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

scholarly journals Properties of the high burnup structure in nuclear light water reactor fuel

2017 ◽  
Vol 105 (11) ◽  
Author(s):  
Thierry Wiss ◽  
Vincenzo V. Rondinella ◽  
Rudy J. M. Konings ◽  
Dragos Staicu ◽  
Dimitrios Papaioannou ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Nuclear Fuel ◽  
Experimental Studies ◽  
Fuel Pellet ◽  
Extended Defects ◽  
Safety Margins ◽  
Fission Gas ◽  
High Burnup ◽  
Formed Structure

AbstractThe formation of the high burnup structure (HBS) is possibly the most significant example of the restructuring processes affecting commercial nuclear fuel in-pile. The HBS forms at the relatively cold outer rim of the fuel pellet, where the local burnup is 2–3 times higher than the average pellet burnup, under the combined effects of irradiation and thermo-mechanical conditions determined by the power regime and the fuel rod configuration. The main features of the transformation are the subdivision of the original fuel grains into new sub-micron grains, the relocation of the fission gas into newly formed intergranular pores, and the absence of large concentrations of extended defects in the fuel matrix inside the subdivided grains. The characterization of the newly formed structure and its impact on thermo-physical or mechanical properties is a key requirement to ensure that high burnup fuel operates within the safety margins. This paper presents a synthesis of the main findings from extensive studies performed at JRC-Karlsruhe during the last 25 years to determine properties and behaviour of the HBS. In particular, microstructural features, thermal transport, fission gas behaviour, and thermo-mechanical properties of the HBS will be discussed. The main conclusion of the experimental studies is that the HBS does not compromise the safety of nuclear fuel during normal operations.

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

FRACTAL MODEL OF FISSION PRODUCT RELEASE IN NUCLEAR FUEL

2012 ◽  
Vol 23 (09) ◽  
pp. 1250057
Author(s):  
GEDIMINAS STANKUNAS
Keyword(s):  
Diffusion Process ◽  
Nuclear Fuel ◽  
Functional Dependence ◽  
Experimental Studies ◽  
Fractional Diffusion ◽  
Fractal Model ◽  
Fuel Pellet ◽  
Operation Conditions ◽  
Fuel Burn ◽  
Fission Gas

A model of fission gas migration in nuclear fuel pellet is proposed. Diffusion process of fission gas in granular structure of nuclear fuel with presence of inter-granular bubbles in the fuel matrix is simulated by fractional diffusion model. The Grunwald–Letnikov derivative parameter characterizes the influence of porous fuel matrix on the diffusion process of fission gas. A finite-difference method for solving fractional diffusion equations is considered. Numerical solution of diffusion equation shows correlation of fission gas release and Grunwald–Letnikov derivative parameter. Calculated profile of fission gas concentration distribution is similar to that obtained in the experimental studies. Diffusion of fission gas is modeled for real RBMK-1500 fuel operation conditions. A functional dependence of Grunwald–Letnikov derivative parameter with fuel burn-up is established.

Fission Product Release Under Supercritical Water-Cooled Reactor Conditions

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
D. Guzonas ◽  
L. Qiu ◽  
S. Livingstone ◽  
S. Rousseau
Keyword(s):  
Supercritical Water ◽  
Fission Products ◽  
Fuel Pellet ◽  
Product Release ◽  
The Core ◽  
Fuel Pellets ◽  
Simulated Fuel ◽  
Core Components ◽  
High Concentrations ◽  
Supercritical Water Cooled Reactor

Most supercritical water-cooled reactor (SCWR) concepts being considered as part of the Generation IV initiative are direct cycle. In the event of a fuel defect, the coolant will contact the fuel pellet, potentially releasing fission products and actinides into the coolant and transporting them to the turbines. At the high pressure (25 MPa) in an SCWR, the coolant does not undergo a phase change as it passes through the critical temperature in the core, and nongaseous species may be transported out of the core and deposited on out-of-core components, leading to increased worker dose. It is therefore important to identify species with a high risk of release and develop models of their transport and deposition behavior. This paper presents the results of preliminary leaching tests in SCW of U-Th simulated fuel pellets prepared from natural U and Th containing representative concentrations of the (inactive) oxides of fission products corresponding to a fuel burnup of 60  GWd/ton. The results show that Sr and Ba are released at relatively high concentrations at 400°C and 500°C.

Axial Fission Gas Transport in Nuclear Fuel Rods

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”.

Crack nucleation in rod-type nuclear fuel pellet

2018 ◽  
Vol 24 (3) ◽  
pp. 668-685
Author(s):  
Vagif Mirsalimov
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactor ◽  
Crack Nucleation ◽  
Singular Integral Equations ◽  
Prefracture Zone ◽  
Fuel Pellet ◽  
Deformation Pattern ◽  
Fuel Pellets ◽  
Nuclear Reactor Fuel ◽  
Material Interaction

A plane problem of fracture mechanics on crack nucleation in a rod-type nuclear fuel pellet is considered. Nuclear reactor fuel pellets in operation may be damaged in various ways; in particular, crack nucleation. We consider a problem for the case of a heat-releasing fuel pellet with cladding: as the heat release intensity increases, zones of heightened stress are formed in the nuclear fuel pellet. The heightened stress will promote the appearance of prefracture bands that are simulated as zones of weakened interparticle bonds of the material. Interaction of prefracture zone faces is simulated by placing bonds between faces that have a specified deformation pattern. The problem of equilibrium of a fuel pellet with prefracture zones is reduced to the solution of a system of singular integral equations. An analysis of the ultimate state of the zone of weakened interparticle bonds of the material is realized on the basis of the criterion of critical opening of prefracture zone faces.

Effect of Dual Surface Cooling on the Temperature Distribution of a Nuclear Fuel Pellet

2018 ◽  
Vol 769 ◽  
pp. 296-310
Author(s):  
Odii Christopher Joseph ◽  
Agyekum Ephraim Bonah ◽  
Bright Kwame Afornu
Keyword(s):  
Heat Transfer ◽  
Temperature Distribution ◽  
Nuclear Fuel ◽  
Solid Fuel ◽  
Nucleate Boiling ◽  
Temperature Management ◽  
Fuel Pellet ◽  
Surface Cooling ◽  
Fuel Pellets ◽  
External Cooling

Heat removal from nuclear reactor core has been one of the major Engineering considerations in the construction of nuclear power plant. At the center of this consideration is the nuclear fuel pellet whose burning efficiency determines the rate of heat transfer to the coolant. This research, focuses on the study of temperature distribution of solid fuel, temperature distribution of annular fuel with external cooling and the temperature distribution of annular fuel with internal and external cooling. We analyzed the different distribution and made a conclusion on the possibility of improving temperature management of Nuclear fuel rod, by designing fuel pellets based on this geometrical and thermal Analysis. To date, a lot of studies has been done on the thermal and geometrical properties of Nuclear fuel pellet, it is observed that annular fuel pellet with simulteneous internal and external cooling can achieve better temperature distribution which leads to high linear heat generation rate, thus generating more power in the design [1]. It has also been observed that annular fuel pellets has low fission gas release [10]. In large LOCA, the peak cladding temperature of annular fuel is about 600 which is significantly less than that of solid fuel (920 ), this is due to the fact that annular fuel cladding has lower initial temperature and the thinner annular fuel can be cooled more efficiently than the solid fuel. One of drawbacks of annular fuel technology is “the fuel gap conductance assymmetry” which is caused by outward thermal expansion, it has a potential effect on the MDNBR (Minimum Departure from Nucleate Boiling Ratio), which is the minimum ratio of the critical to actual heat flux found in the core [10]. In this model, we used the ceramic fuel pellet of UO2 as our case study. All the parameters in this model are assumed parameters of UO2. The Heat Transfer tool (ANSYS APDL) was used to validate the Analytical Model of this research.

Determination of actinide and fission-product isotopes in very-high-burnup spent nuclear fuel

2008 ◽  
Vol 277 (1) ◽  
pp. 59-64 ◽  
Author(s):  
V. S. Sullivan ◽  
D. L. Bowers ◽  
M. A. Clark ◽  
D. G. Graczyk ◽  
Y. Tsai ◽  
...  
Keyword(s):  
Fission Product ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
High Burnup ◽  
Very High

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.

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