Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions
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概要
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The release of volatile fission products from high-burnup UO2 fuel was examined in a steam atmosphere under severe accident conditions as a part of the VEGA program. The effects of fuel oxidation and dissolution were totally evaluated, by comparing the results with those from previous inert, hydrogen and steam atmosphere tests. It was shown that the oxidation of UO2 to UO2+x by steam generally enhances Cs and Kr release. However, the enhancement becomes smaller above the melting temperature of Zircaloy, about 2030 K, likely due to reduction of UO2+x by molten Zircaloy. The burst release of Cs occurs above about 2300 K in the hydrogen atmosphere, while the release rate does not increase so significantly for the examined temperature range (<2800 K) in the steam atmosphere. Analysis of the hydrogen atmosphere test showed that fuel dissolution is apparently connected with the burst release and that a large fraction of Cs is quickly released from the dissolved fuel above 2300 K. It is considered that the fuel dissolution rate in the steam atmosphere is about 1/1000 of that in the hydrogen atmosphere.
- 2007-11-25
著者
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NAGASE Fumihisa
Japan Atomic Energy Agency
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FUKETA Toyoshi
Japan Atomic Energy Agency
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NAKAMURA Takehiko
Japan Atomic Energy Agency
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Kudo Tamotsu
Japan Atomic Energy Agency
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Fuketa T
Japan Atomic Energy Agency
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Nakamura T
Japan Atomic Energy Res. Inst. Ibaraki‐ken
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KIDA Mitsuko
Japan Atomic Energy Agency
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NAGASE Fumihisa
Nuclear Safety Research Center, Japan Atomic Energy Agency
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Nagase Fumihisa
Nuclear Safety Research Center Japan Atomic Energy Agency
関連論文
- Identification of Radial Position of Fission Gas Release in High-Burnup Fuel Pellets under RIA Conditions
- Stress Intensity Factor at the Tip of Cladding Incipient Crack in RIA-Simulating Experiments for High-Burnup PWR Fuels
- Optimized Ring Tensile Test Method and Hydrogen Effect on Mechanical Properties of Zircaloy Cladding in Hoop Direction
- Terminal Solid Solubility of Hydrogen in Hafnium
- Behavior of High Burn-up Fuel Cladding under LOCA Conditions
- Thermal Conductivity Change in High Burnup MOX Fuel Pellet
- Evaluation of Initial Temperature Effect on Transient Fuel Behavior under Simulated Reactivity-Initiated Accident Conditions
- Clad-to-Coolant Heat Transfer in NSRR Experiments
- Radionuclide Release from Mixed-Oxide Fuel under High Temperature at Elevated Pressure and Influence on Source Terms
- Proposal of Simplified Model of Radionuclide Release from Fuel under Severe Accident Conditions Considering Pressure Effect
- Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions
- Releases of Cesium and Poorly Volatile Elements from UO_2 and MOX Fuels under Severe Accident Conditions
- Fracture Behavior of Irradiated Zircaloy-4 Cladding under Simulated LOCA Conditions
- Enhancement of Cesium Release from Irradiated Fuel at Temperature above 2,800 K
- Influence of Pressure on Cesium Release from Irradiated Fuel at Temperatures up to 2,773K
- Behavior of Coated Fuel Particle of High-Temperature Gas-Cooled Reactor under Reactivity-Initiated Accident Conditions
- Fission Gas Release in BWR Fuel with a Burnup of 56GWd/t during Simulated Reactivity Initiated Accident (RIA) Condition
- Measurements of Crystal Lattice Strain and Crystallite Size in Irradiated UO_2 Pellet by X-ray Diffractometry
- Behavior of 60 to 78MWd/kgU PWR Fuels under Reactivity-Initiated Accident Conditions
- Fission Gas Release in Irradiated UO_2 Fuel at Burnup of 45GWd/t during Simulated Reactivity Initiated Accident (RIA) Condition
- Failure Thresholds of High Burnup BWR Fuel Rods under RIA Conditions
- Behavior of Irradiated ATR/MOX Fuel under Reactivity Initiated Accident Conditions
- Fission Gas Release Behavior of High Burnup UO_2 Fuel under Reactivity Initiated Accident Conditions
- Fission Gas Induced Cladding Deformation of LWR Fuel Rods under Reactivity Initiated Accident Conditions
- Hydrogen Generation during Cladding/Coolant Interactions under Reactivity Initiated Accident Conditions
- Behavior of Pre-hydrided Zircaloy-4 Cladding under Simulated LOCA Conditions
- Thermal Stress Analysis of High-Burnup LWR Fuel Pellet Pulse-Irradiated in Reactivity-Initiated Accident Conditions
- Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6 Code
- Investigation of Hydride Rim Effect on Failure of Zircaloy-4 Cladding with Tube Burst Test
- Influence of Hydride Re-orientation on BWR Cladding Rupture under Accidental Conditions
- Influence of Cladding-Peripheral Hydride on Mechanical Fuel Failure under Reactivity-Initiated Accident Conditions
- Effect of Cladding Surface Pre-oxidation on Rod Coolability under Reactivity Initiated Accident Conditions
- RANNS Code Analysis on the Local Mechanical Conditions of Cladding of High Burnup Fuel Rods under PCMI in RIA-Simulated Experiments in NSRR
- Effect of Pre-Hydriding on Thermal Shock Resistance of Zircaloy-4 Cladding under Simulated Loss-of-Coolant Accident Conditions
- Effect of Cooling History on Cladding Ductility under LOCA Conditions
- Oxidation kinetics of Low-Sn Zircaloy-4 at the Temperature Range from 773 to 1, 573 K
- B_4C/Zircaloy Reaction at Temperatures from 1, 173 to 1, 953 K
- Development of Leak Detection System Using High Temperature-Resistant Microphones
- Ring Compression Ductility of High-Burnup Fuel Cladding after Exposure to Simulated LOCA Conditions
- Ring Compression Ductility of High-Burnup Fuel Cladding after Exposure to Simulated LOCA Conditions
- Irradiated Fuel Behavior under Power Oscillation Conditions
- Oxidation kinetics of Low-Sn Zircaloy-4 at the Temperature Range from 773 to 1, 573 K
- Influence of Hydride Re-orientation on BWR Cladding Rupture under Accidental Conditions