Multiphysics coupled modeling of light water reactor fuel performance

A fuel performance code for light water reactors called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed. The CAMPUS code considers heat generation and conduction, oxygen diffusion, thermal expansion, elastic strain, densification, fission product swelling, grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, fuel thermal and irradiation creep, cladding thermal and irradiation creep and oxidation. All the equations are implemented into the COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet with cladding. Comparisons of critical fuel performance parameters for UO₂ fuel using CAMPUS are similar to those obtained from BISON, ABAQUS and FRAPCON. Additional comparisons of beryllium doped fuel (UO₂-10%volBeO) with silicon carbide, instead of Zircaloy as cladding, also indicate good agreement. The capabilities of the CAMPUS code were further demonstrated by simulating the performance of oxide (UO₂), composite (UO₂-10%volBeO), silicide (U₃Si₂) and mixed oxide ((Th_0.9,U_0.1)O₂) fuel types under normal operation conditions. Compared to UO₂, it was found that the UO₂-10%volBeO fuel experiences lower temperatures and fission gas release while producing similar cladding strain. The U₃Si₂ fuel has the earliest gap closure and induces the highest cladding hoop stress. Finally, the (Th_0.9,U_0.1)O₂ fuel is predicted to produce the lowest fission gas release and a lower fuel centerline temperature when compared with the UO₂ fuel. These tests demonstrate that CAMPUS (using the COMSOL platform) is a practical tool for modeling LWR fuel performance.     

Molten fuel radial motion and cladding melting during a PCM event in LWRs

The thermal behavior of the fuel and cladding during off-normal operating conditions, generally termed power-cooling-mismatch (PCM), are of great interest to light water reactor (LWR) safety analysis. During a power-cooling-mismatch event, fuel melting may begin at the center of the rods propagating radially outward. The induced pressure at the center of the rod due to fuel melting, fission gas release, and UO₂ fuel vapor would tend to force such molten fuel to flow through radially open cracks in the outer unmelted portion of the pellet and relocate in the fuel-cladding gap. The zircaloy cladding, which is at high temperature during film boiling, may undergo melting at its inside surface upon being contacted by the extruded molten fuel, eventually resulting in a thermal failure of the cladding.Three topics of interest are analyzed in this paper. First, fuel conditions during a hypothesized PCM accident are assessed with regard to pellet cracking and central fuel melting. Secondly, the transient freezing of a superheated liquid penetrating an initially empty crack, maintained at constant subfreezing temperatures, is studied analytically. The analysis is presented in a dimensionless form, illustrating the effect of the governing parameters, namely the driving pressure, crack shape (that is, a divergent, a parallel wall, or a convergent crack), density ratio, Stefan number for freezing, and steady state crust thickness. The calculational results are used to assess the radial extrusion of molten UO₂ fuel observed in some in-pile tests, in which PCM conditions in a pressurized water reactor were simulated. Thirdly, conditions for potential melting of zircaloy cladding upon being contacted by the extruded molten fuel are investigated analytically. The analytical predictions were consistent with the experimental results from PCM in-pile tests.     

Analysis of MOX fuel behavior in reduced-moderation water reactor by fuel performance code FEMAXI-RM

To assess the feasibility of the 31% Pu-MOX fuel rod design of reduced-moderation water reactor (RMWR) in terms of thermal and mechanical behaviors, a single rod assumed to be irradiated in the core of RMWR up to 106 GWd/tHM has been analyzed by a fuel performance code FEMAXI-RM which is an extended version of FEMAXI-6 code. In the analysis, design specifications of fuel rod and irradiation conditions have been input, and available models of both MOX fuel and UO₂ fuel have been used as appropriate. The results are: fission gas release is several tens of percent, rod internal pressure does not exceed the coolant pressure, and the highest fuel center temperature is 2400 K, while cladding diameter increase caused by pellet swelling is within 1% strain. These predictions suggest that the MOX fuel rod integrity will be held during irradiation in RMWR, though actual behavior of MOX pellet swelling and cladding oxidation require to be investigated in detail.     

三维燃料棒精细化模拟软件FUPAC3D与FUPAC的对比验证研究

为验证基于三维有限元分析平台建立的三维燃料棒精细化模拟软件FUPAC3D在分析评价压水堆燃料棒辐照-热-力耦合行为方面的能力和精度,本文给出了三维FUPAC3D软件采用的热学模型、燃料棒力学模型、裂变气体释放模型以及腐蚀模型,以华龙一号典型燃料棒参数和运行工况作为输入参数,分别使用三维FUPAC3D软件和已工程化应用的1.5维FUPAC软件进行建模分析,并针对2种软件在芯块和包壳温度、包壳应力与应变、芯块与包壳间间隙宽度的计算结果进行对比研究。研究结果表明,FUPAC3D软件与FUPAC软件具有相当的精度,FUPAC3D软件具备压水堆燃料棒辐照-热-力耦合行为的精细化模拟能力。      

Materials chemistry and transport modeling for severe accident analyses in light-water reactors II: Gap processes and heat release

Several consequences of steam starvation of the gas filling the internals of the core of a light-water reactor in the fuel-uncovery phase of a severe accident up to cladding melting are analysed. Emphasis is placed on processes that occur in the H₂-rich gas external to the fuel rod cladding; absorption of oxygen and hydrogen by the cladding; the composition and flow rates of gas in the fuel-cladding gap; and the response of the fuel to these conditions. The transport processes and chemical reactions in the cladding, and the fuel controlled by the behavior of the gas in the gap are modeled for a simple temperature transient characteristic of a severe fuel damage accident in a light-water reactor. Cladding burst is assumed to occur at 1273 K at the midplane elevation of the fuel rod, permitting the gas in the gap to come into contact with that external to the fuel rod. The results of the analysis include the following. Steam ingress is restricted to a few centimeters from the failure site by the gettering action of the metal-water reaction on the cladding inner wall. Hydrogen moves axially into the gap only a few times further than steam by diffusion in the Xe-He mixture. The chief process restricting H₂ ingress is the backflow resulting from thermal expansion of the gas in the fuel rod as the temperature rises. When the protective ZrO₂ scale on the outer surface of the cladding disappears by dissolution in the metal, hydrogen permeation through the cladding wall rapidly replaces the inert gas in the gap with H₂. Hydrogen uptake by the cladding draws gas into the core region from the upper plenum and augments the heat release by the metal-water reaction. Exposure of the fuel to this H₂-rich gas results in minor fuel reduction and accompanying cladding oxidation.     

Evaluation of LWR fuel rod behavior under operational transient conditions

To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding.The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8 × 8 RJ fuel rod temperatures under power ramp conditions.     

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.     

燃料棒包壳辐照蠕变与生长行为模拟研究

根据建立的包壳材料和行为模型,利用ABAQUS对包壳辐照蠕变与生长行为进行相关模拟研究。在设定的稳态工况下,计算了包壳的辐照蠕变,分析了它的应力以及蠕变的关系。通过设置节点集和不同路径方式,分析了包壳的辐照生长现象,表明了ABAQUS在核燃料性能研究方面的适用性。以及利用ABAQUS的二次开发接口,可以把自定义的模型结合到商用的有限元软件中,利用商用有限元软件的优势,快速解决部分核燃料模拟中的问题。     

Fission Gas Induced Cladding Deformation of LWR Fuel Rods under Reactivity Initiated Accident Conditions

Behavior of irradiated fuel rods under power burst conditions by accidental reactivity insertion in light water reactors (LWRs) has been studied in the Nuclear Safety Research Reactor (NSRR). In the experiments, cladding hoop deformation, which reached up to about 10%, was much larger than that of the fresh rods. The current LWR fuel behavior analysis codes, which only take account of the thermal expansion of the fuel pellets for the deformation calculation, under-predicted the plastic deformation of the cladding to be less than about 1%. Fission gas release during the pulse irradiation tests reached as high as 22% in the NSRR irradiated fuel tests. In order to describe these test results, a model of grain boundary fission gases to cause the cladding deformation has been developed and installed in a fuel behavior simulation code, FRAP-T6. In the model, the over-pressurized gases by the pulse irradiation cause grain boundary separation and stress the cladding during the tests. The model assumes that the gases remain in the fuel during the early part of pulse irradiation and are released to the open volume in the rod after the cladding deformation. The model, in combination with a fuel thermal expansion model, GAPCON, which was validated through fresh fuel tests, reproduces the NSRR test results reasonably well.     

A model for fission gas release and gaseous swelling of the uranium dioxide fuel coupled with the FALCON code

A comprehensive model GRSW-A was developed to analyse the processes of fission gas release, gaseous swelling and microstructural evolutions in the uranium dioxide fuel during base irradiation and under transient conditions. The GRSW-A analysis incorporates a number of models published in open literature, as well as some original models that were already published by the authors elsewhere. Consequently, only the most prominent aspects of GRSW-A and its coupling with the FALCON fuel behaviour analysis and licensing code are described in this paper. The analysis of fuel behaviour in the REGATE experiment is presented, which includes the base irradiation of the fuel segment in a PWR to a burn-up of about 50 MWd/kgU, which was followed by a power ramp in the SILOE research reactor. Besides, the generalized data on fission gas release (FGR) in PWR fuel during the base irradiation up to a burn-up of about 70 MWd/kgU is interpreted using coupled FALCON and GRSW-A. Moreover, a mechanistic interpretation of the published data for pellet swelling during the base irradiation up to a burn-up of 100 MWd/kgU is put forward. In all the cases, the coupled FALCON/GRSW-A analysis has shown the improved prediction capability compared to the original FALCON MOD01, which is achieved due to the account for the mutual effect of thermal and, in particular, high-burn-up-assisted mechanisms of fission gas release and swelling under steady-state and transient conditions.     

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.     

A model for analysis of fuel behaviour in transients

For many applications, analysis of fuel element behaviour must take non-linear thermal, and elasto-plastic effects into account. This is particularly true if the fuel undergoes large deformations and rapid temperature transients. To meet this need a multi-dimensional fuel model based on finite element stress and thermal analysis has been developed. The model is solved for the transient temperature distribution by a step-by-step time incremental procedure. The temperature is then introduced into the elasto-plastic analysis as a thermal load and stresses and deformations are calculated. A model for treatment of creep and a special element for the gap between fuel pellet and cladding is incorporated together with semi-empirical procedures for calculating fission gas release, fuel pellet to cladding heat transfer coefficients, etc.The fuel model has been compared with both analytical solutions and in-reactor experimental results. The observed and predicted results are in good agreement.     

热蠕变对UMo/Zr单片式燃料板起泡行为的影响

针对含有气腔的UMo/Zr单片式燃料板,考虑包壳材料的热蠕变效应,将包壳的变形与气腔压力相耦合,发展了一种对燃料板宏观起泡行为进行数值模拟的方法。基于所建立的模拟方法,计算分析了包壳热蠕变和气腔内裂变气体原子数对起泡行为的影响。研究发现,在考虑包壳热蠕变时,若局部开裂区域内的裂变气体原子数为4.0×1017,以鼓泡高度0.1 mm作为起泡阈值的判断标准,所预测出的阈值温度比不考虑热蠕变时低100℃;若局部开裂区内的裂变气体原子数由2.5×1017增加至4.0×1017,则燃料板的起泡阈值温度将可能降低40℃,通过降低包壳材料的热蠕变率可以有效提高燃料板的抗鼓泡能力。      

燃料棒破损热力学模块开发及在严重事故分析程序中的应用

包壳肿胀和破损是严重事故早期阶段的重要现象。包壳形变不仅会造成局部流动堵塞,同时,水蒸气会从破裂处进入包壳气隙,增加包壳被蒸汽氧化的表面积。广泛使用的一体化严重事故分析程序不能分析早期事故过程中燃料棒的热力学行为,判断包壳破裂也只是基于简单的参数模型。本文开发了分析燃料棒热力学行为的FRTMB模块,并集成在严重事故分析程序ISAA中。使用开发的耦合系统ISAA FRTMB分析了CAP1400反应堆直接注射（DVI）管线小破口事故过程中燃料棒的热力学行为,并预计了包壳破裂时间及相应的失效温度。计算结果整体验证了ISAA FRTMB分析瞬态事故过程中燃料棒热力学行为以及判断包壳破裂的适用性和可靠性。     

Thermal Stress Analysis of High-Burnup LWR Fuel Pellet Pulse-Irradiated in Reactivity-Initiated Accident Conditions

For RIA-simulated experiments in the NSRR with high-burnup PWR fuel and BWR fuel, numerical analyses were performed to evaluate the temporal changes of profiles of temperature and thermal stress in pellet induced by pulse power, using the RANNS code. The pre-pulse states of rods were calculated using the fuel performance code FEMAXI-6 along the irradiation histories in commercial reactors and the results were fed to the RANNS analysis as initial conditions of the rod. One-dimensional FEM was applied to the mechanical analysis of the fuel rod, and the calculated cladding permanent strain was compared with the measured value to confirm the validity of the PCMI calculation. The calculated changes in the profiles of temperature and stress in the pellet during an early transient phase were compared with the measured data such as the internal gas pressure rise, cracks and grain structure in the post-test pellet, anddiscussed in terms of PCMI and grain separation. The analyses indicate that the pellet cracking appearances coincided with the calculated tensile stress state and that the compressive thermal stress suppresses the fission gas bubble expansion leading to grain separation.     

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.     

热蠕变对UMo/Zr单片式燃料板起泡行为的影响

For a UMo/Zr monolithic fuel plate with a gas space, a method is developed to simulate the macroscale blister behavior considering the thermal creep effects of the cladding, in which the calculation of cladding deformation is coupled with the gas space pressure. Based on the developed simulation method, the effects of thermal creep strain of cladding and the internal fission gas atom number on the blister behavior are analyzed. The research results indicate that with the thermal creep of cladding considered, if the fission gas atom number is 4.0×1017, the predicted blister threshold temperature will be 100℃ lower than the case without considering the thermal creep of cladding, with the blister threshold temperature set as the temperature at which the blister height reaches 0.1 mm, with the fission gas atom number increasing from 2.5×1017 to 4.0×1017, the blister threshold temperature might decrease by 40℃. The blister threshold temperature of the fuel plates could be improved by using a cladding material with low thermal creep rate.