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 |
ICSBEP2022-2023-HB, 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 |