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.     

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。     

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.     

国产压水堆核电站40 GW•d/tU乏燃料棒金相检验

为评价国产燃料棒在较高燃耗水平下的辐照性能，在中国原子能科学研究院燃料与材料检验设施（303热室）对燃耗为40 GW•d/tU的国产压水堆核电站乏燃料棒进行了金相检验。检验内容包括芯块宏观与微观组织、包壳水侧腐蚀与氢化物分布、芯块-包壳相互作用状况等。金相检验结果表明：40 GW•d/tU燃耗下，芯块未发生明显的轮廓变化，气孔率为3.3%～5.8%，晶粒组织为等轴晶，平均晶粒尺寸为7.2 μm；Zr合金最大水侧氧化膜厚度为23 μm，氢化物分布和含量正常，最大氢含量约为150 μg/g，同时不同部位的包壳氢含量与水侧氧化膜厚度基本呈线性关系，水侧腐蚀处于正常水平；包壳内壁有局部轻微腐蚀，包壳与芯块之间存在间隙，未发生包壳与芯块相互作用情况。     

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.     

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.     

BEAF - a computer program for analysis of light water reactor fuel rod behavior

BEAF - a computer program for analysis of light water reactor fuel rod behavior was developed. The BEAF code, which is appropriate for on-line prediction of fuel rod behavior, can analyze fuel rod thermal and mechanical behaviors using the axisymmetric, plane strain approximation and finite difference method to realize a fast running time.In the mechanical analysis, a new cracked pellet compliance model is introduced, in which pellet cracking and crack healing, pellet initial relocation, modified elastic moduli of a cracked fuel pellet, and stress dependent hot pressing are considered. Adding to those capabilities, fission gas flow and diffusion in the fuel-clad gap are analyzed to take into account the slow fission gas dilution effect on the gap conductance during power ramp.     

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。      

Effect of Fuel Pellet Eccentricity with Cladding on Fuel Rod Thermal Behavior under Reactivity Initiated Accident Condition

Fuel rod behavior under Reactivity Initiated Accident (RIA) conditions has been studied in the Nuclear Safety Research Reactor (NSRR), JAERI. In the experiments, cladding thermal behavior was observed to be influenced by the fuel pellet eccentricity to produce large azimuthal temperature variation in the cladding. The maximum azimuthal cladding temperature difference was measured to be as large as 150°C by thermocouples attached to opposite sides of the cladding around the circumference, though the thermocouples did not always detect the maximum temperature difference around the circumference. The actual temperature differences in the fuel rods subjected to less than 290 cal/g?UO₂ were estimated to be 350°C at maximum based on metallographies. A simple calculation considering gap conductance variations also showed that the maximum temperature difference became 350°C under fully eccentrical condition in the fuel rod subjected to 260 cal/g?UO₂. Moreover, as the rod damage such as cladding deformation, melting and failure occurs unevenly around the circumference due to the fuel pellet eccentricity in general, the fuel pellet eccentricity should influence the fuel rod failure under RIA conditions.     

International Solvent Extraction Conference ISEC ′90

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs.

The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.     

Oxide thickness-dependent transient cladding hoop stress

A few thrice-burned Zry-4 fuel assemblies which were loaded in one of the PWRs operating in Korea were found to be failed due to PCI during a power ramp following a rector trip, while thrice-burned Zr–Nb fuel assemblies and twice-burned Zry-4 ones were intact. To investigate the effect of fuel rod oxide thickness on power ramp-induced cladding hoop stress, three intact fuel rods were selected, which include an intact twice-burned Zry-4 fuel rod, an intact thrice-burned Zr-4 fuel rod and an intact thrice-burned Zr–Nb fuel rod. With the use of a fuel performance analysis code, burnup-dependent steady-state cladding stress and ramp power-dependent cladding stresses at the power-ramped burnup were predicted for the three intact fuel rods. It was found that the cladding oxide thickness has a considerable effect on the ramp power-dependent cladding hoop stresses. In addition, the cladding maximum stress of the thrice-burned Zry-4 fuel rod with 125 μm oxide thickness exceeded an ultimate cladding tensile strength of the Zry-4 cladding when the pellet–clad friction coefficient-dependent cladding stress concentration ratio was considered. However, the thrice-burned Zr–Nb fuel rod with 50 μm oxide thickness was evaluated to have a considerable margin against the power ramp-induced PCI failure.     

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.     

UO2-不锈钢燃料棒的光纤激光切割研究

为验证光纤激光用于燃料组件解体和燃料棒切割的可行性,研究光纤激光用于热物性差别很大的UO2芯块 不锈钢包壳管复合结构的切割和铀芯块的切割质量,本文采用光纤激光切割UO2芯块 316Ti包壳管元件棒,并通过扫描电子显微镜、能谱和X射线衍射对UO2芯块的切断面进行微观表征分析,研究激光切割过程对铀芯块切断的表面微观形貌、元素组成及物相的影响。研究结果表明,光纤激光可用于切割UO2芯块 316Ti包壳管元件棒,激光切割过程虽会造成铀芯块切断面出现大量微孔和碎渣,但不会造成UO2的相变。以上结果表明,光纤激光可用于UO2芯块 316Ti包壳管元件棒的切割,通过后续对激光切割系统的抗辐射屏蔽防护,可应用于乏燃料组件解体和乏燃料棒切割。     

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。     

Analysis of fuel rod behavior under normal operating conditions in Super Fast Reactor

A Super Fast Reactor is a pressure-vessel type, fast spectrum supercritical water-cooled reactor (SCWR) that is presently researched in a Japanese project. A preliminary core has been designed with 1.59E+06 W/m³ of power density [1]. In order to ensure the fuel rod integrity, the fuel rod behaviors under the normal operating conditions are analyzed using FEMAXI-6 code. Three types of the limiting fuel rods, with the maximum cladding surface temperature (MCST), maximum power peak (MPP) and maximum discharge burnup (MDB), are chosen to cover all the fuel rods in the core. The power histories of these fuel rods are taken from the neutronics calculation results in the core design. The available design range of the fuel rod design parameters, such as the initial gas plenum pressure, gas plenum length, grain size and pellet-cladding gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900 °C. (2) Maximum cladding stress in circumstance direction should be less than 100 MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Compressive stress to yield strength ratio should be less than 0.2. (5) Cumulative damage fraction (CDF) on the cladding should be less than 1.0. Finally the improved fuel rod design is proposed.     

Transient Elongation of a Fresh Fuel Rod under Reactivity Initiated Accident Conditions

In-pile experiments of fresh fuel rods under reactivity initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor at the Japan Atomic Energy Research Institute in order to understand the basic pellet cladding mechanical interaction (PCMI) behavior. Rapid fuel pellet expansion due to a power excursion would cause radial and longitudinal deformation of the cladding. This PCMI could be one of the possible incipient failure modes of an embrittled cladding of a high burnup fuel under the RIA conditions.

Basic PCMI behavior was studied by measuring cladding deformation of a fresh fuel rod without complicated irradiation effects. The transient elongation measurements of the fuel with two kinds of gap width indicated not only PCMI-induced cladding elongation, but also reduction of the pellet stack displacement by the cladding constraint. In the tests under a high-pressure and high-temperature condition simulating an operation condition of BWRs, additional ridge-type cladding deformation was generated due to the axial collapse of the cladding. A preliminary analysis for interpretation of the tests was made using a computer code for the transient analysis of fuel rods, FRAP-T6.     

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.     

Author index to volume 47

The URANUS code, a digital computer programme for the thermal and mechanical analysis of integral fuel rods, is described. With this code the fuel rods found in the majority of power reactors can be analyzed. URANUS is built around a quasi two-dimensional analysis of fuel and cladding. The mechanical analysis can accommodate seven components of strain: elastic, time-independent plastic, creep and thermal strains, as well as strains due to swelling, cracking and densification. The heat generation and temperature distribution, cladding/fuel gap closure, pellet cracking and crack healing, fission-gas release, corrosion, O/M-distribution and plutonium redistribution are modelled. Geometric non-linearities (large displacements) are included; steady state or transient loading (pressure, temperature) is possible. In this paper special attention is paid to a theory for determining crack structures. The present status of the URANUS computer programme and a critical comparison with other fuel rod codes as well as sample analyses are given.     

Mechanical design issues and resolutions of a dual cooled fuel for the OPR-1000

A dual cooled fuel is recently brought into focus due to its potential of considerable power uprating. The purpose of present work is to realize the innovative concept of a dual cooled fuel to be a fuel assembly structure compatible with the OPR-1000 system. Under the framework, the critical issues such as the outer cladding thickness and the growth difference of the inner and outer cladding tubes are dealt with in this paper. We designed the thickness of outer cladding tube by using the elastic buckling criterion and safety factor analysis. From the concern of the inner cladding's bowing during irradiation, it was suggested that the outer cladding would grow more than the inner one by applying different microstructures to the inner and outer cladding tubes. It was noted that the gap between fuel rods would not be narrowed further during the different irradiation growth. The structural components such as fuel rod supporting structure, top and bottom end pieces and guide tubes could be designed without a drastic change of those of the conventional fuel. Candidate designs of the components are also presented.     

URANUS — A computer programme for the thermal and mechanical analysis of the fuel rods in a nuclear reactor

The URANUS code, a digital computer programme for the thermal and mechanical analysis of integral fuel rods, is described. With this code the fuel rods found in the majority of power reactors can be analyzed. URANUS is built around a quasi two-dimensional analysis of fuel and cladding. The mechanical analysis can accommodate seven components of strain: elastic, time-independent plastic, creep and thermal strains, as well as strains due to swelling, cracking and densification. The heat generation and temperature distribution, cladding/fuel gap closure, pellet cracking and crack healing, fission-gas release, corrosion, O/M-distribution and plutonium redistribution are modelled. Geometric non-linearities (large displacements) are included; steady state or transient loading (pressure, temperature) is possible. In this paper special attention is paid to a theory for determining crack structures. The present status of the URANUS computer programme and a critical comparison with other fuel rod codes as well as sample analyses are given.