核燃料组件无损检测探测系统设计

燃料组件是反应堆的核心部分,在高温、高压及强中子辐射场等复杂环境条件下,燃料棒中芯块会出现肿胀、变形甚至包壳破裂,严重威胁反应堆的安全运行。为了更好地了解燃料组件在反应堆内的变化,研究高燃耗的燃料组件中燃料棒的中心空洞形成和燃料棒的变形情况,高能X射线无损检测是一种有效的技术手段。由于辐照后核燃料组件自身具有强放射性,探测系统设计中必须考虑减弱燃料组件自身辐射对探测采集的影响,因此组件探测系统中探测器阵列及准直器的优化设计十分必要。经过建模及相关模拟计算,得到了探测器单元最佳尺寸,优化了后准直器的结构设计,为提高燃料组件无损检测系统重建图像的质量提供帮助。     

Improvement of LWR thermal margins by introducing thorium

The use of thorium in pressurized water reactor fuel assemblies is investigated in this paper. The novelty of the reported work is to study a fuel design primarily intended to control the excess of reactivity at beginning of life, and flatten the intra-assembly power distribution rather than converting fertile Th-232 into fissile U-233. The fuel assembly is a traditional 17 × 17 pressurized water reactor fuel design. The majority of the fuel pins contain a mixture of uranium and thorium oxides, while a few fuel pins contain a mixture between uranium and gadolinium oxides. The calculation were performed by two-dimensional transport calculations with the Studsvik Scandpower CASMO-4E code in order to determine the main neutronic properties of the new fuel design, compared with the traditional uranium-based fuel assembly containing gadolinium used as reference. The majority of the neutronic properties of the uranium-thorium-based fuel assembly were similar to the reference fuel assembly. The Doppler and the moderator temperature coefficients of reactivity were found to be appreciably more negative in the uranium-thorium-based design, but still within acceptable limits. One advantage of this new uranium-thorium-based design is a reduction of the pin peak power at beginning of life, because of smaller amount of gadolinium being used. This is important from an operational and safety viewpoint, since the margin to departure from nucleate boiling becomes larger. Consequently, this new type of thorium-based fuel assembly shows advantageous properties for use in power-uprated cores.     

Development and characterization of the control assembly system for the large 2400 MWth Generation IV gas-cooled fast reactor

The present paper is related to the design and neutronic characterization of the principal control assembly system for the reference large (2400 MWth) Generation IV gas-cooled fast reactor (GFR), which makes use of ceramic–ceramic (CERCER) plate-type fuel-elements with (U–Pu) carbide fuel contained within a SiC inert matrix. For the neutronic calculations, the deterministic code system ERANOS-2.0 has been used, in association with a full core model including a European fast reactor (EFR)-type pattern for the control assemblies as a starting point. More specifically, the core contains a total of 33 control (control system device: CSD) and safety (diverse safety device: DSD) assemblies implemented in three banks. In the design of the new control assembly system, particular attention was given to the heat generation within the assemblies, so that both neutronic and thermal–hydraulic constraints could be appropriately accounted for. The thermal–hydraulic calculations have been performed with the code COPERNIC, significant coolant mass flow rates being found necessary to maintain acceptable cladding temperatures of the absorber pins.     

Diagnosis and cure of Dhruva fuel vibration

Dhruva is a high flux research reactor with a nominal thermal power of 100 MW. The fuel for the reactor is in the form of seven-pin cluster of metallic natural uranium clad with aluminium. The optimisation from the physics and thermal hydraulic considerations has resulted in this design of small diameter, long pins arranged hexagonally ensuring a minimum specified clearance between the pins. The clearance is maintained throughout the length by a number of spacers located at regular intervals. This seven pin cluster is assembled inside an aluminium flow tube and the assembly goes into coolant channels made of zircaloy. The fuel assembly is constrained radially (i.e. in the horizontal plane) by the bulges at the two ends of the flow tube.The fuel was endurance tested in an out-of-pile flow test facility for many thousands of hours without any visible damage. However, on loading them in the reactor, many of the fuel pins failed due to fretting wear at the spacer locations. The maximum wear- was on the outer pins near the mid-length of the fuel assembly. The paper gives the details of the measurement and analysis carried out to understand the causes. The solution adopted was to make the supporting bulges flexible - the bottom one by cutting axial slits to obtain a collet type fixture and the top by a sleeve with slits to obtain leaf spring type support. With these design changes, the fuel performs satisfactorily.     

超临界水冷堆双排燃料组件子通道分析

研究基于Cobra-IV程序,开发了适用于超临界水冷堆燃料组件分析的子通道程序.针对超临界水冷堆慢谱双排组件,进行了稳态计算,获取了相关组件热工水力参数.在此基础上,针对单一通道进行了瞬态计算,分析了燃料棒线功率变化和冷却剂流量变化条件下,超临界水冷堆燃料组件的流动和传热的动态响应,为超临界水冷堆组件的优化设计提供了参考.     

Application of X-Ray Computer Tomography for Observing the Deflection and Displacement of Fuel Pins in an Assembly Irradiated in FBR

A nondestructive method making use of X-ray computer tomography (X-ray CT) has been applied to post irradiation examination of fast breeder reactor (FBR) fuel assemblies. In the study, an examination is made of the deflection and displacement of fuel pin in a fuel assembly irradiated to 74.2GWd/t peak burnup in the fast reactor “JOYO.”

In the examination, X-ray CT images of transverse cross sections of fuel pin were obtained at different heights of fuel pin along its axis. Analysis of the resulting images indicated that:

1. The hexagonal wrapper tube had its lateral wall faces slightly bulged outward;

2. The fuel pins loaded in the outermost array were markedly displaced in the direction of wrapper tube, particularly in portions of fuel pin intermediate between positions constrained by wrapping wire.

The latter behavior of fuel pins was substantiated by the contours of fuel pin along its axis, which were derived from cross section images obtained at different levels along axis.

Such fuel pin displacement is surmised to have been caused by thermal stressing of the affected fuel assembly cladding.     

六边形套管型燃料堆芯临界质量测量试验结果验证分析

为验证核设计程序对燃料组件、铍组件和铝组件的计算可靠性,对六边形套管型燃料堆芯（HCTFR）临界质量测量试验数据进行了验证计算和偏差分析。通过分析不同位置铝组件的反应性差异,提出了新的近活性区铝组件计算模型,将铝组件近活性区布置方案的计算偏差从2.2%降低至0.1%,为堆芯核设计程序的工程验证奠定了较好的基础。      

动力堆乏燃料组件燃耗的非破坏性分析

概要综述了用无源和有源非破坏性分析技术测量动力堆乏燃料组件燃耗的基本原理、方法和实验装置。由电离室和裂变室组成的标准叉型探测器具有性能稳定可靠、分析速度快、操作简单、携带方便等优点。当前，它对ＬＷＲ组件的燃耗测量值和申报值的偏差在±１％以内。用高分辨γ谱方法（ＨＲＧＳ）测量组件的燃耗，也能达到同样的精度。根据测量得到的中子计数或γ放射性，可以确定组件中可裂变物质的含量。     

Irradiation Performance of Fast Reactor MOX Fuel Assemblies Irradiated to High Burnups

Mixed oxide fuel assemblies (MFA-1 and MFA-2 assemblies) were irradiated in the fast flux test facility to evaluate the irradiation performance of fast reactor core fuels at high burnups and high fast neutron fluences. The MFA-1 and MFA-2 assemblies achieved respective peak pellet burnups of 147 and 162GWd/t, and resisted to respective peak fast neutron fluences (E > 0:1 MeV) of 21:4 _ 1026 and 23:8 _ 1026 n/m2, without any indication of fuel pin breaching. Structural components of these assemblies were made of modified type 316 stainless steel and 15Cr-20Ni base advanced austenitic stainless steel. Postirradiation examinations of these assemblies revealed dimensional changes of fuel pins and assembly ducts due to irradiation-induced void swelling and irradiation creep, and fuel cladding local oval distortions due to bundle-duct interaction (BDI). The swelling resistance of 15Cr-20Ni base advanced austenitic stainless steel fuel pin cladding was almost the same as that of the modified type 316 stainless steel cladding, while the assembly duct of the former material had a slightly higher swelling resistance than that of the latter material. Analyses of fuel pin bundle deformations indicated that these assemblies likely mitigate BDI mainly by fuel pin bowings and cladding oval distortions.     

使用离散钍铀燃料组件的CANDU6堆物理特性初步研究

为研究钍铀燃料在CANDU6堆中的应用,采用DRAGON/DONJON程序,对使用离散型钍铀燃料37棒束组件的CANDU6堆进行时均堆芯分析。结果表明,组件采用235U富集度为2.5%的铀棒以及第1、2、3圈布置钍棒的37棒束组件,堆芯在8棒束换料、3个燃耗分区的方案下,组件的冷却剂空泡反应性较使用天然铀的37棒束组件（NU-37组件）与采用混合钍铀元件棒的37棒束组件更负;堆芯最大时均通道/棒束功率满足小于6700?kW/860?kW的限值;燃料转化能力比采用NU-37组件时更高;卸料燃耗可到达13400?MW·d/t(U)。研究表明,所设计的离散型钍铀燃料37棒束组件可用于现有CANDU6堆芯,且无需对堆芯结构及控制机构作重大改造;燃料组件和堆芯设计方案可为钍铀燃料在CANDU6堆芯的应用提供参考。      

Experimental validation of radial reconstructed pin-power distributions in full-scale BWR fuel assemblies with and without control blade

Total fission rate measurements have been performed on full-size BWR fuel assemblies of type SVEA-96+ in the zero power reactor PROTEUS at the Paul Scherrer Institute. This paper presents comparisons of reconstructed 2D pin fission rates from nodal diffusion calculations to the experimental results in two configurations: one “regular” (I-1A) and the other “controlled” (I-2A). Both configurations consist of an array of 3 × 3 SVEA-96+ fuel assemblies moderated with light water at 20 °C. In configuration I-2A, an L-shaped hafnium control blade (half of a real cruciform blade) is inserted adjacent to the north-west corner of the central fuel assembly. To minimise the impact of the surroundings, all measurements were done in fuel pins belonging to the central assembly. The 3 × 3 experimental configuration (test zone) was modelled using the core monitoring and design tools that are applied at the Leibstadt Nuclear Power Plant (KKL). These are the 2D transport code HELIOS, used for the cross-section generation, and the 3D, 2-group nodal diffusion code PRESTO-2. The exterior is represented, in the axial and radial directions, by 2-group partial current ratios (PCRs) calculated at the test zone boundary using a 3D Monte Carlo (MCNPX) model of the whole PROTEUS reactor. Sensitivity cases are analysed to show the impact of changes in the 2D lattice modelling on the calculated fission rate distribution and reactivity. Further, the effects of variations in the test zone boundary PCRs and their behaviour in energy are investigated.     

Calculation of the ex-core neutron noise induced by individual fuel assembly vibrations in two PWR cores

Calculation of the neutron noise induced by fuel assembly vibrations in two pressurized water reactor(PWR) cores has been conducted to investigate the effect of cycle burnup on the properties of the ex-core detector noise. An extension of the method and the computational models of a previous work have been applied to two different PWR cores to examine a hypothesis that fuel assembly vibrations cause the corresponding peak in the auto power spectral density(APSD) increase during the cycle. Stochastic vibrations along a random two-dimensional trajectory of individual fuel assemblies were assumed to occur at different locations in the cores. Two models regarding the displacement amplitude of the vibrating assembly have been considered to determine the noise source. Then, the APSD of the ex-core detector noise was evaluated at three burnup steps. The results show that there is no monotonic tendency of the change in the APSD of ex-core detector; however, the increase in APSD occurs predominantly for peripheral assemblies. When assuming simultaneous vibrations of a number of fuel assemblies uniformly distributed over the core, the effect of the peripheral assemblies dominates the ex-core neutron noise.This behaviour was found similar in both cores.     

田湾核电厂长周期混合堆芯稳态热工水力分析

田湾核电厂1、2号机组计划自2014年开始向长周期燃料循环过渡,在AFA型燃料组件组成的堆芯中逐步装入TVS-2M新型燃料组件,经过3个燃料循环的过渡,堆芯将全部装载TVS-2M型燃料组件,以实现长周期燃料循环。燃料组件结构的改变使原堆芯热工水力分析不再适用。本文以长周期燃料循环过渡时期的5种典型堆芯组成情况为例,介绍了VVER机组稳态热工水力分析的程序和方法,对混合堆芯的稳态热工水力特性进行了重新分析。结果表明,混合堆芯稳态设计仍满足热工水力设计准则。     

环形元件超高通量堆堆芯初步概念设计

基于环形燃料元件,提出了一种超高通量堆(UFR)堆芯概念设计。UFR燃料组件设计采用61个燃料元件构成的六角形组件,堆芯采用52盒燃料组件、9盒控制棒组件和厚反射层设计。通过开展堆芯概念设计方案评价,给出了堆芯循环长度、中子注量率、中子能谱、中子空间分布等关键参数。结果表明,在当前的总体参数下所提出的UFR的最大中子注量率可达到1.0×10¹⁶ cm^-2·s^-1。     

Characterisation of radial reaction rate distributions across the 92-pin section of a SVEA-96 Optima2 assembly

Characterisation of the SVEA-96 Optima2 boiling water reactor assembly, in terms of the radial distributions of normalised total fission and ²³⁸U capture rates, is reported at its central elevation, i.e. at the 92-pin section, where the one-third part-length pins are replaced by water. Measurements performed in the PROTEUS facility are compared with MCNPX predictions. The calculation model included the measured locations of the SVEA-96 Optima2 assemblies and sub-assemblies, within the PROTEUS test zone. Predicted and experimental fission and ²³⁸U capture rates are found to agree, respectively, within 3.5% and 4% for all pins. Fission rates in the burnable-absorber UO₂–Gd₂O₃ fuel pins have been predicted without bias using the ENDF/B-VI data library but show an average 1.4% under-prediction with the JEFF-3.1 data library. A slight overestimation of the total fission rate in the pins located at the periphery of the assemblies was observed and has been attributed to an inaccurate modelling of the pin positions. However, there was no systematic bias observed due to the absence of the one-third pins at the corners of the assembly.     

快堆组件稠密棒束数值模拟

为详细研究快堆组件稠密棒束中的冷却剂流动方式,本工作采用Fluent程序对169棒束快堆燃料组件进行了三维数值模拟,并与已公开发表的文献结果进行了对比。由计算结果可知：计算得到的摩擦系数结果在Re为35885～61354时与试验结果符合较好;从中心到外围,横向流和轴向流在不同的方向和位置呈现出不同的流动特性。根据模拟结果可更准确地预测棒束通道内的流动情况,可为今后稠密棒束组件水力学设计和子通道内流量测量试验提供参考。     

高能X射线CT技术在辐照后核燃料组件检测中的发展及应用

定期检测辐照后核燃料组件对保障反应堆安全运行和开展高燃耗下核燃料组件的性能研究具有重要意义。为了能在不拆卸、不破坏燃料组件的情况下更好地观察燃料组件及其内部燃料棒的缺陷及结构变化等信息,高能X射线计算机断层扫描(X射线CT)技术作为一种有效手段可用于辐照后核燃料组件的检测。日本多年来一直致力于该技术的研究工作,成为世界上唯一一个研制出用于辐照后燃料组件检测的高能、高分辨率X射线CT检测装置且应用于快中子反应堆现场检测的国家。为此,本文梳理日本近几十年来相关研究成果,介绍日本原子能研究开发机构(JAEA)研发的燃料组件高能X射线CT装置结构、工作原理、研究现状及部分应用实例,以期对我国核燃料组件无损检测技术的发展提供参考、借鉴。     

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis.     

超高通量快中子试验堆堆芯初步概念设计

针对先进核能系统发展需要,提出了超高通量堆的堆芯概念设计。本文采用板型燃料、正方形燃料组件设计,设置宽流道保证堆芯冷却剂占有较高的体积份额。堆芯采用52盒燃料组件,设置8盒控制棒组件和较厚的反射层。通过堆芯概念设计方案评价,结果表明堆芯循环长度可达100EFPD(等效满功率天),所提出的超高通量堆的最大中子注量率可达到1.08×10¹⁶ cm^-2·s^-1。     

Optimization of the number of spacers in a nuclear fuel bundle with respect to flow-induced vibration

The design of nuclear fuel, which is mostly subjected to loads due to high temperature, pressure and flow under radiation environment, is of immense importance in nuclear engineering. Generally, the nuclear fuel consists of number of fuel pins, which carry the fuel pellets inside. A number of such pins are grouped together to a specific size to form a bundle. Flow-induced vibration in the bundle due to high velocity coolant flow cause interaction between the fuel pins. Commonly, spacers are provided between the fuel pins to maintain the gap between the neighboring pins. Large number of such spacers in the reactor core is considered to be a load on thermal-hydraulics and on physics of reactor control. This paper addresses the work done with an objective to help optimization of the number of spacers required on a typical fuel bundle by experimental and analytical study.