Behaviour of fission gas products in single crystal UO2 during intermediate temperature irradiation

A new mathematical interpretation is presented of fission gas release from monocrystalline uranium dioxide fuel during intermediate temperature irradiation in terms of a defect trap model, knock-out process and diffusion of bubbles. In the present model it is assumed that gas in an intermediate state exists side by side with the dissolved fission gas and that trapped in bubbles. It is assumed also that the isolated gas atoms, being re-dissolved, are immobile.The present model gives a satisfactory interpretation of the relative proportions of isotopes in the steady state fission gas release at diffrent temperatures. The dependence of fractional fission gas release on fission rate is also interpreted; regimes either proportional to fission rate or inversely proportional to fission rate are predicted depending on the fission rate interval considered. Both temperature dependent and temperature independent fission gas release can arise.The presented dynamic method of studying the release of fission gases during irradiadion provides a further test beside the static method of the veracity of the assumed mechanisms. Calculations show that fission gas behaviour becomes more complex for oscillated fission rate in the regime where the fractional release is inversely proportional to the fission rate for the steady state.     

Investigations on swelling and fission gas behaviour in uranium dioxide

UO₂ irradiated at temperatures between 1000 and 2100 K was investigated with respect to fission gas behaviour and swelling. The amount of fission gas was measured in three steps as released fission gas, fission gas retained in bubbles and pores, and fission gas in the fuel matrix. The retained fission gas reaches concentrations up to $\text{1.6 × 10}{}^{\mathrm{?2}}$ gas atoms per uranium atom at temperatures below 1250 K and decreases with increasing temperature. The swelling was evaluated by measuring the volume changes and by immersion density measurements. The maximum fission gas swelling without extensive bubble migration is about 20% at 2000 K. It diminishes to about 5% at 1250 K.     

压水堆燃料包壳破损条件下裂变气体非稳态释放数值模拟研究

压水堆燃料包壳破损后,芯块-包壳间隙内积累的裂变气将释放到冷却剂中,其内部的微观机理还尚不清楚。为了揭示裂变气体释放过程中冷却剂与气体的相互作用规律,基于三维计算流体力学（CFD）方法对该物理过程展开数值模拟,所利用的模型为VOF模型以及k-ε模型。模拟结果表明,包壳破损后冷却剂首先进入芯块-包壳间隙,在芯块-包壳间隙内蒸发,引起芯块-包壳间隙内压强上升,而后裂变气体释放到子通道;裂变气体从芯块-包壳间隙释放到子通道可分为2个阶段。第一阶段:芯块-包壳间隙与子通道间压差较大,气体射流进入子通道,该阶段持续时间较短,裂变气体释放率较大,且变化也较大。第二阶段:芯块-包壳间隙与子通道间压差较小且相对平稳,裂变气体通过破口内涡的对流传质进入子通道,该阶段持续时间较短,裂变气体释放率较小,且相对稳定。      

弥散型燃料的裂变气体行为研究

探讨了弥散型燃料中对辐照肿胀有重要影响的裂变气体的行为机理。裂变气体原子聚集成气泡引起燃料相肿胀,气泡的尺寸分布是影响辐照肿胀的重要因素。决定气泡生长的裂变气体的行为机理主要有：裂变气体原子的产生和热扩散迁移,气泡的成核和聚合长大,气泡内气体原子的重溶,燃料相的辐照亚晶化等过程。燃料中各种尺寸的气泡浓度随时间的变化率可用气泡生长的动力学速率方程组来描述。当裂变密度较高时,辐照产生的缺陷引起燃料相的     

基于弥散燃料颗粒开裂的裂变气体释放模型

根据弥散燃料颗粒开裂后裂变气体的3种释放途径,分别建立了裂纹连通释放模型、气泡连通释放模型以及原子扩散释放模型,综合得到了基于弥散燃料颗粒开裂的裂变气体释放模型,并采用该模型对裂变气体释放量进行了计算。结果表明:裂变气体释放量主要由裂纹连通释放途径贡献;燃耗深度越高,裂变气体释放量的增加速率会越大;随着退火温度的增加,裂变气体释放量迅速增加,而退火时间越长,裂变气体释放量的增加速率越低。通过裂变气体释放量模型计算得到的裂纹宽度与实验观察到的裂纹宽度符合较好,对比结果验证了基于弥散燃料颗粒开裂的裂变气体释放模型的合理性。      

Radiation re-solution of fission gas in uranium dioxide and carbide

Calculations have been performed to estimate the removal rate of fission gas atoms from bubbles due to collisions with energetic fission fragments and recoil cascades. The efficiency of this process was found to be higher than estimated earlier, but is still too low to be responsible for the experimental observations of fission gas bubble destruction during irradiation of oxide fuel. An irradiation experiment to investigate the interaction of fission spikes with free surfaces has enabled a simple theory to be developed which can explain the shrinkage of bubbles and pores by the surface relaxation of a shock wave produced by the passage of a fission fragment. This mechanism occurs in oxides but not carbides because of the faster dispersion of the fission fragment energy and provides the major reason for the difference in gas bubble distributions in oxide and carbide fuel. This process, however, does not remove gas atoms from the bubbles. Since high levels of apparently diffusive fission gas release are observed in oxides, the “effective solubility” of the fission gases required for this release must be sought in phenomena other than the fission spike.     

Molecular dynamics simulation of intragranular Xe bubble re-solution in UO2

The homogeneous re-solution of Xe fission gas bubbles in UO₂ is investigated by combined Monte Carlo and molecular dynamics simulations. Using a binary collision model, based on the Ziegler-Littmark-Biersack potential [J.F. Ziegler, J.P. Biersack, U. Littmark, The Stopping and Range of Ions in Solids, Stopping and Ranges of Ions in Matter, vol. 1, Pergamon Press, New York, 1984], the recoil energy distribution of fission gas atoms is obtained. An extensive library of fission gas atom displacement cascades is then compiled using molecular dynamic simulations. It used for calculating recoil spectrum averaged quantities. The calculations yield a re-solution parameter for homogeneous re-solution and a displacement distribution of fission gas atoms around the fission gas bubbles. The results disagree considerably from past estimates. The importance of channeling and threshold energy for fission gas escape are discussed.     

Molecular dynamics simulation of the pressure-volume-temperature data of xenon for a nuclear fuel

The exact equation of state for the fission gas is necessary for the accurate prediction of the fission gas behavior in a nuclear fuel. However, certain kinds of extrapolating data are used to construct and verify the equations of state for the fission gas because experimental data are very limited at high temperatures and pressures that are encountered in the nuclear fuel. To fill the lack of experimental data for the fission gas, the behavior of Xe gas atoms was investigated by molecular dynamics simulation assuming an exponential-six potential. The molecular dynamics simulation produced reasonable pressure-volume-temperature data for Xe and the simulation results were compared with existing equations of state for Xe.     

Fission gas release from UO2 fuel during low-temperature irradiation

A new mathematical interpretation is presented of fission gas release from UO₂ fuel during low-temperature irradiation in terms of a defect trap model and the knock-out process. In the present model it is assumed that gas in an intermediate state exists side by side with the dissolved fission gas and that trapped in bubbles. The present model gives a satisfactory interpretation of the relative proportion of isotopes in the steady state fission-gas release. The dependence of the fission-gas release rate on the fission rate is also interpreted; regimes either proportional to the square of fission rate or proportional to fission rate are predicted, depending on the fission rate interval considered.     

A comparison between fission gas release data and FEMAXI-IV code calculations

The FEMAXI-IV code is an extension of the earlier version FEMAXI-III. The primary improvement in the new version is the provision for treating the fuel rod behavior during an operational transient. For this purpose, the time-dependent models are used for heat conduction, fission gas release, and mixing of the released gas with the plenum gas.In FEMAXI-IV, the fission gas release model was thoroughly revised from the previous version. It is based on the fission gas release model presented by White and Tucker. The model takes into account the following mechanisms:

• - diffusion of gas atoms to the grain boundary;

• - sweeping of gas atoms by grain growth;

• - precipitation of gas atoms into intragranular gas bubbles;

• - resolution of gas atoms from intragranular and grain boundary gas bubbles;

• - fission gas release due to bubble interconnection.

The model was incorporated into FEMAXI-IV and code calculations were compared with the fission gas release data obtained in the Inter-Ramp and Over-Ramp experiments.This paper describes the fission gas release model involved and results of calculations.     

The fission gas bubble distribution in a mixed oxide fast reactor fuel pin

The fission gas bubble distribution has been studied in a mixed oxide fast reactor fuel pin irradiated in DIDO MTR to 2.8% burn-up at centre and surface temperatures of 2000 and 1000°C. The intragranular fission gas bubbles are very small (<6 nm diameter) and this is a consequence of the high re-solution rate at fast reactor ratings. The bubbles nucleate heterogeneously and linear arrays of bubbles, due to nucleation on fission tracks, are observed up to irradiation temperatures of 1900°C. At 1980°C ~4% of the fission gas produced is present in intragranular bubbles. There is no definite evidence for gas bubble mobility or coalescence. Apart from any effects of columnar grain growth fission gas release in fast reactor fuel pins seems to occur predominantly by the diffusion of single gas atoms, at least up to irradiation temperatures of 2000°C.     

Fuel swelling due to retained fission gas and thermal effects during high temperature transients

Fuel swelling of previously irradiated pressurized-water-reactor-type fuel rods tested under power-cooling-mismatch conditions is due to retained fission gas and thermal effects within the film boiling region. In this paper empirical correlations for fuel swelling are presented, and mechanisms contributing to the observed swelling and the applicability of an analytical fission gas behavior computer code (GRASS-SST) to fuel swelling are evaluated. Major contributors to fuel swelling are fuel melting and expansion, expansion of solid fuel, fission gas bubble coalescence, fission gas diffusion to grain boundaries, and change in surface tension of fuel upon melting. The contributions to fuel swelling of solid fission products and the effects of cladding contraction and wall thinning on rod swelling are also included. The overall empirically-calculated fuel swelling values and the GRASS-SST code calculated values are compared with measured values. The agreement between measured and empirically calculated fuel swelling is generally close. Fuel swelling due to retained fission gas during the film boiling transient, as calculated by the GRASS-SST code, was found to be in good agreement with experimental results.     

Reduction of fission gas release at extended burnups in light water reactors by decreasing fuel temperature

The effects of fuel temperature on fission gas release in light water reactor UO₂ fuel at extended burnups of up to 56 effective full power months (EFPMs) are evaluated using a simple fission gas release mechanistic model. The model is first described and then model validation comparisons are made against experimental fission gas release date. The study shows that by decreasing the maximum operating fuel temperature to below 1200°C, it is possible to reduce the amount of released fission gas at 56 EFPMs to less than that at the current design burnup of 36 EFPMs.     

基于GEM工艺的裂变时间投影室性能模拟研究

基于GEM工艺的裂变时间投影室具有探测效率高和空间分辨率高的特点,可实现裂变产物的多参量测量。本文主要研究基于GEM工艺的裂变时间投影室在不同条件下的测量精度,使用Garfield++软件计算得到裂变时间投影室中不同的裂变产物质量数测量误差约为4～6 u,并通过时间信息的径迹重建研究了裂变碎片在不同工作气体中的角度分辨。研究发现,电子漂移时间长的工作气体中,裂变产物具有更好的角度分辨,并可依此在实验中选择合适的工作气体、气压和漂移电场强度来进行裂变碎片的测量。     

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other hand, in order to accommodate solid fission product swelling and to control fuel clad mechanical interaction of the stiffer fuel, the fuel smear density is reduced to 70%. In addition, plenum height is increased to accommodate for fission gases.     

燃料棒芯块中Kr、Xe的低温释放机理

在Wise反冲机理微观模型的基础上,考虑芯块倒角的影响,给出了倒角表面平均反冲效率的计算模型;击出机理参考Olander的理论;由此建立了一个更为精细的裂变气体低温释放微观模型.使用新建模型进行计算的结果显示,倒角表面的平均反冲效率约为圆柱体芯块表面的两倍;当倒角的表面积与圆柱体芯块表面积相比不可忽略时,裂变气体反冲释放份额的计算应考虑倒角的影响.     

Modeling of fission gas swelling in LWR UO2 fuel under irradiation

An understanding of the behavior of fission gas in uranium dioxide (UO₂) fuel is necessary for the prediction of the performance of fuel rods under irradiation. A mechanistic model for matrix swelling by the fission gas in LWR UO₂ fuel is presented. The model takes into account intragranular and intergranular fission gas bubbles behavior as a function of irradiation time, temperature, fission rate and burn-up. The intragranular bubbles are assumed to be nucleated along the track of fission fragments, which play the dual role of creator and destroyer of intragranular bubbles. The intergranular bubble nuclei is produced until such time that a gas atom is more likely to be captured by an existing nucleus than to meet another gas atom and form a new nucleus. The capability of this model was validated by a comparison with the measured data of fission gas behavior such as intragranular bubble size, bubble density and total fuel swelling. It was found that the calculated intragranular bubble size and density are in reasonable agreement with the measured results in a broad range of average fuel burn-ups 6–83 GW d/tU. Especially, the model correctly predicts the fuel swelling up to a burn-up of about 70 GW d/tU.     

Numerical method for the containment of fission products at a hypothetical accident in an underground nuclear power plant

As one means to expand the siting of nuclear power plants, construction of underground plants is now under study. An underground nuclear power plant has the feature that ground surrounding the underground cavity can contain the fission products of a hypothetical accident.If it is assumed that in a hypothetical reactor accident the cooling system loses its capacity wholly or partially, and gas containing fission products is emitted into the underground cavity. As a result, temperature, gas concentration and gas pressure in the cavity increase and it can be supposed that the gas leaks up to the surface through the ground, and that ground-water contains and carries fission products. The present paper numerically simulates a course of movement as mentioned above by the finite element method and gives the underground containment effect for fission products from a hypothetical accident.     

Investigation of swelling of U-Pu mixed carbide

Mixed carbide fuel samples irradiated in various types of capsules were investigated with respect to fuel swelling and fission gas behaviour. The irradiations were carried out in the FR 2 reactor in Karlsruhe at temperatures between 300 and 1750°C up to 5.5% burnup. The swelling was evaluated by immersion density measurements in carbon tetrachloride. The fission gas determinations were carried out by measuring the released gas and by measuring the retained fission gas.The swelling rate of mixed carbide is a strong function of temperature. At temperatures below 1000°C it is between 1 and 1.5% per % burnup. At temperatures above 1000°C the swelling rate increases with temperature. It is about 3% per % burnup at 1300°C and about 12% per % burnup at 1750°C. The swelling rate at high temperatures decreases with increasing burnup due to a saturation of the fission gas bubble porosity.     

MACROS benchmark calculations and analysis of fission gas release in MOX with high content of plutonium

This paper focuses on the modelling of fission gas release in mixed oxide fuel. In a first part, the irradiation experiment, conducted by the Japanese Nuclear Energy Safety organization on high enriched mixed oxide fuel, is outlined. In a second part, the approach for fission gas release modelling, as implemented in the fuel performance code MACROS, is explained and a comparison between calculated and experimental results is made. The code MACROS is conceived to provide not only integral (rod average) results on fission product and fission gas retention and release, but also to calculate local concentrations (radial profiles). In this way, it is possible to compare results from post-irradiation examinations with calculated profiles.