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.     

Pellet compliance model based on out-of-pile simulation

The thermal and mechanical behavior of fuel rods is significantly influenced by the extent of their relocation and by compliance of the cracked pellets. Movement of the cracked pellet pieces towards the cladding results in softer pellets with crack voids which accommodate some fraction of the thermoelastic pellet deformation and make the pellet more compliant under the restraint of the cladding. It is difficult to model such a pellet compliance independently of experimental observations because the cracked pellet behavior is uncertain by nature.Electrically heated simulation of pellet-cladding mechanical interaction (PCMI) facilitates much quicker and more flexible experimentation than actual in-pile tests. Testing apparatus consists of the simulated fuel rod with hollow UO₂ pellets and a tungsten rod in the center, and a diameter measuring device including three pairs of diameter sensors. Test parameters include the pellet-cladding gap and the cladding thickness. Results show that rods with a smaller gap have a larger increasing rate of cladding diameter. This suggests that a group of cracked pellet pieces induced by thermal stress has an apparent compliance which increases with pellet-cladding gap. Results also show more sensitivity to cladding thickness than those calculated assuming pellets having intrinsic stiffness. This also suggests the compliant nature of cracked pellets.Such a compliant nature can almost be described by reducing the elasticity of the pellet. A simple pellet compliance model was obtained by fitting calculations with measurements to describe a cracked pellet as a uniform axisymmetric body with apparent elasticity.     

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。      

An empirical model of the effects of pellet cracking on the thermal conductivity of UO2 light water reactor fuel

Light water reactor fuel pellet cracking and pellet fragment relocation into the pellet-to-cladding gap during normal operation alters both the fuel thermal conductivity and the thermal resistance of the gap. Uranium dioxide fuel pellet thermal conductivity data from a series of tests being conducted in the Power Burst Facility to evaluate the thermal performance of LWR design fuel are presented. These data indicate that the effective thermal conductivity of a cracked and relocated light water reactor fuel pellet is strongly influenced by the closing or opening of the cracks as the rod power is increased or decreased and is dependent on the initial pellet and cladding dimensions. An empirical correlation is introduced which provides a means for calculating the effective thermal conductivity of cracked and relocated fuel within helium bonded fuel rods. The method also provides a means for estimating the relocated hot pellet-to-cladding gap width as a function of rod power.     

FEMAXI-III,a computer code for fuel rod performance analysis

This paper presents a method of fuel rod thermal-mechanical performance analysis used in the FEMAXI-III code. The code incorporates the models describing thermal-mechanical processes such as pellet-cladding thermal expansion, pellet irradiation swelling, densification, relocation and fission gas release as they affect pellet-cladding gap thermal conductance. The code performs the thermal behavior analysis of a full-length fuel rod within the framework of one-dimensional multi-zone modeling. The mechanical effects including ridge deformation is rigorously analyzed by applying the axisymmetric finite element method. The finite element geometrical model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The 8-node quadratic isoparametric ring elements are adopted for obtaining accurate finite element solutions. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behaviors accurately and stably. The pellet-cladding interaction mechanism is exactly treated using the nodal continuity conditions. The code is applicable to the thermal-mechanical analysis of water reactor fuel rods experiencing variable power histories.     

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.     

Behavior of Irradiated ATR/MOX Fuel under Reactivity Initiated Accident Conditions

Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20MWd/kgHM were conducted at the NSRR in Japan Atomic Energy Research Institute to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Relatively large radial deformation of the fuel rods due to pellet-cladding mechanical interaction occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet.     

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.     

MIPAC, a computer code for fuel performance analysis by the finite element method

An axisymmetric finite element computer code named MIPAC has been developed for analysis of the mechanical interaction behaviour between a fuel pellet and cladding. This computer code can deal with elastoplasticity of the pellet and cladding materials, creep effects for the both materials, pellet-cladding and pellet-pellet contact problems, hot pressing effect of the fuel pellet, fuel pellet cracking, and the cracked pellet's stiffness. A cyclical boundary condition is introduced to deal with one pellet length instead of the full-size fuel rod. The contact problems are solved without a fictitious contact element. In the fuel pellet cracking model the crack opening and closing behaviour under arbitrary power changes can be treated by introducing five kinds of crack modes. Mismatch of irregular crack surfaces is taken into account in the evaluation of the cracked pellet's stiffness. Finally, calculated results are compared with experimental data to show validity of the computer code.     

Analysis of mechanical load on cladding induced by fuel swelling during power ramp in high burn-up rod by fuel performance code FEMAXI-6

Mechanical load on cladding induced by fuel swelling in a high burn-up BWR type rod was analyzed by a fuel performance code FEMAXI-6. The code was developed for the analysis of LWR fuel rod behaviors in normal operation and transient conditions using finite element method (FEM).During a power ramp for the high burn-up rod, instantaneous pellet swelling can significantly exceed the level that is predicted by a “steady-rate” swelling model, causing a large circumferential strain in cladding. This phenomenon was simulated by a new swelling model to take into account the fission gas bubble growth. As a result it was found that the new model can give reasonable predictions on cladding diameter expansion in comparison with PIE data. The bubble growth model assumes that the equilibrium state equation holds for a bubble under external pressure, and simultaneous solution is obtained with both bubble size determination equation and diffusion equation of fission gas atoms. In addition, a pellet-clad bonding model which has been incorporated in the code to assume solid mechanical coupling between pellet outer surface and cladding inner surface predicted the generation of bi-axial stress state in the cladding during ramp.     

压水堆燃料棒三维热-力学性能的精细模拟

压水堆燃料棒工作在复杂的辐照、热和力学环境中,对其性能进行定量评估涉及多种复杂的物理现象。目前常用的燃料性能分析程序一般对结构采用简化的轴对称假设,对辐照肿胀、辐照蠕变和高温蠕变等物理现象以及辐照-热-力等物理场之间的耦合考虑并不充分。基于ABAQUS有限元求解框架,开发了压水堆燃料棒三维热-力学性能的模拟程序,利用程序对压水堆燃料棒进行了稳态分析,以及升功率和反应性引入事故两种瞬态分析。结果表明:辐照引起燃料致密化和肿胀对燃料温度变化有重要影响;芯块应变增加主要是由裂变产物肿胀引起的;芯块几何结构导致包壳应力集中发生在芯块间的交界面处;燃料棒功率的急剧变化会加快芯块表面破裂的进程;反应性引入事故会导致芯块从内部开始破裂,并会引发芯块-包壳的接触。     

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.     

Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6 Code

Capabilities of the FEMAXI-6 code to analyze the behavior of high burnup MOX fuels in LWRs have been evaluated. Coolant conditions, detailed power histories and specifications of the MIMAS-MOX fuel rods, rod 10 and rod 11, of IFA-597.4–7 irradiated in the Halden reactor were input, and calculated rod internal pressures and pellet center temperatures were compared with the measured data for the range of 0-31 MWd/kgUO₂. Some sensitivity studies were conducted mainly with respect to pellet thermal conductivity and swelling rate to investigate the changes in thermal behavior and their effects on fission gas release.

In the irradiation period up to about 23 MWd/kgUO₂, the calculated pellet center temperatures sufficiently agreed with the measured data and also the calculated rod internal pressures reproduced the tendency of an increase in the measured rod internal pressures. These results suggest that fission gas release from MOX fuels can be reasonably predicted by a diffusion process that is modeled in UO₂ pellet grains. On the other hand, the steep increase in the measured rod internal pressures observed at the power ramp around 23 MWd/kgUO₂ cannot be reproduced by FEMAXI-6 and can be regarded as the result of a relatively large amount of gas release, which possibly caused a pellet-cladding-gap closure through pellet gas-bubble swelling.     

Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO₂ temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete and smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.     

UO2燃料裂变气体渗流模型研究

为分析UO₂燃料晶界气泡连通导致裂变气体间歇性释放的动力学过程,从而解决目前扩散模型预测的沿芯块径向释放份额与实验测量不符的问题,采用二维渗流模型模拟UO₂燃料晶界气泡网络的演化及与燃料棒内自由空间连通的释放过程。研究结果表明,渗流模型预测沿芯块径向的裂变气体释放份额在芯块中间部分出现局部峰值,并随着时间向芯块外侧推进,与辐照试验观察到不同燃耗下径向裂变气体分布现象定性符合。因此,本研究建立的渗流模型能够从机理上解释此前扩散模型未能预测的UO₂燃料裂变气体释放份额沿径向非单调分布现象。      

Simple Model for Analysis of Circumferential Ridging Deformation of Fuel Rods

A simple model for analysis of fuel rod ridging is proposed. In this model, a piece of radially cracked pellet is considered as a beam, and cladding as a tube shell, allowing fuel rod ridging to be analyzed by applying the beam and shell theories. Ridging height and contact force between the pellet piece and the cladding tube are expressed in a relatively simple form as a function of elastic constants of the pellet and cladding, temperatures, dimensional parameters, etc. Effects of fuel design parameters on the fuel rod ridging are evaluated using this model.     

A methodology to predict a fission gas release ratio of MOX fuel with heterogeneous microstructure

The objective of this study is to formulate a methodology to predict a fission gas release ratio of MIMAS MOX. An irradiated MIMAS MOX fuel with plutonium rich agglomerates was subjected to elemental analyses by electron probe micro analysis and secondary ion mass spectrometry in order to investigate xenon distribution. The results of the elemental analyses showed that the plutonium rich agglomerates at the periphery of the fuel pellet sample retained a high concentration of xenon as gas bubbles. Then, the results were used as reference data for modification of models in a fuel rod analysis code, FEMAXI-7. Using the modified FEMAXI-7, we applied an approach to prediction of fission gas release ratio of MOX fuel with plutonium rich agglomerates. In the approach, two separated analyses using FEMAXI-7 were performed for the plutonium rich agglomerates and the matrix. Fission gas release ratios obtained from the two analyses were processed through weighted-average with burnup ratios of the plutonium rich agglomerates and the matrix. Finally, the fission gas release ratios were compared with results of rod puncture tests. As a result of the comparison, it was confirmed that the proposed approach could well predict fission gas release ratio of MOX fuel with plutonium rich agglomerates.     

Artificial neural network modeling for fission gas release in LWR UO2 fuel under RIA conditions

A fission gas release (FGR) model was developed by using an artificial neural network method to predict fission gas release in UO₂ fuel under reactivity initiated accident (RIA) conditions. Based on the test data obtained in the CABRI test reactor and nuclear safety research reactor, the model takes into account the effect of the five parameters: pellet average burnup, peak fuel enthalpy, the ratio of peak fuel enthalpy to pulse width, fission gas release during base-irradiation, and grain size of a fuel pellet. The parametric study of the model, producing a physically reasonable trend of FGR for each parameter, shows that the pellet average burnup and the ratio of peak fuel enthalpy to pulse width are two of the most important parameters. Depending on the combination of input values for the five parameters, the application of the model to a fuel rod under typical RIA conditions of light water reactor produces 1.7-14.0% of FGR for the pellet average burnup ranging from 20 to 70 MW d/kg U.