Behavior of 60 to 78MWd/kgU PWR Fuels under Reactivity-Initiated Accident Conditions
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概要
- 論文の詳細を見る
To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) is being studied in the Nuclear Safety Research Reactor (NSRR) program of the Japan Atomic Energy Agency (JAEA). The paper presents recent results obtained from the NSRR power burst experiments with high burnup fuels, and discusses effects of pellet expansion as PCMI (Pellet-Cladding Mechanical Interaction) loading and cladding embrittlement primarily due to hydrogen absorption. Results from the recent four experiments on high burnup (about 60 to 78 MWd/kgU) PWR UO2 rods with advanced cladding alloys showed that the fuel rods with improved corrosion resistance have larger safety margin against the PCMI failure than conventional Zircaloy-4 rods. The tests also suggested that the smaller inventory of inter-granular gas in the pellets with the large grain could reduce the fission gas release during the RIA transient; and high burnup structure in pellet periphery (so-called rim structure) does not have strong effect on reduction of the failure threshold because the PCMI load is produced primarily by solid thermal expansion.
- 社団法人 日本原子力学会の論文
- 2006-09-25
著者
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NAGASE Fumihisa
Japan Atomic Energy Agency
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SUGIYAMA Tomoyuki
Japan Atomic Energy Agency
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FUKETA Toyoshi
Japan Atomic Energy Agency
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Fuketa T
Japan Atomic Energy Agency
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NAGASE Fumihisa
Nuclear Safety Research Center, Japan Atomic Energy Agency
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Sugiyama T
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
- 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
- Behavior of Coated Fuel Particle of High-Temperature Gas-Cooled Reactor under Reactivity-Initiated Accident Conditions
- Application of Proton-conducting Ceramics and Polymer Permeable Membranes for Gaseous Tritium Recovery
- 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
- 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
- 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