核电站燃料棒破损在线探测系统FDD-1

核电站燃料棒破损在线探测系统FDD-1由γ射线探头、γ谱仪、计算机和燃料棒破损状况分析程序组成。将探头对准化容系统管道,测量一回路内水中放射性核素γ剂量变化,监测燃料棒是否发生破损,如破损发生,分析破损性状,给出破损燃料棒的根数、破口大小和破损燃料棒的燃耗(以判断破     

反应堆燃料组件破损检查装置

反应堆在运行中发现冷却剂中的放射性水平增高,在确信燃料棒有破损的情况下,就要停堆,找出破损组件。进而找出破损单棒。然后,或是通过调换破损单棒,使破损组件修复继续参加运行,或是将破损组件另行存放,不再运行。为了积累燃料组件破损检查装置的设计经验,同时为了检查一批存放已久的考验燃料组件,以对燃辩组件的质量做出评价,我们设计安装了燃料组件破损检查装置。并进行了校核模拟试验和正式的检查。     

压水堆电站燃料元件破损定位探测方法介绍

本文主要介绍70年代以来国外压水堆电站燃料元件破损定位探测技术的新发展。利用无损检洲技术(如超声探伤、 发射技术、红外测温等)来定位检测破损燃料元件棒,以实现不解体燃料组件而更换掉组件中破损元件棒,然后将未达燃耗的燃料组件再重新入堆。     

核电站燃料棒破损在线探测系统研制

采用在线检测方法对现役核电站燃料棒的破损情况进行监测可以克服传统化学取样方法不能连续探测和不能及时报告堆内燃料破损情况的不足.本工作研制出核电站燃料棒破损在线探测系统(FDDS-1),通过检测一回路核燃料裂变产物的活度,根据燃料破损性状分析程序FUDAC-1计算出燃料棒的破损根数等参数,给出在线探测报告.     

水堆燃料元件的破损监测、定位和处理（续）

4在燃料组件内鉴别破损的燃料棒 一旦某束燃料组件用上述方法己判定为破损组件，如果条件不具备可将其放入专门容器贮存。在正常情况下就要从这束燃料组件中找出破损的燃料棒，然后抽出换上新棒。鉴别破损燃料棒的第一项技术就是外观检查超声检验。     

反应堆控制棒组件超声检测探头的研制

为了检测核电站反应堆控制棒组件,保障核电站在役检查顺利实施、降低检测成本。对反应堆控制棒组件（RCCA）检测用超声探头进行自主研制,本文详细介绍了15MHz-Φ4mm-FP8 mm RCCA超声探头制作流程,通过对压电晶片、声透镜、背衬3方面详细介绍探头制作工艺。通过对超声探头进行性能测试,测试脉冲周期数为1.5周,频带宽度为105%,在高温环境下仍能保持优良性能。对超声探头进行模拟检验测试,缺陷测试结果清晰可见,满足检验需求,可完全实现国产化替代进口产品。       

水冷反应堆燃料元件的在役检测和处理

水冷反应堆包括轻水堆和重水堆,轻水堆分为压水堆和沸水堆;重水堆分为加压重水堆和加拿大的氘铀堆。国际上把它们归为一类进行研究。本文涉及的破损燃料元件的在役检测和处理包括：反应堆运行时的检测;换料时或换料后的检测;在燃料组件内鉴别破损的燃料棒;燃料组件的监测、拆卸和修复;破损燃料棒拆出后的检测,破损定位与修补。     

水冷反应堆燃料元件的在役检测和处理

水冷反应堆包括轻水堆和重水堆,轻水堆分为压水堆和沸水堆;重水堆分为加压重水堆和加拿大的氘铀堆。国际上把它们归为一类进行研究。本文涉及的破损燃料元件的在役检测和处理包括：反应堆运行时的检测;换料时或换料后的检测;在燃料组件内鉴别破损的燃料棒;燃料组件的监测、拆卸和修复;破损燃料棒拆出后的检测,破损定位与修补。     

破损燃料定位检测方法

本发明提供一种破损燃料定位检测方法。其棒束对定位，实时监测卸料池间的伽玛剂量变化趋势，判断是否符合特征曲线。符合特征曲线，则表明破损燃料还在该通道内，继续对该通道内的剩余乏燃料棒束进行换料。反之，则表明该对燃料棒束含有破损，确定出破损燃料棒束对所在位置。     

基于人工神经网络的压水堆燃料破损状态监测

为提升对核反应堆燃料棒包壳破损的预测能力,建立两个串联的人工神经网络分别判断燃料棒包壳是否破损以及破损程度。通过改变沾污铀质量、增加数据扰动、改变运行功率和使用更少的特征核素进行训练,对用于判断是否破损的神经网络模型和判断破损等级的神经网络进行了性能测试和分析。在沾污铀质量小于0.5 g、数据扰动在30%以内、单棒功率在77 kW到120 kW之间的条件下,第1个人工神经网络能较好地判断出是否破损。第2个神经网络,对于考虑的5种破损程度,判断的精确性较高。与传统的碘同位素比值法相比,神经网络方法响应更快,精度更高。结果表明,人工神经网络可用于预测反应堆燃料包壳是否发生破损以及破损程度。     

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.     

The effect of fuel rod loading speed on spacer grid spring force

The burnup-dependent grid-to-rod gap combined with the fluid-induced vibration may generate grid-to-rod fretting wear-induced fuel failure for some fuel assemblies in a certain burnup range. The grid-to-rod gap is dependent on initial spacer grid spring force, spring force relaxation and cladding creepdown. It is found that the initial spring force is reduced during the fuel rod loading into the fuel assembly skeleton. The extent of the initial spring force loss is strongly dependent on the fuel rod loading speed. Based on the initial spring force loss data obtained from two kinds of fuel rod loading speeds of 0.18 and 0.33 m/s, it can be said that the higher rod loading speed generates the larger initial spring force loss. This is because the higher speed generates the larger overshooting of spring deflection during the fuel rod loading. The extent of overshooting may be affected by axial misalignment of SG cells, spring-to-fuel rod end plug contact angle, ballooning of FR end plug weld region and the extent of gravity-induced FR bowing, combining with the fuel rod loading speed. The rod loading speed of 0.33 m/s is found to produce some spacer grid cells less than a minimum initial spring force requirement of 12 N against the grid-to-rod fretting wear-induced failure. In order to produce initial spacer grid spring force meeting the minimum spring force requirement, it is recommended that the lower rod loading speed be used, combined with axially aligned spacer grid cells and lower contact angle of spring-to-fuel rod end plug.     

WTRLGD—A computer program for the transient analysis of waterlogged fuel rods under the RIA condition

A computer code WTRLGD has been developed to describe the transient internal pressure of a waterlogged fuel rod during power burst and also to predict the possibility of the rod failure in the mode of cladding rupture. The code predicts transient thermal behavior of the fuel rod on the basis of an assumption of axisymmetry, and thermal-hydraulic transients of the internal water on the basis of a homogeneous volume-junction model modified so as to involve the cladding deformation. Calculated transients of the rod pressure are in fairly good agreement with those measured in the NSRR experiments, simulating the fuel rod behavior under an RIA condition. The comparison between calculation and experiment verifies that the code is an effective tool for the prediction of the failure of a waterlogged fuel rod.     

The effect of fuel rod supporting conditions on fuel rod vibration characteristics and grid-to-rod fretting wear

The grid-to-rod fretting wear-induced fuel rod failure observed in PWRs may be caused by excessive fluid-induced vibration and inadequate fuel rod support by the spacer grid spring. In order to simulate in-reactor grid-to-rod fretting wear behaviors, the grid-to-rod fuel rod supporting conditions as a function of time were predicted by taking into account cladding creep rate, initial spacer grid spring deflection, spacer grid spring force relaxation, etc. Based on these grid-to-rod supporting conditions, the fuel rod vibration modes and natural frequencies were calculated with the help of the ANSYS code, while the fuel rod vibration amplitudes were estimated by the Paidoussis’ empirical formula. With these vibration characteristics that depend upon the grid-to-rod supporting conditions, the in-reactor fretting wear axial profile observed on the fuel rod surface are found to be simulated quite well. In addition, key design guidelines for the fuel assembly and the spacer grid are proposed to minimize the grid-to-rod fretting wear that may be utilized to develop an advanced fuel design against fretting wear.     

Unbiasedness Evaluation of Statistical Error in Staged Monte Carlo Calculation

Fuel rod failure behavior has been studied under a reactivity initiated accident condition in Nuclear Safety Research Reactor (NSRR), JAERI. In the studies, inetallurgical observations showed that the incipient fuel rod failure mode was oxygen-induced embrittlement of the cladding independent of the test conditions such as fuel designs and cooling environments except for pressurized and waterlogged fuels. Development of the oxidation layers and embrittlement of β-Zry were quantitatively evaluated through the metallurgical examinations. A diffusion equation of oxygen was solved under a finite system with moving boundary conditions to obtain the oxygen concentration and evaluate the cladding embrittlement. The calculation showed that the wall thinning due to the cladding melt is needed for the complete embrittlement because the wall thinning enhances the oxygen concentration in the β-Zry, which well explain the experimental results. Therefore the failure threshold energy is determined by the cladding melting temperature. The failure threshold derived from this study is expected to be applicable to predicting the fuel rod failure behavior in computer analyses and also useful to evaluate the failure threshold energy for the new types of fuel rod.     

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.     

A model for predicting pellet-cladding interaction-induced fuel rod failure

A model for predicting pellet-cladding mechanical-interaction-induced fuel rod failure is presented. Cladding failure is predicted by explicitly modelling the formation and propagation of radial cladding cracks by the use of non-linear fracture mechanics concepts in a finite element computational framework. The failure model is intended for implementation in finite element fuel performance codes in which local pellet-clad interaction is modelled. Crack initiation is supposed to take place at pre-existing cladding flaws, the size of which is estimated by simple probabilistic concepts, and the subsequent crack propagation is assumed to be due to either iodine-induced stress corrosion cracking or ductile fracture. The novelty of the outlined approach is that the development of cladding cracks which may ultimately lead to fuel rod failure can be treated as a dynamic and time-dependent process. The influence of complex or cyclic loading, ramp rates and material creep on the failure mechanism can thereby be investigated. The presented failure model has been incorporated in the ABB Atom transient fuel performance code. Numerical results from some applications of the code are used to illustrate the usefulness of the model.     

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.     

基于MOOSE框架的五方程两相流分析程序开发

基于多物理场耦合框架MOOSE,采用五方程两相流模型开发了模块化程序ZEBRA,实现了高阶时间、空间离散格式两相流动传热问题的求解。采用Bartolomei开展的垂直圆管过冷沸腾实验对ZEBRA进行验证,在不同热流密度、质量流密度、压力工况下,将程序计算值与实验值进行了数值验证和计算分析。结果表明：ZEBRA中五方程模型预测值与实验值符合良好,沸腾起始点和空泡份额的预测合理,表明ZEBRA初步具备了处理两相流问题的能力。     

Effects of Rod Pre-Pressurization on Light Water Reactor Fuel Behavior during Reactivity Initiated Accident Conditions

Previously pressurized (pre-pressurized) fuel rod tests recently performed in the Nuclear Safety Research Reactor (NSRR) investigate the effects of initial internal pressure on fuel rod behavior during reactivity initiated accident (RIA) conditions. A single PWR type fuel rod was contained within a waterfilled, ambient temperature and ambient pressure capsule. The fuel rod was then heated by the pulsing operation of the NSRR.

Results from the tests show that the effect of pre-pressurization was significant for the fuel rods with initial internal pressure of 0.8 MPa and above, and fuel rod failure occurred from rupture of the cladding with lower threshold energy deposition for failure as the initial internal pressure was increased. The cladding rupture was governed mainly by the cladding temperature rise, not by the rod internal pressure rise during the transient. The relationships between cladding burst pressure and cladding burst temperature and between cladding strain and cladding temperature at cladding rupture obtained in the present study under an RIA condition agree with the results obtained from various in- and ex-reactor experiments under a LOCA condition, although the obtained time-averaged strain rate of the Zircaloy cladding was much greater than that in a LOCA condition.