| 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-0402 |
Implementation of the Control Rod Movement Option by means of Control Variables in RELAP5/PARCS v2.7 Coupled Code |
| 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-0415 |
TRACE (V 5.0 Patch 2) Validation Based on the RELAP5-Calculation of FIX-III LOCA Experiments NO. 5052, 4011, 3051 |
| NUREG/IA-0416 |
Implementation of Advanced Multigroup Nodal and Pin Power Reconstruction Methods into PARCS 3.1 |
| NUREG/IA-0417 |
Post-Test Thermal-Hydraulic Analysis of PKL Tests F1.1 and F1.2 |
| 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-0443 |
Research Reactor 'MARIA' Primary Cooling Loop Transient Analysis Using RELAP5 Mod 3.3 |
| 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-0446 |
Assessment of Channel Coolant Voiding in RD-14M Test Facility using TRACE |
| 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 |
| NUREG/IA-0450 |
The Development and Application of Kuosheng (BWR/6) Nuclear Power Plant TRACE/SNAP Model |
| NUREG/IA-0451 |
The Establishment and Assessment of Kuosheng (BWR/6) NPP Dry-storage System TRACE/SNAP Model |
| NUREG/IA-0452 |
Spent Fuel Pool Safety Analysis of TRACE in Chinshan NPP |
| NUREG/IA-0453 |
Benchmarking of a Generic CANDU Reactor with PARCS, MCNP and RFSPP |
| NUREG/IA-0454 |
Modelling of ROCOM Mixing Test 2.2 with TRACE v5.0 Patch 3 |
| NUREG/IA-0455 |
Analysis of the Control Rod Drop Accident (CRDA) for Lungmen ABWR |
| NUREG/IA-0456 |
BEPU Analysis and Benchmark with IIST 2% SBLOCA Experiment Using TRACE/DAKOTA |
| NUREG/IA-0457 |
Assessment of Critical Subcooled Flow Through Cracks in Large and Small Pipes Using TRACE and RELAP5 |
| NUREG/IA-0458 |
RELAP5/MOD3.3 Analysis of Event with Actuation of Safety Injection System at the Loss of External Power |
| NUREG/IA-0459 |
EPR Medium Break LOCA Benchmarking Exercise Using RELAP5 and CATHARE |
| NUREG/IA-0460 |
Model 3D Cores for PWR Using Vessel Components in TRACEv5.OP3 |
| NUREG/IA-0461 |
TRAC-BF1 to TRACE Model Semi-Automatic Conversion. PBTT Example |
| NUREG/IA-0462 |
Uncertainty and Sensitivity Investigations with TRACE-SUSA and TRACE-DAKOTA by Means of Post-test Calculations of NUPEC BFBT Experiments |
| NUREG/IA-0463 |
(Availability of) An International Report on Safety Critical Software for Nuclear Reactors by the Regulator Task Force on Safety Critical Software (TF-SCS) |
| NUREG/IA-0464 |
RELAP5/MOD3.3 Model Assessment and Hypothetical Accident Analysis of Kuosheng Nuclear Power Plant with SNAP Interface |
| NUREG/IA-0465 |
Fuel Rod Performance Uncertainty Analysis During Overpressurization Transient for Kuosheng Nuclear Power Plant with TRACE/ FRAPTRAN/ DAKOTA Codes in SNAP Interface |
| NUREG/IA-0466 |
International Agreement Report – Analysis of the OSU-MASLWR 001 and 002 Tests by Using the TRACE Code |
| NUREG/IA-0467 |
RELAP5 Analysis of Mitigation Strategy for Extended Blackout Power Condition in PWR |
|
NUREG/IA-0468
|
Validation of RELAP5 Model of Ringhals 4 Against a Load Step Test at Uprated Power |
| NUREG/IA-0469 |
Development of a Coupled TRACE/PARCS Model for KKL and Benchmark Against the Turbine Trip Test |
| NUREG/IA-0470 |
Nuclear Regulatory Authority Experimental Program to Characterize and Understand High Energy Arcing Fault (HEAF) Phenomena |
| NUREG/IA-0471 |
Fuel Rod Behavior and Uncertainty Analysis by FRAPTRAN/TRACE/DAKOTA Code in Maanshan LBLOCA |
| NUREG/IA-0472 |
RELAP5/MOD3.3 Model Assessment of Maanshan Nuclear Power Plant with SNAP Interface |
| NUREG/IA-0473 |
Feedwater Line Break Analysis Using RELAP5/MOD3.3 for Steam Generator Blowdown Load Assessment |
| NUREG/IA-0474 |
Steam Line Break Analysis Using RELAP5/MOD3.3 for Steam Generator Blowdown Load Assessment |
| NUREG/IA-0475 |
TRACE/RELAP5 Comparative Calculations For Double-Ended LBLOCA and SBO |
| NUREG/IA-0476 |
Main Steam Line Break Analysis for Lungmen ABWR |
| NUREG/IA-0477 |
Thermal Hydraulic and Fuel Rod Mechanical Combination Analysis of Kuosheng Nuclear Power Plant with RELAP5 MOD3.3/FRAPTRAN/Python in SNAP Interface |
| NUREG/IA-0478 |
TRACE/SNAP Model Establishment of Chinshan Nuclear Power Plant for Ultimate Response Guideline |
| NUREG/IA-0479 |
RELAP5 and TRACE Calculations of LOCA in PWR |
| NUREG/IA-0480 |
TRACE Assessment for Effect of Spacer Grid in RBHT Reflood Heat Transfer Experiments |
| NUREG/IA-0481 |
Evaluation of TRACE Spacer Grid Model with FLECHT-SEASET Reflood Test |
| NUREG/IA-0482 |
Using TRACE, MELCOR, CFD, and FRAPTRAN to Establish the Analysis Methodology for Chinshan Nuclear Power Plant Spent Fuel Pool |
| NUREG/IA-0483 |
Loss of Flow Analysis of Maanshan Nuclear Power Plant with RELAP5/SNAP |
| NUREG/IA-0484 |
PACTEL Small Break LOCA Experiment SBL-30 Calculation with TRACE Code |
| NUREG/IA-0485 |
TRACE VVER-440/V-213 Model Validation |
| NUREG/IA-0486 |
Simulation of the G3.1 Experiment at PKL Facility Using RELAP5/Mod3.3 |
| NUREG/IA-0487 |
Simulation of the PKL-G7.1 Experiment in a Westinghouse Nuclear Power Plant Using RELAP5/Mod3.3 |
| NUREG/IA-0488 |
Simulation of the LSTF-PKL Counterpart G7.1 Test at PKL Facility Using TRACE 5 |
| NUREG/IA-0489 |
RELAP5 Model of a CANDU-6 (Embalse) Nuclear Power Plant: Application to a Turbine Trip Event |
| NUREG/IA-0490 |
TRACE VVER-1000/V-320 Model Validation |
| NUREG/IA-0491 |
Assessment of the Wall Film Condensation Model with Non-condensable Gas in RELAP5 and TRACE for Vertical Tube and Plate Geometries |
| NUREG/IA-0492 |
Assessment of TRACE V5.0 Patch 4 Code Against PWR PACTEL Loop Seal Clearing Experiment |
| NUREG/IA-0493 |
The Ultimate Response Guideline Simulation and Study for Lungmen (ABWR) Nuclear Power Plant Using RELAP5/SNAP |
| NUREG/IA-0494 |
RELAP5 and TRACE Simulation of Hot Leg Break LOCA Experiment on LSTF |
| NUREG/IA-0495 |
Assessment of NEPTUN Reflooding Experiments 5050 and 5052 with TRACE V5.0 Patch 5 |
| NUREG/IA-0496 |
The Analysis and Study of ELAP Event and Mitigation Strategies Using TRACE Code for Maanshan PWR |
| NUREG/IA-0497 |
IBLOCA Analysis for Vandellòs-NPP Using RELAP5/MOD3.3. Sensitivity Calculations to EOP Actions |
| NUREG/IA-0498 |
Core Exit Temperature Response during an SBLOCA Event in the Ascó NPP |
| NUREG/IA-0499 |
Post-Test Calculation of the PKL-2 Test G7.1 Using RELAP5/MOD3.3 |
| NUREG/IA-0500 |
Post-Test Calculation of the ROSA-2 Test 3 Using RELAP5/MOD3.3 |
| NUREG/IA-0501 |
Investigation of the Loop Seal Clearing Phenomena for the ATLAS DVI Line and Cold Leg SBLOCA Tests Using MARS-KS and RELAP5/MOD3.3 |
| NUREG/IA-0502 |
Post-Test Analysis of Cold Leg Small Break 4.1% at PSB-VVER Facility using TRACE V5.0 |
| NUREG/IA-0503 |
Post-Test Analysis of ROSA-2 Test 2 (IBLOCA) with TRACE |
| NUREG/IA-0504 |
Assessment of TRACE 5.0 Against ROSA-2 Test 3 Counterpart Test to PKL |
| NUREG/IA-0505 |
Assessment of TRACE 5.0 Against ROSA-2 Test 5, Main Steam Line Break with Steam Generator Tube Rupture |
| NUREG/IA-0506 |
Using SNAP/RADTRAD and HABIT to Establish the Analysis Methodology for Maanshan PWR |
| NUREG/IA-0507 |
Natural Circulation (Interruption) Analysis with MELCOR 2.2 during Asymmetric Cooldown Transients |
| NUREG/IA-0508 |
Validation of RELAP5/MOD3.3 Friction Loss and Heat Transfer Model for Narrow Rectangular Channels |
| NUREG/IA-0509 |
LBLOCA Uncertainty Analysis of Maanshan Nuclear Power Plant with RELAP5/SNAP and DAKOTA |
| NUREG/IA-0510 |
MELCOR-ASTEC Crosswalk of the Accident at Fukushima-Daiichi Unit 1: Phase I Analysis |
| NUREG/IA-0511 |
Simulation of ROSA-2 Test-2 Experiment: Application to Nuclear Power Plant |
| NUREG/IA-0512 |
Simulation of ROSA-2 Test 3 Counterpart with TRACE5 – Application to Nuclear Power Plant |
| NUREG/IA-0513 |
Semiscale S-NC-02 and S-NC-03 Natural Circulation Tests Performed by RELAP5/MOD3.3 Patch05 |
| NUREG/IA-0514 |
Customization of XTV Graphics Output in TRACE v5.0 Patches 5, 4 & 3 |
| NUREG/IA-0515 |
Analyses of an Unmitigated Station Blackout Transient in a Generic PWR–900 with ASTEC, MAAP and MELCOR Codes |
| NUREG/IA-0516 |
LOCAs With Loss of One Active Emergency Cooling System Simulated by RELAP5 |
| NUREG/IA-0517 |
Analysis of Maanshan Station Blackout Accident and Rescue Procedures under Different Tube Plugging Situations with TRACE |
| NUREG/IA-0518 |
PWR PACTEL Small Break LOCA Experiment SBL-50 Calculation with TRACE Code |
| NUREG/IA-0519 |
Survey of Member Countries' Nuclear Power Plant Fire Protection Regulations by the OECD Nuclear Energy Agency (NEA) Fire Incidents Records Exchange (FIRE) Database Project – Topical Report No. 2 |
| NUREG/IA-0520 |
Simulation with RELAP5/MOD3.3 of an Integral-Effect Test on Loop-Seal Clearing in the Upper Plenum Test Facility During Test A5 |
| NUREG/IA-0521 |
Analysis with TRACE Code of PKL III Tests G1.2. Study on Heat Transfer Mechanisms in the SG in Presence of Nitrogen, Steam and Water as a Function of the Primary Coolant Inventory in Double Loop Operation |
| NUREG/IA-0522 |
RELAP5 and TRACE Constitutive Relations Comparison |
| NUREG/IA-0523 |
Evaluation for 4-Inch Cold Leg Top-Slot Break LOCA in ATLAS Facility with RELAP5 Mod3.3 Patch5 |
| NUREG/IA-0524 |
TRACE VVER-440/V-213 Model Cross-Code Validation |
| NUREG/IA-0525 |
TRACE VVER-1000/V-320 Model Cross-Code Validation |
| NUREG/IA-0526 |
Simulation of Total Loss of Feedwater LOFT LP-FW-1 Test using RELAP5/MOD3.3 |
| NUREG/IA-0527 |
Analysis of Main Steam Line Break Accident for 3-Loop PWR with RELAP5/MOD3.3 Code |
| NUREG/IA-0528 |
Uncertainty Analysis of Main Steam Line Break Accident for Maanshan PWR with RELAP5/DAKOTA |
| NUREG/IA-0529 |
Simulations of the BEAVRS PWR with SCALE and PARCS |
| NUREG/IA-0530 |
Analysis with TRACE Code of PKL III Tests G1.1 & G1.1a. Study on Heat Transfer Mechanisms in the SG in Presence of Nitrogen, Steam and Water as a Function of the Primary Coolant Inventory in Single Loop Operation |
| NUREG/IA-0531 |
RELAP5 and TRACE Simulation of Bethsy 9.1b Test with Accuracy Quantification |
| NUREG/IA-0532 |
MELCOR – DAKOTA Coupling for Uncertainty Analyses, in a SNAP Environment/Architecture |
| NUREG/IA-0533 |
RELAP5, TRACE and APROS Model Benchmark for the IAEA SPE-4 Experiment |
| NUREG/IA-0534 |
Assessment of Condensation Heat Transfer Models of TRACE V5.0 Patch 5 Using PASCAL Tests |
| NUREG/IA-0535 |
Using VARSKIN for Hot Particles Ingestion Dosimetry Evaluation |
| NUREG/IA-0536 |
RELAP5 Simulation of Total Loss of Feedwater in Two-Loop PWR |
| NUREG/IA-0537 |
Plant Application with TRACE Code of the PKL III G1 Test Series. Study on Heat Transfer Mechanisms in the SG in Presence of Nitrogen, Steam and Water as a Function of the Primary Coolant Inventory in Single & Double Loop Operation |
| NUREG/IA-0538 |
Natural Circulation Assessment of a PWR Loss of Off-site Power with RELAP5/MOD 3.2 |
| NUREG/IA-0540 |
Assessment of TRACE5.0 Code Against ATLAS Test A5.2. Counterpart Test to LSTF |