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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-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-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-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-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 |
Page Last Reviewed/Updated Friday, May 25, 2012
