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Publications Resulting from International Agreements

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Publications resulting from international agreements and overseen by NRC staff. Other International Agreements may be available in ADAMS.

Document Identifier Title
NUREG/IA-0001 Assessment of TRAC-PD2 Using SUPER CANNON and HDR Experimental Data
NUREG/IA-0002 Heat Transfer Processes During Intermediate and Large Break Loss-of-Coolant Accidents (LOCAs)
NUREG/IA-0003 Influence of the Wetting State of a Heated Surface on Heat Transfer and Pressure Loss in an Evaporator Tube
NUREG/IA-0004 Thermal Mixing Tests in a Semiannular Downcomer With Interacting Flows from Cold Legs
NUREG/IA-0005 Assessment of RELAP5/MOD2, Cycle 36, Against FIX-II Split Break Experiment No. 3027.
NUREG/IA-0006 Assessment of RELAP5/MOD2 Against Marviken Jet Impingement Test 11 Level Swell
NUREG/IA-0007 Assessment of RELAP5/MOD2 Against Critical Flow Data From Marviken Tests JIT 11 and CFT 21.
NUREG/IA-0008 Assessment Study of RELAP-5 MOD-2 Cycle 36.01 Based on the DOEL-2 Steam Generator Tube Rupture Incident of June 1979
NUREG/IA-0009 Assessment of RELAP5/MOD2 Against 25 Dryout Experiments Conducted at the Royal Institute of Technology
NUREG/IA-0011 TRAC-PF1 MOD1 Post Test Calculations of the OECD LOFT Experiment LP-SB-1
NUREG/IA-0012 RELAP/MOD2 Calculations of OECD-LOFT Test LP-SB-01
NUREG/IA-0013 RELAP5/MOD2 Calculation of OECD-LOFT Test LP-SB-03
NUREG/IA-0014 Analysis of the THETIS Boil Down Experiments Using RELAP5/MOD2.
NUREG/IA-0015 Assessment of Interphase Drag Correlations in the RELAP5/MOD2 and TRAC-PF1/MOD2 Codes
NUREG/IA-0016 Assessment of RELAP5/MOD2, Cycle 36.04 Against FIX-II Guillotine Break experiment No. 5061
NUREG/IA-0018 RELAP5/MOD2 Assessment, OECD-LOFT Small Break Experiment LP-SB-03
NUREG/IA-0019 TRAC-PF1/MOD1 Post-Test Calculations of the OECD [Organisation for Economic Co-operation and Development] LOFT Experiment LP-SB-2
NUREG/IA-0020 Assessment Study of RELAP5/MOD2, CYCLE 36.04 Based on Spray Start-up Test for DOEL-4
NUREG/IA-0021 RELAP5/MOD2 Calculations of OECD LOFT Test LP-SB-2
NUREG/IA-0022 TRAC-PF1/MOD1 Post-Test Calculations of the OECD LOFT Experiment LP-SB-3
NUREG/IA-0024 Application of RELAP5/MOD3.1 Code to the LOFT Test L3-6
NUREG/IA-0025 RELAP5/MOD3 Subcooled Boiling Model Assessment
NUREG/IA-0027 TRAC-PF1/MOD1 Calculations of LOFT experiment LP-02-6
NUREG/IA-0028 Review of LOFT [Loss-of-Fluid Test] Large Break Experiments [OECD LOFT project]
NUREG/IA-0029 Assessment of RELAP5/MOD2, Cycle 36.04 Against FIX-II Split Break Experiment No. 3051
NUREG/IA-0030 Assessment of RELAP5/MOD2 Code Using Loss of Offsite Power Transient Data of KNU [Korea Nuclear Unit] No. 1 Plant
NUREG/IA-0031 ICAP [International Code Assessment and Applications Program] Assessment of RELAP5/MOD2, Cycle 36.05 Against LOFT [Loss of Fluid Test] Small Break Experiment L3-7
NUREG/IA-0032 Assessment of RELAP5/MOD2, Cycle 36-04 Using LOFT [Loss of Fluid Test] Large Break Experiment L2-5
NUREG/IA-0033 Assessment of RELAP5/MOD2, Cycle 36.04 Against LOFT Small Break Experiment L3-6
NUREG/IA-0034 Assessment Study of RELAP5/MOD2 Cycle 36.04 Based on Pressurizer Safety and Relief Valve Tests
NUREG/IA-0036 Analysis of LOBI Test BLO2 (Three Percent Cold Leg Break) with RELAP5 Code
NUREG/IA-0037 Assessment of RELAP5/MOD2, Cycle 36.04 Against LOFT Small Break Experiment L3-5
NUREG/IA-0038 Assessment of TRAC-PF1/MOD1 Against an Inadvertent Feedwater Line Isolation Transient in the Ringhals 4 Power Plant
NUREG/IA-0040 Boil-Off Experiments with the EIR-NEPTUN Facility: Analysis and Code Assessment Overview Report
NUREG/IA-0041 Assessment of TRAC-PF1/MOD1 Against an Inadvertent Steam Line Isolation Valve Closure in the Ringhals 2 Power Plant
NUREG/IA-0042 Dispersed Flow Film Boiling: An Investigation of the Possibility to Improve the Models Implemented in the NRC Computer Codes for the Reflooding Phase of the LOCA
NUREG/IA-0043 Assessment Study of RELAP5/MOD2 Cycle 36.04 Based on the DOEL-4 Manual Loss of Load Test of November 23, 1985
NUREG/IA-0044 Assessment Study of RELAP5/MOD2 Cycle 36.05 Based on the Tihange-2 Reactor Trip of January 11, 1983
NUREG/IA-0045 Assessment of RELAP5/MOD2 Using LOCE Large Break Loss-of-Coolant Experiment L2-5
NUREG/IA-0046 Assessment of RELAP5/MOD2 Using Semiscale Large Break Loss-of-Coolant Experiment S-06-3
NUREG/IA-0047 Assessment of RELAP5/MOD2 Cycle 36.04, Against the Loviisa–2 Stuck-Open Turbine By-Pass Valve Transient on September 1, 1981
NUREG/IA-0049 Thermal-Hydraulic Post-Test Analysis of OECD LOFT LP–FP–2 Experiment
NUREG/IA-0050 TRAC–PF1 Code Assessment Using OECD LOFT LP–FP–1 Experiment
NUREG/IA-0051 Assessment Study of RELAP5/MOD2 Cycle 36.05 Based on the DOEL 4 Reactor Trip of November 22, 1985
NUREG/IA-0052 An Analysis of Semiscale Mod–2C S–FS–1 Steam Line Break Test Using RELAP5/MOD2
NUREG/IA-0064 Analysis of Semiscale Test S–LH–1 Using RELAP5/MOD2
NUREG/IA-0065 Analysis of Semiscale Test S–LH–2 Using RELAP5/MOD2
NUREG/IA-0066 RELAP5/MOD2 Analysis of LOFT Experiment L9–4
NUREG/IA-0067 Recirculation Suction Large Break LOCA Analysis of the Santa Maria De Garoña Nuclear Power Plant Using TRAC–BF1 (G1J1)
NUREG/IA-0068 Assessment of the "One Feedwater Pump Trip Transient" in Cofrentes Nuclear Power Plant With TRAC–BF1
NUREG/IA-0069 Assessment of RELAP5/MOD2 Cycle 36.04 Using LOFT Intermediate Break Experiment L5–1
NUREG/IA-0070 Assessment of RELAP5/MOD2 Cycle 36.04 with LOFT Large Break LOCE L2–3
NUREG/IA-0071 Analysis of the UPTF Separate Effects Test 11 (Steam-Water Countercurrent Flow in the Broken Loop Hot Leg) Using RELAP5 /MOD2
NUREG/IA-0072 LOFT Input Dataset Reference Document for RELAP5 Validation Studies
NUREG/IA-0073 Time Step and Mesh Size Dependencies in the Heat Conduction Solution of a Semi-Implicit, Finite Difference Scheme for Transient Two-Phase Flow
NUREG/IA-0074 RELAP5/MOD2 Post-Test Calculation of the OECD LOFT Experiment LP-SB-1
NUREG/IA-0075 RELAP5/MOD2 Analysis of a Postulated "Cold Leg SBLOCA" Simultaneous to a "Total Black-Out" Event in the José Cabrera Nuclear Station
NUREG/IA-0087 RELAP5/MOD2 Post-Test Calculation of the OECD LOFT Experiment LP–SB–2
NUREG/IA-0088 Post-Test-Analysis and Nodalization Studies of OECD LOFT Experiment LP–02–6 With RELAP5/MOD2 CY36–02
NUREG/IA-0089 Post-Test-Analysis and Nodalization Studies of OECD LOFT Experiment LP–LB–1 With RELAP5/MOD2 CY36–02
NUREG/IA-0090 Assessment of RELAP5/MOD2 Using the Test Data of REWET-II Reflooding Experiment SGI/R
NUREG/IA-0091 Assessment of RELAP5/MOD2 Against a Natural Circulation Experiment in Nuclear Power Plant Borssele
NUREG/IA-0092 Assessment of RELAP5/MOD2 Computer Code Against the Net Load Trip Test Data From Yong–Gwang, Unit 2
NUREG/IA-0093 RELAP5/MOD3 Assessment for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads
NUREG/IA-0094 Assessment of RELAP5/MOD3 Against Twenty-Five Post-Dryout Experiments Performed at the Royal Institute of Technology
NUREG/IA-0095 RELAP5 Assessment Using LSTF Test Data SB–CL–18
NUREG/IA-0096 Numerics and Implementation of the UK Horizontal Stratification Entrainment Off-Take Model Into RELAP5/MOD3
NUREG/IA-0099 RELAP5 Assessment Using Semiscale SBLOCA Test S–NH–1
NUREG/IA-0100 Assessment of CCFL Model of RELAP5/MOD3 Against Simple Vertical Tubes and Rod Bundle Tests
NUREG/IA-0103 Assessment of BETHSY Test 9.1.b Using RELAP5/MOD3
NUREG/IA-0104 RELAP5/MOD3 Assessment Using the Semiscale 50% Feed Line Break Test S–FS–11
NUREG/IA-0105 Assessment of RELAP5/MOD3 Version 5m5 Using Inadvertent Safety Injection Incident Data of Kori Unit 3 Plant
NUREG/IA-0106 Assessment of PWR Steam Generator Modelling in RELAP5/MOD2
NUREG/IA-0107 Assessment of RELAP5/MOD2 Against a Load Rejection From 100% to 50% Power in the Vandellos II Nuclear Power Plant
NUREG/IA-0108 Assessment of RELAP5/MOD2 Against a Turbine Trip From 100% Power in the Vandellos II Nuclear Power Plant
NUREG/IA-0109 Assessment of RELAP5/MOD2 Against a 10% Load Rejection Transient from 75% Steady State in the Vandellós II Nuclear Power Plant
NUREG/IA-0110 Assessment of RELAP5/MOD2 Against a Main Feedwater Turbopump Trip Transient in the Vandellos II Nuclear Power Plant
NUREG/IA-0112 Assessment of RELAP5/MOD2 Against ECN-Reflood Experiments
NUREG/IA-0113 Preliminary Assessment of PWR Steam Generator Modelling in RELAP5/MOD3
NUREG/IA-0114 Assessment of RELAP5/MOD3 With the LOFT L9–1/L3–3 Experiment Simulating an Anticipated Transient With Multiple Failures
NUREG/IA-0116 Assessment of RELAP5/MOD3/V5m5 Against the UPTF Test No. 11 (Countercurrent Flow in PWR Hot Leg)
NUREG/IA-0118 Analysis of LOFT Test L5–1 Using RELAP5/MOD2
NUREG/IA-0119 Assessment and Application of Blackout Transients at Asco Nuclear Power Plant with RELAP5/MOD2
NUREG/IA-0120 Assessment of the Turbine Trip Transient in Cofrentes NPP with TRAC–BF1
NUREG/IA-0121 Assessment of a Pressurizer Spray Valve Faulty Opening Transient at Asco Nuclear Power Plant with RELAP5/MOD2
NUREG/IA-0122 Assessment of MSIV Full Closure for Santa Maria De Garoila Nuclear Power Plant Using TRAC-BFl (G1J1)
NUREG/IA-0123 Application of Full Power Blackout for C. N. Almaraz with RELAP5/MOD2
NUREG/IA-0124 Assessment of RELAP5/MOD2 Against a Pressurizer Spray Valve Inadverted Fully Opening Transient and Recovery by Natural Circulation in Jose Cabrera Nuclear Station
NUREG/IA-0125 Assessment of RELAP5/MOD2 Computer Code Against the Natural Circulation Test Data from Yong–Gwang Unit 2
NUREG/IA-0126 2D/3D Program Work Summary Report
NUREG/IA-0127 Reactor Safety Issues Resolved by the 2D/3D Program
NUREG/IA-0128 International Code Assessment and Applications Program: Summary of Code Assessment Studies Concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC–B
NUREG/IA-0129 An Assessment of the CORCON-MOD3 Code Part I: Thermal-Hydraulic Calculations
NUREG/IA-0130 Assessment of RELAP5/MOD3.1 With the LSTF SB-SG-06 Experiment Simulating a Steam Generator Tube Rupture Transient
NUREG/IA-0131 Assessment of RELAP5/MOD3 Using BETHSY 6.2TC 6-Inch Cold Leg Side Break Comparative Test
NUREG/IA-0132 Improvements to the RELAP5/MOD3 Reflood Model and Uncertainty Quantification of Reflood Peak Clad Temperature
NUREG/IA-0133 Development, Implementation, and Assessment of Specific Closure Laws for Inverted-Annular Film-Boiling in a Two-Fluid Model
NUREG/IA-0134 Assessment of RELAP5/MOD3.1 for Gravity-Driven Injection Experiment in the Core Makeup Tank of the CARR Passive Reactor (CP-1300)
NUREG/IA-0135 Post-Test Analysis of PIPER-ONE PO-IC-2 Experiment by RELAP5/MOD3 Codes
NUREG/IA-0137 A Study of Control Room Staffing Levels for Advanced Reactors
NUREG/IA-0139 Assessment of RELAP5/MOD3.2 Using LOFT Large Break LOCA Test, LP–02–6
NUREG/IA-0140 Developmental Assessment of RELAP5/MOD3.1 with Separate-Effect and Integral Test Experiments: Model Changes and Options
NUREG/IA-0141 Result of BETHSY Test 9.1.b Using RELAP5/MOD3
NUREG/IA-0142 Installation of RELAP5/MOD3.2 on 80486 and Pentium Based Personal Computers
NUREG/IA-0143 Assessment of RELAP5/MOD3.2 With the LSTF Experiment Simulating a Loss of Residual Heat Removal Event During Mid-Loop Operation
NUREG/IA-0144 Assessment of RELAP5/MOD3.2 With the Semiscale Natural Circulation Experiment, S–NC–8B
NUREG/IA-0145 RELAP5 Assessment Against PACTEL Experimental Data
NUREG/IA-0146 Implementation and Assessment of Improved Models and Options in TRAC-BF1
NUREG/IA-0147 Assessment of RELAP5/MOD3.2 for Steam Condensation Experiments in the Presence of Noncondensibles in a Vertical Tube of PCCS
NUREG/IA-0148 Assessment of RELAP5/MOD3.1 Using LSTF Ten-Percent Main Steam-Line-Break Test Run SB-SL-01
NUREG/IA-0150 Study of Transients Related to AMSAC Actuation, Sensitivity Analysis
NUREG/IA-0151 Verification of RELAP5/MOD 3 With Theoretical and Numerical Stability Results on Single-Phase, Natural Circulation in a Simple Loop
NUREG/IA-0152 RELAP5/MOD3.2 Post Test Analysis and Accuracy Quantification of Lobi Test BL–34
NUREG/IA-0153 RELAP5/MOD3.2 Post Test Analysis and Accuracy Quantification of Lobi Test BL–44
NUREG/IA-0154 RELAP5/MOD3.2 Post Test Analysis and Accuracy Quantification of SPES Test SP-SB-03
NUREG/IA-0155 RELAP5/MOD3.2 Post Test Analysis and Accuracy Quantification of SPES Test SP-SB-04
NUREG/IA-0156 Data Base on the Behavior of High Burnup Fuel Rods with Zr-1%Nb Cladding and U02 Fuel (VVER Type) under Reactivity Accident Conditions
NUREG/IA-0157 Contrast of RELAP5/MOD3.2 Results From Different Computing Platforms
NUREG/IA-0160 Analysis of the Critical Flow Model in TRAC-BF1
NUREG/IA-0162 Test LOBI–BL06: Post-Test Analysis and RELAP5/MOD3.2.1 Code Performance Assessment
NUREG/IA-0163 A Study of the Dispersed Flow Interfacial Heat Transfer Model of RELAP5/MOD2.5 and RELAP5/MOD3
NUREG/IA-0164 Modification of USNRC's FRAP–T6 Fuel Rod Transient Code for High Burnup VVER Fuel
NUREG/IA-0165 Modification of IPSN's SCANAIR Fuel Rod Transient Code for High Burnup VVER Fuel
NUREG/IA-0166 RELAP5/MOD3.2 Assessment Using GERDA Small Break Test, 1605AA
NUREG/IA-0167 Assessment Study of RELAP5/MOD3.2 Based on the Kalinin NPP Unit-1 Stop of Feedwater Supply to the Steam Generator No. 4
NUREG/IA-0168 Assessment of RELAP5/MOD3.2 for Thermohydraulic Processes in Heated Rod Bundles with Tight Lattice at CKTI Test Facility
NUREG/IA-0169 Analysis of KS-1 Experimental Data on the Behavior of the Heated Rod Temperatures in the Partially Uncovered VVER Core Model Using RELAP5/MOD3.2
NUREG/IA-0170 RELAP5/MOD3.2 Post Test Calculation of the PKL-Experiment PKLIII-B4.3
NUREG/IA-0171 Simulation of LOCA 6" and LOCA 2" Transients in the RHR of a PWR Under Low Power Conditions Using RELAP5/MOD3.2
NUREG/IA-0172 Assessment of RELAP5/MOD3.2 Against a Main Steam Isolation Valve Closure at TRILLO I Nuclear Power Plant
NUREG/IA-0173 Simulation of a Station Black-Out in a PWR Under Midloop Conditions Using RELAP5/MOD3.2
NUREG/IA-0174 Study of Unusual Occurrence of a Partial Core Uncovery in an SBLOCA Scenario
NUREG/IA-0175 Analysis of Pin-by-Pin Effects for LWR Rod Ejection Accident
NUREG/IA-0176 Post-Test Analysis of P5 Experiment in PANDA Facility With TRAC-BF1 Code
NUREG/IA-0177 Assessment of a Reactor Coolant Pump Trip for TRILLO NPP with RELAP5/MOD3.2
NUREG/IA-0178 Cofrentes NPP (BWR/6) ATWS (MSIVC) Analysis with TRAC-BF1: 1D vs. Point Kinetics and Containment Response
NUREG/IA-0179 A Standardized Methodology for the Linkage of Computer Codes: Application to RELAP5/MOD3.2
NUREG/IA-0180 Application of RELAP5/MOD3.1 to ATWS Analysis of Control Rod Withdrawal From 1% Power Level
NUREG/IA-0181 Assessment of RELAP5/MOD3.2 for Reflux Condensation Experiment
NUREG/IA-0182 Application of RELAP5/MOD3.2 to the Loss-of-Residual-Heat-Removal Event Under Shutdown Condition
NUREG/IA-0183 Analysis of the LOBI Experiment Test BT–56 Using the RELAP5/MOD3.2 Code
NUREG/IA-0184 In-Tube Steam Condensation in the Presence of Air
NUREG/IA-0185 Development and Validation of a Transition Boiling Model for the RELAP5/MOD3 Reflood Simulation
NUREG/IA-0186 Analysis of the RELAP5/MOD3.2.2beta Critical Flow Models and Assessment Against Critical Flow Data From the Marviken Tests
NUREG/IA-0187 RELAP5/MOD3 Analysis of BETHSY Test 6.9c: Loss of RHRS: SG Manway Open
NUREG/IA-0188 RELAP5/MOD3.2 Validation Using BETHSY Test 6.9a
NUREG/IA-0189 Improvements of RELAP5/MOD3.2.2 Models for the CANDU Plant Analysis
NUREG/IA-0190 Nowadays Tools for Graphical Post-Processing of TRAC-BF1 Results
NUREG/IA-0191 A Tool for Drawing With Excel
NUREG/IA-0192 Assessment of RELAP5/MOD3.2.2 Gamma With the LOFT L9-3 Experiment Simulating an Anticipated Transient Without Scram
NUREG/IA-0193 Assessment of Single Recirculation Pump Trip Transient in Santa Maria de Garona Nuclear Power Plant With TRAC-BF1/MOD1, Version 0.4
NUREG/IA-0194 Analysis of Inadvertent Pressurizer Spray Valve Opening Real Transient with RELAP5/MOD3.2
NUREG/IA-0195 LBLOCA Analysis in a Westinghouse PWR 3-Loop Design Using RELAP5/MOD3
NUREG/IA-0196 Analysis of PANDA Experiments P3 and P6 Using RELAP5/MOD3.2
NUREG/IA-0197 Assessment of RELAP5/MOD3.2-NPA3.4 Against an Inadvertent Closure of all Three MSIV's in VANDELLOS-II Nuclear Power Plant
NUREG/IA-0198 Assessment of RELAP5/MOD3 With the SNUF Test Simulating Hot Leg Break LOCA in the View of Mass and Energy Release Analysis
NUREG/IA-0199 Mechanical Properties of Unirradiated and Irradiated Zr-1% Nb Cladding: Procedures and Results of Low Temperature Biaxial Burst Tests and Axial Tensile Tests
NUREG/IA-0200 Assessment Study on the PMK-2 Total Loss of Feedwater Experiment Using RELAP5 Code
NUREG/IA-0201 Description and RELAP5 Assessment of the PMK-2 CAMP-CLB Experiment: 2% Cold Leg Break Without HPIS With Secondary Bleed
NUREG/IA-0202 Analyses of KS Test Data on the Heated Rod Bundle Temperature Behavior in RBMK-1500 Core Model Under Stop and Recovery Flow Using RELAP5/MOD3.2 and RELAP5/MOD3.2.2 GAMMA
NUREG/IA-0203 Assessment of RELAP5/MOD3.2.2γ Against Flooding Database in Horizontal-to-Inclined Pipes
NUREG/IA-0204 OLKILUOTO 2 RELAP5/MOD3.2.1.2 Analysis of the Reactor Scram on June 13, 1997
NUREG/IA-0207 RELAP5/MOD3.2.2 Gamma Assessment For Down To Top Reflooding Process At VVER Like 37-Rod Bundle
NUREG/IA-0206 Simulation of the Propagation of Pressure Waves in Piping Systems with RELAP5/MOD 3.2.2: Comparison of Computed and Measured Results
NUREG/IA-0208 Analysis of the VTI Test Data on the Behavior of the Heated Rod Temperatures in the Partially Uncovered VVER-440 Core Model Using RELAP5/MOD3.2.2 Gamma
NUREG/IA-0209 Adaptation of USNRC's FRAPTRAN and IRSN's SCANAIR Transient Codes and Updating of MATPRO Package for Modeling of LOCA and RIA Validation Cases with Zr-1%Nb (VVER type) Cladding
NUREG/IA-0210 In-Tube Steam Condensation in the Presence of Air Under Transient Conditions
NUREG/IA-0211 Experimental Study of Embrittlement of Zr-1%Nb VVER Cladding under LOCA-Relevant Conditions
NUREG/IA-0212 Kalinin VVER-1000 Nuclear Power Station Unit 1 PRA (Beta Project): Executive Summary
NUREG/IA-0213 Experimental Study of Narrow Pulse Effects on the Behavior of High Burnup Fuel Rods with Zr-1%Nb Cladding and UO2 Fuel (VVER Type) under Reactivity-Initiated Accident Conditions
NUREG/IA-0215 Spatial Effects and Uncertainty Analysis for Rod Ejection Accidents in a PWR
NUREG/IA-0216 International HRA Empirical Study
NUREG/IA-0217 Investigations of the VVER-1000 Coolant Transient Benchmark I with the Coupled Code System RELAP5/PARCS
NUREG/IA-0219 Estimation of Operator Action Time Windows by RELAP5/MOD3.3
NUREG/IA-0220 Quantitative Code Assessment with Fast Fourier Transform Based Method Improved by Signal Mirroring
NUREG/IA-0221 Reactor Trip Analysis at Krško Nuclear Power Plant
NUREG/IA-0222 Analysis of RELAP5/MOD3.3 Prediction of 2-Inch Loss-of-Coolant Accident at Krško Nuclear Power Plant
NUREG/IA-0223 Assessment of RELAP5/MOD3.3 against Single Main Steam Isolation Valve Closure Events at the Krško Nuclear Power Plant
NUREG/IA-0224 An Assessment of TRACE V5 RC1 Code Separator Model with the Westinghouse Model Boiler 2 Experiments
NUREG/IA-0225 Analyzing Operator Actions During Loss of AC Power Accident with Subsequent Loss of Secondary Heat Sink
NUREG/IA-0226 Assessment of the Turbine Trip Transient in Santa María de Garoña Nuclear Power Plant with TRACE version 4.16
NUREG/IA-0227 IJS Animation Model for Krško NPP
NUREG/IA-0228 Assessment of RELAP5/MOD3.3Beta Code for the LOFT Experiment L9-1/L3-3
NUREG/IA-0229 RELAP5/MOD3.3 Assessment against New PMK Experiments
NUREG/IA-0230 An Assessment of TRACE V5 RC1 Code Against UPTF Counter Current Flow Tests
NUREG/IA-0231 An Assessment of TRACE V4.160 Code Against PACTEL ATWS-10 – 13 and ATWS-20 – 21 Pressurizer Experiments
NUREG/IA-0232 Validation of the CHAN-Component in TRACE Using BWR Full-Size Fine-Mesh Bundle Tests
NUREG/IA-0233 Assessment of TRACE 4.160 and 5.0 against RCP Trip Transient in Almaraz I Nuclear Power Plant
NUREG/IA-0234 Analysis of a Loss of Normal Feedwater Transient at the Ringhals-3 NPP Using RELAP5/Mod3.3
NUREG/IA-0235 Numerical Analysis of Mixing Factors in the RPV of VVER-440 Reactor Using the TRACE Code
NUREG/IA-0236 Analysis and Computational Predictions of CHF Position and Post-CHF Heat Transfer
NUREG/IA-0237 An Assessment of TRACE V4.160 Code Against PACTEL LOF-10 Experiment
NUREG/IA-0238 RELAP5/MOD3 Horizontal Off-Take Model for Application to Reactor Headers of CANDU Type Reactors
NUREG/IA-0239 Development of Horizontal Off-Take Model for Application to Reactor Headers of CANDU Type Reactors
NUREG/IA-0240 Sensitivity Analyses of a Hypothetical 6 Inch Break, LOCA in Ascό NPP using RELAP/MOD3.2
NUREG/IA-0241 Assessment of the TRACE Code Using Transient Data from Maanshan PWR Nuclear Power Plant
NUREG/IA-0242 Qualification of the Three-Dimensional Thermal Hydraulic Model of TRACE using Plant Data
NUREG/IA-0243 Development of a Vandellòs II NPP Model using the TRACE Code: Application to an Actual Transient of Main Coolant Pumps Trip and Start-up
NUREG/IA-0244 Assessment of TRACE 5.0 Against ROSA Test 6-2, Vessel Lower Plenum SBLOCA
NUREG/IA-0245 Assessment of TRACE 5.0 against ROSA Test 6-1, Vessel Upper Head SBLOCA
NUREG/IA-0246 RELAP5/MOD3.3 Assessment against PMK Test T3.1 – LBLOCA with Nitrogen in PRZ
NUREG/IA-0247 RELAP5 Simulation of Darlington Nuclear Generating Station Loss of Flow Event
NUREG/IA-0248 Post-Test Analysis of Hot Leg 2x25% Break at PSB-VVER Facility Using TRACE V5.0 Code
NUREG/IA-0249 Loss of External Load Analysis with RELAP5/MOD3.3 Patch 03 Code
NUREG/IA-0250 Simulation of the F2.1 Experiment at PKL Facility Using RELAP5/MOD3
NUREG/IA-0251 Improvement of RELAP5/MOD3.3 Reflood Model Based on the Assessments against FLECHT-SEASET Tests
NUREG/IA-0252 The development and verification of TRACE model for IIST experiments
NUREG/IA-0253 Development of a Computer Tool for In-Depth Analysis and Post Processing of the RELAP5 Thermal Hydraulic Code
NUREG/IA-0254 Suitability of Fault Modes and Effects Analysis for Regulatory Assurance of Complex Logic in Digital Instrumentation and Control Systems
NUREG/IA-0255 Coupled RELAP/PARCS Full Plant Model – Assessment of a Cooling Transient in Trillo Nuclear Power Plant
NUREG/IA-0256 Simulation of PKL Loss of RHRS Experiment E3.1 with RELAP5 and TRACE Codes – Application to a PWR NPP Model
NUREG/IA-0257 Simulation of PKL Loss of RHRS Experiment F2.2 Run 2 with RELAP5 and TRACE Codes – Application to a PWR NPP Model
NUREG/IA-0401 Assessment of Two-Phase Critical Flow Models Performance in RELAP5 and TRACE against Marviken Critical Flow Tests
NUREG/IA-0403 Full Scale Loop Seal experiments with TRACE V5 Patch 1
NUREG/IA-0404 The Development and Assessment of TRACE Model for Maanshan Nuclear Power Plant LOCA
NUREG/IA-0405 Coupling the RELAP Code with External Calculation Programs (Shared Memory Version)
NUREG/IA-0406 Post-Test Calculations on Steam Cool-Down Test QUENCH-04 with RELAP5, SCDAP/RELAP5, and TRACE
NUREG/IA-0407 Proposal for the Development and Implementation of an Uncertainty and Sensitivity Analysis Module in SNAP
NUREG/IA-0408 IJS Procedure for Converting Input Deck from RELAP5 to TRACE
NUREG/IA-0409 Post-Test Calculation of the ROSA/LSTF Test 3-1 using RELAP5/mod3.3
NUREG/IA-0410 Post-Test Calculation of the ROSA/LSTF Test 3-2 using RELAP5/mod3.3
NUREG/IA-0411 Simulation of the Experimental Series F2.2 at PKL Facility Using RELAP5/Mod 3.3
NUREG/IA-0412 Assessment of TRACE 5.0 Against ROSA Test 3-2, High Power Natural Circulation
NUREG/IA-0413 Assessment of TRACE 5.0 Against ROSA Test 3-1, Cold Leg SBLOCA
NUREG/IA-0414 Comparison of the U.S. NRC PARCS Core Neutronics Simulator Against In-Core Detector Measurements for LWR Applications
NUREG/IA-0416 Implementation of Advanced Multigroup Nodal and Pin Power Reconstruction Methods into PARCS 3.1
NUREG/IA-0418 Application of TRACE V5.0 P2 to Natural Circulation Reactor Safety Analysis
NUREG/IA-0419 Analysis with TRACE Code of ROSA Test 1.1: ECCS Water Injection Under Natural Circulation Condition
NUREG/IA-0420 Analysis with TRACE Code of Rosa Test 1.2: Small LOCA in the Hot-Leg with HPI and Accumulator Actuation
NUREG/IA-0421 Improvements and Validation of the System Code TRACE for Lead and Lead-Alloy Cooled Fast Reactors Safety-Related Investigations
NUREG/IA-0422 Transient Analysis of the Research Reactor MARIA MC Fuel Elements Using RELAP5 Mod 3.3
NUREG/IA-0423 Analysis with TRACE Code of PKL-III Test F 1.2
NUREG/IA-0424 RELAP5 Extended Station Blackout Analyses
NUREG/IA-0425 TRACE5 Assessment of 100% Direct Vessel Injection Line Break in ATLAS Facility
NUREG/IA-0426 Simulation of LSTF Upper Head Break (OECD/NEA ROSA Test 6.1) with TRACE Code.  Application to a PWR NPP Model
NUREG/IA-0427 Application of TRACE V5.0 P2 to China Domestic PWR LBLOCA Analysis
NUREG/IA-0428 Performing Uncertainty Analysis of IIST Facility SBLOCA by TRACE and DAKOTA
NUREG/IA-0429 Analysis of Loss of Feedwater Heater Transients for Lungmen ABWR by TRACE/PARCS
NUREG/IA-0430 TRACE Simulation of SBO Accident and Mitigation Strategy in Maanshan PWR
NUREG/IA-0431 The FSAR Transients Analysis of Lungmen ABWR Using TRACE/PARCS
NUREG/IA-0432 Analysis of the Test OECD-PKL2 G7.1 with the Thermal-Hydraulic System Code TRACE
NUREG/IA-0433 RELAP5/MOD3.3 RELEASE Pre & Postprocessor
NUREG/IA-0434 The Development and Application of TRACE/PARCS Model for Lungmen ABWR
NUREG/IA-0435 Assessment of RELAP5/MOD3.3 and TRACE V5.0 Computer Codes against LOCA Test Data from PSB-VVER Test Facility
NUREG/IA-0436 Assessment of LONF ATWS for Mananshan PWR Using TRACE Code
NUREG/IA-0437 Sensitivity Study of the DEG LBLOCA Transient on the Counter-Current Flow Limitation by Using TRACE
NUREG/IA-0438 ATWS Analysis of Lungmen ABWR for MSIV Closure Transient
NUREG/IA-0439 TRACE Analysis on Heat Removal Decrease Accidents for AP1000
NUREG/IA-0440 The Alternate Mitigation Strategies Study of Chinshan BWR/4 by Using the LOCA and SBO Analysis of TRACE
NUREG/IA-0441 Assessment Against ACHILLES Reflood Experiment with TRACE V5.0 Patch3
NUREG/IA-0442 RELAP5/MOD3.3 analysis of steam generator tube rupture (SGTR) accident for NPP Krško
NUREG/IA-0444 Simulation of LSTF Hot Leg Break (OECD/NEA ROSA-2 Test 1) with TRACE Code: Application to a PWR NPP Model
NUREG/IA-0445 The Establishment and Assessment of Chinshan (BWR/4) Nuclear Power Plant TRACE/SNAP Model
NUREG/IA-0447 RELAP5/MOD3.3 Assessment by Comparison with PKL III G3.1 Experiment (small break in the main steam line)
NUREG/IA-0448 Uncertainty Analysis for Maanshan LBLOCA by TRACE and DAKOTA
NUREG/IA-0449 Post-Test Analysis of Upper Plenum 11% Break at PSB-VVER Facility using TRACE V5.0 and RELAP5/MOD3.3 Code
Page Last Reviewed/Updated Tuesday, August 12, 2014