The mechanistic prediction of transient fission-gas release from LWR fuel

The steady-state and transient gas release and swelling subroutine (GRASS-SST) is a mechanistic computer code for the prediction of fission-gas behavior in UO₂-base fuels. GRASS-SST treats fission-gas release and fuel swelling on an equal basis and simultaneously treats all major mechanisms that influence fission-gas behavior. The GRASS-SST transient analysis has evolved through comparisons of code predictions with the fission-gas release and physical phenomena that occur during reactor operation and transient direct-electrical heating (DEH) testing of irradiated light-water reactor fuel. The GRASS-SST steady-state analysis has undergone verification for end-of-life fission-gas release and intragranular bubble-size distributions. The results of GRASS-SST predictions for transient fission-gas release during DEH tests are in good agreement with experimental data. Comparisons of GRASS-SST predictions of gas release and bubble-size distributions with the results of DEH transient tests indicate that (1) coalescing bubbles do not have sufficient time to grow to equilibrium size during most transient conditions, (2) mobilities of fission-gas bubbles in UO₂ are enhanced during nonequilibrium conditions if the excess pressure in the bubble is sufficient to generate an equivalent stress greater or equal to the yield stress of the surrounding matrix, and (3) channel formation on grain surfaces and coalescence of the channels with each other and with the tunnels of gas along the grain edges can contribute to grain-boundary separation and/or the rapid, long-range interconnection of porosity. The phenomena of grain-boundary separation and/or long-range interconnection of porosity provides an important release mechanism for fission gas that has moved out of the grains of irradiated fuel.     

FRAPCON-2程序改进和校验

通过改进FRAPCON-2程序中的燃料导热系数模型和裂变气体释放模型,使之能对高燃耗的燃料进行性能分析计算。并利用Halden堆IFA 597.3 ROD8的试验数据对程序进行了验证。结果表明,改进后的程序所计算出的参数（如燃料温度和裂变气体释放份额）均与实测值符合很好,对程序的改进是成功的。     

U_3Si_2Al弥散型燃料的辐照肿胀研究

介绍了U3Si2 Al弥散型燃料的辐照肿胀机理。将弥散型燃料的芯体视为连续基体中的微型燃料元件 ,应用裂变气体的行为机理描述燃料相中的气泡形成过程。研究结果表明 :燃料相的肿胀引起燃料颗粒和金属基体之间的力学相互作用 ,金属基体能抑制燃料颗粒的辐照肿胀。在一定辐照条件下 ,本模型对燃料元件辐照肿胀的预测值与测量值相符     

Verification of FEMAXI-7 code by using irradiation test in Halden reactor for He-pressurization effect on FGR of BWR fuels under power transient

Power ramp test for He-pressurization effect on fission gas release (FGR) of about 42GWd/tUO₂ boiling water reactor (BWR) fuel rods was analyzed by the fuel performance code FEMAXI-7. The experimental data were obtained with the two rods, which were base irradiated in the Halden reactor for 12 years (IFA-409), then subjected to the power ramp tests (IFA-535) to investigate the He-pressurization effect. The FEMAXI-7 calculations were performed by inputting rod specifications and experimental conditions in both the baseand test irradiations. The results showed that the calculations reasonably followed the trends of measured cladding elongation and FGR during the power ramp test, depending on the pellet temperature and fission gas atoms diffusion rate. Based on the calculated results, the reason that no apparent He-pressurization effect was observed in the experiment was considered to be caused by insufficient gas communication during strong pellet–clad mechanical interaction (PCMI) and enhanced gap thermal conductance by the solid–solid contact due to gap closure.     

PECITIS-II, a computer program to predict the performance of collapsible clad UO2 fuel elements

The codes devised and used in India for the design of fuel for their Pressurized Heavy Water Reactor (PHWR) programme are described. The scheme includes the use of collapsible fuel cladding for improved neutron economy.This code is made with reference to collapsible clad UO₂ fuel elements. This evaluates sheath strain and fission gas pressure. The fuel expansion is calculated by a two zone model which assumes that above a certain temperature the UO₂ deforms plastically and below that temperature it cracks radially and behaves as an elastic solid; the plastic core is under compression. The pellet clad gap conductance is calculated by using a modified Ross and Stoute model considering the effects of fuel and clad thermal expansion, fission gas release, dilution of filler gas and irradiation swelling. Stress relaxation of the sheath and its effect on fuel sheath contact pressure is also considered for arriving at the end result.     

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

An engineering code to predict the irradiation behavior of U–Zr and U–Pu–Zr metallic alloy fuel pins and UO₂–PuO₂ mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel–clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios.FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal–fuel version is called FEAST-METAL, and is described in this paper. The oxide–fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel–clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors.FEAST-METAL was benchmarked against the open-literature EBR-II database for steady state and furnace tests (transients). The results show that the code is able to predict important phenomena such as clad strain, fission gas release, clad wastage, clad failure time, axial fuel slug deformation and fuel constituent redistribution, satisfactorily.     

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     

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.     

Crack formation,crack healing and porosity redistribution during irradiation of UO2 and (U,Pu) O2

The irradiation-induced void volume redistribution in the fuel was analysed. The radial crack volume and porosity distributions, the central radii and the radial gap width were measured after irradiation and compared with the calculated values. Short-time (He-loop experiments in the FR2 reactor), medium-time (bundle irradiation in the BR2 reactor) and long-time (trefoil-irradiation in the DFR reactor) irradiated fuel pins were examined. The model of pore migration, used in the computer code SATURN-la, is based on the evaporation-condensation mechanism. Measured swelling rates were extrapolated to higher temperatures and used. The crack volume distribution was calculated on the basis of a multifractured fuel model. One can conclude from the comparison between calculated and measured void volume distributions that several mechanisms redistribute void volume. These are crack formation, crack healing, migration of sinter pores and fission gas bubbles, gas swelling, evaporation-condensation phenomena in the region of the central void, irradiation-induced sintering and increase in diameter of the cladding.     

Computer modeling of steady state fission gas behavior in carbide fuels

A model for the simulation of long-term, steady-state fission gas behavior in carbide fuels is formulated. It is assumed that fission gas release occurs entirely through gas atom diffusion to grain boundaries and cracks. Fission gas bubbles are assumed to remain stationary and to grow as the net result of gas atom precipitation into the bubbles from the matrix solid and gas atom re-solution from the bubbles into the matrix. Furthermore, assuming that local gas atom redistribution process in the immediate neighborhood of a bubble is very rapid, the bubble size is assumed to correspond to the equilibrium size that maintains exact balance between the rate of gas atom re-solution and that of gas atom precipitation.The model also treats the effect of attachment between bubbles and second-phase precipitates; the experimentally observed faster growth rate of precipitate bubbles is simulated using a reduced re-solution parameter for precipitate bubbles. With the grain matrix assumed to be spherical, the model allows the computation of the radial distribution of the intragranular bubbles and the gas atom concentration in the matrix.The flux of gas atoms arriving at the grain boundary is computed. The continual growth of grain boundary bubbles, resulting from the accumulation of gas atoms on the grain boundary, leads to grain boundary interlinkage and all gas atoms that subsequently reach the grain boundary are assumed to be released. Similarly, all gas atoms generated following the interlinkage of intragranular bubbles are also assumed to be immediately released.Application of the model indicates that fission gas swelling is largely due to intragranular bubbles. Grain boundary bubbles, although very large in size, contribute little to fission gas swelling and the contribution from gas atoms in solid solution in the matrix is even less significant.Physical parameters entering the model were assigned numerical values that closely represent the physical characteristics of the irradiation samples. Careful comparisons between the results of sensitivity studies and the experimental data readily identify the re-solution parameter to have the strongest influence on the results predicted by the code and that the grain size, and not the temperature, is the dominant factor affecting gas release.When allowance is made for the uncertainties of the experimental data, the predicted fission gas swelling also correlates well with experiment. The spread in the fuel swelling data, however, indicates that fuel cracking, and not fission gas swelling alone, very often contributes significantly to the fuel external dimensional changes. The linear fission gas swelling rate prediceted by the model exhibits almost a linear variation with temperature. This result correlates well with the linear swelling rate obtained from experimental swelling data if immersion density data alone are used, in order to eliminate the sources of uncertainties associated with fuel cracking.     

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.     

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.     

A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

A model for the release of fission gas from irradiated UO₂ fuel is presented. It incorporates the relevant physical processes: fission gas diffusion, bubble and grain boundary movement, intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO₂ fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW m⁻¹, burnups between 10 and 300 MW h(kg U)⁻¹, and power histories including constant, high-to-low and low-to-high power periods.The predictions of the model are shown to be most sensitive to fuel power (temperature), the choice of diffusion coefficient for fission gas in UO₂, and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth.     

The response of carbide fuel elements to fast thermal transients using the UNCLE-T-BUBE code

The response of fuel elements to fast thermal transients have great implications to the safety of LMFBR's. In this article, fission gas swelling and release, and clad stress and strain are computed for a carbide fuel element during several fast thermal transients as a function of steady stae power and percent burnup. The computations are made with the UNCLE-T-BUBE code which allows for equilibrium and nonequilibrium fission gas bubbles. In some of the transients, the code UNCLE-T-BUBE predicts fuel-clad gap closure, attended with a high clad hoop stress, whereas UNCLE-T does not. It is also found that allowing for nonequilibrium fission gas bubbles strongly affects fuel swelling and clad strain but has negligible effect on gas release.     

Modeling the behavior of Xe,I, Cs,Te, Ba,and Sr in solid and liquefied fuel during severe accidents

This paper describes the primary physical/chemical models recently incorporated into a mechanistic code (FASTGRASS) for the estimation of fission product release from fuel, and compares predicted results with test data. The theory of noble gas behavior is discussed in relation to its effect on the release behavior of I, Cs, Te, Ba, and Sr. The behavior of these fission products in the presence of fuel liquefaction/dissolution and grain-growth phenomena is presented, as is the chemistry of Sr, Ba, I, and Cs.Comparison of code predictions with data indicates the following trends. Fission product release behavior from solid fuel strongly depends on fuel microstructure, irradiation history, time at temperature, and internal fuel rod chemistry. Fuel liquefaction/dissolution, fracturing, and oxidation also exert a pronounced effect on release during fuel rod degradation. For low burnup fuel (e.g., TMI-2), appreciable fission product retention in previously liquefied fuel can occur due to the low concentration of fission products, and the limited growth of bubbles in the liquefied material.Many of the calculations described in this paper were made with a version of FASTGRASS developed for use on a personal computer (IBM compatibile).     

Simulation of pellet-cladding thermomechanical interaction and fission gas release

This paper summarizes the present status of a computer code that describes some of the main phenomena occurring in a nuclear fuel rod throughout its life. Temperature distribution, thermal expansion, elastic and plastic strains, creep, mechanical interaction between pellet and cladding, fission gas release, gas mixing, swelling, and densification are modeled. The modular structure of the code allows for the incorporation of models to simulate different phenomena and material properties. Collapsible rods can be also simulated.The code is bidimensional, assumes cylindrical symmetry for the rod and uses the finite element method to integrate the differential equations. The stress–strain and heat conduction problems are nonlinear due to plasticity and to the temperature dependence of the thermal conductivity. The fission gas inventory is calculated with a diffusion model, assuming spherical grains and using a one-dimensional finite element scheme. Pressure increase, swelling and densification are coupled with the stress field.Good results are obtained for the simulation of the irradiation tests of the first argentine prototypes of MOX fuels, where the bamboo effect is clearly observed, and of the FUMEX series for the fuel centerline temperature, the inside rod pressure and the fractional gas release.     

Benchmarking results of the COMETHE code

The validation of a code for fuel rod behaviour prediction requires a comparison of its results with corresponding experimental data. Benchmarking of the COMETHE code has been done in parallel with its development, but more time has been spent on that work than in the development of the models themselves. Three experiments are presented; they have been selected from amongst those used by BN for the calibration as being good examples of various features:

1. (1) The ELP2 experiment, performed in the EL3 reactor by CEA-Saclay and related to fuel restructuring. Results show that behaviour is very well modelled in COMETHE.

2. (2) The BR3/VN post-irradiation data, which show a large sensitivity of the fission gas release to the power level and reveal that coupling between the fission gas release model and the gaseous swelling model is beneficial.

3. (3) The BM01 low density fuel BN pin, irradiated in the FBR RAPSODIE: close agreement is found between the cracking pattern computed by the “pivot model” and the experimental cold state results.

A lot of the benchmarking results arise from the EPRI RP 397 Fuel Rod Modeling Code Evaluation Project.     

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.