| 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 |
| csni2044 | ATLAS-2 PROJECT, 2nd Phase of 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 |
| csni2041 | BSAF-2, Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station Project, Phase 2 |
| 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 |
| 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 |
| csni2043 | HYMERES-1, Hydrogen Mitigation Experiments for Reactor Safety Project, phase 1 |
| 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 | ICSBEP2021-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-M501/D10, Belgonucleaire Beznau-1 PWR irradiated MOX Fuel Rod M501/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, 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-1, 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, Fission Gas Release at High Burn-up, Effect of a Power Cycling |
| nea-1510 | IFPE/HBEP, 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 & IFA-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, (JAEA) 6 rods |
| nea-1860 | IFPE/IFA-519.9, Three PWR rods irradiated to 90 MWd/kg UO2 |
| nea-1549 | IFPE/IFA-533.2, Fuel Thermal Behaviour at High Burnup, Halden Reactor |
| nea-1684 | IFPE/IFA-534.14, fission gas release as a function of burnup at high power (52-55 MWd/kg) |
| nea-1548 | IFPE/IFA-535.5&6, Fission Gas Release, Power Ramps, High Burnup Fuel |
| nea-1547 | IFPE/IFA-562.1, 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 & 650.2, LOCA testing at Halden, Two experiments, IFA-650 series |
| nea-1921 | IFPE/IFA-650.9-10-11, LOCA testing at Halden, 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, 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-1, Fission gas release from high-burnup water reactor fuel |
| nea-1502 | IFPE/RISOE-2, Fuel Performance Data from Transient Fission Gas Release |
| nea-1493 | IFPE/RISOE-3, Fuel Performance Data from 3rd Risoe Fission Gas Release |
| nea-1722 | IFPE/ROPE-1, 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, 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) |
| nea-1715 | IRPHE-JAPAN, Reactor Physics Experiments carried out in Japan |
| 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-1662 | IRPHE/JOYO MK-II, JOYO MK-II core management and characteristics database |
| nea-1765 | IRPHE2022/23-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-1747 | MONTE-CARLO-WS-2005, Proceedings of Monte Carlo Criticality Calculations & TRIPOLI-IV Workshops 2005 |
| nea-1926 | N-THERMALISATION, Notes on the scattering of thermal neutrons |
| 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-1933 | Proceedings of the 12th International Conference on Nuclear Criticality Safety (ICNC2023), 1-6 Oct. 2023, Sendai |
| 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 |
| nea-1925 | TCOFF, Thermodynamic Char. of Fuel Debris and Fission Products based on Scenario Analysis of Severe Accident Progression |
| csni2016 | THAI, Thermal-hydraulics, Hydrogen, Aerosols, Iodine (ThAI) Project, First phase (2007-2009) |
| csni2031 | THAI-2, Thermal-hydraulics, Hydrogen, Aerosols, Iodine (ThAI) Project, Second phase (2011-2014) |
| csni2045 | THAI-3, Thermal-hydraulics, Hydrogen, Aerosols, Iodine (ThAI) Project, Third phase (2016-2019) |
| 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/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 |
