Test Type CMA CMA FMA IET CMA SET N/A CMA Associated Parameters (b)(4)

Evaluation The parameter identified is sufficient to adequately model the RCP bearing and seal flow in the rimar coolants stem. The abilit of thel....

Cb_)<4_) ___ ~

, X4) result in a limited upper o erance,m, wI e assesse tn a e.

The parameters identified are sufficient to adequately model pressurizer relief valve flow in the rimar coolant s stem. The a bi lit of the ICb)(4)

)(4) res'-u-.lt,....,~n-a-~

upper to erance,m,t w, The parameters identified are sufficient to adequately model the RCS heat structure latent heat stora e and release in the rimar coolants stem.

The ability of the....

Cb_)(_4) __________________ ~

ICbX4)

I result in a limited 95/95 upper tolerance limit will be assessed in Table 5.

PNNL agrees that S-RELAP5 correctly calculates the flashin of coolant in the reactor vessel u er head. The abilit of the )(4)

(b)(4) result in a limited 95/95 upper tolerance limit will be assessed in Table 5.

The parameter identified is sufficient to ade uatel model the heat addition from RCP operation. The ability of the

)(4) jCb)(4) resu tn a ImI e upper o erance limit will be assessed in Table 5.

The parameters identified are sufficient to adequately model choking and critical flow in the rima coolants stem. The abili of the~)(4)

Cb)( 4)

,_re_s_u..,...lt....,.i n_ a _ ___,

limited 95/95 upper tolerance limit will be assessed in Table 5.

The parameter identified is sufficient to adequately model the RCS metal thermal expansion in the primary coolant system. The ability of the (b)(4) result upper o erance,m, wI e assesse,n a e.

The parameters identified in conjunction with the primary-to-secondary heat transfer and internal recirculation addressed in SEC-2 and SEC-4 are sufficient to adequately model the stored energy in the secondary coolant OFFIOl,lcl USE ONLY PNNL-34230 30

Process/Phenomenon SEC-2. Primary/

Secondary Heat Transfer SEC-3. Liquid Separation in the Steam Generator Separators SEC-4. Internal Recirculation SEC-5. Liquid Mist Deposition on the Steam Generator Dryer Hardware SEC-6. Moisture Carryover out the Steam Generator Exit Nozzle SEC-7. Steam Generator Water Level Measurement SEC-8. MFW Addition and Regulation Technical Evaluation Test Type IET SET CMA CMA N/A SET CMA CMA Associated Parameters X4)

PNNL-34230 Evaluation system. The ability of the )(4) r x4) lre....

su......,...t ~m-a----..-,m

--,-,-,te---.-==-=5~u-p_p_e_r..,..to""Te_r_a_n-ce-,-m..,.,,t-w~,,.,....,..-e__,

assessed in Table 6.

~-------------~resu tin a 1m1te upper tolerance limit will be assessed in Table 6.

The parameter identified is sufficient to adequately model the liquid separation in the steam generator se arators in the seconda coolant system. The ability of the._Cb_)(_4) ________________

~

ICb)(4)

~esult in a limited 95/95 upper tolerance limit will be assessed in Table 6.

The parameter identified is sufficient to adequately model the internal recirculation in the seconda coolants stem. The abilit of the ~~Cb-X-4)--~I CbX4) result in a hm,ted upper o erance 1m1 w1 Framatome stated l,'.b)(4)

I (b)(4)

(b)(4)

I PNNL finds this acceptable but notes that ICb)(4)

(b)(4) b)(4)

I PNNL agrees that S-RELAP5 correctly calculates the moisture carryover out the steam generator exit nozzle. This parameter is either calculated correctly or conservatively.

The parameters identified are sufficient to adequately model the steam generator water level measurement in the seconda coolant s stem. The ability of the Cb)(4)

Cb)(4) esult in a limited 95/95 upper tolerance limit will be assessed in Table 6 The parameters identified are sufficient to adequately model the MFW addition and regulation in the secondary coolant system. The ability of the (b)(4) result in a limited 95/95 upper tolerance limit will be assessed in Table 6 31 6FFle l)!(L USE 6NLY

Associated Process/Phenomenon Test Type Parameters Evaluation SEC-9. MFW Flow Isolation SEC-10. Flashing of FMA MFW Downstream of CMA Isolation Valve SEC-11. AFW Injection CMA and Regulation SEC-12. Steam SET Generator Exit Nozzle or lnline Venturi/Orifice Choking/Flow Restriction SEC-13. Main Steam FMA Line Flow Resistance SEC-14. Main Steam CMA Line Flow Isolation (MSIV closure)

SEC-15. Steam CMA Generator / Main Steam Line Pressure Measurement SEC-16. Secondary CMA Relief Valve Flow Technical Evaluation (b)(4)

The parameters identified are sufficient to adequately model the MFW flow isolation in the secondary coolant system. The ability of the [oX4)

I (b)(4) result in a limited 95/95 upper tolerance limit will be assessed in Table 6 The parameter identified is sufficient to adequately model the flashing of MFW downstream of the isolation valve in the secondary coolant system.

The ability of the o)(4)

~-------------------~

~)(4) jresult in a limited 95/95 upper tolerance limit will be assessed m Table 6.

The parameters identified are sufficient to adequately model the AFW in'ection and re ulation in the seconda coolant s stem. The abili of the (b)(4) result in a limited 95/95 upper tolerance limit will be assessed in Table 6 The parameters identified are sufficient to adequately model the steam generator exit nozzle or inline Venturi/orifice choking/flow restriction in the secondary coolant system. The ability of the..._!Cb"'"'X__,.4) _________

ICbX4)

I result in a limited 95/95 upper tolerance hm1t will be assessed m Table 6 The parameter identified is sufficient to adequately model the main steam line flow resistance in the secondary coolant system. The ability of the P,)C4) tesult in a limited 95/95 upper tolerance limit will be assessed in Table 6.

The parameters identified are sufficient to adequately model the main steam line flow isolation in the secondary coolant system. The ability of the (b)(4) result in a limited 95/95 upper tolerance limit will be assessed in Table 6 The parameters identified are sufficient to adequately model the steam generator and main steam line pressure measurement in the seconda coolant system. The ability of the Cb)C4)

P,)C 4) lre~rTn"">>T!n'rnfn7'!'-ui-='"Tii'1=1""1?',==l'5"'1i~

will be assessed in Table 6 The parameters identified are sufficient to adequately model the secondary relief valve flow in the secondary coolant system. The ability of the QfifilGIAb YSli ONLY PNNL-34230 32

Associated Process/Phenomenon Test Type Parameters Evaluation SEC-17. Secondary FMA Relief Valve Inlet Piping Flow Resistance SEC-18. Secondary FMA Heat Structure Heat Storage & Release SEC-19. Choking /

SET Critical Flow SEC-20. Flashing of IET Tube Rupture Effluent FMA CMA SEC-21. Boiler Region IET Mixture Level SEC-22.Steam N/A Generator secondary metal thermal expansion Technical Evaluation

bX4)

)(4)

~------------------------~

result in a limited 95/95 upper tolerance limit will be assessed in Table 6 The parameter identified is sufficient to adequately model the secondary relief valve inlet pi in flow resistance in the seconda coolants stem.

The abilit of the (b)(4)

b)(4)

':n--,=-::::-r.::==::-:r,l"Tl!'"l,n"l:,"-:-S:==-=====-==-:-:,:-:n-,::-:--====c::!

in a e.

The parameters identified are sufficient to adequately model the secondary heat structure heat stora e and release in the secondar coolants stem.

The ability of the X4)

~ )(4)

I res~u"""lt,....,i_n_a-,1.,...im"""'i,.,...te_d.,...9""'5=/=9=5-u_p_p_e_r t,-o.,..le_r_a-nc-e---,-,-lim_,.,.it-w"""il,.,...I,-be---~

assessed in I able 6.

The parameters identified are sufficient to adequately model the choking and critical flows in the seconda coolants stem. The abilit of the erance ImI wI PNNL agrees that S-RELAP5 correctly calculates the flashing of tube rupture effluent. This parameter is either calculated correctly or conservatively.

PNNL agrees that S-R ;=-=-"---'"-"'-=--'-=.......,='-'===-:a-=-=-=='--'-=='-'-'-'=='-'=-.

level. The ability of the pi)(4) resul *'"="=-r==,...,...,,..,....,.--;-;u-=p=-pe=-=r=-=or.:ec:-ra=-=n=-=c:c,:e::-r.::1m=1:r:--:cw.,,..,, r-c-:e:c---

assessed in Table 5.

This approach is considered acceptable (b)(4)

(bX4)

(b)(4)

OFFlel)!cl USE OHL¥ PNNL-34230 33

Process/Phenomenon TG-1. Operation of Turbine Valves TG-2. Post-trip Power to RCPs HUM-1. Operator action Test Type CMA CMA CMA GCN-1. Delayed FMA Neutron Fraction, (Peff) CMA PTO GCN-2. Mean FMA Generation Time (1) from the peff/i component of the lnhour Equation for point kinetics.

GCN~. Dop~er FMA Reactivity PTO GCN-4. Moderator Reactivity GCN-5. Scram Worth GCN-6. Shutdown Margin Technical Evaluation PTO PTO FMA PTO Associated Parameters (6)(4)

Evaluation The parameters identified are sufficient to ade uatel model the o eration of turbine valves. The abilit of the )(4)

)(4) esult in a limited 95/95 upper o erance ImI wI e assesse in a e 7.

The parameter identified is sufficient to adequately model the post-trip power to RCPs. The abilit of the ICb)(4)

Cb)( 4) resu lt..,.,n_a.,.,h-m...,.,,t,_e....

d..,9'""5'""/9""'5~up_p_e_r-.to_,l,...e-ra_n_c_e...,.I~1 m-.i,-t ~

wI e assesse In a e The parameter identified is sufficient to adequately model the operator action. The ability of the CbX4)

P,)(4) lresu...

1t"'"'in-"-a-li-m-it_e_d_9~5~/9~5-up_p_e_r_t_ol-e-ra_n_c_e_l_im_i_t w-ill_b_e _ _,

assessed in Table 8.

The parameters identified are sufficient to adequately model the delayed neutron fraction. The ability of thel(b)(4)

I P,)C4)

I result in a limited 95/95 upper tolerance limit will be assessed In Table 9.

The parameters identified are sufficient to adequately model the mean generation time. The ability of the t;h.__

)(4_) ____________

P,>(4)

I result in a limited 95/95 upper tolerance limit will be assessed in Table 9.

The parameter identified is sufficient to adequately model the doppler reactivity. The ability of the (6)(4)

Cb)(-1) result in a limited 95/95 upper tolerance limit WI The parameters identified are sufficient to ade uatel model the moderator reactivit. The abilit of the

)(4)

(6)(4) result in a limited 95/95 upper tolerance limit wI e assesse in a e.

The parameter identified is sufficient to adequately model the scram worth.

The ability of the

)(4)

~-------------------~

[oX4)

I result in a limited 95/95 upper tolerance limit will be assessed in Table 9.

The parameter identified is sufficient to adequately model the shutdown margin. The ability of the o )(4)

QFFIGIAb YSli QNb¥ PNNL-34230 34

Process/Phenomenon GCN-7. Rod Worths GCN-8. Boron Concentrations GCN-9. Boron Worth GCN-10. Power Asymmetries GC N-11. Ex core Flux GCN-12. Decay Heat GCN-13. Xenon Reactivity Test Type FMA PTO FMA PTO FMA PTO FMA CMA FMA PTO CMA PTO FMA CMA GCN-14. Multi-FMA Dimensional Kinetics PTO LCN-15. Allowed Initial PTO Power Peaking FMA Technical Evaluation Associated Parameters

)(4)

PNNL-34230 Evaluation ICb)C4)

I result in a limited 95/95 upper tolerance limit will be assessed in Table 9.

The parameter identified is sufficient to adequately model the rod worths.

The ability of the CbX4)

ICb)C4) lresu'°Tr",'1nc:-:=a-.=1m~,r:e=-i...,...,,...,_-:-=-==:c-=~==-=-i=,r-:-:-.,,.,,-,c,-:::--=:,:c-::-:::=-=--::r-'

in Table 9.

The parameters identified are sufficient to adequately model the boron concentration. The ability of the[o)(4)

I IC6)(4)

!result in a limited 95/95 upper tolerance limit will be assessed in Table 9.

The parameter identified is sufficient to ade The ability of the (b)(-t)

P,)(4)

I resu\\i-r:1-=n--::a::-r.-::1 m:::c1"'e,,..,,,...,l'TC"l.-n::---:-:-:::==-===:::-:::-=-:r-:-:7.TT'"=-:::-:::===--'

in Table 9.

The parameter identified is sufficient to ade uatel ower asymmetries. The ability of the (b)(4)

~ )(4)

I res~u~t~m-a-

,-m~,t-e-~~u_p_p_e_r-to~ e-ra_n_c_e_1~m-,t,.....

will be assessed m Table 9.

The parameter identified is sufficient to adequately model the excore flux.

The ability of the )(4) k\\,)(4)

I resu~ 1;;::;n:--:a:;-r.;1m=1 e=--nrrrn::--;-;-;:=~==~r.;:;:;';T";'.;7,'TT"=--::v:==':'-;:f in Table 9.

The parameter identified is sufficient to adequately model the decay heat.

The ability of the

)(4)

"',-'--'--'--~~~~~---~--~~~~-----,J

!CbX4) lresult in a limited 95/95 u er tolerance limit will be assessed m Table 9. Pl\\JNL finds the use of b )(4) acceptable within EM2 and th (b)(4)

EM1.

PNNL agrees that ARTEMIS provides an adequate solution for the 30 kinetics. )(4) 35 OFFIGIAL ~SE ONLY

Associated Process/Phenomenon Test Type Parameters CMA CbX4)

LCN-16. Event Specific FMA Power Distributions CMA LCN-17. Energy Deposition Factor LCN-18. Rod and Assembly Bow Penalties CMA CMA TH-1. Nonuniform Core FMA Inlet Flow Distribution SET CMA TH-2. Nonuniform Flow FMA Resistance in the Core SET TH-3. Fuel Assembly Bow TH-4. Fuel Rod Bow TH-5. Crossflow Resistance Technical Evaluation FMA CMA FMA CMA FMA Evaluation

[o)(4)

!result in a limited 95/95 upper tolerance limit will be assessed in Table 10.

The parameters identified are sufficient to adequately model the event specific power distributions. The ability of )(4)

bX4) result in a limited 95/95 upper o erance ImI wI e assesse in a e The parameter identified is sufficient to adequately model the energy deposition factor. The ability of the ICb)C4)

\\'b)(4)

!result in a limited 95/95 upper tolerance 1mit will be assessed in Table 10.

The parameters identified are sufficient to adequately model the boron rod and assembly bow penalties. The ability of the 4

CbX4) result in a limited 95/95 upper tolerance limit will be assessed in Table 10.

The parameter identified is sufficient to adequately model the nonuniform core inlet flow distribution. The ability of the..::[o..:.:.)<4"'"') _________

!Cb)(4)

I result in a limited 95/95 upper tolerance limit will be assessed in Table 11.

Framatome states Cb)C4)

(b)(4)

(b)(4) equate ca cu a I0n o non-uni w (4)

4)

The parameter identified is sufficient to adequately model the fuel assembly bow. The ability of the ~'b_X4_) ________________ ~

!CbX4)

!result in a limited 95/95 upper tolerance limit will be assessed in Table 11.

The parameter identified is sufficient to adequately model the fuel rod bow.

The ability of theP,>C4)

I

!Cb)C4)

~esult in a limited 95/95 upper tolerance limit will be assessed in Table 11.

PNNL agrees that COBRA-FLX rovides an ade uate solution for the crossflow resistance. CbX4)

CbX4)

OFFIOIAL ~SE ONLY PNNL-34230 36

Process/Phenomenon Test Type TH-6. Diversion Crossflow TH-7. Grid-Induced Flow Mixing FMA SET FMA TH-8. Turbulent Energy FMA Exchange TH-9. Two-Phase Flow FMA

/ Voiding SET TH-10. Spacer Grid Flow Resistance TH-11. Bare Rod Friction Technical Evaluation FMA FMA Associated Parameters

)(4)

Evaluation PNNL agrees that C~llil~~m!:'i.l.Qte.s....an.ao:~J.at!Ll.!lli!tm.D.lQL,the diversion crossflow. Cb)(4)

X4)

PNNL agrees that COBRA-FLX provides an adequate solution for the grid-induced flow mixino. IC6)(4)

I CbX4)

PNNL agrees that COBRA-FLX provides an adequate solution for turbulent energy exchange. (6)(4) r X4J

)(4)

PNNL agrees that COBRA-FLX provides an adequate solution for the two-phase flow and voiding. b )(4)

(6)(4) b )(4)

PNNL agrees that COBRA-FLX rovides an ade uate solution for the spacer grid flow resistance. b )(4)

(6)(-1) e b )(4)

~-,,------.I result in a limited 95/95 upper tolerance limit will be assessed PNNL agrees that COBRA-FLX provides an adequate solution for the bare rod friction. (6)(4)

(6)(4)

(6)(4)

QFFIGIAb Uilii ONLY PNNL-34230 37

Process/Phenomenon Test Type TH-12. DNB / CHF FMA Correlation IET TH-13. DNB Propagation N/A THAP-1. Manufacturing CMA Tolerances THAP-2. Cladding FMA Surface Heat Transfer THAP-3. Radial and Axial Nodalization THAP-4. Impact of TH Parameters on Nodal Model Results FRR-1. Normal Operation Power History and Fluence History FRR-2. Fuel Pellet Radial Power Profile Technical Evaluation FMA IET FMA IET FMA SET Associated Parameters

)(4)

Evaluation The parameter identified is sufficient to adequately model the DNB/CHF correlation. The abilit of the !Cb)(4)

I X4) result in a limited 95/95 upper tolerance limit The ARITA methodolo uses a conservative a CbX4) ib)(4)

PNNL agrees that this approach Is conserva Ive an accep a e or Is application The parameters identified are sufficient to adequately model the effect of manufacturin tolerances on the thermal h draulic solution. The abili of the Cb)(4)

~

esult in a limited 95/95 upper tolerance limit will be assessed in Table PNNL agrees that empirical correlations in COBRA-FLX rovide an ade uate solution for the claddin surface heat transfer. X4)

(b)(4)

(b)(4)

PNNL agrees that the method for determinin radial and axial nodalization in COBRA-FLX is acceptable )(4)

Framatome's approach is..--....__--------------,---'

(b)(4)

IS approach is reasonable.

ICb)(4) r )(4)

I PNNL agrees

)(4)

PNNL agrees with Framatome that )C4)

X4)

QfifilGIAb Y&~ ONLY PNNL-34230 38

Process/Phenomenon FRR-3. Fuel Rod Axial Power Distribution FRR-4. Transient Power Conditions FRR-5. Transient Coolant Conditions FRR-6. Fission Gas Release from Pellets FRR-7. Heat Conduction through Pellet FRR-8. Pellet-Cladding Gap Conductance Test Type IET IET CMA IET IET CMA IET SET FRR-9. Pellet-IET Cladding Gap Closure Technical Evaluation Associated Parameters

)(4)

Evaluation The parameters identified are sufficient to ade uatel model the transient power conditions. The ability of the (b)(4)

(b)(4) resu t in a ImIte upper to erance ImIt w, e assesse tn a le 13.

Framatome states, )(4)

The parameter identified is sufficient to ade model the fission as release from ellets. The abili of the (b)(4)

(b)(4) res~u~t ~tn-a-,-m~, -e- -~u_p_p_e_r~to~e-ra~nce ImItwI The parameters identified are sufficient to adequately model the heat conditions through the pellet. The ability of the

)(4)

(b)(4) esult in a limited upper to erance ImI wI e assesse tn a e The parameters identified are sufficient to adequately model the pellet-claddin a conductance. The abilit of the l ~(b_)(4_) ________ ~

(b)(4) result in a limited 95/95 upper o erance ImI wI The parameters identified are sufficient to adequately model the pellet-cladding gap closure. The ability of the._~_X4....,.>-,--,.,....-,--.,.....,....,....,...,.-------------'

j(b)(4)

Iresult in a limited 95/95 upper tolerance limit will be assessed in Table 13.

OFFIOl,lcl ~SE QNLY PNNL-34230 39

Process/Phenomenon FRR-10. Cladding Corrosion, Crud Deposition, and Cladding Hydriding FRR-11. Burnup of Fuel Rod Associated Test Type Parameters SET 1>X4)

IET FMA FRR-12. Manufacturing CMA Uncertainties Technical Evaluation Evaluation The parameter identified is sufficient to adequately model the cladding corrosion and CRUD Deposition. [1i)(4)

)(4)

(b)(4)

The abilit of the i ;:;

(PNNL agrees thatti,x,,

(4)

The parameters identified are sufficient to adequately model the effect of manufacturing uncertainties on the fuel thermal-mechanical solution. The ability of th (b)(4)

(b)(4) esu in a Im1 e e

assesse in able 13.

QfifilGIAb Uili 0Nb¥ PNNL-34230 40

PNNL-34230 3.8.2 Non-parametric Uncertainty Analysis Process For each of the target scenarios that have been identified, the ARITA methodology can be used to make a non-arametric statistical statement. For each scenario, a population of event results is defined Cb)(4)

A Monte Carlo statistical a roach is used to le from this o ulation Cb)<4>

--~--- sing e I en 1 Ie or eac arge scenario.

e accep ance en ena or each a r g e scenario are verified by ensuring that the 95/95 value for each figure of merit does not exceed predefined limits (Table 1 ).

For events that ICb><4)

I the ARITA methodology uses the Wilks method (15], a non-parametric approach that provides the ability to calculate the number of realizations needed to determine the desired 95% probability level at the specified 95% confidence level.

This rocess can treat a lar e number of uncertainties simultaneously. For events that Cb)C4)

ARITA has been modified accordin to RAl-8 to demonstrate that (bX4)

Within the Wilks method, the resulting figure of merit from each realization is ordered from least to greatest (x1 is greatest, x2 is second greatest, etc.) and based on the number of samples run, the 95/95 tolerance limit can be taken as the figure of merit from the realization with the appropriate order statistic from Table 4.

Table 4. Wilks Method Required Number of Observations for Each Order Statistic for at least 95/95 Tolerance Selected Order Statistic for at least 95/95 Tolerance Limit X 1 X2 X3 X4 XS X6 X7 Xa X9 X10 Required Observations (N) 59 93 124 153 181 208 234 260 286 311 Each of the three ARITA EMs use this statistical rocess. CbX4>

(b)(4)

(b)(4)

Once this has been established, a random initial seed is chosen and, from that seed, a random number generator is used to select random values from the input parameter distributions. These values are then input to the appropriate location within the EM for each realization.

Technical Evaluation 8FFIGIAL Y&i! QNbY 41

PNNL-34230 In the original version of ARITA, it was proposed to 1Cb)C4)

(b)(4)

I PNNL and NRC asked Framatome to iustifv 1Cb>C4)

(b)(4)

(bX4)

I The following simulation process is used to perform the analysis of each target scenario using the ARITA methodology:

1.

(1,)(4)

2.

3.

4.

5.

6.

For two of the SAFDLS, FCM and TCS, GALILEO performs a process to~X4)

Technical Evaluation 42

PNNL-34230 (bX-1) l(b)(4)

IThis information was reviewed and PNNL concludes that this (bX-t) acceptable Cb)(4)

PNNL finds the non-parametric uncertaintv analvsis process to be valid as lono as j(b)(4)

(bX4) is (bX-t)

I Therefore part of the review of the parameters in Section

,____--~----~__,I 3.8.3 will confirm r'°)<4)

~------------~

3.8.2.1 Parameters with [....

1>X_ 4_) _____

Several parameters where l(bX4)

I (b)(4)

I P,X4)

IPNNL and NRC asked Framatome to justify these sampling ranges in RAl-11. For the issue of j(b)(4) 1,)(4)

(b)(4)

I (b)(4)

(b)(4)

I (b)(4)

(b)(4)

Technical Evaluation OFFICIAL YSE ONLY 43

PNNL-34230 (b)(4)

Cb><4)

I PNNL has reviewed this approach and has determined that it is an acceptable treatment ofl(b)(4)

I PNNL's review of the event initiators is shown in Table 14.

Cb)(4)

IPNNL has reviewed this approach and has determined that 1t Is an acceptable treatment otICb)(4)

I 3.8.2.2 Other conservatisms In the ARITA methodology, following the determination of an UTL for TCS P,X4)

(b)(4)

CbX4)

PNNL

'-------~-------~----------------'

concludes that the (b)(4) is a reasonable conservatism based on experience with the NRC fuel p'-e~ o_r_m_a_n-ce-co~~...iST[16] which is similar to GALILEO.

3.8.2.3 Application caveats Framatome has requested that ARITA be approved for applications with mixed cores and reconstituted fuel assemblies.

For mixed cores, which is a core loaded with assemblies having different spacer grids, CHF correlations, and cross-sectional geometry all assemblies in the core are analyzed with the

6)(4) tb)<4)

I-PNNL has reviewed this material and agrees that the ARITA method rs acceptable tor modeling Framatome fuel in mixed core applications.

Regarding reconstituted fuel assemblies ARITA contains appropriate models to analyze fuel assemblies that have previously been irradiation are reconstituted using inert rods inserted into Technical Evaluation OFFIGIAL YSE QNbY 44

PNNL-34230 selected fuel rod locations. This is typically done because an assembly has had one or more leaky fuel rods removed and replaced with a non-fueled rod. Framatome clarified in RAl-83 that the ARITA reconstitution methodology is expected to be applicable to future fuel designs because the current fuel design process ensures that the new fuel will fit into this methodology.

Framatome further clarified that the ARITA methodolo for use of inert rods~ (4)

I

)(4)

P NL has reviewed this material and agrees that the ARITA reconstitution methodology is acceptable.

3.8.3 Uncertainty Parameters for Non-Parametric Uncertainty Analysis The uncertainties used in ARITA for the uncertainty analysis is generally divided into three types.

(b)(-l)

Framatome documented in the response to RAl-9 how each uncertainty parameter would be treated includina ICb)C4)

(b)(4)

KbX4)

I The following tables rovide documentation of each uncertaint as well as PNNL's review of the acceptability of the :b)C4)

These tables also identify (b)(4)

(b)(4)

Table 14 provides the same assessment of the uncertainty parameters associated with the event initiators.

Technical Evaluation OFFIOl~L USE ONLY 45

PNNL-34230 Table 5. Evaluation of PCS Uncertainty Parameters Number Description Type Parameters Evaluation (b)(4)

(b)(4)

(b)(4)

I I ne use of a conservative loX4)

(b)(4) ps acceptable. ICb)(4)

(b)(4)

Kb)(4)

I these will neea m oe rev1ewea ana approvea m me ume OT app11cauon.1(b)(4)

(b)(4)

(b)(4)

Technical Evaluation 46 OFFle l)!(L USE OHL¥

Number Description Type (b)(4)

'hX4) ri,)(4) libl(4)

I (b)(4) r )(4)

I (b)(4)

)(4)

Technical Evaluation Parameters PNNL-34230 Evaluation tbX4)

I Boron injection concentration from the RWST is defined by the TS for the LCO. p,)(4)

Cb)C4)

The use o a conserva 1ve,_Cb_)(_4) ___________ __, 1s accep a e.

(b)(4)

I land may oecome unoorarea.1(b)(4)

(b)( 4)

ICb)(4) 11 ne assumpuon 01 ICbX4) 11s accep1arne as u 11s 1s a conservative j(b)(4)

(b)(4) lis an engineered safety feature for actuation of the C.l.,l.,;:,. ti,)(4)

KbX4)

~ )(4)

I Use of a conservative.._

l1Cb_X4_) ________ __,

p,)C4)

Hs acceptable. However, because sensitivity studies may be needed to determine the conservative direction specific to an event ~

)(4)

'b)(4)

ECCS. The actuation time delay for ECCS can influence FoM for clad DNB, FCM and clad strain. (b)(4) jCb)(4)

ISRP 15.2 and 15.4 do not challenge this system. I his system does not provide primary protection for SRP 15.3. SRP 15.1 47 OFFIGIAb Uili ONLY

Number Description Type jCb)(4)

(b)(4) i:h)(4)

(b)(4)

Technical Evaluation PNNL-34230 Parameters Evaluation and 15.2.8 may challenge this system, and this s stem does protection for SRP 15.5 and 15.6. (b)(4)

(b)(4)

(b)(4) e Ime o app Ica I0n.

I

~an influence the FoM for clad DNB, FCM and clad stram. lCb)(4)

I r x*l I

fb)(4)

~ SRP 15.2 (except for 15.2.8), 15.3 and 15.4

'ao not challenge this system. SRP 15.1 and 15.2.8 may challenge this system, and this system provides primary protection for SRP 15.5 and 15.6. Use of a conservative jCbX4) ps acceptable. (b)(4)

(b)(4) is considered acceptable ~:b-)(-4).!:a::c::==============t (b)(4)

Is ess c a engmg o o sa e y margins )(4)

(b)(4)

FoM or clad DNB, F M and clad strain. (b)(4)

(b)(4)

(b)(4)

IS system. SRP

. an

.. may c a enge Is sys em, an Is sys em provides primary protection for SRP 15.5 and 15.6.

(b)(4)

(b)(4) is considered acceptable. )(4) will have QfifilGIAb Uili 0Nb¥ 48

PNNL-34230 Number Description Type Parameters Evaluation a small impact of FoM l(b)(4)

(b)(4)

(b)(4)

(b)(4)

I The use ofj(bX4) jlS considered acceptable. vmer conservatisms are included tor l(bX4)

I fiiX4)

I The

'-----------------:-:-us=--=e=-=of,--a::.==)(4=)================;-i;:-s--:c=-=o-=n-=-si;:;d;::e:-:re=--=d.-a=--=c=--=c=-=e-=p.='

ta'ble.

Other co~n-se~rv-a:-:-tis_m s_a-re---:-in-c-:-lu-d;-e-.d--;fo-r--;(b=)=(4=)====----------..:.._---~

(bX4)

)(4) will have small

)(4)

(b)(4)

Technical Evaluation 49 OFFICl1S.L USE ONL i

PNNL-34230 Number Description Type Parameters Evaluation considered acceptable because other conservatisms l(b)(4) f6)<4)

I ~-------~

r,)(4)

I 1 )(4) l(b)(4)

I kb)(4) 11 ne use of a conservative ~b)(4)

I l(b)(4)

I (b)(4)

Cb)(4) r X4J I

The RCS trip setpoint and uncertainty t:tiX4)

KbX4)

(bX4)

!The use of a conservative l(bX4)

!b)(4) 1s considered acceptable.

The RCS actuation response time (b)(4)

(b)(4)

(b)(4) e use o a conserva 1ve "'"(b""")(....;.4) ________

~(b~)<4~) ___

~ is considered acceptable.

i X4J I

SG tube plugging l(bX4)

CbX4)

(bX4)

I ' ne use OT conservauve1y l(bX4) 11s cons1aerea acceprao1e. ICbX4)

(b)(4)

I I

Technical Evaluation 50 QfifilGIAb Uili 0Nb¥

PNNL-34230 Number Description Type Parameters Evaluation I

(b)(4)

~ )(4)

I as 1nd1cated 1s considered acceptable.

RCP performance affects the total RCS pressure drop and heat transfer between the primary and secondary side. t'.l>X4)

I (b)(4)

!(b)(4)

I The use of a conservatively biased coast-down is considered acceptable.

r )(4)

(b)(4)

Technical Evaluation 51 OFFIOIAL ~SE ONLY

PNNL-34230 Number Description Type Parameters Evaluation (b)(4)

(b)(4) 1andis conservauve J(bX4)

(b)(4) 11 ne use or conservat1ve1y ICbX4)

(b)(4) ps considered acceptable.

(b)(4)

Kl,)(4)

(b)(4)

(b)(4)

I flow is considered acceptable. An 1mpllcat1on of using the cold bypass flow as a l1m1t at hot cond1t1ons 1s that this variance 1s greater than ICb)(4)

I This is considered a conservative assumpt1on.l(b)(4) l'.b)(4)

K6X4)

Technical Evaluation 52 OFFICIJlcl USE OHLY

PNNL-34230 Number Description Type Parameters Evaluation (6)(4)

This is considered acceptable. Justification of the excore ration factor uncertainty will be reviewed and approved at the time of application.

P,)<4>

I (6)(4)

IC6)(4)

I This is consIaerea acceptaoIe. IC6)(4)

I I

(6)(4)

IC6)(4) iThus, the performance of the coolant loop temperature measurement system can influence the timing of reactor trip, and core power level at trip for some plants.

(6)(4)

SRP 15.1 may challenge this system, and for SRP 15.4 this system is considered a secondary influence. SRP 15.2 (with the exception of 1.5.2.4 ), and SRP 15.3 are not expected to challenqe this system. 1(6)(4)

(6)(4)

(6)(4) 11 his is considered acceptable. I(6)(4)

I I

Technical Evaluation 53 OFFIOl,lcl USE OP4LY

PNNL-34230 Number Description Type Parameters Evaluation (b)(4)

(bX4)

X4) us, t e u

u y tern can influence the timing of reactor trip, and core power level at trip for some plants.

(bX4)

(b)(4) 1s cons1

,,_(b-)(4_) ______

.,..1_s_c...,.o...,.ns=1....,...,...er--e--.-a--c--c...,.e-p..,.able.

f:1,)(4) p,)(4) lis considered

- -1Cb)(4) 11s avvt:fJldu11:: d::> mis 1Sl(b)(4)

I l:1,X4)

(b)(4)

ICb)(4) 11s considered Technical Evaluation 54 OFFIGIAb USE ONLY

PNNL-34230 Number Description Type Parameters Evaluation acceptable. :bJ(4) is acceptable as this is P,)(4)

(b)(4) l (b)(4)

I RCS coolant flow rate is dependent on performance of the RCPs and the flow rate measurement uncertainty. (b)(4)

- - [ X')

I (b)(4)

(b)(4)

Technical Evaluation

_ )(4)

RCS flow and flow measurement uncertainty are defined by plant licensing documentation. The use of a sampling approach using +/-4a truncated distribution (b)(4)

I is considered acceptable. A (b)(4) 1is acceptable as this is rt,)(4)

(b)(4) l(b)(4)

I This is considered a

...,,c~o~ns~e~r-va=n=,v~e-a~s~s-u=m~p=rn~o=n-. ----------~

OFFIOIAL ~Si ONLY 55

PNNL-34230 Number Description Type Parameters Evaluation (b)(4) l(b)(4) 1Control modes are defined by plant licensing documentation. l(b)(4)

(b)(4)

I When the (6)(4)

I is modeled in automatic mode, uiv 4)

I (b)(4) conservative modeling ~o X4)

[ When the (6)(4) 11s modeled m manual mode, the system is (6)(4)

(b)(4)

The use of conservatively lo )C4)

(6)(4) 11s considered acceptable 1(6)(4)

(b)(4)

t,)(4) r )(4)

~ -)(4_> _________ ~1's considered acceptable.

ICb)(4)

I Control)

ICb)(4)

I is considered acceptable.

Technical Evaluation 56 QFFIGIAL USE OP4LY

PNNL-34230 Number Description Type Parameters Evaluation l(b)(4)

(bX4)

P,>C4)

I is considered acceptable. P,X4) lis acceptable as this ISl(b)(4) 1(b>c4>

I l(b)(4)

I

'.b)(4) l(b)(4) 1 is considered acceptable. )(b)(4)

Is accep1ao1e as 1(bX4) l(b)(4)

I j:b)(4)

I (bX4)

(b)(4)

I conservative modeling P,X4)

I (b)(4)

(b)(4) 11s cons1aerea acceprnrne Technical Evaluation 57 OFFICIJlcl USE ONLY

PNNL-34230 Number Description Type Parameters Evaluation ICb)(4)

I (b)(4) r )(4)

I ICb)C4) 1 conservauve moae11ng or me Kl,)(4)

~ X4) 1 cons1aerea acceptao1e (b)(4)

CbX4)

I (b)(4)

I

~)(4)

(b)(4) 1conservative modeling 1Cb)C4)

(b)(4)

(b)(4) 1is considered acceptable tb)C4)

,b)(4)

(b)(4)

Technical Evaluation 58 OFFle l)!(L USE OHLY

PNNL-34230 Number Description Type Parameters Evaluation (b)(4)

I conservative modeling j;ii)(4)

(b)(4)

This tb)(4)

I (b)(4)

CbX4) 1conservative modeling ti,)(4)

(b)(4) tb)(4) 1is considered acceptable jCbX4)

ICbX4)

I (b)(4)

Technical Evaluation 59 Of Flel)!(L USE ONLY

PNNL-34230 Number Description Type Parameters Evaluation (b)(4)

(b)(4) 11s considered acceptable.

(b)(4) 1~)(42 ps considered acceptable. µ>)(4)

(6)(4) 11s considered acceptable. l(bX4) ps acceptable as this 1s l(b)(4) t:6)(4)

(b)(4)

I Pressurizer phenomena affects RCS pressure.kl,X4J P,)(4)

!provides a discussion ofl(b)(4)

(b)(4)

(b)(4) 1conservatively demonstrate l(b)(4)

!The approaches presented t:i,)(4) 1are considered acceptable.

(b)(4) 1s considered acceptable.

r )(4)

)(4)

Surge line flow resistances are (b)(--1)

<b)!4)

This approach

=-'-'--------------------------'

is considered acceptable.

(b)(4)

Technical Evaluation 60 Qf'.f'.ICIJ\\L I ISE ONLY

Number Description Type (bX4)

(bX4)

(bX4)

Technical Evaluation PNNL-34230 Parameters Evaluation

!(b)(4)

I I rns system has no significant influence for SRP 15.3 and 15.4 (except for 15.4.6). SRP 15.1 and 15.2 may have small influences )(4)

=-::-~

--:-:--::-:::--::--:------;===============:::;

SRP 15.4.6, 15.6.2 and 15.6.3 have[o)(4)

X4)

. Th.~e-us_e_o_f~c_o_n-se-rv-at-iv_e_ly~r,- )-(4-) _____

(bX4) is considered acceptable as this j(b)(-4)

(b)(4)

(b)(4)

I is considered acceptable as thiS!(b)(4)

(b)(4)

(b)(4) ps considered acceptable. This is becausel(bX4)

(bX4) bll4) 11s c,1.,1.,t::1,1caoie as u u::, 1s ~'bX4)

(b)(-4)

I OFFlel)!(L USE OHLY 61

Number Description Type r X4J j

Technical Evaluation Parameters PNNL-34230 Evaluation (b)(4) which is considered conservative (b)(4)

(b)(4)

(b)(4)

)(4)

'4)

Is cons, Is Is conserva Ive (b)(4)

RCP bearin an sea ea

)(4)

(b)(4) these flows cou ave an m uence on an event Igure o merit (b)(4) is is not a significant contributor for SRP Sections 15.1, 15.2, 15.3, 15.4 and 15.5. These may be influential for SRP 15.6.2 and 15.6.3. 1(b)(4)

'-:'":-;-;:-:-:-:-:-;::==========:

!ibX4) j1s considered acceptable as this is l;l,X4)

(4

!Cb)(4)

~:::::::::::=====::::=...j Is considered acceptable as this is P,)(4)

PORV are used for control of RCS:-p::-:r=-=e-=-s-=-su:--::r:-::e--;:l(b=)=(4=)====:

SRP 15.1 does not challenge operation of the P"<T"7....,.....,,..""7"7-....----:n.....,..---:n-,---------.--::15_5 may be exacerbated by operation of the PORV.ICb)(4)

I (b)(4) 1*SRP 15.6.1 is the driving force for !Cb)(4)

I

____________..._ __ ~I and for SRP 15.6.::S !ibX4)

(b)(4)

(b )(4)

Ino sIgn1t1cance for these two events. The use Ol!(b)(4)

~)(4)

!(b)(4)

I PORV are used for control of RCS pressure fo<4) ib)(4)

)(4)

SRP 15.1 does not challenge operation of the PORV. SRP 15.2, 15.3, 15.4, and 15.6 may be exacerbated b operation of the PORV. SRP 15.6.1 is the drivin force for RCS (b)(4)

(b)(4)

(b)(4)

Is cons, even s as Is )(4)

(b)(4) 62 OFFICIJlcl USE OHLY

PNNL-34230 Number Description Type Parameters CbX4)

Evaluation See the evaluation for (b)(4)

See the evaluation for I

PSV setpointslCb)(4)

(b)(4)

(b)(4)

(b)(4)

The PSV are used is to miti ate RCS over ressure)(b)(4)

Cb)( 4)

SR!J,,,

'5-,1,...,.5....,_ 1.-,-do_e_s-no....,t,....c.... h-a...

ll-en_g_e __

~

operation o t e

. an 15.5.1 ma challen e the PSV to protect RCS overpressure. Cb)(4)

'b)(4) l(b)(4)

['b)(4)

I1s considered acceptable as this....

Cb'""')(4...;..) _____

I l

(b)(4)

I (b)(4)

(b)(4) is considered acceptable as this

,...._ ____________________ __. does not influence the._ICb_X4_) ___________ _.I t=oM.

r X4) _ _ _ _

____.Ir"

This approach is considered acceptable !CbX4)

In addition, conservatisms )(4)

~~--------------------'

I e I as indi 1

cated is considered acceptable.

Technical Evaluation OFFICl1S.L USE ONL i 63

Number Description (b)(4)

CbX4)

Type Parameters PNNL-34230 Evaluation Events that may result in flashing of the coolant in the RV UH are characterized by a severe RCS depressurization caused by cooldown of the RCS coolant (i.e.,

steam line break) or a substantial loss of RCS inventory (i.e., steam generator

.__ ____________________ ___, tube rupture). (bX4)

,--)(-4)--~~----------------------~--,

cons,

e.

RCP operation introduces heat into the RCS, increasing secondary system heat removal. lb )(4)

(b)(4)

KbX4)

I approacn was recommenaea tbX4)

(b)(4)

I for conservatism. 1 rns approach is considered acceptable as this Is conservative !Cb)(4)

The HEM critical flow modellCb)(4)

Cb)C4) 1s used in tne smgIe-pnase vapor cno1<mg (b)(4)

(b)(4)

!approach is considered acceptable as m,s IS l(b)(4)

I ti,)(4)

I in auumon, conservansms p,)(4)

I Technical Evaluation 64 OFFICIAL I ISE ONLY

PNNL-34230 Number Description Type Parameters Evaluation (b)(4)

Subcooled and two-phase critical flow 1:1')<4)

(b)(4) nis approacn Is consIaerea acceptaoIe as mis Is l(b)(4)

)(4) and conservatisms arel<bl(4)

~ )(4)

(b)(4)

(b)(4)

(b)(4) I The RCSl(b)C4)

I componentl(b)(4)

(b)(4)

(b)(4)

!This approach is considered acceptable as t;bX4) l(b)(4) l(b)C4) 1In aaamon, conservatisms l(b)(4) tJ,)(4)

I Technical Evaluation 65 OFFICIJlcl USE OHLY

PNNL-34230 Table 6. Evaluation of SEC Uncertainty Parameters Number Description Type Parameters Evaluation (6)(4)

The secondarv side stored enerav is defined bv IC6)(4)

(6)(4)

I SG mass inventory, (6)(4) tbX4)

!small influence on FoM for SRP 15.1, 15.2, 15.3, 15.4, 15.5 and 15.6.

(6)(4) fhl/4)

Iseconaary sIae smrea energy. 1(6)(4)

(6)(4)

~b}('t)

IIs consIaerea accep{aoIe. l(b)(4)

CbX4)

(6)(4)

CbX4)

(b)(4)

Technical Evaluation 66 OFFICIJlcl USE OHLY

PNNL-34230 Number Description Type Parameters Evaluation (b)(4) l(b)(4)

I (b)(4)

I is considered acceptable. l'.b)(4) ri,)(4)

!will need to be reviewed and approved at the time ot app11cat1on. ~oX4)

I i X4J I

KbX4)

P,X4) lis considered acceptable !CbX4)

I ib)(4) 1conservat1ve1y o,asea ICbX4)

I (b)(4) l'.b)(4)

I n,x4) iassoc1ated measurement uncertamtyri,)(4)

KhX4)

(b)(4)

I is considered acceptable. 1Cb)(4)

'b)(4) ib)(4) lis considered acceptable.

(b)(4)

I is acceptable ICbX4)

I i:b)(4)

I (b)(4) is considered acceptable Technical Evaluation 67 QFFIGIAb YSli QNbY

PNNL-34230 Number Description Type

)(4)

Parameters Evaluation RAl-43 reviews the process involved in modeling a steam generator. Cb)(4)

ICbX4) lheat transfer correlation (6)(4) 00~

IBU~

or su coo e eat trans er

)(4) is used for nucleate boiling heat transfer G(b)(4)

I

!CbX4)

I is used for steam heat transfer. ('o)C4)

I foX4)

!T his 1s considered acceptable.

.---__.__ro_,vides justification ICb)(4)

PNNL found that..,,th-e- re-fe_r_e_n-ce-s=1te-d~ to-r-t~he_s_e_v_a-1-ue-s~1s_n_o_t_w_e~II--~

in either the paper Framatome sited or in the references used by that paper. However, PNNL does find these values reasonable based on an independent review of data.

PNNL's independent review of )(4)

)(4) rama ome proposa 1s cons e

RAl-43 rovides *ustification..,.,...,-,,---------------------,

l

~X4J

.....==========================================-'----. representing the uncertainty._ICb_)(4_) ______ __,

r

)(4)

I RAl-43 discusses [liquid separation in the Steam Generator Separators. lCbX4)

____.r

,-.l(b-)(-4) __________________

..,.li_s_a_c_c_e-pt_a_b_le-.----~

SEC-4a Technical Evaluation SG Biased to Recirculation achieve Flow Path Form BE (G)

Losses See RAl-43 reviews the process for defining the [SG internal recirculation jCb)(4) k\\,)(4)

I Target SG recirculation flowjCb)(4)

(6)(4) 0FFIOl,lcl ~SE ONLY 68

Number Description SEC-8a MFW Flow

)(4)

I.,,

Technical Evaluation Type Parameters Biased (G) Set to limiting control state I

PNNL-34230 Evaluation (bX4) l._(b_)(4_)____,,---.-.-----------------'I fhis approach is considered acceptable.

Main Feedwater (MFW) (bX4) response 1s

!(b)(4)

I which will determine the MFW flow response during the event. Nominal feedwater temperature as a function of~[b)-(4-) ~

b,)r4) land the associated measurement uncertainty is=i'b~)1~-11~-~

' )(4)

(b)(4)

MFW flow l;b)(4)

(b)(4)

(b)(4) ib)(4)

(b)(4) e 1me o app 1ca 10n.

MFW treatment (bX4)

(b)(4)

(6)(4)

IS IS (b )(4)

Kb)<-<>

I conservatively IC6)(4)

I (b)(4)

I setpoint definea l(b)(4) 69 OFFIOl,lcl USE OHLY

PNNL-34230 Number Description Type Parameters Evaluation

~

conservatively l(bX4)

I is considered acceptable.

.-l(b-)(-4)-----------------------.1 MFW flow isolation is de endent on actuation set oints and isolation/control valve

. response time. )(4)

~---------------------..... Cb)(4) the actuation setpoints and response times of the l(b)(4)

MFW isolation valves (MFIVs) and control valves. (b)(4)

(bX4) response time is defined by (b)(4)

(b)(4) 1s cons, ere accepta e.

Isolation valves are provided on the MFW supply to the steam generators. In the event these valves close it is possible that flashing can occur in this piping during some events. This can impact MSS pressure response and steam generator level.

This piping may be modeled explicitly, or a conservative estimate of FW added to the steam generator is assumed. Cold, best estimate dimensions are used to generate model input. The MSS pipinQ and component dimensions are defined from plant licensing documentation. P,)(4)

I tt,)(4) 1 is considered acceptable. rt,)(4)

I (b)C4)

I

""'(b"")C"""4) ________________

___,1rev1ewea ana approvea at the time of application.

Auxiliary feedwater (AFW) injection and regulation may be important for event response hindering recovery by delayed or early and/or preferential deliver of fluid to the SGs. AFW injection and regulation is dependent on RPS and ESFAS actuation setpoints.~'b)(4) 1(6)(4)

I and fluid temperature. 11-1(b-)(4_) ____ ---rl a=c=:1u=a=11=1o=n'""'s""'e=i1p=-o=,=m=:s"T1(b-.)-(4_) __

l(bX4)

At-W l(b)(4) l(b)(4)

I (b)(4)

I (b)(4) lis defined 1(6)(4)

I conservatively

~(b--)(4~)- ---'-'-----'-----,-1,,s-c_o_n_s~,aT""e-re-,a-.--a-cc_e_p"T>ta....,1 D~le-. __

r

)(4) j(b)(4)

~1/h=)=r.i=) _______________

..... 1conservat1ve1y l~(b_X4_> ____ ~

~--------------------~ (b)(4) is considered acceptable.

(b)(4)

(b)(4)

Technical Evaluation Qf:f:ICIJ\\L Uili 0NLY 70

PNNL-34230 Number Description Type Parameters Evaluation e

(4)

I (b)(4)

(b)(4) 11s considered acceptable. l:h)C4) r )(4)

RAl-69 clarifies that ~o )(4)

I single-phase vapor flow b)(4)

(b)(--1) r (4)

I two-phase choked flow l(b)(4)

RAl-70 clarified that [two-phase choke flow j:1,)(4)

(b)(4) b)(4)

I conservative [Cb)(4)

(b)(4) b)(--1) jd1scussed in KAl-9.

These approaches are considered acceptable.

I AAl-43 identifies Technical Evaluation 71 QFFIGIAb l:lili QNb¥

Number Description Type Technical Evaluation Parameters PNNL-34230 Evaluation is considered acceptable.

4 onservat1ve y X4) 1s considered accepta..,.,....e,...._-------------------~

r X*l

)(4) closure fj,)(4)

(b )(4)

CbX4)

(b)(4) rev1ewe Seconda

'1,)(4)

(b)(4)

I MSIV conservatively, X4) conservatively

...... -':--;:---;----....,....,,-;---,..------'-----___::~

is considered acceptable. (b)(4) e,me o app,ca 10,.,,.n.,..,.-,,--------------.

• ressure relief is de endent Cb)(4) is considered acceptable will n '="=....,....,......,....,,...,,..,,.,..,....,..,...,..,.,,...,,,....~,......,,.=~~c-=-r--'

e 1me o app,ca 10n.

72 QFFIGIAL ~SE OP4LY

PNNL-34230 Number Description Type Parameters Evaluation 1,)(4)

,~2(42 I MSSV flowlCbX4)

IMSSV flow ICb)C4) iMSSV flow ICb)C4)

(b)(4)

ICb)C4) 11s acceptable, l(b)(4)

CbX4)

IS acceptable when this system is performing a safety related function.

l (b)(4)

I Slowdown 1s the difference between set ressure and reseatm ressure of the MSSV.

)(4)

(b)(4)

---

~

(b)(4) 1s cons, j(b)(4)

~---===================-~

l (b)(4)

I (b)(-4) fb>c4) 1,s unimportant relative to the figures of merit. !(6)(4)

---

~ (b)(4) 1')(4) conserva 1ve y )(4)

I~

If r

ese w, nee e 1me o app 1ca 10n.

X4) 1,s acceptable l

(b)(4)

I (b)(4)

.=-~-)(=4)=====-=-=~==.-----------~lis considered acceptable.

ICbX4)

-=====================::::::;'

r X*J I The SEC piping r,x,,

l ------------'* rx*,

Technical Evaluation QFFIGIAb Uili 0NLY 73

PNNL-34230 Number Description Type Parameters Evaluation his approach is considered acceptable as the effects conservatisms X4)

(bX4)

~ )(-l)

I

[hY4) jUOUnamg ~ )(4)

I r X4J I

1,)(4)

See evaluation o X

4

)

(bX4)

See evaluation o (b)(4)

(b)(4) jCb)C4)

I This approach 1s considered acceptable but may require some additional justification specific to an application.

Technical Evaluation 74 QfifilCIJ\\L Uili ONLY

PNNL-34230 Table 7. Evaluation of TG Uncertainty Parameters Number Description Type Parameters Evaluation l

(b)(4)

I TCV/TGV ICb)(4)

(b)(4) b)(4) 11s considered acceptable P,)(4)

_________ ___.,__:i>~)C4~) ____________

1conservatisms 1....,,(b_)(4.... ) ________

-c (b)(4) lare conservatively ICbX4)

(b)(4) 1conservat1vely

!=ti,=)(4=)========::::::...-----'========'---------=-~

~i X' _) -------~I

>X4)

Technical Evaluation (b)(4)

(b)(4) are e ine acceptable.

Turbine Control Valve (TCV)/Turbine Governor Valve Conservative lant 1censm ocumentat1on. his a es roach is cons, ere tL..

ib;...;.

)(4..... )_~-~--~~~~-~~~~___.lthese will need to be reviewed and approved at the time of application.

(b)(4)

I

~(b)(4)

I Turbine Control Valve (TCV)/Turbine Governor Valve (TGV) µ>)(4)

I Conservative (b)(4)

(b)(4)

I TCV/TGV ICb)(4)

OFFICIJlcl USE OHLY 75

PNNL-34230 Number Description Type Parameters (b)(4)

Evaluation 1>)(4)

This approach is considered acceptable.

l(b)(4) lthe RCPs continue to receive power following a turbine-generator trip or loss of offsite power, this may influence DNB margins immediately following the trip. Continued operation of the RCPs may also affect 1>)(4)

(b)(4)

(b)(4)

This general approach is considered acceptable, (b)(4) ese wI nee Table 8. Evaluation of HUM Uncertainty Parameter Number Description Type Parameters Evaluation Some plants and/or events may rely upon additional operator actions to address potential single failures or to terminate the event. Any operator action credited during the first 10 minutes of the transient must be justified, considering the plant licensing basis and the plant operating procedures. (b)(4)

Cb)(4) perator actions are etermme y @ ant IcensIng ocumentat1on.

general approach is considered acceptable, ~ )(4)

IS ICb)(4)

Ithese will need to be reviewed and approved at the time of application.

Table 9. Evaluation of GCN Uncertainty Parameters Number Description (b)(4)

Technical Evaluation Type Parameters Evaluation Delayed neutron fraction is computed using ARTEMIS and validated indirectly through physics startup measurements utilizing a reactimeter. This approach is considered acceptable.

QfifilGIAb Y&li ONLY 76

PN NL-34230 Number Description Type Parameters Evaluation

)(4)

(6)(4)

(6)(4) rac,on X4) or conservatism. (6)(4)

e.

(6)(4)

Mean generation th'.1m= e~,s:--:-co.,.,m= p,..,,ut.:-:e....,,....,uc-::s..,..,m:--:g:----x-.....,......-1'1'?'<.=(b=X=4)==========---.:....This approach is considered acceptable~.------------~

lmean generation time. jCbX4)

This approach is considered acceptable (6)(4)

OTC are l,....

Cb-)(4_) _ ____::::...._ __ -==----____::==--__,

(6)(4) estimate using ARTEMIS and compared to measured data ~(6)(4)

The approach is considered acceptable.

r X4) i

~

I ~

~M=-=T::::C:.p,=)=(4=)=========.! =T-:-he-a-pp_r_o_a-:ch:--:-is_c_o_n_s,:-:.d:-e-re-:d:-a-c-c-ep-:t:-a-:-b:--'

le.

(6)(4)

I The approach 1s considered acceptable.

(6)(4)

(b)(--1)

Technical Evaluation 77 OFFIOl,tcl USE OHL¥

Number Description Type (b)(4)

Technical Evaluation PNNL-34230 Parameters Evaluation Reference in ANP-10338P-A mentions ICbX4) l(ANP-10297P-A, Revision 0, Supplement 1, "The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results Topical Report," June 2015.) !(bY4)

(b)(4)

This approach is considered acceptable.

ARTEMIS code is validated with total rod worth measurements during startup testing at HZP. )(4)

(b)(4)

(b)(4)

e.

Is approac Is cons, ered acceptable.

(b)(4) individual bank worthlCbX4)

(b)(4)

(b)(4)

Acceptance criteria is at 15%. This approac

.ICb=)(=4)=============

:;;:;:;:-;-:--:=:-::-:-:;::-;-:---:-:-::'.I critical boron concentration l(l,~_X4_) __

~

~ICb_>C_4> ______ ~IThis approach is considered acceptable.

Difference between measured and calculated must fall within ICb)(-1)

ICb)(4) f-RIT A methodology uses ICb)(4)

I This approach is considered acceptable.

)(4) to represent a conservative (4)

(4) critical boron

)(4)

)(4)

IS (b)(4) asymmetric conditions due to ICbX4) l(b)(-1) las~y_m_m

_ e_tr_,c_p_o_w_e_r_s~h_a_p-es-.--------~

Qf:f:ICIJ\\L Uili 0NLY 78

Number Description Type Parameters (b)(4)

Table 10. Evaluation of LCN Uncertainty Parameters Number Description Type Parameters CbX4)

Technical Evaluation PNNL-34230 Evaluation Asymmetries in core are a result of either asymmetrical rod configuration (such as a dropped rod) or differences in the reactor coolant inlet temperature between loops (possibly caused by a MSLB). )(4)

(bX4)

The most severe asymmetry is for the HZP post-trip MSLB event _{..,1i""'

)<4~)-----,--,-----,--,----,-~1 (1')(4)

!This approach is considered acceptable.

(b)(4)

'b)(4) 11 nis approach Is considered acceptable.

Decay heat based on l:b)(4l I

l(b)(4)

I ARTEMIS decay heat model initialized with ICb)(-4)

I CbX4)

Iwhich is conservative ICb)(4)

I (bX4)

I l(b)(4)

I This approach is considered acceptable.

)(4)

I Evaluation ICb)(4)

I based on the Plant Technical Specification p,

~~>(4_) ____ ~

f::l,)(4)

I This approach is considered a._c_c_e_p~ta~10~,,e-.-------------'

I,.,

(bX4)

I reviewed from

.'t,;=.--:;;:;-;:;-;:;-::;--;::;-:-::-;::-:-:-nri;-;:::::==========-----'

ANP-10297 Section 8. 3 1(b)(4)

I 79 QFFIGIAL ~SE QNLY

PNNL-34230 Number Description Type Parameters Evaluation ICb)(4) lis acceptable to determine l~Cb_X4_) ____ ~

w(4)

(4)

The approach is considered acceptable.

)(4)

)(4)

Technical Evaluation has no conservative (6)(4)

Similar approach ! ~Cb~

)(4~2--~l I his approach is considered acceptable.

jC6)(4)

I is acceptable given that t,)(4)

~

1Cb_)c4_) _~~his approach is considered acceptab're

.,.,e,...._----------~

~

Cb_X4_) _____________ --r--=--,--,-,-----!1 axial offset. ~@,=)(4_) --~

~(6_)(4_) ____________ ~ 1 within the allowed range. This approach is considered acceptable.

on supporting ocumentat1on.

ICbX4)

I Primarily a function of the water to fuel ratio (mostly constant for PWRs and no change for events with fixed geometries).

Value is defined in plant or other licensing documentation and confirmed,...(b-X4_)__,

OFFIOl,lcl USE ONLY 80

Number Description Type Parameters (b)(4)

Table 11. Evaluation of TH Uncertainty Parameters Number Description Type Parameters (b)(4)

Technical Evaluation PNNL-34230 Evaluation

~h>~X4~) __

~ lwith MCNP code calculations. This approach is considered acceptable.

The rod bow procedure sited has been approved for use and is found to be acceptable for this application. lCbX4) of the ~

.... _)(4_) ____________

b,)(4)

!Section 9.1.3.5 was vague. Framatome clarified that

~ )(4)

~ X4_) ____________________ ~

Iwas reviewed and found to be acceptable)

This clarified procedure !CbX4)

I is acceptable.

Evaluation (b)(4)

I how inlet flow ICb)(4)

(b)(4)

I (b)(4)

IIs acceptable (b)(4) 1 conservative l!b)(4)

(b)(4) 1ix 4) I Rod Bo~ X4)

(b)(4)

[o)(4)

I conservatively r,)(4) conservative P,)(4)

I The approach described in ANP-10339P RAI responses is acceptable and will result in a conservative prediction of Critical Heat Flux. ICb)C4)

I (b)(4) ti,)(4)

I Plant specific fuel information will be used in the evaluation for non-Framatome fuel.

I I

81 0FiFilCIAb ~SE OPJLY

Number Description Type CbX4)

PNNL-34230 Parameters Evaluation This parameter determines how the Departure from Nucleate Boiling Ratio (DNBR) margin and limit is calculated because they are both based on the critical heat flux predicted by these correlations. DNBR margin is the relation of the DNBR to the limit. Positive margin is favorable and negative margin is unfavorable.

The AR IT A meth.,;;.o..;;;.do

..;;.;l..:..o...,_,_

av.L t:11>-'-X'-'4) ________

____.l-=-ca.;;;.;l..:..c.;;;.ulc;;;a..;.;te;_;_;_he;;_;a:;;..;t....;.f;_;;;luc....;x....;;a;.;_n..:..d....;t....;.he..;;.,

critical heat flux. t,i,)(4)

(b)(4)

(b)(4) 11s an acceptable method. ICb)(4)

(b)(4)

I a reasonable 95/95 calculated DNBR will be achieved. k:6)(4)

(b)(4) rx*,

I ICb)(4) l t:Secause the CHF correlation is fuel assembly specific, a different correlation will e used for different analysis or in a single analysis for a mixed core. In each of these correlations

)(4)

)(4)

This is consistent with other approved methodology and will provide a reasonable result given the relative spread of CHF data.

Table 12. Evaluation of THAP Uncertainty Parameters Number Description Type Parameters Evaluation I "

ICb)(4)

I Hot Channel Factor (HCF) do Technical Evaluation 82 OFFICIJlcl USE OHLY

PNNL-34230 Number Description Type Parameters Evaluation CbX4)

CbX4)

CbX4)

CbX4) 1conservative!Cb)(4)

I (b)(4)

IHCF ICb)(4)

(b)(4)

I is an acceptable representation of these factors l:b)(4)

CbX4)

CbX4)

Table 13. Evaluation of FRR Uncertainty Parameters Number Description Type Parameters Evaluation (b)(4)

(b)(4)

bX4)

I Framatome has provided ICb)(4)

Technical Evaluation 83 QfifilGIAb YSli ONLY

PNNL-34230 Number Description Type Parameters Evaluation rbX4) j(bX4)

I (RAl-5). The ARITA methodology ~)(4)

I j(b)(4)

I The use of r,)(4)

I is acceptable.

l(bX4)

I The calculational procedure was reviewed and tound to be acceptable.

This j(bX4)

!was reviewed and accepted during the acceptance of the GALILEO code and method. ICbX4)

Kb)C4) rb)(4)

!are also acceptable for use within the ARITA methodology.

This t:1'X4) lthat was accepted for GALILEO (see FRR-7d).!(b)(4)

I l(bX4) 1:ire acceptable for use within l(b)C4)

I the ARITA methodology.

Thisl(b)(4) r,,as reviewed and acce12ted during the acce12tance of the GALILEO code and method. ~)(4)

I J'.b)(4)

~)(4)

I are also acceptable for use within the ARITA methodology.

It is conservative j(bX4) l(bX4) acceptable for use within the ARITA methodology.

~re also This ~X--t)

I was reviewed and accepted durinQ the acceptance of the GALILEO code and method. r")C4) ib1r.1,

b,.--t 1are also acceptable for use within the ARITA me1noao1ogy.

This approach 1(b)(4)

I gap conductivity j(b)(4) 1is consistent withl(b)(4) i(b)(4) kt,)(4)

I Gap conductivity is impacted by a large I

Technical Evaluation 84 OFFIOIAL ~SE ONLY

Number Description Type Technical Evaluation PNNL-34230 Parameters Evaluation number of fuel performance parameters @)C4)

X4) is acceptable for use within the ARITA methodology.

gap conductivity (b)(4) is (bX4) 1s more an a equa e o ound j(bX4)

I Gap conductivity is impacted by a large number of fuel performance parameters P,)C4)

I (b)(4)

]._(b_)C4.,...)--,r-r-------~I acceptable for use within the ARITA methodology.

It is conservative (b>C4>

are also accepta e or use wit

)(4)

(b)(4) me o oogy.

This P,)C4) lwas reviewed and accepted during the acceptance of the GALILEO code and method. ](b)(4)

(b)(4)

(b)(4)

I are also acceptable for use within the ARITA methodology.

This P,)C4)

I was reviewed and accepted during the acceptance of the GALILEO code and method. (b)(4)

b)(4) j'.b)(4)

I are also acceptable for use within the ARITA methodology.

This ['b)(4)

I was reviewed and accepted during the acceptance of the GALILEO code and method.l(bX4)

OFFIOl,lcl USE ONLY 85

Number Description Type (b)(4)

Technical Evaluation Parameters PNNL-34230 Evaluation

)(4)

'1,)(4) are also acceptable for use within the ARITA

~m_e_t~o~ o-o_g_y __ ----~

This ["'X4) 1was reviewed and accepted during the acceptance of the GALILEO code and method. ICb)(4)

(b)(4)

~(b_X4_) _______ ~ 1 are also acceptable for use within the ARITA methodology.

The collection of process data f>)<4) f1,X4)

!was reviewed and accepted during the acceptance of the GALILEO code and method. piX4)

)(4) are a so accepta e or use wit m The collection of process data (b)(4) j(b)(4) 1was reviewed and accepted during the acceptance of the GALILEO code and method. l(b)(4)

(b)(4)

I are also acceptable for use within the ARITA methodology.

The collection of process data (b)(4) f0 )(4)

I was reviewed and accepted during the acceptance of the GALILEO code and method. P,X4)

(b)(4)

(b)(4) are also acceptable for use within

'-:------:-=

---:---.----:---:---------~

the ARITA methodology.

The collection of process data ~)(4_) ________________

~

l(b )( 4) twas review~.arJJQ..a~illl!~QilllllQLillL _ _,

acceptance of the GALILEO code and method. Cb)C4) 86 QFFIGIAb YSli QNbY

PNNL-34230 Number Description Type Parameters Evaluation (b)(4)

(b)(4) are also acceptable for use within e

me o oogy.

)(4)

The collection of process data Cb)C4)

(b)(4) was reviewed and accepted during the acceptance of the GALILEO code and method.

)(4)

b)(4)

~1,""'X4""'")~,,__.--___,,,--.,......... _______ ____,! are also acceptable for use within the ARITA methodology.

The collection of process data Cb)C4)

ICb)(4) jwas reviewed and accepted during the acceptance of the GALILEO code and method. ICbX4) r"

bX4) f are also acceptable tor use w1tnm the ARITA methodology.

The collection of process data ~x4_-) ________________

~

j'.b)C4)

I was reviewed and accepted during the acceptance of the GALI LEO code and method )(4)

(b)(4)

[:b)(4) jare also acceptable for use within the ARITA methodology.

The collection of process data f>)(4)

P,)(4)

I was reviewed and accepted during the acceptance of the GALILEO code and method. !Cb)(4)

(b)(4)

The collection of process data :b)(4)

!Cb)(4)

I was reviewed and accepted during the acceptance of the GALILEO code and method. =kk~X~4) _________

(b)(4)

Technical Evaluation 87 OFFIGIAb blili 0NLY

Number Description Type (b)(4)

Technical Evaluation PNNL-34230 Parameters Evaluation

~_X_4)-,-:'"'="=--_,,,--..,_,..---------'lare also acceptable for use within the ARITA methodology.

The collection of process data jCbX4)

(b)(4)

I was reviewed and accepted durinq the acceptance of the GALILEO code and method. p,)(4)

(b )(4)

(b)(4)

I are also acceptable for use within the ARITA methodology.

(b)(4)

I was reviewed and accepted durinq the acceptance of the GALILEO code and method. l::b)C4)

CbX4) l(b)(4) ps also acceptable for use within the ARITA methodology.

The use of the manufacturing tolerance ICb)C4)

I l(b)(4)

I was reviewed and acceeted during the acceetance of the GALILEO code and method. jCbX4)

I X'J (b)(4) 1aie also acceptable tor use within the ARI I A methodology.

The collection of process data Yl>X4)

I ICb)(4)

I was reviewed and accepted during the acceptance of the GALILEO code and method. j(b)(4)

(b)(4)

KJ,)(4)

I are also acceptable for use within the ARITA methodology.

The collection of process data P,X4)

[o)(4) 1was reviewed and accepted during the acceptance of the GALILEO code and method. l!b)(4)

(b)(4)

QFFIGIAb YSli QNbY I

I I

I 88

Number Description Type Parameters PNNL-34230 Evaluation

.,.,.!Cb_X4_)-:-==-=-

---

,,-----,---,--------~f are also acceptable for use within the ARITA methodology.

Table 14. Event Initiators for Transients performed with ARITA SRP#

Event 15.1.1 Decrease in Feedwater Temperature 15.1.2 Increase in Feedwater Flow 15.1.3 Increase in Steam Flow 15.1.4 Inadvertent Opening of a I

Steam Generator Technical Evaluation Event Initiator p)(4)

Treatment PDF/Range/

Truncation or Bias OFFICIAi I !SE QNI Y Evaluation (b)(4)

(b)(4)

IPNNL concludes thisj(bX4) ps acceptable.

(bX4)

PNNL concludes thisj(b)(4) js acceptable.

(b)(4)

(b)(4)

IPNNL concIuaesI(b)(4)

I (b)(4)

Its acceptable.

(b)(4)

CbX4)

IPNNL concludes ICb)(4)

I (b)(4) 1s acceptable.

rt,)(4) 89

SRP #

Event Event Initiator Relief or Safety

)(4)

Valve 15.1.5 Main Steam Line Break 15.2.1 Loss of External Load 15.2.2 Turbine Trip 15.2.3 Loss of Condenser Vacuum Technical Evaluation Treatment PDF/Range/

Truncation or Bias Qf:f:ICIJ\\L Uili QNL¥ PNNL-34230 Evaluation

b)(4)

!Cb)(4)

I PNNL concludes this Is acceptable.

b)(4)

PNNL concludes s accep a e.

PNNL concludes 90

SRP #

Event Event Initiator 15.2.4 Closure of p)(4)

Main Steam Isolation Valve 15.2.6 Loss of Nonemergency AC Power to the Station Auxiliaries 15.2.7 Loss of Normal Feedwater Flow 15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment 15.3.1 Loss of Forced Reactor Coolant Flow Technical Evaluation Treatment PDF/Range/

Truncation or Bias Qf:f:ICIJ\\L Uili QNL¥ PNNL-34230 Evaluation

'.b)(4)

(b)(4) ti PNNL concludesl(b)(4) s acceptable.

(b)(4)

PNNL concludes this is acceptable.

b)(4)

K\\>)(4) iPNNL concludes this ~

1s acceptarne.

(b)(4)

b)(4)

IPNNL concludes thisl(b)(4) ps acceptable.

~)(4)

(b)(4)

IPNNL concludes this is accepiao1e.

1>)(4) 91

SRP#

Event Including Trip of Pump Motor 15.3.3 Reactor Coolant Pump Rotor Seizure 15.3.4 Reactor Coolant Pump Shaft Break 15.4.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition 15.4.2 Uncontrolled Control Rod Assembly Technical Evaluation Event Initiator Treatment PDF/Range/

Truncation or Bias 6FFlel)!(L USE 6NLY PNNL-34230 Evaluation

)(4)

PNNL concludes this is acceptable.

CbX4)

X4)

PNNL concludes this

,s accepta e.

(bX4)

(b)(4) accep a e.

(b)(4) b)(4)

IPNNL concludes this is acceptable.

)(4)

)(4)

PNNL 92

SRP #

Event Withdrawal at Power 15.4.3 Control Rod Misoperation (Single Rod W/D) 15.4.3 Control Rod Misoperation (Dropped Rod/Bank)

Technical Evaluation Event Initiator (b)(4)

Treatment PDF/Range/

Truncation or Bias OFFICIJlcl USE ONLY PNNL-34230 Evaluation oosition is between the bite oosition and (b)(4)

I I

I I

b)(4)

IPNNL concludes this 1s acceptarne.

'1JX4)

PNNL concludes this is acceptable.

'1,)(4) l(b)(4)

I PNNL

\\.,VI 11.,1uu,:,;:, ll "" IS Ql.,l.,<:,1-'lCIUI<:,.

93

SRP #

Event Event Initiator 15.4.3 Control Rod CbX4)

Misoperation (Misaligned Rod) 15.4.4 Startup of an Inactive Loop or Recirculation Loop at an Incorrect Temperature 15.4.6 Inadvertent Decrease in Boron Concentration in the Reactor Coolant System (PWR) 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position 15.5.1 Inadvertent Operation of ECCS Technical Evaluation Treatment PDF/Range/

Truncation or Bias PNNL-34230 Evaluation p)(4)

ICbX4)

IPNNL concludes this is I

accepta61e.

'.b)(4)

ICb)(4)

IPNNL concludes this is acceptable.

e (4)

I

~ )(4)

IPNNL concludes this bias is acceptable.

ri,)(4)

PNNL concludes thisr'X4) Jis acceptable.

rt,)(4) 94 0FFl61Ab l:Jili 0~1LY

SRP #

Event 15.5.2 Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve 15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment Technical Evaluation Event Initiator (b)(4)

Treatment PDF/Range/

Truncation or Bias QFFIGIAb YSli QNb¥ PNNL-34230 Evaluation

~1,)(4) r NNL I

\\.,VI " *" "

u,1::; 1::; i::lvvO::f.JlCIUIO:,.

(b)(4)

~X4>

f NNL considers this ICb)C4)

I ccepmme.

(b)(4) 1,)(4)

IPNNL concludes this is acceptable.

(b)(4)

(b)(4) iPNNL considers thisl'.b)(4)

I acceptable.

b)(4) lCb)(4)

I PNNL concludes this~

is acceptable.

L X4)

I l(b)(4) acceptable. F NNL concludes this is 95

SRP#

Event Event Initiator 15.6.3 Rad iolog ica I (b)(4)

Consequences of Steam Generator Tube Failure Technical Evaluation Treatment PDF/Range/

Truncation or Bias OFFICIAL USE ONLY PNNL-34230 Evaluation

b)(4)

(b)(4)

I PNNL concludes this,s acceptable.

96

PNNL-34230 3.8.4 Application of Non-Parametric Statistics for Licensing Analyses As clarified in RAl-16, Framatome is requesting generic NRC approval relative to the capability of the ARITA methodology to analyze the acceptance criteria identified in Table 1 for each event in Table 2. Any changes to the acceptance criteria will be justified in a plant submittal implementing the ARITA methodology.

Framatome further clarified in RAl-17 that the ARITA methodology should not be rigidly defined, but rather should be seen as a methodology that provides a framework within which to perform safety analyses, the specific details of which are to be reviewed and approved in the plant specific License Amendment Request (LAR) implementing the ARITA method. It is PNNL's judgement that approval of such a methodology will put a greater technical review burden on the review of approval of the LAR implementing the ARITA method than previous methods had.

There are a number of items that must be adequately described in the ARITA methodology for the application of the ARITA uncertainty analysis for plant licensing analyses. Many of these items were not adequately described in the original L TA and were addressed through the RIA process. These items and PNNL's review of their acceptability are described in the sections below.

3.8.4.1 Disposition of events The ARITA LTR has provided a list of target scenarios and which EM will be used to address each scenario (Table 2). However, some of the discussion in the L TR seems to indicate that each event would not always be explicitly modeled, but that it would be possible to develop a composite case that would bound multiple events. RAl-18 requested more information on the increase in steam flow events ( 15.1.1-15.1.4) and RAl-19 requested more information on the process that is used to decide that some events are more limiting than others.

Framatome clarified that they do intend to identify limiting events in some cases including cases with increase in steam flow. Framatome will rely on results from (bX-1)

(b)(4)

The results of this disposition will be described for each plant in the LAR to use ARITA.

PNNL concludes that it is generally acceptable to identify limiting events, but a sound technical basis for the conclusion that one event is more limiting than another should be included in the LAR.

3.8.4.2 Cycle-specific changes that require ARITA re-analysis The uncertainty analyses described in the preceding sections will be used in the cycle specific safety analysis that is performed prior to each reload. In many cases a general ARITA analysis will be performed in advance of each cycle and as part of the cycle-specific specific safety-analysis it will be confirmed that the analysis of record is applicable. However, it is acknowledged that it is possible for plant configuration and operation to deviate sufficiently from the underlying bases and assumptions of the safety analysis of record such that the analysis is no longer applicable. Because of this, PNNL and NRC asked in RAl-13 for Framatome to provide a methodology to determine what changes will trigger a re-analysis using ARITA.

Technical Evaluation 9FFIGIAL Y&i! 9Nb¥ 97

PNNL-34230

t,)(4)

Cb)C4)

I If a plant has the potential to deviate from the assumed operation for an extended period of time, the licensee is responsible for informinq Framatome so that the impact on the safety analvsis can be assessed. jCb)(4)

I KbX4) b)(4)

I PNNL concludes that the general approach to dispositioning changes is acceptable.

3.8.4.3 Discrepancies between plant licensing basis and the standard review plan As discussed in the start of this section Framatome has requested that the ARITA methodology should not be rigidly defined, but rather should be seen as a methodology that provides a framework within which to perform safety analyses. This is necessary as the licensing basis of each plant is not always consistent with the standard review plan. PNNL and NRC asked Framatome in RAl-15 and RAl-21 to provide more detail on how ARITA will be used to perform these analyses. Consistent with their requested use of ARITA, Framatome will apply ARITA to the plant licensing basis and in areas where the licensing basis and the standard review plan are different, the justification for the application of ARITA will be provided in the LAR, which again is an increased regulatory review burden for the LAR review.

3.8.4.4 Events that challenge multiple figures of merit There are some events that could challenge more than one figure of merit. Additionally, due to the stochastic nature of ARITA, it is possible that each realization may predict failure by exceeding different figures of merit. Therefore, PNNL and NRC asked Framatome in RAl-8 to clarify if ARITA would make a statistical statement on individual acceptance criteria or a general statement that no acceptance criteria would be exceeded. After discussion it was concluded that f~r,event~ t~at ICbX4)

I r )(4)

The ARITA methodology has been modified as described in RAl-8 to also make jCb>C4)

I (6)(4)

PNNL has reviewed this approach and has determined that it is acceptable to make f X4)

(b)(4)

(6)(4)

I Technical Evaluation Ol=l=ICIAL IISE ONI Y 98

PNNL-34230 3.8.4.5 Reanalysis when limits are not met While performing a safety analysis using AR~I_T_A_t_he_r_e_i_s_a~----------~

that a SAFDL will be exceeded. In this cas Cb)C4) 1 Cb)C4>

I PN NL and '":-N=R=-cc=-a-s.,...ke-d-:--=-fo_r_c-=-1a-r=ifi=-,c-a.,.,.tio_n_ o_n....,.h_o_w_t:-:--h..,..is...,.Cb_)_C4)---~

Cb)C4)

I (RAl-12). Due to the stochastic nature of ARITA, it is possible to get slightly different upper tolerance levels from different analyses that are started with different random seeds. Therefore, it should be ensured that if a change is made, the acceptability of that change is not just based in variability within the stochastic process.

Framatome stated that in the event the analysis shows failure to meet acceptance criteria,P,)C4)

(b)(4)

(b)(4)

I PNNL has reviewed this approach and finds it acceptable to ensure that the design change will result in acceptable SAFDL performance and not stochastic variation in the method.

Technical Evaluation OFFICIAL YSE ONLY 99

PNNL-34230 4.0 Conclusions The ARITA LTR has been reviewed and the following conclusions have been made.

The figures of merit (FoM) and acceptance criteria are identified for this submittal (Section 3.2).

The necessary documentation to review the L TR has been provided (Section3.3)

Reactor applicability of code has been provided (Section3.4)

Transient and accident scenarios that the code and methodology will be used to perform safety analyses on have been provided (Section 3.4)

The four main phenomenological evaluation models in ARITA; thermal hydraulics (S-RELAP5), neutronics (ARTIMIS Nodal and ARTIMIS THM), systems modeling (S-RELAP5), and thermal-mechanical fuel performance (GALILEO) have been reviewed and have been found to be acceptable (Section 3.5)

The identification of target accident scenarios and phenomena identification and ranking is acceptable (Section 3.6)

The ARITA code assessment relative to applicable data has been reviewed and found to be acceptable (Section 3.7)

The ARITA Uncertainty analysis including; treatment of uncertainties for highly ranked phenomena, the non-parametric uncertainty analysis process, the uncertainty parameters for the non-parametric uncertainty analysis, and the application of the non-parametric statistics for licensing analyses have been reviewed and have been found to be acceptable within the scope of application that was requested (Section 3.8)

The following limits of application are recommended for ARITA

1. Framatome requested that the ARITA methodology not be rigidly defined, but rather should be seen as a methodology that provides a framework within which to perform safety analyses. Therefore, the specific details of the modeling selections made which deviate from the general structure presented in the ARITA L TR should be reviewed and approved in the plant specific License Amendment Request (LAR) implementing the ARITA method.

2. Framatome shall use the uncertainty parameters given in the tables in Section 3.8.3 where values are given. Were values and parameters are taken from P,X4)

I (b)(4) adequate justification for these values should be included for review and approval in the plant specific License Amendment Request (LAR) implementing the ARITA method.

3. When developing plant specific distributions for uncertainty parameters the following guidance should be adhered to

a. :t>X4)

b.

C.

d.

Conclusions OFFIGIAL USE ONLY 100

PNNL-34230

4. The PIRT discussed in Section 3.6.2 provided useful information regarding the identification of important parameters but should not be considered part of the approved ARITA methodology for future submittals Conclusions OFFIOl~L YGE ONLY 101

PNNL-34230 5.0 References

1. ANP-10339P, Revision 0, "ARITA -ARTEMIS/RELAP Integrated Transient Analysis Methodology Topical Report," August 2018

2. NUREG-0800, "Standard Review Plan for Light Water Reactors, Chapter 15," March 2007.

3. NUREG-0800, "Standard Review Plan for Light Water Reactors, Chapter 4.2," March 2007.

4. Letter, Jonathan G. Rowley (NRC) to Gary Peters (Framatome Inc.), "Request for Additional Information Regarding Framatome Inc. Topical Report ANP-10339P, Revision 0, 'ARITA-ARTEMIS/RELAP Integrated Transient Analysis Methodology' (EPID: L-2018-TOP-0034),"

December 2019

5. Letter, Jonathan G. Rowley (NRC) to Gary Peters (Framatome Inc.), "Request for Additional Information Regarding Framatome Inc. Topical Report ANP-10339P, Revision 0, 'ARITA-ARTEMIS/RELAP Integrated Transient Analysis Methodology'," (EPID: L-2018-TOP-0034),

April 2020.

6. ANP-10339Q2P, Revision 0, "Response to Request for Additional Information - ANP-10339P," March 2020.

7. ANP-10339Q3P, Revision 0, "Response to Request for Additional Information -ANP-10339P," July 2020.

8. ANP-10339Q4P, Revision 0, "Response to Request for Additional Information -ANP-10339P," November 2020

9. ANP-10339Q5P, Revision 0, "Response to Request for Additional Information -ANP-10339P," June 2021

10. ANP-10339O6P, Revision 0, "Response to Request for Additional Information - ANP -

10339," June 2022

11. NUREG-0800, "Standard Review Plan for Light Water Reactors," March 2007

12. U. S. Nuclear Regulatory Commission NRC Regulatory Guide 1.203, " Transient and Accident Analysis Methods".

13. ANP-10339O1 P, Revision 0, "Response to Request for Supplemental Information -ANP-10339P," March 2019

14. ANP-10323P, Revision 1, "GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors," June 2018.

15. Wilks, S.S., "Determination of Sample Sizes for Setting Tolerance Limits," Ann. Math. Stat.,

Vol. 29, No. 2, pp. 599-601, June 1958.

16. Geelhood, K.J et al, "FAST-1.1: A Computer Code for the Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients," PNNL-32770, April 2022.

17. Geelhood, K.J. et al, "FRAPCON-4.0: A Computer Code for the Calculation of Steady-State, Thermal-Mechanical Behavior of Oxide Fuel Rods for High Burnup," PNNL-19418, Vol.1,

Rev.2, September 2015.

References 01-lrlCIAI. USE O~ILY 102

Pacific Northwest National Laboratory 902 Battelle Boulevard P.O. Box 999 Richland, WA 99354 1-888-375-PNNL (7665) www.pnnl.gov