CZT探测器测量核燃料组件铀丰度实验研究

采用CZT探测器、数字谱仪、准直器等组成了1套便携式CZT探测器铀丰度测量装置。该装置可对燃料组件铀丰度进行测定,以便确定相应铀产品丰度符合规定要求。实验研究中,对几类燃料组件丰度进了测量,建立了CZT探测器测量燃料组件铀丰度的方法。现场测量结果表明,铀丰度测量结果相对偏差小于3%,方法简单可靠,装置简便,能满足核材料保障监督和核设施现场测量中的需求。     

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。     

核燃料组件高能X射线无损探测技术研究

正燃料组件是反应堆堆芯的关键部件,其性能质量直接关系到反应堆的安全性、经济性和先进性,核燃料组件在反应堆运行期间会在裂变产物、高温、中子辐照等因素的综合作用下发生开裂、肿胀,严重的情况下会出现破碎,这些将对反应堆的安全运行造成威胁,因此对核燃料组件的定期检测尤为重要。辐照后核燃料组件存在很强的放射性,对其进行检测存在一定的技术难度,通过国外的经验及相关研究,利用高能X射线直线加速器射线源进行核燃料组件的无损探测是一种有效的检测手段。中国原子能科学研究院"十二五"期间在核能     

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。     

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。     

放射性废物桶γ检测探测器准直器模拟设计

介绍了在放射性废物桶γ检测中基于MC模拟的阵列探测器准直器的设计。利用MCNP程序建立探测系统模型,计算了在圆形、方形、菱形准直孔,以及铅、钨、不锈钢准直材料条件下,碲锌镉探测器对体源和点源的探测效率,分析了准直器形状和材料对探测效率的影响。结果表明:相比方形和菱形准直孔,圆形准直孔的探测效率相对较高;相比铅和钨,不锈钢准直时出现多路射线干扰,且铅和钨的准直效果相当。     

WWER-1000燃料组件特点及棒弯曲分析

本文根据WWER-1000反应堆的设计特点及其运行实践,阐述了WWER-1000燃料组件的设计特点,并与西方压水堆燃料组件进行了相应的比较.重点分析论述了WWER-1000反应堆燃料棒弯曲的特点,以及在热工水力和燃料组件设计中是如何考虑棒弯曲效应的,进行了燃料棒弯曲对临界热流密度影响实验的研究.结果表明:WWER-1000燃料组件在整个运行寿期内的性能是可以保证的.     

芯块年龄对燃料棒富集度检测的影响研究

为保证核电站运行安全,反应堆装入核燃料组件的²³⁵U富集度具有严格的设计要求。因此,反应堆燃料棒均必须经过100%富集度及装料均匀性无损检测。已有富集度检测包括有源法和无源法,通过对有源法、无源法两种富集度检测方法中UO₂芯块年龄现象的研究,发现有源检测法在较低活度下,受UO₂芯块年龄影响,使得燃料棒富集度无法正确检测,需采用被动放置等待的方式解决芯块年龄问题。无源法可在检测中直接校正芯块年龄,满足生产检测要求,已得到工程化应用。     

水堆燃料元件的破损监测、定位和处理

本文简介反应堆破损燃料元件的监测、定位和处理；反应堆运行时的监测与定位；换料时或换料后的监测；在燃料组件内鉴别破损的燃料棒；破损燃料棒拆出后的监测与定位；燃料组件的监测、拆卸和修复等方面在国际上的研究开发现状。     

Detection System for the Fuel Assembly Nondestructive Testing

Fuel assemblies are the central components of a reactor. The core fuel pellets in the fuel pins will swell and deform and the fuel cladding may even break under the complex environment of high temperature, high pressure and intense neutron radiation field, which threats the safety of the reactor. To better understand the changes in the behavior of the fuel assembly in the reactor and study the central void formations and deformations of fuel pins in fuel assemblies to high burn-up, high-energy X-ray non-destructive testing is an effective technical means. Irradiated nuclear fuel assembly has a strong radioactivity, it is necessary to optimize the design of the detector system and the collimator to reduce the effect from gamma rays emitted from the irradiated fuel assembly during detection system designing phase. Through modeling, estimating and optimization, the optimal size of the detector unit is obtained and the collimator design is optimized which can lay the foundation to improve the quality of the reconstructed images of the fuel assembly nondestructive system.     

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。