Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6 Code
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概要
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Capabilities of the FEMAXI-6 code to analyze the behavior of high burnup MOX fuels in LWRs have been evaluated. Coolant conditions, detailed power histories and specifications of the MIMAS-MOX fuel rods, rod 10 and rod 11, of IFA-597.4-7 irradiated in the Halden reactor were input, and calculated rod internal pressures and pellet center temperatures were compared with the measured data for the range of 0–31 MWd/kgUO2. Some sensitivity studies were conducted mainly with respect to pellet thermal conductivity and swelling rate to investigate the changes in thermal behavior and their effects on fission gas release.In the irradiation period up to about 23 MWd/kgUO2, the calculated pellet center temperatures sufficiently agreed with the measured data and also the calculated rod internal pressures reproduced the tendency of an increase in the measured rod internal pressures. These results suggest that fission gas release from MOX fuels can be reasonably predicted by a diffusion process that is modeled in UO2 pellet grains. On the other hand, the steep increase in the measured rod internal pressures observed at the power ramp around 23 MWd/kgUO2 cannot be reproduced by FEMAXI-6 and can be regarded as the result of a relatively large amount of gas release, which possibly caused a pellet-cladding-gap closure through pellet gas-bubble swelling.
- 社団法人 日本原子力学会の論文
- 2007-08-25
著者
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FUKETA Toyoshi
Japan Atomic Energy Agency
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UDAGAWA Yutaka
Japan Atomic Energy Agency
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SUZUKI Motoe
Japan Atomic Energy Agency
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Udagawa Yutaka
Research Laboratory For Nuclear Reactors Tokyo Institute Of Technology
関連論文
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- 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
- FEMAXI-6 Code Verification with MOX Fuels Irradiated in Halden Reactor
- 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
- Behavior of 60 to 78MWd/kgU PWR Fuels under Reactivity-Initiated Accident Conditions
- 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
- 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
- Effects of Fuel and Coolant Temperatures and Neutron Fluence on CANDLE Burnup Calculation
- Influence of Hydride Re-orientation on BWR Cladding Rupture under Accidental Conditions
- Oxidation of zircaloy-4 under high temperature steam atmosphere and its effect on ductility of cladding.