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。     

板型燃料元件内部缺陷的无损检测

本文使用超声波水浸单探头反射法对不贴紧缺陷进行了检测,根据缺陷反射波的形状及在时域上的位置对缺陷的深度和类型进行了分析评价。使用²⁴¹Am源59.5 keV的单能窄缝γ射线,对芯体分布的均匀性进行了检测。根据特定检测频率下包壳厚度与涡流阻抗变化的对应关系对包壳厚度进行了分析评价。通过超声、射线和涡流检测方法对板型燃料元件内部缺陷检测并给予评价,保证产品质量。     

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。     

用于高能X射线成像的CdWO4闪烁探测器探测灵敏度的研究

高能X射线成像系统对小截面探测器的探测灵敏度提出了很高的要求，分析了由CdWO4晶体耦合光电二极管所组成的探测器单元探测灵敏度的决定因素，采用蒙－卡计算、经验估算等对其在高能加速器和^60Co源下的探测灵敏度进行了估算，并和测量值进行比较。二者吻合较好，证明了估算方法的正确性。     

Long-Cycle and High-Burnup Fuel Assembly for the VHTR

For a prismatic VHTR fuel assembly, a physics study has been performed to maximize the fuel performance in terms of the cycle length and the discharge burnup for a given fuel enrichment. The relationship between the fuel performance and the fuel configurations has been investigated in terms of the TRISO packing fraction, diameter of the fuel kernel, fuel management, and moderating power of the fuel block. Both a typical low-enrichment uranium fuel (LEU) and a fuel made of transuranics (TRU) from LWR spent fuel are considered in this paper. It is shown that in order to obtain a long refueling cycle and a high burnup at the same time, the fuel loading needs to be increased together with the moderating power of the fuel block. Three ways are considered for a higher moderation of the fuel block: a larger pitch of the coolant hole pattern, an extra graphite thickness in the fuel block, and a higher graphite density. The impact of the increased pitch on the fuel temperature is also evaluated with a thermal analysis code. We have shown that long refueling cycles and high burnups can be achieved simultaneously for both LEU and TRU fuels.     

“核-光转换”中子探测系统物理结构的模拟研究

"核-光转换"中子探测器是以惰性气体为介质将裂变碎片能量转换为光辐射的裂变室,拥有电离探测器所没有的优点:不需要供电电源;信号传输方式采用光导或光纤,而不是绝缘电缆;对伽马辐射极不灵敏;输出信号较大,可以避免在探测器附近使用前置放大器。根据"核-光转换"中子测量系统的特点,采用Geant4模拟了铀裂变靶厚度、惰性气体成分、腔体材料等对到达惰性气体的裂变碎片和可见光的影响,给出了NOC结构设计的最佳参数和中子能量响应。     

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.     

5×1024×10高速核辐射能谱获取系统研制

在磁约束聚变实验装置的物理实验中，为了测量高温等离子体箱射的某特定能量区域的X射线或其他核箱射能谱的时、空分布，需要在托卡马克的板向和环向方向布置多台X射线探测器，每次放电可同时获得数十个能谱，为了不牺牲能谱中的重要信息(如线辐射)．以及能量分辨的需要，要求每个能谱有1024道，每个谱采集的时间分为：40、80、160、240ms，可自由设定，每道容量16位，每次放电可同时获得50个能谱，可扩展到80个能谱。     

钍基重水核能系统燃料的物理特性研究

对全铀CANFLEX燃料和含钍CANFLEX燃料的物理特性进行了研究。用WIMS-AECL程序计算了参考栅元的冷却剂空泡反应性、燃耗等参数。通过与铀燃料的对比，展示了含钍燃料在安全性和经济性上的特点还特别介绍WIMS-AECL程序在使用过程中的参数选择方法，     

~(14)C-碳酸钠测量方法的改进

本文介绍用液体闪烁技术测量Na_2~(14)CO_3或NaH~(14)CO_3的一种改进方法。在一定量的蒸馏水中加入适量的Na_2CO_3载体和NaOH,再加一定量的待测Na_2~(14)CO_3溶液,制成供测量用的放射性浓度适中的Na_2~(14)CO_3贮备液。在甲苯/Triton X-100闪烁系统(PPO4g,POPOP 0.25g,Triton X-100 165ml,加甲苯至500ml)中加入DL-α-苯乙胺,最后加入Na_2~(14)CO_3贮备液,制备样品。用样品道比法测定效率。每日测量一次,在连续测量的9天中,未发现~(14)C损失,也未发现存在化学发光,探测效率约为83％。     

Development of a Small-Aperture Slit System for a High Collimator Ratio at the Thermal Neutron Radiography Facility in JRR-3

A small-aperture slit system has been developed and installed to enhance the collimator ratio (“L/D”) of the thermal neutron radiography facility (TNRF) in JRR-3. The degree of unsharpness on neutron images is reduced by increasing the L/D. The small-aperture slit system increased the L/D by creating a small aperture size (“D”). Image sharpness improved when the aperture size was reduced to below 10mm by 10mm in the TNRF. On the other hand, there was almost no difference in unsharpness on images obtained above 10mm by 10mm in aperture size. These results indicate that an aperture size of less than 10mm by 10mm should be used for high-spatial-resolution imaging at the TNRF. The beam area of the small-aperture slits was relatively small in comparison with that of a conventional large collimator, though gradually increasing with increasing aperture size. Even with an aperture size of 5mm by 5 mm, the practical beam area for imaging examinations corresponded to around 25mm by 20 mm, which is enough area to carry out high-spatial-resolution imaging.     

Characteristics of a Geological Disposal System for the Increasing Burn-up of Spent Nuclear Fuel in Korea

The characteristics of a geological disposal system that can accommodate increasingly higher burn-up levels of spent fuel were assessed based on the Korea reference disposal system concept. First, a status investigation that included a projection of spent fuel quantity versus burn-up was carried out to demonstrate the trend toward higher burn-up levels. Next, the main features of the Korea reference disposal system were introduced. Finally, the disposal tunnel length, excavation volume, and raw materials (e.g., a cast insert, copper, bentonite and backfill) necessary for a disposal system were comprehensively analyzed to define the characteristics and overall effects on geological disposal at increasingly higher burn-up levels. Our study determined that it is reasonable to use a canister containing 4 spent fuel assemblies with burn-up levels up to 50GWD/MTU, while a canister containing 3 spent fuel assemblies can accommodate burn-up levels beyond 50GWD/MTU. A remarkable increase of 33% in disposal tunnel length and that of 30% in excavation volume were observed as the burn-up increased from 50 to 60GWD/MTU. However, this was offset by a reduction of 17% in raw materials used in canister fabrication. Therefore, it seems that spent fuel at increasingly higher burn-up levels is not a serious concern for deep geological disposal in Korea.     

A new value for the half-life of Be by Heavy-Ion Elastic Recoil Detection and liquid scintillation counting

The importance of ¹⁰Be in different applications of accelerator mass spectrometry (AMS) is well-known. In this context the half-life of ¹⁰Be has a crucial impact, and an accurate and precise determination of the half-life is a prerequisite for many of the applications of ¹⁰Be in cosmic-ray and earth science research. Recently, the value of the ¹⁰Be half-life has been the centre of much debate. In order to overcome uncertainties inherent in previous determinations, we introduced a new method of high accuracy and precision. An aliquot of our highly enriched ¹⁰Be master solution was serially diluted with increasing well-known masses of ⁹Be. We then determined the initial ¹⁰Be concentration by least square fit to the series of measurements of the resultant ¹⁰Be/⁹Be ratio. In order to minimize uncertainties because of mass bias which plague other low-energy mass spectrometric methods, we used for the first time Heavy-Ion Elastic Recoil Detection (HI-ERD) for the determination of the ¹⁰Be/⁹Be isotopic ratios, a technique which does not suffer from difficult to control mass fractionation. The specific activity of the master solution was measured by means of accurate liquid scintillation counting (LSC). The resultant combination of the ¹⁰Be concentration and activity yields a ¹⁰Be half-life of T_1/2 = 1.388 ± 0.018 (1 s, 1.30%) Ma. In a parallel but independent study (Chmeleff et al. [11]), found a value of 1.386 ± 0.016 (1.15%) Ma. Our recommended weighted mean and mean standard error for the new value for ¹⁰Be half-life based on these two independent measurements is 1.387 ± 0.012 (0.87%) Ma.     

双层屏蔽厚度γ反解技术研究

分析高分辨(能谱获取测量对象特定信息在军控核查领域有重要应用,根据多组不同能峰强度比等信息反解给出源外屏蔽材料几何信息有望有效消除未知包装容器对容器内部材料特性认识的影响。分析建立了反解源外双层屏蔽厚度的方法,实验测量反解了钚源和~(152)Eu点源外双层屏蔽材料的厚度,~(152)Eu点源的解析结果明显优于钚源,分析了实验解析结果与实际厚度值之间存在一定偏差的原因。     

核能制氢技术的发展

氢是清洁能源,有非常好的应用前景.但氢是二次能源,需要利用一次能源来生产.以可持续的方式(原料来源丰富、无温室气体排放)实现氢的大规模生产是实现氢广泛利用的前提.核能是清洁的一次能源,核电已经成为世界电力生产的主要方式之一.正在研发的第四代核能系统除了要使核电生产更经济和更安全之外,还要为实现核能在发电之外的领域的应用...     

Precise Measurement of the Gamma-ray Emission Probabilities for 116m In with Half-life of 54.15 min

The γ-ray emission probabilities for ^116mIn with half-life of 54.15 min have been measured by a 4πβ(pc)-γ(HPGe) coincidence apparatus using a live-timed two-dimensional data acquisition system. Eleven γ-ray peaks of ^116mIn were recognised and the uncertainties of the measured emission probabilities for principal γ-rays were found to be less than 1%. On the other hand, the relative intensities for 18 energy γ-rays were determined with improved uncertainties by an ordinary γ-ray spectrometer. The decay parameters were determined with uncertainties of approximately 1%using these values and the evaluated values of weak γ-ray intensities and the internal conversion coefficients.     

An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the ‘ASME PRA standard’ also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's action time by using the MARS platform. The typical example case is a small break loss of coolant accident without the high pressure safety injection system, which is one of the most important accident sequences in the PSA of the KSNP. Under the given accident sequence, the operator has to perform a recovery action known as a fast cooldown operation. This study focuses on two aspects regarding an operator's action; one is how they can operate it under some restrictions; the other is how much time is available to mitigate this accident sequence. To assess these aspects, this study considered: (1) the operator's action model and (2) the starting time of the operation. To show an effect due to an operator's action, three kinds of control models (the best-fitting, the conservative, and the proportional-integral) have been assessed. This study shows that the developed method and the platform are useful tools for this type of problem and they can provide a valuable insight related to an operator's actions.     

Capabilities of TRANSURANUS code in simulating power ramp tests from the IFPE database

Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. OECD NEA sets up the “International Fuel Performance Experiments (IFPE) database”, a public domain database on nuclear fuel performance experiments with the purpose of model development and code validation. The objective of the activity (performed in the framework of the IAEA CRP FUMEX-III project) is to investigate the pellet-clad interaction mechanism and the capability of TRANSURANUS code in simulating the phenomena, processes occurring in the fuel rod during the power ramps, with focus on the parameters influencing the cladding failures. The experimental database adopted is the Studsvik PWR Super-Ramp subprogram, part of the IFPE database, which consists of 28 pressurized water reactor fuel rods power ramped at burnup from 28 to 45 MWd/kgU. Relevant results by TRANSURANUS are presented in connection with the experimental evidences. Focus is given on the PCI/SCC failure, demonstrating that the failure threshold, available in TRANSURANUS, results conservative both in case of KWU and W rods.