Analysis of damage to RBMK core components as a result of an unanticipated accident

The results of a computational analysis of the damage to RBMK core components occurring during unanticipated accidents are presented. The deformation and scale of damage to fuel elements, non-bearing fuel-assembly rods, and channel pipes are examined for three accidents. The calculations show that all fuel elements become unsealed in the accidents examined. Substantial deformation of cladding, resulting in blocking of the flow section of a RBMK fuel channel, does not occur. The central carrying rod (tube) is likely to break with standard and central clamping of fuel element bundles in an assembly. Rupture of fuel channels is unavoidable in all unanticipated accidents studied. __________ Translated from Atomnaya énergiya, Vol. 104, No. 4, pp. 212–216, April, 2008.     

Analytical study on effects of BWR fuel spacer on droplet deposition

In a boiling water nuclear reactor (BWR), liquid film dryout may occur on a fuel rod surface when the fuel assembly power exceeds the critical power. The spacers supporting fuel rods affect on the thermal-hydraulic performance of the fuel assembly. The spacer is designed to enhance critical power significantly. If spacer effects for two-phase flow could be estimated analytically, the cost and time for the development of the advanced BWR fuel would be certainly decreased. The final goal of this study is to be able to analytically predict the critical power of a new BWR fuel assembly without any thermal-hydraulic tests. Initially, we developed the finite element code to estimate spacer effects on the droplet deposition. Then, using the developed code, the spacer effects were estimated for various spacer geometries in a plane channel and one subchannel of BWR fuel bundle. The estimated results of the spacer effects showed a possibility to analytically predict the critical power of a BWR fuel assembly.     

Analytical model for calculation of the thermo hydraulic parameters in a fuel rod assembly

The paper presents the procedure of the cellular calculation of thermo hydraulic parameters of a single-phase gas flow in a fuel rod assembly. The procedure is implemented in the DARS program. The program is intended for calculation of the distribution of the gaseous coolant parameters and wall temperatures in case of arbitrary, geometrically specified, arrangement of the rods in fuel assembly and in case of arbitrary, functionally specified in space, heat release in the rods.In mathematical model the flow cross-section of the channel of intricate shape is conventionally divided to elementary cells formed by straight lines, which connect the centers of rods. Within the limits of a single cell the coolant parameters and the temperature of the corresponding part of the rod surface are assumed constant. The entire fuel assembly is viewed as a system of parallel interconnected channels.Program DARS is illustrated by calculation of a temperature mode of 85-rod assembly with spacers of wire wrapping on the rods.     

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。     

破损燃料组件热室检查技术研究

燃料组件破损直接影响了反应堆的安全运行,分析燃料组件破损原因是燃料组件研发改进的重要环节。通过破损燃料组件水下解体、破口位置定位、破口试样取样等关键技术的研究,建立了破损燃料组件热室检查方法。研究结果表明,该技术路线合理,检查方法可行,为热室条件下开展燃料元件破损检查提供了技术途径。?      

Results of a long-term wet storage demonstration test with intact and operational defective LWR fuel rods

28 spent fuel rods — 18 intact and 10 operational defective rods — were included in the storage test program. Within 7 years the spent fuel rods were inspected four times. To characterize the spent fuel rods the following methods were applied during pool inspections: visual inspection, profilometry, eddy current testing, and oxide thickness recording.Summarizing the results of the intermediate and of the final inspection it has to be concluded that — as predicted — no change exceeding the detection limit could be found either at the intact or at the operational defective fuel rods. These results must be regarded as conservative because handling of the different spent fuel rods during inspection provided additional and atypical loads — especially for the operational defective spent fuel — in comparison with the long term storage of complete fuel bundles.The results of this carefully documented demonstration test has shown agreement with the theoretical analysis and with the overall experience available from pool storage that wet spent LWR-fuel storage can be performed without any problems even for extended periods of time.     

燃料棒周向导热对超临界水堆堆芯子通道分析的影响

本文在子通道程序的燃料棒模型中引入三维导热方程,使该模型能用来模拟燃料棒的周向导热情况。采用改造后的子通道程序对混合谱超临界水堆设计中的两种燃料组件结构进行计算分析,研究燃料棒周向导热对超临界水堆燃料组件子通道分析的影响。结果表明：热谱组件的子通道计算中,燃料棒周向导热的影响不能忽略;快谱组件的子通道计算中,燃料棒周向导热的影响基本可忽略。     

Grid-to-rod flow-induced impact study for PWR fuel in reactor

Dynamic contact impact from hydraulic flow-induced fuel assembly vibration is the source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR). To support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. It is an essential and effective way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.     

反应堆乏燃料组件的破损检测

为了解决某反应堆101盒乏燃料组件外运送贮,对乏燃料组件破损检测方法进行了研究,在已有技术的基础上,根据自身的需求设计、加工了新的采样系统,设计了工作流程,并给出了测量数据的判断依据。该采样系统可以在水下实现一次对一批乏燃料组件进行逐个取样,每个样品对应一个确定的组件,检测效率高。测量破损检测样品101个,根据测量结果判断出1个样品含有85Kr,其余样品不含85Kr或是85Kr的含量低于该条件下仪器的探测限,表明101盒乏燃料组件中有1盒组件包壳存在破损情况,其余组件包壳未破损。     

Modele de comportement mecanique d'element combustible dans un assemblage PWR

Irradiated fuel in pressurized water reactors (PWRs) frequently displays rod bowing, due to two kinds of assymmetry. The first originates in the fabrication of the sheath, causing eccentricity, ovalization and thickness non-uniformity. The other comes from in-pile fuel element conditions such as off-line grids compressing the rods, circumferential thermal gradients on the sheath, and pellet-clad interactions.The MAC code was developed for parametric studies of some of these effects. It shows that:In the case of fuel rods undergoing compressive forces by the spacer grids the usual friction forces are unable to bow the rods significantly, except when the rods are blocked by the spacer grid springs.In some assembly configurations, the temperature difference between adjacent rods is able to bow them, requiring an increase in number of spacer grids.Localized pellet-clad interactions may cause significant bowing, particularly when they occur near the grids.     

大晶粒UO2燃料芯块堆内试验组件的研制

为了开发高性能的压水堆燃料,研制了大晶粒燃料芯块。试验燃料芯块具有高的235U富集度、小直径和大晶粒尺寸的特点。通过堆内辐照试验可以对不同制造工艺的燃料芯块进行评价和筛选,以便确定燃料制造工艺。为了在中国原子能科学研究院池式研究堆中随堆考验,设计了一种试验组件,包含四根双包壳的燃料棒。双包壳燃料棒是在外包壳内装入两根单包壳燃料棒。试验组件直接由反应堆一次循环水冷却,不设专门的冷却回路。试验组件上安装了多种堆芯测量传感器,包括燃料中心温度热电偶、自给能中子探测器和冷却剂出、入口温度热电偶,可以在线监测燃料试验参数。描述了大晶粒UO2燃料芯块的研制、试验燃料组件的研制和检验。     

Fuel Assembly Hydraulic Test of Research Program on In-Pile Irradiation Test for Advanced UO2 Pellets

In order to ensure that the irradiation test for fuel assembly is safe, it is necessary to determine that the coolant velocity on the surface of fuel rods and hydraulic conditions. There is no flow meter on the fuel assembly, both out-of-pile and in-pile hydraulic test have been completed and the flow rate of coolant oassed through the fuel assembly is determined in terms of test results.     

A new method to predict Grid-To-Rod Fretting in a PWR fuel assembly inlet region

Grid-To-Rod Fretting (GTRF) is one of the main causes of leaking fuel in a Pressurized Water Reactor (PWR). GTRF is caused by grid-to-rod gap, secondary flow, and axial/lateral turbulence caused pressure fluctuations within the fuel assembly, which produces rod vibration and wear. The cross flow and vortex shedding phenomenon produce low frequency vibration forces on fuel rods. In some plants, leaking fuel has been detected at the fuel inlet region of fuel assembly designs that do not have Protective Grid (P-grid) which, in addition to providing debris protection, also provides lateral stability against vibration. In order to understand the root cause of the fuel leaks, a thorough investigation of the flow field at the fuel inlet region is required. Leaking fuel has also been detected in the fuel inlet region in transition cores. In the transitional core arrangement, there are different fuel assembly designs next to each other. Due to the structure difference, there will be cross flow between fuel assemblies, which may be the initiating factor for fuel leaks.A method based on Computational Fluid Dynamics (CFD) has been developed in Westinghouse to predict the GTRF in the fuel inlet region. The fuel inlet region consists of the lower core plate, the bottom nozzle, the fuel rods, the thimble rods, the P-grid, and the bottom grid. This study employed CFD to investigate the unsteady forces on the fuel rods under typical reactor in-core conditions. Two fuel assembly (FA) inlet regions with and without the P-grid were simulated. The time history of the unsteady force components on fuel rods was recorded. Fast Fourier Transform (FFT) analyses were carried out for the force history. Compared to the data from operating plants, the new method predicted synchronized excitation forces on the rods that leaked in real operation. The CFD results also demonstrated the advantage of using the P-grid. GTRF at the fuel inlet region can be significantly reduced when the P-grid is used in Westinghouse fuel assembly designs.     

Performance improvement analysis of boiling water reactors by incorporation of hydride fuel

The feasibility of improving the neutronic characteristics of boiling water reactors (BWR) by using U–Zr hydride fuel is studied. Several modified BWR fuel assembly designs are considered. These include designs in which hydride fuel rods replace water rods only, replace water rods and a fraction of the oxide fuel rods, replace oxide fuel in the upper half of all the fuel rods, and replace all the oxide fuel in the assembly. It is found that replacement of at least half of the oxide fuel rods in the fuel assembly by U–ZrH_1.6 fuel might simultaneously improve the performance of BWR in three ways: (a) Increasing the energy extracted per fuel assembly and the cycle length by up to 10%. (b) Reducing the uranium ore and SWU requirements by approximately 10%. (c) Reducing the negative void coefficient of reactivity by, at least, 50%. It is also found that replacement of all the oxide fuel by hydride fuel opens interesting new options for the design of BWR fuel assemblies. The net result might be simplified assembly designs that can generate significantly more energy while featuring small negative void coefficient of reactivity. U–ThH₂ fuel appears to be even more promising than U–ZrH_1.6. For the potential benefits from hydride fuel to be realized, a clad material that is not permeable to hydrogen and is not as neutron absorbing as stainless steel needs to be developed.     

破损燃料组件修复后的物理特性分析

破损燃料组件修复后再次入堆使用是必须进行安全评估,以确保核安全。本文以采用AFA3G燃料组件的CPR1000机组为研究对象,对装入反应堆后的正常燃料组件和修复燃料组件的核物理和功率分布进行分析评估。结果表明:燃料组件内更换一根燃料棒对燃料组件反应性的影响很小,该影响可以忽略。更换不锈钢棒的数量越大,燃料组件反应性变化幅度越大。随着燃耗的加深,燃料组件反应性变化幅度也增大。修复的燃料组件虽然在换棒位置局部区域发生功率畸变,相对功率略微的升高,但离换棒位置较远的燃料棒的相对功率没有变化,换棒不会导致组件内功率峰发生象限的偏移。     

Pressurized heavy water reactor fuel behaviour in power ramp conditions

In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.     

基于涡流技术的燃料棒氧化膜测量信号有效性评估与统计

应用统计学中的中位数、四分位距概念并结合实践经验,研究了燃料棒氧化膜涡流测量信号在线及离线处理流程的优化方法,提出了使用标定信号检验结合氧化膜轴向变化趋势图进行在线的燃料棒氧化膜信号有效性评估,以减少事后复测带来的频繁吊运燃料组件的问题。同时,在对燃料棒氧化膜涡流信号进行离线统计时发现,相对平均数统计,中位数统计可以排除极端信号的干扰,燃料棒轴向长约150mm区域段的氧化膜厚度数据的四分位距多数分布于0～3μm。结果表明,使用中位数可以准确、直观地代表各个区域段的氧化膜厚度水平。     

Specific fuel rod thermomechanical studies for the Advanced Plutonium fuel assembly concept

In order to burn plutonium in PWRs, an innovative assembly has been proposed: the Advanced Plutonium fuel Assembly (APA) concept. This heterogeneous assembly is based on the standard 17x17 geometry, replacing 144 classical UO₂ fuel rods by 36 large fuel rods containing plutonium in an inert matrix. To accommodate the high power and burn-up, a thin annular geometry has been proposed. However, the design of such assembly is a challenge due to the highly innovative fuel rod concept. Regarding the fabrication, different processes are under investigation: concentric claddings with annular ceramic pellets (cercer), ceramic/metal fuel (cermet) metallurgically bonded to the claddings, vibrocompacted fuel. This paper focuses on thermomechanical studies performed for the two first concepts, in view of a future selection between the different processes. The analysis concerns temperature profiles, fuel-cladding gap evolutions and cladding buckling risk. Regarding those parameters, the conclusion is that the cermet matrix presents satisfactory conditions: operating at very low temperatures and no buckling risk.     

Analysis of stress applied to a ruptured cladding tube under horizontal vibration

ABSTRACT

After the termination of a loss-of-coolant accident (LOCA), the reactor continues to be cooled for a long term until fuel assemblies are withdrawn from the reactor core. The fuel cladding tube degrades in strength due to high-temperature oxidation during a LOCA event. It is important to confirm that fuel rods exposed to LOCA conditions can withstand earthquakes during the long-term cooling in terms of preserving the coolable geometry of the reactor core. Finite element method analyses were performed to estimate the deformation of fuel rods in a fuel assembly under vibrations simulating an earthquake as well as the stress applied to the fuel cladding tube with a rupture opening. The localized stress at the rupture opening in the analyses was compared with the strength assessed through bending tests of the cladding tube samples that were ruptured and oxidized to less than 15% equivalent cladding reacted (ECR) in advance. As the result, the fuel rods are expected to be prevented from fracture due to bending at earthquakes during the post-LOCA cooling unless the oxidation of cladding tubes exceeds the limit defined in the current Japanese LOCA criteria, 15% ECR and a deflection of the fuel rodexceeds approximately 40 mm.