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Spotlight
Publications Resulting from International Agreements
Publications resulting from international agreements and overseen by NRC staff. Other International Agreements may be available in ADAMS.
Document Identifier | Title |
|---|---|
Assessment of TRAC-PD2 Using SUPER CANNON and HDR Experimental Data | |
Heat Transfer Processes During Intermediate and Large Break Loss-of-Coolant Accidents (LOCAs) | |
Influence of the Wetting State of a Heated Surface on Heat Transfer and Pressure Loss in an Evaporator Tube | |
Thermal Mixing Tests in a Semiannular Downcomer With Interacting Flows from Cold Legs | |
Assessment of RELAP5/MOD2, Cycle 36, Against FIX-II Split Break Experiment No. 3027. | |
Assessment of RELAP5/MOD2 Against Marviken Jet Impingement Test 11 Level Swell | |
Assessment of RELAP5/MOD2 Against Critical Flow Data From Marviken Tests JIT 11 and CFT 21. | |
Assessment Study of RELAP-5 MOD-2 Cycle 36.01 Based on the DOEL-2 Steam Generator Tube Rupture Incident of June 1979 | |
Assessment of RELAP5/MOD2 Against 25 Dryout Experiments Conducted at the Royal Institute of Technology | |
TRAC-PF1 MOD1 Post Test Calculations of the OECD LOFT Experiment LP-SB-1 | |
RELAP/MOD2 Calculations of OECD-LOFT Test LP-SB-01 | |
RELAP5/MOD2 Calculation of OECD-LOFT Test LP-SB-03 | |
Analysis of the THETIS Boil Down Experiments Using RELAP5/MOD2. | |
Assessment of Interphase Drag Correlations in the RELAP5/MOD2 and TRAC-PF1/MOD2 Codes | |
Assessment of RELAP5/MOD2, Cycle 36.04 Against FIX-II Guillotine Break experiment No. 5061 | |
RELAP5/MOD2 Assessment, OECD-LOFT Small Break Experiment LP-SB-03 | |
TRAC-PF1/MOD1 Post-Test Calculations of the OECD [Organisation for Economic Co-operation and Development] LOFT Experiment LP-SB-2 | |
Assessment Study of RELAP5/MOD2, CYCLE 36.04 Based on Spray Start-up Test for DOEL-4 | |
RELAP5/MOD2 Calculations of OECD LOFT Test LP-SB-2 | |
TRAC-PF1/MOD1 Post-Test Calculations of the OECD LOFT Experiment LP-SB-3 | |
Application of RELAP5/MOD3.1 Code to the LOFT Test L3-6 | |
RELAP5/MOD3 Subcooled Boiling Model Assessment | |
TRAC-PF1/MOD1 Calculations of LOFT experiment LP-02-6 | |
Review of LOFT [Loss-of-Fluid Test] Large Break Experiments [OECD LOFT project] | |
Assessment of RELAP5/MOD2, Cycle 36.04 Against FIX-II Split Break Experiment No. 3051 | |
Assessment of RELAP5/MOD2 Code Using Loss of Offsite Power Transient Data of KNU [Korea Nuclear Unit] No. 1 Plant | |
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 | |
Assessment of RELAP5/MOD2, Cycle 36-04 Using LOFT [Loss of Fluid Test] Large Break Experiment L2-5 | |
Assessment of RELAP5/MOD2, Cycle 36.04 Against LOFT Small Break Experiment L3-6 | |
Assessment Study of RELAP5/MOD2 Cycle 36.04 Based on Pressurizer Safety and Relief Valve Tests | |
Analysis of LOBI Test BLO2 (Three Percent Cold Leg Break) with RELAP5 Code | |
Assessment of RELAP5/MOD2, Cycle 36.04 Against LOFT Small Break Experiment L3-5 | |
Assessment of TRAC-PF1/MOD1 Against an Inadvertent Feedwater Line Isolation Transient in the Ringhals 4 Power Plant | |
Boil-Off Experiments with the EIR-NEPTUN Facility: Analysis and Code Assessment Overview Report | |
Assessment of TRAC-PF1/MOD1 Against an Inadvertent Steam Line Isolation Valve Closure in the Ringhals 2 Power Plant | |
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 | |
Assessment Study of RELAP5/MOD2 Cycle 36.04 Based on the DOEL-4 Manual Loss of Load Test of November 23, 1985 | |
Assessment Study of RELAP5/MOD2 Cycle 36.05 Based on the Tihange-2 Reactor Trip of January 11, 1983 | |
Assessment of RELAP5/MOD2 Using LOCE Large Break Loss-of-Coolant Experiment L2-5 | |
Assessment of RELAP5/MOD2 Using Semiscale Large Break Loss-of-Coolant Experiment S-06-3 | |
Assessment of RELAP5/MOD2 Cycle 36.04, Against the Loviisa–2 Stuck-Open Turbine By-Pass Valve Transient on September 1, 1981 | |
Thermal-Hydraulic Post-Test Analysis of OECD LOFT LP–FP–2 Experiment | |
TRAC–PF1 Code Assessment Using OECD LOFT LP–FP–1 Experiment | |
Assessment Study of RELAP5/MOD2 Cycle 36.05 Based on the DOEL 4 Reactor Trip of November 22, 1985 | |
An Analysis of Semiscale Mod–2C S–FS–1 Steam Line Break Test Using RELAP5/MOD2 | |
Analysis of Semiscale Test S–LH–1 Using RELAP5/MOD2 | |
Analysis of Semiscale Test S–LH–2 Using RELAP5/MOD2 | |
RELAP5/MOD2 Analysis of LOFT Experiment L9–4 | |
Recirculation Suction Large Break LOCA Analysis of the Santa Maria De Garoña Nuclear Power Plant Using TRAC–BF1 (G1J1) | |
Assessment of the "One Feedwater Pump Trip Transient" in Cofrentes Nuclear Power Plant With TRAC–BF1 | |
Assessment of RELAP5/MOD2 Cycle 36.04 Using LOFT Intermediate Break Experiment L5–1 | |
Assessment of RELAP5/MOD2 Cycle 36.04 with LOFT Large Break LOCE L2–3 | |
Analysis of the UPTF Separate Effects Test 11 (Steam-Water Countercurrent Flow in the Broken Loop Hot Leg) Using RELAP5 /MOD2 | |
LOFT Input Dataset Reference Document for RELAP5 Validation Studies | |
Time Step and Mesh Size Dependencies in the Heat Conduction Solution of a Semi-Implicit, Finite Difference Scheme for Transient Two-Phase Flow | |
RELAP5/MOD2 Post-Test Calculation of the OECD LOFT Experiment LP-SB-1 | |
RELAP5/MOD2 Analysis of a Postulated "Cold Leg SBLOCA" Simultaneous to a "Total Black-Out" Event in the José Cabrera Nuclear Station | |
RELAP5/MOD2 Post-Test Calculation of the OECD LOFT Experiment LP–SB–2 | |
Post-Test-Analysis and Nodalization Studies of OECD LOFT Experiment LP–02–6 With RELAP5/MOD2 CY36–02 | |
Post-Test-Analysis and Nodalization Studies of OECD LOFT Experiment LP–LB–1 With RELAP5/MOD2 CY36–02 | |
Assessment of RELAP5/MOD2 Using the Test Data of REWET-II Reflooding Experiment SGI/R | |
Assessment of RELAP5/MOD2 Against a Natural Circulation Experiment in Nuclear Power Plant Borssele | |
Assessment of RELAP5/MOD2 Computer Code Against the Net Load Trip Test Data From Yong–Gwang, Unit 2 | |
RELAP5/MOD3 Assessment for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads | |
Assessment of RELAP5/MOD3 Against Twenty-Five Post-Dryout Experiments Performed at the Royal Institute of Technology | |
RELAP5 Assessment Using LSTF Test Data SB–CL–18 | |
Numerics and Implementation of the UK Horizontal Stratification Entrainment Off-Take Model Into RELAP5/MOD3 | |
RELAP5 Assessment Using Semiscale SBLOCA Test S–NH–1 | |
Assessment of CCFL Model of RELAP5/MOD3 Against Simple Vertical Tubes and Rod Bundle Tests | |
Assessment of BETHSY Test 9.1.b Using RELAP5/MOD3 | |
RELAP5/MOD3 Assessment Using the Semiscale 50% Feed Line Break Test S–FS–11 | |
Assessment of RELAP5/MOD3 Version 5m5 Using Inadvertent Safety Injection Incident Data of Kori Unit 3 Plant | |
Assessment of PWR Steam Generator Modelling in RELAP5/MOD2 | |
Assessment of RELAP5/MOD2 Against a Load Rejection From 100% to 50% Power in the Vandellos II Nuclear Power Plant | |
Assessment of RELAP5/MOD2 Against a Turbine Trip From 100% Power in the Vandellos II Nuclear Power Plant | |
Assessment of RELAP5/MOD2 Against a 10% Load Rejection Transient from 75% Steady State in the Vandellós II Nuclear Power Plant | |
Assessment of RELAP5/MOD2 Against a Main Feedwater Turbopump Trip Transient in the Vandellos II Nuclear Power Plant | |
Assessment of RELAP5/MOD2 Against ECN-Reflood Experiments | |
Preliminary Assessment of PWR Steam Generator Modelling in RELAP5/MOD3 | |
Assessment of RELAP5/MOD3 With the LOFT L9–1/L3–3 Experiment Simulating an Anticipated Transient With Multiple Failures | |
Assessment of RELAP5/MOD3/V5m5 Against the UPTF Test No. 11 (Countercurrent Flow in PWR Hot Leg) | |
Analysis of LOFT Test L5–1 Using RELAP5/MOD2 | |
Assessment and Application of Blackout Transients at Asco Nuclear Power Plant with RELAP5/MOD2 | |
Assessment of the Turbine Trip Transient in Cofrentes NPP with TRAC–BF1 | |
Assessment of a Pressurizer Spray Valve Faulty Opening Transient at Asco Nuclear Power Plant with RELAP5/MOD2 | |
Assessment of MSIV Full Closure for Santa Maria De Garoila Nuclear Power Plant Using TRAC-BFl (G1J1) | |
Application of Full Power Blackout for C. N. Almaraz with RELAP5/MOD2 | |
Assessment of RELAP5/MOD2 Against a Pressurizer Spray Valve Inadverted Fully Opening Transient and Recovery by Natural Circulation in Jose Cabrera Nuclear Station | |
Assessment of RELAP5/MOD2 Computer Code Against the Natural Circulation Test Data from Yong–Gwang Unit 2 | |
2D/3D Program Work Summary Report | |
Reactor Safety Issues Resolved by the 2D/3D Program | |
International Code Assessment and Applications Program: Summary of Code Assessment Studies Concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC–B | |
An Assessment of the CORCON-MOD3 Code Part I: Thermal-Hydraulic Calculations | |
Assessment of RELAP5/MOD3.1 With the LSTF SB-SG-06 Experiment Simulating a Steam Generator Tube Rupture Transient | |
Assessment of RELAP5/MOD3 Using BETHSY 6.2TC 6-Inch Cold Leg Side Break Comparative Test | |
Improvements to the RELAP5/MOD3 Reflood Model and Uncertainty Quantification of Reflood Peak Clad Temperature | |
Development, Implementation, and Assessment of Specific Closure Laws for Inverted-Annular Film-Boiling in a Two-Fluid Model | |
Assessment of RELAP5/MOD3.1 for Gravity-Driven Injection Experiment in the Core Makeup Tank of the CARR Passive Reactor (CP-1300) | |
Post-Test Analysis of PIPER-ONE PO-IC-2 Experiment by RELAP5/MOD3 Codes | |
A Study of Control Room Staffing Levels for Advanced Reactors | |
Assessment of RELAP5/MOD3.2 Using LOFT Large Break LOCA Test, LP–02–6 | |
Developmental Assessment of RELAP5/MOD3.1 with Separate-Effect and Integral Test Experiments: Model Changes and Options | |
Result of BETHSY Test 9.1.b Using RELAP5/MOD3 | |
Installation of RELAP5/MOD3.2 on 80486 and Pentium Based Personal Computers | |
Assessment of RELAP5/MOD3.2 With the LSTF Experiment Simulating a Loss of Residual Heat Removal Event During Mid-Loop Operation | |
Assessment of RELAP5/MOD3.2 With the Semiscale Natural Circulation Experiment, S–NC–8B | |
RELAP5 Assessment Against PACTEL Experimental Data | |
Implementation and Assessment of Improved Models and Options in TRAC-BF1 | |
Assessment of RELAP5/MOD3.2 for Steam Condensation Experiments in the Presence of Noncondensibles in a Vertical Tube of PCCS | |
Assessment of RELAP5/MOD3.1 Using LSTF Ten-Percent Main Steam-Line-Break Test Run SB-SL-01 | |
Study of Transients Related to AMSAC Actuation, Sensitivity Analysis | |
Verification of RELAP5/MOD 3 With Theoretical and Numerical Stability Results on Single-Phase, Natural Circulation in a Simple Loop | |
RELAP5/MOD3.2 Post Test Analysis and Accuracy Quantification of Lobi Test BL–34 | |
RELAP5/MOD3.2 Post Test Analysis and Accuracy Quantification of Lobi Test BL–44 | |
RELAP5/MOD3.2 Post Test Analysis and Accuracy Quantification of SPES Test SP-SB-03 | |
RELAP5/MOD3.2 Post Test Analysis and Accuracy Quantification of SPES Test SP-SB-04 | |
Data Base on the Behavior of High Burnup Fuel Rods with Zr-1%Nb Cladding and U02 Fuel (VVER Type) under Reactivity Accident Conditions | |
Contrast of RELAP5/MOD3.2 Results From Different Computing Platforms | |
Analysis of the Critical Flow Model in TRAC-BF1 | |
Test LOBI–BL06: Post-Test Analysis and RELAP5/MOD3.2.1 Code Performance Assessment | |
A Study of the Dispersed Flow Interfacial Heat Transfer Model of RELAP5/MOD2.5 and RELAP5/MOD3 | |
Modification of USNRC's FRAP–T6 Fuel Rod Transient Code for High Burnup VVER Fuel | |
Modification of IPSN's SCANAIR Fuel Rod Transient Code for High Burnup VVER Fuel | |
RELAP5/MOD3.2 Assessment Using GERDA Small Break Test, 1605AA | |
Assessment Study of RELAP5/MOD3.2 Based on the Kalinin NPP Unit-1 Stop of Feedwater Supply to the Steam Generator No. 4 | |
Assessment of RELAP5/MOD3.2 for Thermohydraulic Processes in Heated Rod Bundles with Tight Lattice at CKTI Test Facility | |
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 | |
RELAP5/MOD3.2 Post Test Calculation of the PKL-Experiment PKLIII-B4.3 | |
Simulation of LOCA 6" and LOCA 2" Transients in the RHR of a PWR Under Low Power Conditions Using RELAP5/MOD3.2 | |
Assessment of RELAP5/MOD3.2 Against a Main Steam Isolation Valve Closure at TRILLO I Nuclear Power Plant | |
Simulation of a Station Black-Out in a PWR Under Midloop Conditions Using RELAP5/MOD3.2 | |
Study of Unusual Occurrence of a Partial Core Uncovery in an SBLOCA Scenario | |
Analysis of Pin-by-Pin Effects for LWR Rod Ejection Accident | |
Post-Test Analysis of P5 Experiment in PANDA Facility With TRAC-BF1 Code | |
Assessment of a Reactor Coolant Pump Trip for TRILLO NPP with RELAP5/MOD3.2 | |
Cofrentes NPP (BWR/6) ATWS (MSIVC) Analysis with TRAC-BF1: 1D vs. Point Kinetics and Containment Response | |
A Standardized Methodology for the Linkage of Computer Codes: Application to RELAP5/MOD3.2 | |
Application of RELAP5/MOD3.1 to ATWS Analysis of Control Rod Withdrawal From 1% Power Level | |
Assessment of RELAP5/MOD3.2 for Reflux Condensation Experiment | |
Application of RELAP5/MOD3.2 to the Loss-of-Residual-Heat-Removal Event Under Shutdown Condition | |
Analysis of the LOBI Experiment Test BT–56 Using the RELAP5/MOD3.2 Code | |
In-Tube Steam Condensation in the Presence of Air | |
Development and Validation of a Transition Boiling Model for the RELAP5/MOD3 Reflood Simulation | |
Analysis of the RELAP5/MOD3.2.2beta Critical Flow Models and Assessment Against Critical Flow Data From the Marviken Tests | |
RELAP5/MOD3 Analysis of BETHSY Test 6.9c: Loss of RHRS: SG Manway Open | |
RELAP5/MOD3.2 Validation Using BETHSY Test 6.9a | |
Improvements of RELAP5/MOD3.2.2 Models for the CANDU Plant Analysis | |
Nowadays Tools for Graphical Post-Processing of TRAC-BF1 Results | |
A Tool for Drawing With Excel | |
Assessment of RELAP5/MOD3.2.2 Gamma With the LOFT L9-3 Experiment Simulating an Anticipated Transient Without Scram | |
Assessment of Single Recirculation Pump Trip Transient in Santa Maria de Garona Nuclear Power Plant With TRAC-BF1/MOD1, Version 0.4 | |
Analysis of Inadvertent Pressurizer Spray Valve Opening Real Transient with RELAP5/MOD3.2 | |
LBLOCA Analysis in a Westinghouse PWR 3-Loop Design Using RELAP5/MOD3 | |
Analysis of PANDA Experiments P3 and P6 Using RELAP5/MOD3.2 | |
Assessment of RELAP5/MOD3.2-NPA3.4 Against an Inadvertent Closure of all Three MSIV's in VANDELLOS-II Nuclear Power Plant | |
Assessment of RELAP5/MOD3 With the SNUF Test Simulating Hot Leg Break LOCA in the View of Mass and Energy Release Analysis | |
Mechanical Properties of Unirradiated and Irradiated Zr-1% Nb Cladding: Procedures and Results of Low Temperature Biaxial Burst Tests and Axial Tensile Tests | |
Assessment Study on the PMK-2 Total Loss of Feedwater Experiment Using RELAP5 Code | |
Description and RELAP5 Assessment of the PMK-2 CAMP-CLB Experiment: 2% Cold Leg Break Without HPIS With Secondary Bleed | |
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 | |
Assessment of RELAP5/MOD3.2.2γ Against Flooding Database in Horizontal-to-Inclined Pipes | |
OLKILUOTO 2 RELAP5/MOD3.2.1.2 Analysis of the Reactor Scram on June 13, 1997 | |
RELAP5/MOD3.2.2 Gamma Assessment For Down To Top Reflooding Process At VVER Like 37-Rod Bundle | |
Simulation of the Propagation of Pressure Waves in Piping Systems with RELAP5/MOD 3.2.2: Comparison of Computed and Measured Results | |
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 | |
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 | |
In-Tube Steam Condensation in the Presence of Air Under Transient Conditions | |
Experimental Study of Embrittlement of Zr-1%Nb VVER Cladding under LOCA-Relevant Conditions | |
Kalinin VVER-1000 Nuclear Power Station Unit 1 PRA (Beta Project): Executive Summary | |
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 | |
Spatial Effects and Uncertainty Analysis for Rod Ejection Accidents in a PWR | |
International HRA Empirical Study | |
Investigations of the VVER-1000 Coolant Transient Benchmark I with the Coupled Code System RELAP5/PARCS | |
Estimation of Operator Action Time Windows by RELAP5/MOD3.3 | |
Quantitative Code Assessment with Fast Fourier Transform Based Method Improved by Signal Mirroring | |
Reactor Trip Analysis at Krško Nuclear Power Plant | |
Analysis of RELAP5/MOD3.3 Prediction of 2-Inch Loss-of-Coolant Accident at Krško Nuclear Power Plant | |
Assessment of RELAP5/MOD3.3 against Single Main Steam Isolation Valve Closure Events at the Krško Nuclear Power Plant | |
An Assessment of TRACE V5 RC1 Code Separator Model with the Westinghouse Model Boiler 2 Experiments | |
Analyzing Operator Actions During Loss of AC Power Accident with Subsequent Loss of Secondary Heat Sink | |
Assessment of the Turbine Trip Transient in Santa María de Garoña Nuclear Power Plant with TRACE version 4.16 | |
IJS Animation Model for Krško NPP | |
Assessment of RELAP5/MOD3.3Beta Code for the LOFT Experiment L9-1/L3-3 | |
RELAP5/MOD3.3 Assessment against New PMK Experiments | |
An Assessment of TRACE V5 RC1 Code Against UPTF Counter Current Flow Tests | |
An Assessment of TRACE V4.160 Code Against PACTEL ATWS-10 – 13 and ATWS-20 – 21 Pressurizer Experiments | |
Validation of the CHAN-Component in TRACE Using BWR Full-Size Fine-Mesh Bundle Tests | |
Assessment of TRACE 4.160 and 5.0 against RCP Trip Transient in Almaraz I Nuclear Power Plant | |
Analysis of a Loss of Normal Feedwater Transient at the Ringhals-3 NPP Using RELAP5/Mod3.3 | |
Numerical Analysis of Mixing Factors in the RPV of VVER-440 Reactor Using the TRACE Code | |
Analysis and Computational Predictions of CHF Position and Post-CHF Heat Transfer | |
An Assessment of TRACE V4.160 Code Against PACTEL LOF-10 Experiment | |
RELAP5/MOD3 Horizontal Off-Take Model for Application to Reactor Headers of CANDU Type Reactors | |
Development of Horizontal Off-Take Model for Application to Reactor Headers of CANDU Type Reactors | |
Sensitivity Analyses of a Hypothetical 6 Inch Break, LOCA in Ascό NPP using RELAP/MOD3.2 | |
Assessment of the TRACE Code Using Transient Data from Maanshan PWR Nuclear Power Plant | |
Qualification of the Three-Dimensional Thermal Hydraulic Model of TRACE using Plant Data | |
Development of a Vandellòs II NPP Model using the TRACE Code: Application to an Actual Transient of Main Coolant Pumps Trip and Start-up | |
Assessment of TRACE 5.0 Against ROSA Test 6-2, Vessel Lower Plenum SBLOCA | |
Assessment of TRACE 5.0 against ROSA Test 6-1, Vessel Upper Head SBLOCA | |
RELAP5/MOD3.3 Assessment against PMK Test T3.1 – LBLOCA with Nitrogen in PRZ | |
RELAP5 Simulation of Darlington Nuclear Generating Station Loss of Flow Event | |
Post-Test Analysis of Hot Leg 2x25% Break at PSB-VVER Facility Using TRACE V5.0 Code | |
Loss of External Load Analysis with RELAP5/MOD3.3 Patch 03 Code | |
Simulation of the F2.1 Experiment at PKL Facility Using RELAP5/MOD3 | |
Improvement of RELAP5/MOD3.3 Reflood Model Based on the Assessments against FLECHT-SEASET Tests | |
The development and verification of TRACE model for IIST experiments | |
Development of a Computer Tool for In-Depth Analysis and Post Processing of the RELAP5 Thermal Hydraulic Code | |
Suitability of Fault Modes and Effects Analysis for Regulatory Assurance of Complex Logic in Digital Instrumentation and Control Systems | |
Coupled RELAP/PARCS Full Plant Model – Assessment of a Cooling Transient in Trillo Nuclear Power Plant | |
Simulation of PKL Loss of RHRS Experiment E3.1 with RELAP5 and TRACE Codes – Application to a PWR NPP Model | |
Simulation of PKL Loss of RHRS Experiment F2.2 Run 2 with RELAP5 and TRACE Codes – Application to a PWR NPP Model | |
Assessment of Two-Phase Critical Flow Models Performance in RELAP5 and TRACE against Marviken Critical Flow Tests | |
Implementation of the Control Rod Movement Option by means of Control Variables in RELAP5/PARCS v2.7 Coupled Code | |
Full Scale Loop Seal experiments with TRACE V5 Patch 1 | |
The Development and Assessment of TRACE Model for Maanshan Nuclear Power Plant LOCA | |
Coupling the RELAP Code with External Calculation Programs (Shared Memory Version) | |
Post-Test Calculations on Steam Cool-Down Test QUENCH-04 with RELAP5, SCDAP/RELAP5, and TRACE | |
Proposal for the Development and Implementation of an Uncertainty and Sensitivity Analysis Module in SNAP | |
IJS Procedure for Converting Input Deck from RELAP5 to TRACE | |
Post-Test Calculation of the ROSA/LSTF Test 3-1 using RELAP5/mod3.3 | |
Post-Test Calculation of the ROSA/LSTF Test 3-2 using RELAP5/mod3.3 | |
Simulation of the Experimental Series F2.2 at PKL Facility Using RELAP5/Mod 3.3 | |
Assessment of TRACE 5.0 Against ROSA Test 3-2, High Power Natural Circulation | |
Assessment of TRACE 5.0 Against ROSA Test 3-1, Cold Leg SBLOCA | |
Comparison of the U.S. NRC PARCS Core Neutronics Simulator Against In-Core Detector Measurements for LWR Applications | |
TRACE (V 5.0 Patch 2) Validation Based on the RELAP5-Calculation of FIX-III LOCA Experiments NO. 5052, 4011, 3051 | |
Implementation of Advanced Multigroup Nodal and Pin Power Reconstruction Methods into PARCS 3.1 | |
Post-Test Thermal-Hydraulic Analysis of PKL Tests F1.1 and F1.2 | |
Application of TRACE V5.0 P2 to Natural Circulation Reactor Safety Analysis | |
Analysis with TRACE Code of ROSA Test 1.1: ECCS Water Injection Under Natural Circulation Condition | |
Analysis with TRACE Code of Rosa Test 1.2: Small LOCA in the Hot-Leg with HPI and Accumulator Actuation | |
Improvements and Validation of the System Code TRACE for Lead and Lead-Alloy Cooled Fast Reactors Safety-Related Investigations | |
Transient Analysis of the Research Reactor MARIA MC Fuel Elements Using RELAP5 Mod 3.3 | |
Analysis with TRACE Code of PKL-III Test F 1.2 | |
RELAP5 Extended Station Blackout Analyses | |
TRACE5 Assessment of 100% Direct Vessel Injection Line Break in ATLAS Facility | |
Simulation of LSTF Upper Head Break (OECD/NEA ROSA Test 6.1) with TRACE Code. Application to a PWR NPP Model | |
Application of TRACE V5.0 P2 to China Domestic PWR LBLOCA Analysis | |
Performing Uncertainty Analysis of IIST Facility SBLOCA by TRACE and DAKOTA | |
Analysis of Loss of Feedwater Heater Transients for Lungmen ABWR by TRACE/PARCS | |
TRACE Simulation of SBO Accident and Mitigation Strategy in Maanshan PWR | |
The FSAR Transients Analysis of Lungmen ABWR Using TRACE/PARCS | |
Analysis of the Test OECD-PKL2 G7.1 with the Thermal-Hydraulic System Code TRACE | |
RELAP5/MOD3.3 RELEASE Pre & Postprocessor | |
The Development and Application of TRACE/PARCS Model for Lungmen ABWR | |
Assessment of RELAP5/MOD3.3 and TRACE V5.0 Computer Codes against LOCA Test Data from PSB-VVER Test Facility | |
Assessment of LONF ATWS for Mananshan PWR Using TRACE Code | |
Sensitivity Study of the DEG LBLOCA Transient on the Counter-Current Flow Limitation by Using TRACE | |
ATWS Analysis of Lungmen ABWR for MSIV Closure Transient | |
TRACE Analysis on Heat Removal Decrease Accidents for AP1000 | |
The Alternate Mitigation Strategies Study of Chinshan BWR/4 by Using the LOCA and SBO Analysis of TRACE | |
Assessment Against ACHILLES Reflood Experiment with TRACE V5.0 Patch3 | |
RELAP5/MOD3.3 analysis of steam generator tube rupture (SGTR) accident for NPP Krško | |
Research Reactor 'MARIA' Primary Cooling Loop Transient Analysis Using RELAP5 Mod 3.3 | |
Simulation of LSTF Hot Leg Break (OECD/NEA ROSA-2 Test 1) with TRACE Code: Application to a PWR NPP Model | |
The Establishment and Assessment of Chinshan (BWR/4) Nuclear Power Plant TRACE/SNAP Model | |
Assessment of Channel Coolant Voiding in RD-14M Test Facility using TRACE | |
RELAP5/MOD3.3 Assessment by Comparison with PKL III G3.1 Experiment (small break in the main steam line) | |
Uncertainty Analysis for Maanshan LBLOCA by TRACE and DAKOTA | |
Post-Test Analysis of Upper Plenum 11% Break at PSB-VVER Facility using TRACE V5.0 and RELAP5/MOD3.3 Code | |
The Development and Application of Kuosheng (BWR/6) Nuclear Power Plant TRACE/SNAP Model | |
The Establishment and Assessment of Kuosheng (BWR/6) NPP Dry-storage System TRACE/SNAP Model | |
Spent Fuel Pool Safety Analysis of TRACE in Chinshan NPP | |
Benchmarking of a Generic CANDU Reactor with PARCS, MCNP and RFSPP | |
Modelling of ROCOM Mixing Test 2.2 with TRACE v5.0 Patch 3 | |
Analysis of the Control Rod Drop Accident (CRDA) for Lungmen ABWR | |
BEPU Analysis and Benchmark with IIST 2% SBLOCA Experiment Using TRACE/DAKOTA | |
Assessment of Critical Subcooled Flow Through Cracks in Large and Small Pipes Using TRACE and RELAP5 | |
RELAP5/MOD3.3 Analysis of Event with Actuation of Safety Injection System at the Loss of External Power | |
EPR Medium Break LOCA Benchmarking Exercise Using RELAP5 and CATHARE | |
Model 3D Cores for PWR Using Vessel Components in TRACEv5.OP3 | |
TRAC-BF1 to TRACE Model Semi-Automatic Conversion. PBTT Example | |
Uncertainty and Sensitivity Investigations with TRACE-SUSA and TRACE-DAKOTA by Means of Post-test Calculations of NUPEC BFBT Experiments | |
(Availability of) An International Report on Safety Critical Software for Nuclear Reactors by the Regulator Task Force on Safety Critical Software (TF-SCS) | |
RELAP5/MOD3.3 Model Assessment and Hypothetical Accident Analysis of Kuosheng Nuclear Power Plant with SNAP Interface | |
Fuel Rod Performance Uncertainty Analysis During Overpressurization Transient for Kuosheng Nuclear Power Plant with TRACE/ FRAPTRAN/ DAKOTA Codes in SNAP Interface | |
International Agreement Report – Analysis of the OSU-MASLWR 001 and 002 Tests by Using the TRACE Code | |
RELAP5 Analysis of Mitigation Strategy for Extended Blackout Power Condition in PWR | |
Validation of RELAP5 Model of Ringhals 4 Against a Load Step Test at Uprated Power | |
Development of a Coupled TRACE/PARCS Model for KKL and Benchmark Against the Turbine Trip Test | |
Nuclear Regulatory Authority Experimental Program to Characterize and Understand High Energy Arcing Fault (HEAF) Phenomena | |
Fuel Rod Behavior and Uncertainty Analysis by FRAPTRAN/TRACE/DAKOTA Code in Maanshan LBLOCA | |
RELAP5/MOD3.3 Model Assessment of Maanshan Nuclear Power Plant with SNAP Interface | |
Feedwater Line Break Analysis Using RELAP5/MOD3.3 for Steam Generator Blowdown Load Assessment | |
Steam Line Break Analysis Using RELAP5/MOD3.3 for Steam Generator Blowdown Load Assessment | |
TRACE/RELAP5 Comparative Calculations For Double-Ended LBLOCA and SBO | |
Main Steam Line Break Analysis for Lungmen ABWR | |
Thermal Hydraulic and Fuel Rod Mechanical Combination Analysis of Kuosheng Nuclear Power Plant with RELAP5 MOD3.3/FRAPTRAN/Python in SNAP Interface | |
TRACE/SNAP Model Establishment of Chinshan Nuclear Power Plant for Ultimate Response Guideline | |
RELAP5 and TRACE Calculations of LOCA in PWR | |
TRACE Assessment for Effect of Spacer Grid in RBHT Reflood Heat Transfer Experiments | |
Evaluation of TRACE Spacer Grid Model with FLECHT-SEASET Reflood Test | |
Using TRACE, MELCOR, CFD, and FRAPTRAN to Establish the Analysis Methodology for Chinshan Nuclear Power Plant Spent Fuel Pool | |
Loss of Flow Analysis of Maanshan Nuclear Power Plant with RELAP5/SNAP | |
PACTEL Small Break LOCA Experiment SBL-30 Calculation with TRACE Code | |
TRACE VVER-440/V-213 Model Validation | |
Simulation of the G3.1 Experiment at PKL Facility Using RELAP5/Mod3.3 | |
Simulation of the PKL-G7.1 Experiment in a Westinghouse Nuclear Power Plant Using RELAP5/Mod3.3 | |
Simulation of the LSTF-PKL Counterpart G7.1 Test at PKL Facility Using TRACE 5 | |
RELAP5 Model of a CANDU-6 (Embalse) Nuclear Power Plant: Application to a Turbine Trip Event | |
TRACE VVER-1000/V-320 Model Validation | |
Assessment of the Wall Film Condensation Model with Non-condensable Gas in RELAP5 and TRACE for Vertical Tube and Plate Geometries | |
Assessment of TRACE V5.0 Patch 4 Code Against PWR PACTEL Loop Seal Clearing Experiment | |
The Ultimate Response Guideline Simulation and Study for Lungmen (ABWR) Nuclear Power Plant Using RELAP5/SNAP | |
RELAP5 and TRACE Simulation of Hot Leg Break LOCA Experiment on LSTF | |
Assessment of NEPTUN Reflooding Experiments 5050 and 5052 with TRACE V5.0 Patch 5 | |
The Analysis and Study of ELAP Event and Mitigation Strategies Using TRACE Code for Maanshan PWR | |
IBLOCA Analysis for Vandellòs-NPP Using RELAP5/MOD3.3. Sensitivity Calculations to EOP Actions | |
Core Exit Temperature Response during an SBLOCA Event in the Ascó NPP | |
Post-Test Calculation of the PKL-2 Test G7.1 Using RELAP5/MOD3.3 | |
Post-Test Calculation of the ROSA-2 Test 3 Using RELAP5/MOD3.3 | |
Investigation of the Loop Seal Clearing Phenomena for the ATLAS DVI Line and Cold Leg SBLOCA Tests Using MARS-KS and RELAP5/MOD3.3 | |
Post-Test Analysis of Cold Leg Small Break 4.1% at PSB-VVER Facility using TRACE V5.0 | |
Post-Test Analysis of ROSA-2 Test 2 (IBLOCA) with TRACE | |
Assessment of TRACE 5.0 Against ROSA-2 Test 3 Counterpart Test to PKL | |
Assessment of TRACE 5.0 Against ROSA-2 Test 5, Main Steam Line Break with Steam Generator Tube Rupture | |
Using SNAP/RADTRAD and HABIT to Establish the Analysis Methodology for Maanshan PWR | |
Natural Circulation (Interruption) Analysis with MELCOR 2.2 during Asymmetric Cooldown Transients | |
Validation of RELAP5/MOD3.3 Friction Loss and Heat Transfer Model for Narrow Rectangular Channels | |
LBLOCA Uncertainty Analysis of Maanshan Nuclear Power Plant with RELAP5/SNAP and DAKOTA | |
MELCOR-ASTEC Crosswalk of the Accident at Fukushima-Daiichi Unit 1: Phase I Analysis | |
Simulation of ROSA-2 Test-2 Experiment: Application to Nuclear Power Plant | |
Simulation of ROSA-2 Test 3 Counterpart with TRACE5 – Application to Nuclear Power Plant | |
Semiscale S-NC-02 and S-NC-03 Natural Circulation Tests Performed by RELAP5/MOD3.3 Patch05 | |
Customization of XTV Graphics Output in TRACE v5.0 Patches 5, 4 & 3 | |
Analyses of an Unmitigated Station Blackout Transient in a Generic PWR–900 with ASTEC, MAAP and MELCOR Codes | |
LOCAs With Loss of One Active Emergency Cooling System Simulated by RELAP5 | |
Analysis of Maanshan Station Blackout Accident and Rescue Procedures under Different Tube Plugging Situations with TRACE | |
PWR PACTEL Small Break LOCA Experiment SBL-50 Calculation with TRACE Code | |
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 | |
Simulation with RELAP5/MOD3.3 of an Integral-Effect Test on Loop-Seal Clearing in the Upper Plenum Test Facility During Test A5 | |
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 | |
RELAP5 and TRACE Constitutive Relations Comparison | |
Evaluation for 4-Inch Cold Leg Top-Slot Break LOCA in ATLAS Facility with RELAP5 Mod3.3 Patch5 | |
TRACE VVER-440/V-213 Model Cross-Code Validation | |
TRACE VVER-1000/V-320 Model Cross-Code Validation | |
Simulation of Total Loss of Feedwater LOFT LP-FW-1 Test using RELAP5/MOD3.3 | |
Analysis of Main Steam Line Break Accident for 3-Loop PWR with RELAP5/MOD3.3 Code | |
Uncertainty Analysis of Main Steam Line Break Accident for Maanshan PWR with RELAP5/DAKOTA | |
Simulations of the BEAVRS PWR with SCALE and PARCS | |
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 | |
RELAP5 and TRACE Simulation of Bethsy 9.1b Test with Accuracy Quantification | |
MELCOR – DAKOTA Coupling for Uncertainty Analyses, in a SNAP Environment/Architecture | |
RELAP5, TRACE and APROS Model Benchmark for the IAEA SPE-4 Experiment | |
Assessment of Condensation Heat Transfer Models of TRACE V5.0 Patch 5 Using PASCAL Tests | |
Using VARSKIN for Hot Particles Ingestion Dosimetry Evaluation | |
RELAP5 Simulation of Total Loss of Feedwater in Two-Loop PWR | |
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 | |
Natural Circulation Assessment of a PWR Loss of Off-site Power with RELAP5/MOD 3.2 | |
New Functionality of TRACE: The 3DPost-Processing for the VESSEL Component in SALOME Platform | |
Assessment of TRACE5.0 Code Against ATLAS Test A5.2. Counterpart Test to LSTF | |
Multi-scale Coupling of TRACE and SUBCHANFLOW based on the Exterior Communication Interface (ECI) | |
Multi-Scale Coupling of Trace and TrioCFD with the Interface for CodeCoupling (ICoCo) | |
Implementation of droplet breakup mode in TRACE to improve the prediction of reactor core reflood conditions | |
RELAP5, TRACE and APROS Model Benchmark for the IAEA SPE-2 Experiment | |
Post-Test Analysis of PKL III Test H2.2 Run 2 (SBO) with TRACE | |
Assessment of a PWR Control Rod Drop Transient with 3D Neutronic-Thermalhydraulic Coupled Codes RELAP5/ PARCSv2.7 and TRACEv5.0P3/PARCSv3.0 | |
Uncertainty and Sensitivity Analysis of Hot Leg LOCA in Two-Loop PWR Using RELAP5 Version 3.3lj | |
Assessment of TRACE V5.0 Patch 7 Using OECD-ATLAS2 B3.2 Test | |
Maanshan PWR FLEX Program Enhance with RELAP5/MOD 3.3 | |
Modelling Guidelines for CCFL Representation During IBLOCA Scenarios of PWR Reactors | |
Assessments of 3D Components in System Codes Against Separate Effect Tests | |
Analysis of CRDM Nozzle Break at the ATLAS Facility with 3D Components in MARS-KS and TRACE | |
Study on the Effect of Dissolved Air During Hydroaccumulator Injection | |
TRACE simulations of LOCAs Together with the Complete Loss of One Emergency Core Cooling Function in Two-Loop PWR | |
Kuosheng BWR Decommissioning SBO Analysis with RELAP5/MOD 3.3 | |
TRACE/RELAP5 Calculation of NPP Krško SGTR Accident Under Realistic and SRP Conditions | |
TRACE Nodalization Performance in PSB-VVER SB-LOCA Benchmark | |
Chinshan BWR Decommissioning SBO Analysis with RELAP5/MOD 3.3 | |
Assessment of the Capability of the TRACE Code to Calculate Hydraulic Piping Loads During Fluid Transients | |
RELAP5 and TRACE Simulation of BETHSY 4.1a TC Test |
Page Last Reviewed/Updated Thursday, August 27, 2026
Page Last Reviewed/Updated Thursday, August 27, 2026