Integral Experiments Data, Databases, Benchmarks and Safety Joint Projects
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csni1015 ACHILLES, Heat Transfer in PWR Core During LOCA Reflood Phase
nea-1913 AEROSOL-SCIENCE, Aerosol Science: Theory and Practice with Special Applications to the Nuclear Industry
nea-1657 ANL-BPB, Argonne National Laboratory Code Center Benchmark Problem Book
csni2039 ATLAS PROJECT, The Advanced Thermal-hydraulic Test Loop for Accident Simulation Project
csni0076 BETHSY/6.9C, Loss of residual heat removal system during mid-loop operation
csni0062 BETHSY/9.1B, Cold Leg Break Test
csni2018 BIP PROJECT, Behaviour of Iodine Project
csni2036 BIP-2, Behaviour of Iodine Project Phase 2
csni2040 BSAF-1, Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station Project, Phase 1
csni2005 BUBBLER CONDENSER, Bubbler Condenser Project
csni1023 CORA-13, Experiment on severe fuel damage, core degradation and quench
csni1024 CORA-W2, Experiment on Severe Fuel Damage for a VVER-type PWR
nea-1681 CRITICALITYACCIDENTS, A Review of Criticality Accidents, 2000 Revision, LA-13638 in PDF format
csni0071 DOEL2/SGTR, Steam Generator Tube Rupture incident at the DOEL 2, Westinghouse 2 loop PWR
nea-1875 EACRP-D2O-LATTICES, Compilation of reactor physics measurements in HWRs lattices
csni1026 ERSEC, investigation of the reflooding phase of a Loss of Coolant Accident
csni1020 FALCON/ISP1-ISP2, fission product and aerosol transport in primary coolant system and in the containment
csni1019 FARO/L-14, Test L-14 on fuel coolant interaction and quenching
ests0022 FIREDATA, Nuclear Power Plant Fire Event Data Base
csni0058 FIST/4DBA1, BWR/4-218 Simulated Double-Ended Large Break Test
csni0057 FIST/6IB1, BWR/6 System Responses to Intermediate Break in Recirculation Suction Line LINE
csni0054 FIST/6MSB1, BWR/6 Main Steamline Double-Ended Break Test
csni0056 FIST/6PMC1, BWR/6-128 Isolation Valve (MSIV) Closure without Power Scram
csni0055 FIST/6SB1, BWR/6 Simulated Recirculation Line Break
csni0053 FIST/6SB2C, BWR/6 Recirculation Suction Line Break Test
csni0059 FIST/T1QUV, Simulated Failure to Maintain Water Level in BWR/6-218
csni0060 FIST/T23C, Simulated Failure to Maintain Water Level in BWR/6-218
csni0001 FIX-II/2032, BWR Pump Trip Experiment 2032, Simulation Mass Flow and Power Transients
csni0049 FIX-II/3025, BWR FIX-II Pump Trip Experiment 3025, Immediate Split Size Break
csni0050 FIX-II/3061, BWR FIX-II Pump Trip Experiment 3061, Large Split Break
csni0051 FIX-II/5052, BWR FIX-II Pump Trip Experiment 5052, Guillotine Break Simulation
csni0052 FIX-II/6261, BWR FIX-II Pump Trip Experiment 626, Transient Dryout Tests
csni1008 G2/716, Westinghouse G2 Loop Test Facility
csni1009 G2/718, Westinghouse G2 Loop Test Facility
csni1010 G2/736, Westinghouse G2 Loop Test Facility
nea-1827 GANAPOL-ABNTT, Analytical Benchmarks for Nuclear Engineering Applications, Case Studies in Neutron Transport Theory
csni2038 HEAF, High Energy Arcing Fault Events
csni0000 I.T.D., CSNI Integral Test Facility Validation Matrix
nea-1823 ICRS1, Proceedings of the First Radiation Shielding Symposium, Cambridge, UK 1958
nea-1486 ICSBEP2020-HANDBOOK, International Criticality Safety Benchmark Experiment Handbook
nea-1664 IFPE DATABASE, International Fuel Performance Experiments Database
nea-1594 IFPE/AEAT-IMC, Onset Gas Release and Grain Face Venting Rates in Fuels
nea-1596 IFPE/AECL-BUNDLE, Fission Gas Release and Burnup Analysis, PHWR Fuel
nea-1799 IFPE/AEKI-EDB-E110, Experimental Database of E110 Claddings under Accident Conditions
nea-1788 IFPE/BN-MOX-M109/D3, Belgonucleaire Beznau-1 PWR irradiated MOX Fuel Rod M109/D3
nea-1863 IFPE/BN-MOX-M510/D10, Belgonucleaire Beznau-1 PWR irradiated MOX Fuel Rod M510/D10
nea-1560 IFPE/BR3-HBFRHCP, BR-3 High Burnup Fuel Rod Hot Cell Program
nea-1705 IFPE/CAGR-UOX-SWELL, Fuel swelling Data Obtained from the AGR/Halden Ramp Test Programme
nea-1858 IFPE/CANDU-FIO-130, CANDU experiment FIO-130 Fuel Behaviour under LOCA Conditions
nea-1783 IFPE/CANDU-FIO-131, CANDU experiment FIO-131 Fuel Behaviour under LOCA Conditions
nea-1777 IFPE/CANDU-IRDMR, In-Reactor Diameter Measuring RIG EXP-FIO-118 and EXP-FIO-119 Fuel Behaviour under LOCA Conditions
nea-1615 IFPE/CEA-DEFECT FUEL, Experiments Irradiated at CEA Grenoble
nea-1626 IFPE/CNEA-MOX-RAMP, CNEA Power Ramp Irradiations with (PHWR) MOX Fuels
nea-1595 IFPE/CONTACT REV.1, PWR Fuel Performance Tests Siloe Reactor
nea-1806 IFPE/DEFEX, Studsvik DEFEX BWR fuel secondary defect formation as a consequence of primary defects
nea-1807 IFPE/DEFEX-II DEMO, BWR fuel primary defect and conditions leading to secondary failure of the cladding by hydriding
nea-1597 IFPE/DEMO-RAMP-I&II, Pellet Clad Interaction Behaviour, Fast Power Ramping
nea-1645 IFPE/EFE-RO, Experimental Fuel Elements RO89 and RO51 in TRIGA 14 MW Reactor (INR-Pitesti)
nea-1841 IFPE/EXP-BDL-406, performance of natural UO2 fuel irradiated at low linear powers in NRU
nea-1774 IFPE/FMDP-MOX4-5, Weapons-Derived MOX Fuel DOE FMDP Test Irradiations Capsules 4 & 5, Advanced Test Reactor (ATR)
nea-1599 IFPE/FUMEX-I, Data from OECD Halden Reactor Project for FUMEX-1 (Fuel Modelling at Extended Burnup)
nea-1720 IFPE/FUMEX-II/CASE27, 7 idealised cases for functional dependence of FGR predictions
nea-1625 IFPE/GAIN, Gadolinia Doped UO2 Fuel Behaviour Experiment
nea-1736 IFPE/GBGI, Grain-Bubble Gas Interlinkage
nea-1697 IFPE/HATAC R1, Fission Gas Release at High Burn-up, Effect of a Power Cycling
nea-1510 IFPE/HBEP REV.1, Battelle's High Burn-Up Effects Programme for Fuel Performance
nea-1546 IFPE/IFA-429, Fission Gas Release, Thermal Behaviour U02 Fuel, Halden Reactor
nea-1488 IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden
nea-1729 IFPE/IFA-507-TF3-TF5, Database For Transient Temperature Experiment Ifa-507
nea-1629 IFPE/IFA-508 & 515, PCMI Behaviour of Thin Cladding Rods, JAERI and HRP
nea-1778 IFPE/IFA-514/565, LWR MOX Fuel Irradiation Tests - HBWR Irradiation with the Instrument Rig, IFA-514/565 (JAEA) 6 rods
nea-1860 IFPE/IFA-519.9, Three PWR rods irradiated to 90 MWd/kg UO2
nea-1549 IFPE/IFA-533, Fuel Thermal Behaviour at High Burnup, Halden Reactor
nea-1684 IFPE/IFA-534.14REV1, fission gas release as a function of burnup at high power (52-55 MWd/kg)
nea-1548 IFPE/IFA-535, Fission Gas Release, Power Ramps, High Burnup Fuel
nea-1547 IFPE/IFA-562, Pellet Surface Roughness Effect on Thermal Performances and PCMI
nea-1803 IFPE/IFA-585, In-Reactor Creep Behaviour of Zircaloy-2 and Zircaloy-4 under Variable Loading Conditions
nea-1773 IFPE/IFA-591, JAEA Power Ramp Tests of MOX Fuel Rods IFA-591
nea-1772 IFPE/IFA-597-MOX, Hollow and solid MOX rods experiments
nea-1685 IFPE/IFA-597.3, centre-line temperature, fission gas release and clad elongation at high burn-up (60-62 MWd/kg)
nea-1861 IFPE/IFA-629.1, The Re-irradiation of MIMAS-MOX Fuel in IFA-629.1
nea-1862 IFPE/IFA-650.1 & .2, LOCA testing at Halden, Two experiments, IFA-650 series
nea-1555 IFPE/INTER-RAMP, Fast Power Ramps Failures of Unpressurised Fuel Rods
nea-1532 IFPE/KOLA-3, WWER-440 Fuel Performance Data from KOLA-3 NPP, FGR
nea-1766 IFPE/KOLA-3-MIR-RAMP, KOLA-3 MIR test temperature during ramp, FGR and pressure at EOL, Bu up to 55 MWd/kgUO2
nea-1710 IFPE/MT4-MT6A-LOCA, Large-break LOCA in-reactor fuel bundle materials tests at NRU
nea-1758 IFPE/NFIR-1, Clad creepdown, power history effect on fission product distribution (6 PWR rods 40-64 MWd/kg in BR-3)
nea-1741 IFPE/NOVOVORONEZH, operation factor data of the Novovoronezh VVER-1000 fuel assembly 4108 rods
nea-1724 IFPE/NSRR-FK1-2-3, Behaviour of 3 BWR segments FK-1, FK-2 & FK-3 under RIA test conditions in NSRR
nea-1622 IFPE/OSIRIS R3, 4 PWR Rods Irradiated in the CEA Osiris Reactor
nea-1556 IFPE/OVER-RAMP, Pellet Clad Interaction Failure Analysis, Power Ramps
nea-1776 IFPE/PRIMO-BD8, Belgonucleaire and SCK-CEN PRIMO Ramped MOX Fuel Rod BD8
nea-1696 IFPE/REGATE L10.3, FGR and Fuel Swelling during power transient at medium burn-up (SILOE reactor)
nea-1634 IFPE/RISOE-I, Fission gas release from high-burnup water reactor fuel
nea-1502 IFPE/RISOE-II, Fuel Performance Data from Transient Fission Gas Release
nea-1493 IFPE/RISOE-III, Fuel Performance Data from 3rd Risoe Fission Gas Release
nea-1722 IFPE/ROPE-I, BWR, 4 rods, Ringhals, investigates clad creep-out from Studsvik (1986-1993)
nea-1723 IFPE/ROPE-II, PWR rod over pressure experiment from Studsvik
nea-1310 IFPE/SOFIT, WWER-440 Fuel Thermal Performance and Fission Gas Release
nea-1623 IFPE/SPC-RE-GINNA, Full Length and Segmented Fuel Rodlet Irradiation in PWR
nea-1809 IFPE/STEED-I, Stored Energy / Enthalpy Determination from Studsvik
nea-1557 IFPE/SUPER-RAMP, PCI Failure Threshold for PWR and BWR Fuels
nea-1648 IFPE/TRANS-RAMP, Fuel behaviour data from PWR/BWR TRANS-RAMP I, II, IV experiments
nea-1536 IFPE/TRIBULATION R1, Fuel Rod Behaviour at High Burnup
nea-1738 IFPE/US-PWR-16X16LTA, Lead Test Assembly Extended Burnup Demonstration Program
nea-1677 IFPE/ZAPOROSHYE-V1K, Zaporoshye VVER1000 fuel behaviour data (4-8 cycles, Burnup about 50 MWd/kgUO2)
iaea1415 IRPHE-KNK-II-ARCHIVE, KNK-II fast reactor documents, power history and measured parameters
nea-1660 IRPHE-SNEAK, KFK SNEAK Fast Reactor Experiments, Primary Documentation
nea-1876 IRPHE-VENUS-RECYCLE, Plutonium Recycling Physics Project Critical Experiments
nea-1661 IRPHE-ZEBRA, AEEW Fast Reactor Experiments, Primary Documentation
nea-1687 IRPHE/B&W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments
nea-1765 IRPHE2019-HANDBOOK, International Handbook of Evaluated Reactor Physics Benchmark Experiments
nea-1726 IRPhE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents
nea-1728 IRPhE-HTR-ARCH-01, Archive of HTR Primary Documents
nea-1764 IRPhE-TAPIRO-ARCHIVE, Fast neutron source reactor primary documents, reactor physics experiments
nea-1739 IRPhE/AVR, High Temperature Reactor Experience, Archival Documentation
nea-1759 IRPhE/BERENICE, effective delayed neutron fraction measurements
nea-1713 IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility
nea-1714 IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility
csni1018 IVO-THERMAL MIXING, study mixing of emergency cooling water with primary water during LOCA accident
csni1028 IVO/AIR-WATER-CCFL, Air/water countercurrent flow limitation experiments with full-scale fuel bundle structures
csni1027 IVO/LOOP-SEAL, IVO-Loop Seal Facility (Air/Water), Two-phase behaviour of a PWR cold leg loop seal during LOCA accidents
nea-1811 JDL-IMPORTANCE, Adjoint Function: Physical Basis of Variational & Perturbation Theory in Transport & Diffusion Problems
nea-1843 JDL-REACTOR-KINETICS, Nuclear Reactor Kinetics and Control
nea-1844 JDL-THERMODYNAMICS, Thermodynamics: Frontiers and Foundations
csni0004 LEIBSTADT/STP-2001, BWR/6, Reactor Core Isolation Cooling System Test
csni0034 LOBI/A1-04R, Loop for Blowdown Investigation, PWR Double-Ended Cold-Leg Break Test
csni0035 LOBI/A1-06, Loop for Blowdown Investigation, PWR Double-Ended 2A Cold-Leg Break
csni0036 LOBI/A1-66, Loop for Blowdown Investigation, PWR Double-Ended 2A Cold-Leg Break
csni0037 LOBI/A2-77A, Loop for Blowdown Investigation, PWR MOD2 Small Leak Programme Experiment
csni0038 LOBI/A2-81, Loop for Blowdown Investigation. PWR MOD2 1% Cold-Leg Break
csni0003 LOBI/B-R1M, Loop for Blowdown Investigation, PWR Single-Ended Cold-Leg Break Experiment B
csni0074 LOBI/BT-00, Simulation of a Loss of Feedwater Transient (LOFW)
csni0017 LOFT/L2-3, Loss of Fluid Test, 2nd NRC L2 Large Break LOCA Experiment
csni0016 LOFT/L2-5, Loss of Fluid Test, 3rd NRC L2 Large Break LOCA Experiment
csni0022 LOFT/L3-5, Loss of Fluid Test, 5th NRC L3 Small Break LOCA Experiment
csni0018 LOFT/L3-6, Loss of Fluid Test, 6th NRC L3 Small Break LOCA Experiment
csni0021 LOFT/L3-7, Loss of Fluid Test, 7th NRC L3 Small Break LOCA Experiment
csni0020 LOFT/L6-7, Loss of Fluid Test, Anticipated Transients with Multiple Failures
csni0070 LOFT/L8-2, Severe Core Transient Experiment
csni0019 LOFT/L9-3, Loss of Fluid Test, Anticipated Transients with Multiple Failures
csni0010 LOFT/LP-02-6, Loss of Fluid Test, 1st OECD Large Break Experiment
csni0012 LOFT/LP-FP-1B, Loss of Fluid Test, Fission Product Release Experiment
csni0013 LOFT/LP-FP-2, Loss of Fluid Test, Fission Product Release from Fuel
csni0007 LOFT/LP-FW-1, Loss of Fluid Test, PWR Response to Loss-of-Feedwater Transient
csni0002 LOFT/LP-LB-1, Loss of Fluid Test, Large-Break LOCA Experiment
csni0008 LOFT/LP-SB-1, Loss of Fluid Test, Small Hot Leg Break LOCA, Early Pump
csni0009 LOFT/LP-SB-2, Loss of Fluid Test, Small Hot Leg Break LOCA, Delayed Pump
csni0011 LOFT/LP-SB-3, Loss of Fluid Test, Cold Leg Break LOCA, No High Pressureinjection System (HPIS)
csni0080 MARVIKEN-ATT, Marviken Aerosol Transport Test experiments
csni1001 MARVIKEN-CFT, Marviken Full Scale Critical Flow Tests
csni0078 MARVIKEN-FSCB-I, Marviken Full Scale Containment Blowdown experiments Series I
csni0079 MARVIKEN-FSCB-II, Marviken Full Scale Containment Blowdown experiments Series II
csni1033 MARVIKEN-JIT, Marviken Full Scale Jet Impingement Tests experiments
csni2008 MASCA, In-vessel phenomena during severe accidents
csni2010 MASCA-2, In-vessel phenomena during severe accidents
csni2003 MCCI PROJECT, Molten Core Concrete Interaction Project
csni2017 MCCI-2 PROJECT, Melt Coolability and Concrete Interaction Phase 2 Project
nea-1706 MMRW, Canadian and early British Energy Reports on Nuclear Reactor Theory (1940-1946)
nea-1792 MMRW-BOOKS, Legacy books on slowing down, thermalization, particle transport theory, random processes in reactors
nea-1874 NEACRP-H2O-LATTICES, Compilation of reactor physics measurements in LWRs lattices
csni1011 NEPTUN/5007, PWR LOCA Cooling Heat Transfer Tests for Loft, Boil-Off Experiments
csni1012 NEPTUN/5050, PWR LOCA Cooling Heat Transfer Tests for Loft, Reflood Test
csni1013 NEPTUN/5052, PWR LOCA Cooling Heat Transfer Tests for Loft, Reflood Test
csni2014 OLHF, Sandia Lower Head Failure of the reactor pressure vessel Project
csni0014 OTIS/220100, Once-Through Integral Systems, 2-Phase Natural Circulation and Reflux
csni0015 OTIS/220402, Once-Through Integral Systems, Cold Leg Small Break LOCA
csni0061 PACTEL-ITE06, VVER-440 natural circulation stepwise coolant inventory reduction
csni2004 PAKS PROJECT, the fuel behaviour in accident conditions on the basis of analyses of the PAKS-2 event
csni1014 PATRICIA/GV-6, Steady State Steam Generator Test Facility
csni1002 PDHT-HP, Post Dryout Heat Transfer Experiments, Upflow and Downflow Conditions
csni1003 PDHT-LP, Low Pressure Post Dryout Loop, Upflow Conditions
csni1025 PHEBUS/B9+, Degradation of a PWR Type Core during a severe fuel damage
csni1021 PHEBUS/test-218, Behaviour of a Fuel Rod Bundle during a Large Break LOCA Transient with a two Peaks Temperature History
csni0048 PIPER-1/PO-SB-7, SBLOCA Simulation of Break in BWR-6 Plant at PIPER1 GE BWR Simulator
csni2001 PKL-1, Experimental data on boron dilution and loss of residual heat removal in mid-loop operation (during shutdown)
csni2013 PKL-2, Solving thermal hydraulic safety issues for current PWR and new PWR design concepts
csni2032 PKL-3, Beyond-design-basis accidents and accidents from cold shut-down condition in PWR
csni0072 PKL/K9, Refill and Reflood Experiment in a Simulated PWR Primary System (PKL)
csni2035 PLASMA PROJECT, Plant Safety Monitoring and Assessment System
nea-1789 PMK2-VVER440-REPORTS, Final reports on the PMK-2 projects for VVER Safety Studies
csni2006 PRISME, Fire and smoke propagation tests
csni2042 PRISME-2, Fire and smoke propagation tests Phase 2
csni2200 PSB-VVER, Computer code validation for transient analysis of VVER and RBMK reactors project
nea-1780 PWR-MOX/UOX-TRANS, OECD/NEA US-NRC PWR MOX/UO2 Core Transient Benchmark
nea-1828 Proceedings of PHYSOR'90 conference: Physics of Reactors, Operation, Design and Computation, Marseille, 23-27 April 1990
nea-1912 Proceedings of the 7th International Conference on Nuclear Criticality Safety (ICNC2003), 20-24 Oct. 2003, Tokai-Mura
csni2300 RASPLAV, Refine accident management strategies during a reactor core meltdown
nea-1873 REACTORPHYSICS-62-91, Archive of Reactor Physics Reports and Summaries of [N]EACRP (1962-1991)
nea-1814 REACTORSHIELDING-NMS, Reactor Shielding for Nuclear Engineers by N. M. Schaeffer
csni1022 REBEKA, Behaviour of a Fuel Bundle Simulator during a Specified Heatup and Flooding Period Results
csni1029 REWET, PWR LOCA accidents experiments
nea-1835 ROCKWELL-RSDM, Reactor Shielding Design Manual by Rockwell T. III
csni2009 ROSA PROJECT, resolve issues in thermal-hydraulics analyses relevant to LWR during design basis events
csni2021 ROSA-2, Rig-of-safety Assessment Project
csni0039 ROSA-III/912, BWR Rig of Safety Assessment for LOCA, 5% Split Break Test
csni0040 ROSA-III/916, BWR Rig of Safety Assessment for LOCA, 50% Split Break Test
csni0041 ROSA-III/922, BWR Rig of Safety Assessment for LOCA, 5% Split Break Test
csni0047 ROSA-III/923, BWR Rig of Safety Assessment for LOCA
csni0042 ROSA-III/926, BWR Rig of Safety Assessment for LOCA, 20% Double-Ended Break
csni0043 ROSA-III/952, BWR Rig of Safety Assessment for LOCA, Reference MSL Break Test
csni0044 ROSA-III/971, BWR Rig of Safety Assessment LOCA, Loss of Offsite Power Transient
csni0045 ROSA-III/984, BWR Rig of Safety Assessment for LOCA, 2.8% Split Break Test
csni0046 ROSA-IV/SB-CL-18, Large Scale Test Facility, 5% Cold-Leg Break Test
csni0073 ROSA-IV/SB-CL-27, Large Scale Test Facility, Gravity-Driven Safety Injection
csni1000 S.E.T., CSNI Separate Effects Test Facility Validation Matrix
nea-1694 SATIF/CYCLO-RADSAFE, Health Physics and Radiological Safety of Cyclotrons 10-250 MeV
csni2019 SCIP PROJECT, Studsvik Cladding Integrity Project
csni0027 SEMISCALE/S-06-3(LV), Thermal and Hydraulic Phenomena with LOCA in U-Tube PWR
csni0028 SEMISCALE/S-06-3(SV), Thermal and Hydraulic Phenomena with LOCA in U-Tube PWR
csni0023 SEMISCALE/S-IB-3, Semiscale MOD-2A, 21.7% Communication Cold Leg Break LOCA Experiment
csni0024 SEMISCALE/S-PL-3E-LV, Scaled Reference 4-Loop PWR Experiment, Loss of Offsite Power Simulation
csni0025 SEMISCALE/S-PL-3E-SV, Scaled Reference 4-Loop PWR Experiment, Loss of Offsite Power Simulation
csni0026 SEMISCALE/S-UT-1, Semiscale MOD-2A, 10% Communication Cold Leg Break LOCA Experiment
csni0077 SEMISCALE/TEST1011, Large break LOCA blowdown starting from isothermal conditions of the primary coolant loop
csni2028 SERENA PROJECT, Steam Explosion Resolution for Nuclear Applications Project
csni2020 SETH-2, SESAR Thermal-hydraulics Project
csni2002 SETH/PANDA, Three-dimensional gas flow distributions relevant to in-reactor containments under accidents conditions
csni2000 SETH/PKL, Countermeasures for two types of PWR accidents
csni2030 SFP, Experimental data relevant for hydraulic and ignition phenomena of prototypic water reactor fuel assemblies
nea-1552 SINBAD ACCELERATOR, Shielding Benchmark Experiments
nea-1553 SINBAD FUSION, Neutronics Benchmark Experiments
nea-1517 SINBAD REACTOR, Shielding Benchmark Experiments.
csni1017 SMD/12R305C, Steady state critical flow in nozzles, medium to high pressure conditions
csni0075 SPES/SP-FW-02, Total Loss Feedwater with Emergency Feed Water Delayed in SPES facility
csni2033 STEM, Source Term Evaluation and Mitigation (STEM) Project
csni2007 STEX-II, International Steam Explosion Experimental Data Base
csni0005 TBL/311, Two Bundle Loop Facility, Small Break in Recirculation Line
csni2016 THAI, Thermal-hydraulics, Hydrogen, Aerosols, Iodine (ThAI) Project
csni2031 THAI-2, Thermal-hydraulics, Hydrogen, Aerosols, Iodine (ThAI) Project.
csni1016 THETIS, Single Phase Cooling, Forced and Gravity Reflood, Level Swell Experiments
csni0029 TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA
csni0030 TLTA/6432, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA
csni2012 TMI-VIP, Three Mile Island Reactor Pressure Vessel investigation Project
nea-1682 U3-U5-PU9-CRITICALS, Critical Dimensions of Systems containing U235, Pu239, and U233
csni1007 UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA
csni1004 UPTF/TEST5A/RUN063, Steam/Water Flow Phenom.Blowdown PWR Cold Leg Break LOCA
csni1005 UPTF/TEST8A/RUN112, Flow Patterns in Hot or Cold Leg, PWR Large Break LOCA
csni1006 UPTF/TEST8B/RUN111, Flow Patterns in Hot or Cold Leg, PWR Large Break LOCA
nea-1398 ZZ 3DLWRCT, 3-D LWR Rod Ejection and Rod Withdrawal Benchmarks
nea-1731 ZZ BFBT, OECD/NEA-US/NRC NUPEC BWR Full-size Fine-mesh Bundle Tests Benchmark
nea-1401 ZZ BUC-1/BENCHMARK, NEACRP Benchmark Specifications for Burnup Criticality Calculation
nea-1551 ZZ BWRSB-FORSMARKS, Stability Benchmark Data from BWR FORSMARKS 1 and 2
nea-1454 ZZ BWRSB-RINGHALS1&2, Stability Benchmark Data from BWR RINGHALS-1
nea-1640 ZZ BWRTT, BWR Turbine Trip Transient Benchmark Based on Peach-Bottom 2
nea-1606 ZZ ECN-BUBEBO, ECN-Petten Burnup Benchmark Book, Inventories, Afterheat
nea-1848 ZZ KALININ3, KALININ-3 Coolant Transient Benchmark
nea-1881 ZZ OSKARSHAMN 2, Oskarshamn-2 (O2) BWR Stability Benchmark
nea-1746 ZZ PBMR-400, OECD/NEA PBMR Coupled Neutronics/Thermal Hydraulics Transient Benchmark - The PBMR-400 Core Design
nea-1849 ZZ PSBT, NUPEC PWR Sub-channel Bundle Tests Benchmark
nea-1607 ZZ PWR-AXBUPRO-GKN, Measured Axial Burnup Profiles, NPP Neckarewstheim
uscd1219 ZZ PWR-AXBUPRO-SNL, Computed Axial Burnup Profile Database for PWR
nea-1554 ZZ PWR-MSLB, PWR Main Steam-Line Break Benchmarks, Coupled Neutronics Thermal-Hydraulics
nea-1769 ZZ UAM-LWR, Uncertainty Analysis in Modelling, Coupled Multi-physics and Multi-scale LWR analysis
nea-1693 ZZ V1000CT-1&2, VVER-1000 Main Coolant Pump Switching-on, Coolant Mixing Tests, Main Steam-Line Break Benchmarks
nea-1610 ZZ WPNCS BENCHM REP, Published Articles and Reports on Criticality Safety
nea-1505 ZZ WPPR-1-A/B and ZZ WPPR-2-CYC1, Pu Recycling Benchmark Results
nea-1434 ZZ WPPR-FR-MOX/BNCMK, Benchmark on Pu Burner Fast Reactor