Fission gas release from irradiated mixed-oxide fuel pellet during simulated reactivity-initiated accident conditions: Results of BZ-3 and BZ-4 tests

The fuel pellet-cladding interaction (PCI) of liquid-metal fast breeder reactor (LMFBR) fuel elements or fuel rods at unsteady state is analyzed and discussed based on experimental results. In the analyses, the heat generation, fuel restructuring, temperature distribution, gap conductance, irradiation swelling, irradiation creep, fuel burnup, fission gas release, fuel pellet cracking, crack healing, cladding cracking, yield failure and fracture failure of the fuel elements are taken into consideration. To improve the sintered (U,Pu)O2 fuel performance and reactor core safety at high temperature and fuel burnup, it is desirable to (a) increase and maintain the ductility of cladding material, (b) provide sufficient gap thickness and plenum space for accommodating fission gas release, (c) keep ramps-power increase rate slow and gentle, and (d) reduce the intensity and frequency of transient PCI in order to avoid intense stress fatigue cracking (SFC) and stress corrosion cracking (SCC) due to fission product compounds CsI, CdI2, Cs2Te, etc. at the inner cladding surface of the fuel elements during PCI.

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Modelling fission gas behaviour in mixed oxide fuel under normal and off-normal conditions in fast reactors

Nuclear Materials for Fission Reactors ◽

10.1016/b978-0-444-89571-4.50048-7 ◽

1992 ◽

pp. 308-311

Author(s):

L. Väth

Keyword(s):

Mixed Oxide ◽

Oxide Fuel ◽

Fast Reactors ◽

Mixed Oxide Fuel ◽

Fission Gas

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Modelling of fine fragmentation and fission gas release of UO2 fuel in accident conditions

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2019030 ◽

2019 ◽

Vol 5 ◽

pp. 11 ◽

Cited By ~ 4

Author(s):

Lars O. Jernkvist

Keyword(s):

External Pressure ◽

Gas Content ◽

Gas Release ◽

Fuel Pellets ◽

Fission Gas ◽

Fuel Fragmentation ◽

Accident Conditions ◽

Reactor Accidents ◽

Operating History ◽

Fission Gas Release

In reactor accidents that involve rapid overheating of oxide fuel, overpressurization of gas-filled bubbles and pores may lead to rupture of these cavities, fine fragmentation of the fuel material, and burst-type release of the cavity gas. Analytical rupture criteria for various types of cavities exist, but application of these criteria requires that microstructural characteristics of the fuel, such as cavity size, shape and number density, are known together with the gas content of the cavities. In this paper, we integrate rupture criteria for two kinds of cavities with models that calculate the aforementioned parameters in UO2 LWR fuel for a given operating history. The models are intended for implementation in engineering type computer programs for thermal-mechanical analyses of LWR fuel rods. Here, they have been implemented in the FRAPCON and FRAPTRAN programs and validated against experiments that simulate LOCA and RIA conditions. The capabilities and shortcomings of the proposed models are discussed in light of selected results from this validation. Calculated results suggest that the extent of fuel fragmentation and transient fission gas release depends strongly on the pre-accident fuel microstructure and fission gas distribution, but also on rapid changes in the external pressure exerted on the fuel pellets during the accident.

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Mixed-oxide fuel pellet fabrication and fuel rod assembly for Dresden 1, Batch 7 reload

10.2172/4306313 ◽

1974 ◽

Author(s):

L. Hower, Jr.

Keyword(s):

Mixed Oxide ◽

Fuel Pellet ◽

Oxide Fuel ◽

Mixed Oxide Fuel ◽

Fuel Rod

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Axial Fission Gas Transport in Nuclear Fuel Rods

Defect and Diffusion Forum ◽

10.4028/www.scientific.net/ddf.283-286.262 ◽

2009 ◽

Vol 283-286 ◽

pp. 262-267

Author(s):

M.T. del Barrio ◽

Luisen E. Herranz

Keyword(s):

Nuclear Fuel ◽

Gas Transport ◽

Fuel Pellet ◽

Gas Release ◽

Gas Mixing ◽

Fuel Performance ◽

Fuel Rods ◽

Fuel Rod ◽

Fission Gas ◽

Fission Gas Release

Fission of fissile uranium or plutonium nucleus in nuclear fuel results in fission products. A small fraction of them are volatile and can migrate under the effect of concentration gradients to the grain boundaries of the fuel pellet. Eventually, some fission gases are released to the rod void volumes by a thermally activated process. Local transients of power generation could distort even further the already non-uniform axial power and fission gas concentration profiles in fuel rods. Most of the current fuel rod performance codes neglects these gradients and the resulting axial fission gas transport (i.e., gas mixing is considered instantaneous). Experimental evidences, however, highlight axial gas mixing as a real time-dependent process. The thermal feedback between fission gas release, gap composition and fuel temperature, make the “prompt mixing assumption” in fuel performance codes a key point to investigate due to its potential safety implications. This paper discusses the possible scenarios where axial transport can become significant. Once the scenarios are well characterized, the available database is explored and the reported models are reviewed to highlight their major advantages and shortcomings. The convection-diffusion approach is adopted to simulate the axial transport by decoupling both motion mechanisms (i.e., convection transport assumed to be instantaneous) and a stand-alone code has been developed. By using this code together with FRAPCON-3, a prospective calculation of the potential impact of axial mixing is conducted. The results show that under specific but feasible conditions, the assumption of “prompt axial mixing” could result in temperature underestimates for long periods of time. Given the coupling between fuel rod thermal state and fission gas release to the gap, fuel performance codes predictions could deviate non-conservatively. This work is framed within the CSN-CIEMAT agreement on “Thermo-Mechanical Behaviour of the Nuclear Fuel at High Burnup”.

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