Behaviour of fission gas in the rim region of high burn-up UO2 fuel pellets with particular reference to results from an XRF investigation

1999 ◽  
Vol 264 (1-2) ◽  
pp. 99-112 ◽  
Author(s):  
M Mogensen ◽  
J.H Pearce ◽  
C.T Walker
Keyword(s):  
Fuel Pellets ◽  
Fission Gas

Grain Subdivision Fission Gas Swelling Model for UO2

MRS Advances ◽  
2016 ◽  
Vol 1 (35) ◽  
pp. 2465-2470
Author(s):  
Thomas Winter ◽  
Richard Hoffman ◽  
Chaitanya S. Deo
Keyword(s):  
Grain Size ◽  
Grain Boundaries ◽  
Gas Diffusion ◽  
Fine Grain ◽  
Fuel Pellets ◽  
Model And Simulation ◽  
Nucleation Sites ◽  
Fission Gas ◽  
High Burnup ◽  
Swelling Model

ABSTRACTUnder high burnup UO2 fuel pellets can experience high burnup structure (HBS) at the rim also known as rim effect. The HBS is exceptionally porous with fine grain sizes. HBS increases the swelling further than it would have achieved at a larger grain size. A theoretical swelling model is used in conjunction with a grain subdivision simulation to calculate the swelling of UO2. In UO2 the nucleation sites are at vacancies and the bubbles are concentrated at grain boundaries. Vacancies are created due to irradiation and gas diffusion is dependent on vacancy migration. In addition to intragranular bubbles, there are intergranular bubbles at the grain boundaries. Over time as intragranular bubbles and gas atoms accumulate on the grain boundaries, the intergranular bubbles grow and cover the grain faces. Eventually they grow into voids and interconnect along the grain boundaries, which can lead to fission gas release when the interconnection reaches the surface. This is known as the saturation point. While the swelling model used does not originally incorporate a changing grain size, the simulation allows for more accurate swelling calculations by introducing a fractional HBS based on the temperature and burnup of the pellet. The fractional HBS is introduced with a varying grain size. Our simulations determine the level of swelling and saturation as a function of burnup by combining an independent model and simulation to obtain a more comprehensive model.

Identification of Radial Position of Fission Gas Release in High-Burnup Fuel Pellets under RIA Conditions

2010 ◽  
Vol 47 (2) ◽  
pp. 202-210 ◽  
Author(s):  
Hideo SASAJIMA ◽  
Tomoyuki SUGIYAMA ◽  
Toshinori CHUTO ◽  
Fumihisa NAGASE ◽  
Takehiko NAKAMURA ◽  
...  
Keyword(s):  
Radial Position ◽  
Gas Release ◽  
Fuel Pellets ◽  
Fission Gas ◽  
High Burnup ◽  
Fission Gas Release

scholarly journals Modeling Axial Relocation of Fragmented Fuel During Loss of Coolant Conditions by Using ABAQUS

2020 ◽  
Author(s):  
Zehua Ma ◽  
Koroush Shirvan ◽  
Wei Li ◽  
Yingwei Wu
Keyword(s):  
Thermal Stresses ◽  
Element Model ◽  
Irradiation Creep ◽  
Irradiation Hardening ◽  
Fuel Pellets ◽  
Loss Of Coolant Accident ◽  
Fission Gas ◽  
Loss Of Coolant ◽  
Single Balloon ◽  
Nuclear Fuel Pellets

Abstract In a light-water reactor, during normal operating condition, the UO2 nuclear fuel pellets undergo fragmentation primarily due to presence of thermal stresses, fission gas development and pellet-clad mechanical interaction. Under Loss of Coolant Accident (LOCA) conditions, a portion of fuel fragments can freely move downwards to the ballooning region due to the significant cladding deformation. The fuel relocation can localize the heat load and in turn accelerate the cladding balloon and burst process. Cladding burst is of great concern because of the potential for fuel dispersal into coolant and clad structural stability. In our work, we built up a finite element model considering cladding balloon, fuel relocation and its resultant thermal feedback during LOCA condition with ABAQUS. The clad balloon model includes phase transformation, swelling, thermal and irradiation creep, irradiation hardening and annealing and other important thermal-mechanical properties. The mass of relocation model was verified against the analytical cases of single balloon and twin balloons. The cladding balloon model combined with fuel thermal conductivity degradation was verified against fuel performance code, FRAPTRAN. Finally, with the evolution of pellet-cladding gap, the fuel mass relocation was calculated and compared against the IFA-650.4 transient test from the Halden reactor.

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

2019 ◽  
Vol 5 ◽  
pp. 11 ◽  
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.

Study of Pore Modality of Irradiated UO2 Fuel Assembly

2017 ◽  
Vol 727 ◽  
pp. 693-697
Author(s):  
Yan Hua Peng ◽  
Fei Wang ◽  
Wei Zhu ◽  
Yan Jiang ◽  
Hong Kui Tang ◽  
...  
Keyword(s):  
Grain Boundary ◽  
Fuel Assembly ◽  
High Performance ◽  
Local Area ◽  
Transgranular Crack ◽  
Fuel Pellets ◽  
Micro Cracks ◽  
Fission Gas ◽  
Main Factors ◽  
Irradiated Fuel

UO2 fuel pellets may be swelling and recrystallization during irradiation. Density, dimension and distribution of pores are main factors to induce irradiation swelling, especially the size distribution of pellet pores plays an important role. 4×4-4 fuel assembly was a high performance fuel assembly which was self-designed and manufactured, the average burn-up of the fuel assembly was 42GWd/tU.For studying the effect of irradiation on pore modality, the specimens of irradiated UO2 fuel pellets were taken from 4×4-4 fuel assembly after dismantling, microscopic structure and distribution of pores for UO2 fuel assembly by scanning electron microscopy were studied in this paper. The results showed that there were many cracks in fuel pellets, most micro-cracks were transgranular crack. The release rate of fission gas with burn up were augmentation, which was consistent with the porosity of diversity burp up fuel rod. Pores were distributed non-uniformly in irradiated fuel pellets, gathered at local area and more obvious connectivity of pores. The size of pores after the irradiation was between 0.2~1.2 μm, and mostly distributed at 0.3μm ~0.6 μm; The pores at grain boundary of two adjacent grains was less, the pores at grain boundary were distributed by the way of triangle or quadrilateral. The size of pores was increased than pre-irradiation, but ratio of pores and density of pores were decreased obviously, the phenomenon of irradiation densification was occurred in fuel pellets after irradiated. Recrystallization and Rim structure effect were not found.

Characterization of Low-Gas-Release LHR Fuels by Transmission Electron Microscopy

1988 ◽  
Vol 127 ◽  
Author(s):  
L. E. Thomas ◽  
R. J. Guenther
Keyword(s):  
Electron Microscopy ◽  
Transmission Electron Microscopy ◽  
Fission Products ◽  
Solid Solution Alloy ◽  
Gas Release ◽  
Spent Fuel ◽  
Fuel Pellets ◽  
Transmission Electron ◽  
Fission Gas ◽  
Waste Repositories

ABSTRACTAnalytical transmission electron microscopy (AEM) was performed on light-water reactor (LWR) spent fuel samples from different radial locations in several low-gas release fuels. The examinations revealed high densities of sub-micrometer gas and solid aggregates. Precipitate coarsening and a change in the nature of the fission gas precipitates was observed near the centers of the fuel pellets. The fission-product phases were: 1) ε-ruthenium (Mo-Ru-Tc-Rh-Pd solid-solution alloy); 2) unidentified gases in intra- and intergranular bubbles from fuel edge to mid-radius locations; 3) dense, highly pressurized xenon/krypton “particles” associated with intragranular °-phase near the fuel centers. All other fission products apparently remained in solution. The nature and distribution of these phases are likely to affect fuel behavior in waste repositories and may be important for understanding fission gas behavior.

scholarly journals Iodine release from high-burnup fuel structures: Separate-effect tests and simulated fuel pellets for better understanding of iodine behaviour in nuclear fuels

MRS Advances ◽  
2021 ◽  
Author(s):  
Janne Heikinheimo ◽  
Teemu Kärkelä ◽  
Václav Tyrpekl ◽  
Matĕj̆ Niz̆n̆anský ◽  
Mélany Gouëllo ◽  
...  
Keyword(s):  
Fission Product ◽  
Nuclear Fuel ◽  
Chemical Species ◽  
Outer Edge ◽  
Fuel Pellet ◽  
Fuel Pellets ◽  
Simulated Fuel ◽  
Fission Gas ◽  
High Burnup ◽  
Iodine Release

Abstract Iodine release modelling of nuclear fuel pellets has major uncertainties that restrict applications in current fuel performance codes. The uncertainties origin from both the chemical behaviour of iodine in the fuel pellet and the release of different chemical species. The structure of nuclear fuel pellet evolves due to neutron and fission product irradiation, thermo-mechanical loads and fission product chemical interactions. This causes extra challenges for the fuel behaviour modelling. After sufficient amount of irradiation, a new type of structure starts forming at the cylindrical pellet outer edge. The porous structure is called high-burnup structure or rim structure. The effects of high-burnup structure on fuel behaviour become more pronounced with increasing burnup. As the phenomena in the nuclear fuel pellet are diverse, experiments with simulated fuel pellets can help in understanding and limiting the problem at hand. As fission gas or iodine release behaviour from high-burnup structure is not fully understood, the current preliminary study focuses on (i) sintering of porous fuel samples with Cs and I, (ii) measurements of released species during the annealing experiments and (iii) interpretation of the iodine release results with the scope of current fission gas release models. Graphical abstract

Enhanced Accident Tolerance of Thoria-Based Nuclear Fuels

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
B. Szpunar ◽  
J. A. Szpunar
Keyword(s):  
Thermal Conductivity ◽  
Nuclear Fuel ◽  
Nuclear Reactor ◽  
Harsh Environment ◽  
Nuclear Fuels ◽  
High Melting Point ◽  
Excessive Heat ◽  
Fuel Pellets ◽  
Fission Gas ◽  
Higher Temperature

Abstract Many factors need to be investigated before alternative nuclear fuel can be adapted for service in the harsh environment of a nuclear reactor. Urania, used conventionally as a nuclear fuel, has a low thermal conductivity, which degrades with increasing stoichiometric deviation. Thoria-based fuel has been considered as an alternative fuel, since it does not oxidize and has a high melting point and higher thermal conductivity. Simulations have shown that the fuel melting observed in urania fuel rods during an accident with steam ingress should not be observed (or will be delayed) in thoria as its thermal conductivity remains high enough to dissipate excessive heat in the center of the fuel pellets. The thermal gradient also remains low and therefore thermal stress is reduced, which should improve the longevity of the fuel. Thoria also has some other desirable properties as our calculations predict a significantly higher temperature of oxygen lattice premelting than urania. Furthermore, we found that the diffusion of fission gas, e.g., helium, is strongly affected by oxygen diffusion and therefore is slower in thoria for the temperatures where the oxygen lattice premelts in urania, but not in thoria.

Sign in / Sign up

Export Citation Format

Share Document