In Pile-fission Gas Release and Swelling of UO2 Fuel in Light Water Reactors

In Pile-fission Gas Release and Swelling of UO2 Fuel in Light Water Reactors PDF Author: Moustafa S. El-Koliel
Publisher:
ISBN:
Category :
Languages : en
Pages : 25

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In Pile-fission Gas Release and Swelling of UO2 Fuel in Light Water Reactors

In Pile-fission Gas Release and Swelling of UO2 Fuel in Light Water Reactors PDF Author: Moustafa S. El-Koliel
Publisher:
ISBN:
Category :
Languages : en
Pages : 25

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Fission Gas Behaviour in Water Reactor Fuels

Fission Gas Behaviour in Water Reactor Fuels PDF Author:
Publisher: Paris, France : Nuclear Energy Agency, Organisation for Economic Co-operation and Development
ISBN:
Category : Science
Languages : en
Pages : 572

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Book Description
Communicates the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors.

Pellet-clad Interaction in Water Reactor Fuels

Pellet-clad Interaction in Water Reactor Fuels PDF Author:
Publisher: OECD Publishing
ISBN:
Category : Business & Economics
Languages : en
Pages : 562

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Book Description
This publication sets out the findings of an international seminar, held in Aix-en-Provence, France in March 2004, which considered recent progress in the field of pellet-clad interaction in light water reactor fuels. It also reviews current understanding of relevant phenomena and their impact on the nuclear fuel rod under the widest possible conditions, and about both uranium-oxide and mixed-oxide fuels.

Fission Gas Induced Fuel Swelling in Low and Medium Burnup Fuel During High Temperature Transients. [PWR].

Fission Gas Induced Fuel Swelling in Low and Medium Burnup Fuel During High Temperature Transients. [PWR]. PDF Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :

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Book Description
The behavior of light water reactor fuel elements under postulated accident conditions is being studied by the EG and G Idaho, Inc., Thermal Fuels Behavior Program for the Nuclear Regulatory Commission. As a part of this program, unirradiated and previously irradiated, pressurized-water-reactor type fuel rods were tested under power-cooling-mismatch (PCM) conditions in the Power Burst Facility (PBF). During these integral in-reactor experiments, film boiling was produced on the fuel rods which created high fuel and cladding temperatures. Fuel rod diameters increased in the film boiling region to a greater extent for irradiated rods than for unirradiated rods. The purpose of the study was to investigate and assess the fuel swelling which caused the fuel rod diameter increases and to evaluate the ability of an analytical code, the Gas Release and Swelling Subroutine - Steady-State and Transient (GRASS-SST), to predict the results.

In-pile Fission-gas-release Behavior of Alumina-coated UO2 Particles Irradiated to High Burnup

In-pile Fission-gas-release Behavior of Alumina-coated UO2 Particles Irradiated to High Burnup PDF Author: Gilbert E. Raines
Publisher:
ISBN:
Category : Fission gases
Languages : en
Pages : 46

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In-Pile Fission-Gas Release from UO2

In-Pile Fission-Gas Release from UO2 PDF Author: JB. Melehan
Publisher:
ISBN:
Category : Complex compounds
Languages : en
Pages : 10

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Book Description
A program has been conducted and is continuing at Battelle Memorial Inst. under the sponsorship of the United States Atomic Energy Commission to study the release of fission gas from UO2 and to relate the release to material properties, both chemical and structural. Work in the program has involved the preparation and characterization of sintered polycrystalline and fused single-crystal UO2 samples, the construction of a beam-tube loop facility for the irradiation of the UO2 under controlled temperature and atmosphere, and the collection and analysis of released fission gases during irradiation. Irradiation experiments have been conducted in purified helium-hydrogen atmosphere at temperatures from 500 to 1500 F, with fission gases continuously collected and analyzed for krypton and xenon during irradiation of natural uranium specimens. Irradiation exposures of the specimens have been in the region of 1018 nvt.

Viability of Inert Matrix Fuel in Reducing Plutonium Amounts in Reactors

Viability of Inert Matrix Fuel in Reducing Plutonium Amounts in Reactors PDF Author: International Atomic Energy Agency
Publisher: IAEA
ISBN:
Category : Business & Economics
Languages : en
Pages : 100

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Book Description
The reactors around the world have produced more than 2000 tonnes of plutonium, contained in spent fuel or as separated forms through reprocessing. Disposition of fissile materials has become a primary concern of nuclear non-proliferation efforts worldwide. There is a significant interest in IAEA Member States to develop proliferation resistant nuclear fuel cycles for incineration of plutonium such as inert matrix fuels (IMFs). This publication reviews the status of potential IMF candidates and describes several identified candidate materials for both fast and thermal reactors: MgO, ZrO2, SiC, Zr alloy, SiAl, ZrN; some of these have undergone test irradiations and post irradiation examination. Also discussed are modelling of IMF fuel performance and safety analysis. System studies have identified strategies for both implementation of IMF fuel as homogeneous or heterogeneous phases, as assemblies or core loadings and in existing reactors in the shorter term, as well as in new reactors in the longer term.

Fission Gas Release from Fuel at High Burnup

Fission Gas Release from Fuel at High Burnup PDF Author: Ralph O. Meyer
Publisher:
ISBN:
Category : Fission gases
Languages : en
Pages : 68

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Fission-product Release from UO2

Fission-product Release from UO2 PDF Author: W. B. Cottrell
Publisher:
ISBN:
Category : Nuclear merchant ships
Languages : en
Pages : 148

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Fission Gas Release from UO{sub 2+x} in Defective Light Water Reactor Fuel Rods

Fission Gas Release from UO{sub 2+x} in Defective Light Water Reactor Fuel Rods PDF Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 14

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Book Description
A simplified semi-empirical model predicting fission gas release form UO{sub 2+x} fuel to the fuel rod plenum as a function of stoichiometry excess (x) is developed to apply to the fuel of a defective LWR fuel rod in operation. The effect of fuel oxidation in enhancing gas diffusion is included as a parabolic dependence of the stoichiometry excess. The increase of fission gas release in a defective BWR fuel rod is at the most 3 times higher than in an intact fuel rod because of small extent of UO2 oxidation. The major enhancement contributor in fission gas release of UO{sub 2+x} fuel is the increased diffusivity due to stoichiometry excess rather than the higher temperature caused by degraded fuel thermal conductivity.