Studies on Production Planning of Dispersion Type U3Si2-Al Fuel in Plate-Type Fuel Elements for Nuclear Research Reactors

Several fuel plants that supply nuclear research reactors need to increase their production capacity in order to meet the growing demand for this kind of nuclear fuel. After the enlargement of the production capacity of such plants, there will be the need of managing the new production level. That level is usually the industrial one, which poses challenges to the managerial staff. Such challenges come from the fact that several of those plants operate today on a laboratorial basis and do not carry inventory. The change to the industrial production pace asks for new actions regarding planning and control. The production process based on the hydrolysis of UF6 is not a frequent production route for nuclear fuel. Production planning and control of the industrial level of fuel production on that production route is a new field of studies. The approach of the paper consists in the creation of a mathematical linear model for minimization of costs. We also carried out a sensitivity analysis of the model. The results help in minimizing costs in different production schemes and show the need of inventory. The mathematical model is dynamic, so that it issues better results if performed monthly. The management team will therefore have a clearer view of the costs and of the new, necessary production and inventory levels.

Studies on Capacity Expansion of Fuel Plants for Nuclear Research Reactors

The demand for nuclear fuel for research reactors is rising worldwide. Thus, the production facilities of this kind of fuel need reliable guidance on how to augment their production in order to meet the increasing demand efficiently and safely. We proposed a specific procedure for increasing production capacity. That procedure was tested with data from a real plant, which produces plate-type fuel elements loaded with LEU U3Si2-Al fuel. The test was made by means of discrete event simulation, and the results indicated the proposed procedure is efficient in raising production capacity.

Neutronic design investigation of a liquid injection-based second shutdown system for a typical research reactor using MCNPX

Safety systems, built on state-of-the-art technology, are essential for achieving acceptable levels of plant safety to minimize hazards to the reactor and the general public. The second shutdown system(SSS) as an engineered safety feature and a part of the reactor protection system(RPS) is a means for rapidly shutting down a nuclear reactor, keeping it in a subcritical state and serving as a backup to the first shutdown system(FSS). In this research, one SSS with two types of optimum chamber designs is proposed that take into account the main current characteristic features of the Tehran research reactor with improvements over earlier designs. They are based on a liquid neutron absorber injection that is preferably different, diverse, and independent from the FSS based on the rod drop mechanism. The major design characteristics of this SSS with two different chambers were investigated using MCNPX 2.6.0 code. The performed calculations showed that the designed SSS is a reliable shutdown system, assuring an appropriate shutdown margin and injection time, with no significant effects on the effective delayed neutron fraction while causing minimal variations to the core structure. Further, the reasonable financial cost and the prolongation of the operation cycle are additional advantages of this design.

Thermal Hydraulic Analysis Improvement for the IEA-R1 Research Reactor and Fuel Assembly Design Modification

This paper presents the sequence of activities to improve the thermal hydraulic analysis of the IEA-R1 research reactor to operate in safe conditions after power upgrade from 2 to 5 MW and core size reduction from 30 to 24 fuel assemblies. A realistic analysis needs the knowledge of the actual operation conditions (heat flow, flow rates) beyond the geometric data and the uncertainties associated with manufacturing and measures. A dummy fuel assembly was designed and constructed to measure the actual flow rate through the core fuel assemblies and its pressure drop. First results showed that the flow distribution over the core is nearly uniform. Nevertheless, the values are below than the calculated ones and the core bypass flow rate is greater than those estimated previously. Based on this, several activities were performed to identify and reduce the bypass flow, such as reduction of the flow rate through the sample irradiators, closing some unnecessary secondary holes on the matrix plate, improvement in the primary flow rate system and better fit of the core components on the matrix plate. A sub-aquatic visual system was used as an important tool to detect some bypass flow path. After these modifications, the fuel assemblies flow rate increased about 13%. Additional tests using the dummy fuel assembly were carried out to measure the internal flow distribution among the rectangular channels. The results showed that the flow rate through the outer channels is 10% - 15% lower than the internal ones. The flow rate in the channel formed between two adjacent fuel assemblies is an estimated parameter and it is difficult to measure because this is an open channel. A new thermal hydraulic analysis of the outermost plates of the fuel assemblies takes into account all this information. Then, a fuel design modification was proposed with the reduction of 50% in the uranium quantity in the outermost fuel plates. In order to avoid the oxidation of the outermost plates by high temperature, low flow rate, a reduction of 50% in the uranium density in the same ones was shown to be adequate to solve the problem.

Current status and technology development tendency of research reactors in China

The current status and development history of domestic and abroad research reactors (RRs) are mentioned.The representative RRs and their respective technology characteristics are introduced.The utilizations of China's RRs,mainly included as nuclear engineering technology,basic research applications of nuclear technology,teaching and personnel training,are explained.

核反应堆用钼铼合金结构材料研究进展

难熔金属钼具有熔点高、高温力学性能优异、导热性良好等特点,加之其良好的抗辐照肿胀能力及与液态金属的相容性,使其成为第四代高温核裂变反应堆、聚变堆等先进核反应堆重要的候选材料,用以满足高温、强腐蚀、大剂量辐照等苛刻环境下结构件的制备需求。但金属钼具有本征室温脆性、加工难和焊接性能差等缺点,严重限制了其应用推广。在金属钼中加入铼元素,形成“铼效应”,不仅可以显著改善钼的室温塑性和加工性能,降低塑-脆转变温度,而且还能提升材料焊接性能和抗蠕变性能,已经成为先进核反应堆结构材料的研究热点。本文从钼铼合金的成分设计、材料制备、焊接性能及核环境应用评价研究四个方面总结了国内外近年来的研究进展,分析了钼铼合金在先进反应堆工程应用中存在的问题,以期为高性能钼铼合金结构材料的开发提供参考。

西安脉冲反应堆

西安脉冲堆是我国设计、建造的第一座实用性多功能脉冲反应堆。它具有固有安全性高、用途广泛、结构简单及运行维护方便的特点;既能稳态运行,又能以脉冲或方波方式运行,稳态额定功率2MW,最大脉冲峰功率4200MW。西安脉冲堆设置有多种实验辐照装置,可以辐照生产放射性同位素,进行中子活化分析、中子照相、单晶硅中子辐照掺杂、材料辐照加工及辐照试验研究,开展核物理、中子物理、放射化学等科学理论研究以及人材培训。近2年的试运行和实验应用表明,西安脉冲堆已经展现出良好的应用特性和广阔的应用前景。

轻水堆两相流机理研究的若干关键问题、现状与发展前沿

两相流是轻水堆热工水力的重要研究主题。正常工况和事故工况下的两相流流动和传热特性关乎核反应堆系统的安全性和经济性。在流动过程中相界面结构不断发生演化,同时两相之间存在着复杂的动量和质量、能量传递,这使得两相流动成为最复杂的流动现象之一。本文回顾了反应堆两相流的发展历程,重申了领域内的若干关键问题,总结各个问题的研究现状,并展望领域研究前景、提出学科发展建议。

我国高放废物地质处置研究

文章提出我国高放废物地质处置拟采用处置库选址和场址评价一特定场址地下实验室一处置库“三步曲”式技术路线。计划目标是于 2 0 3 0～ 2 0 40年前后建成我国的高放废物地质处置库。处置对象是玻璃固化块、超铀废物和部分乏燃料 ,处置库为竖井一坑道型 ,候选围岩为花岗岩 ,位于饱和带中。已初步选定甘肃北山地区为重点预选区。该区地处戈壁 ,地壳稳定 ,人烟稀少 ,地质条件和水文地质条件有利。现已试验获取预选区大量深部地质环境参数。确定使用膨润土作为处置库的回填材料 ,已获得一批放射性核素在花岗岩和膨润土中的吸附、扩散数据 ,建立了模拟处置库温度、压力和氧化还原条件的实验装置。高放废物地质处置场址评价、放射性核素地球化学行为、回填材料研究和环境评价研究正在深入进行 。

与核供热堆耦合的海水淡化系统及蒸发工艺研究

在清华大学核能技术设计研究院建立的大型双塔布置4效竖直蒸发管高温多效海水淡化(VTE-MED)实验装置上,对多塔VTE-MED系统的工艺流程和多孔孔板阻力系数等关键技术进行了实验研究。该系统最高设计咸水温度120℃,日产淡水约4.3吨。实验结果表明:对多塔海水淡化系统的设计,应注意塔间流量匹配,输送流体管道保温等问题;针对系统启动,应考虑设置启动水箱等。此外,根据阻力系数实验结果对多孔孔板提出了修改方案。

模块式小型反应堆研发现状及前景分析

目前的核电站大多数为装机容量较大的反应堆,有些甚至达到了175万千瓦的单机容量,大型堆核电站一次性投入成本高、建造周期长,且难以适应小型电网的需求,而小型反应堆恰好可以解决这一问题,全球掀起了模块式小型核电机组的开发热潮。文章主要介绍了模块式小型反应堆的研发现状、用途,并分析了小型堆与常规堆的优劣势及其发展前景。

核反应堆中防护包壳的研究进展

综述了核反应堆中锆合金防护型包壳的研究进展。石墨或硅氧烷涂层包壳以及内衬锆或铜阻挡型包壳的广泛采用有效地阻止了PCI的产生 ,硅氧烷涂层的防护效果较石墨大 ,但制作这种包壳的成本较高 ;铜阻挡层防护作用较锆阻挡层低 ,已很少被采用。在这几种防护方法中 ,无论采用哪种方法都有效地阻止了PCI的发生 ,但有关防护机制和涂层及阻挡层锆合金包壳制作方法的研究还很少。为此 ,从涂层对裂变产物碘的影响和涂层及内衬金属与锆管内表面的结合等方面提出了今后的研究方向。

CARR控制棒物理设计计算

中国先进研究堆(CARR)是一座高性能、多用途新型研究堆。CARR的控制棒物理计算是一个难点,本文采用Monte Carlo方法计算CARR的控制棒物理特性。物理计算的结果直接指导控制棒的结构设计和加工。

研究堆辐射监测系统故障检修方法

介绍了包括16Nγ监测、元件破损监测、主冷却器破损监测、工艺间γ监测、低放惰性气体β监测在内的辐射监测系统的几种探测模式及其原理。通过分析探测器故障和处理组件故障这两类故障的原因,发现辐射监测系统在运行中故障的产生分布有一定的规律,总结出了典型的故障检修方法,为同类辐射监测系统的检修维护以及设计提供借鉴。

CARR冷中子源物理方案研究

冷中子源是中国先进研究堆(CARR)作为应用平台开展中子散射实验的重要系统。本文建立了可开展冷中子源装置氢系统物理方案研究的计算程序和方法。对影响核发热和冷中子增益的多种因素进行了计算和优化选择,确定了氢系统物理方案。分析结果表明:符合CARR冷中子源自身特点且在国际上具有创新意义的月牙形冷包结构,在核发热量较小的条件下获得了较好的冷中子增益。

某研究堆一回路冷却剂泄漏事件的审评

研究堆与核动力堆的使用目的不同,决定了其在系统设计和固有安全性方面有较大的不同。针对国内某研究堆一回路冷却剂泄漏事件的审评,根据该堆的设计特点,阐述了审评者的关注问题及技术观点,并分析了研究堆与核动力堆在事件审评方面的差异。

空间核反应堆电源热工水力特性研究综述

随着空间探索领域的快速发展,研究高功率、安全、可靠的空间核反应堆电源将变得愈发重要。本文针对国内外空间核反应堆电源的热工水力关键问题,即空间堆系统稳态和事故瞬态研究、堆芯单冷却剂通道及全堆芯的三维流动换热、静态与动态热电转换装置分析、热工水力特性试验研究等进行研究,分析了空间核反应堆电源热工水力研究的趋势。本文结果可为空间核反应堆电源设计分析及热工水力安全特性研究提供帮助和指导。

反应堆压力容器主螺栓孔抛光机改进研究

针对现有反应堆压力容器主螺栓孔抛光机静态就位困难、旋转电机易烧毁的问题,对刷具组件的结构进行分析,提出一种增设橡胶弹簧的改进方案,解决了抛光机静态就位困难的问题;对旋转电机的散热问题进行分析,利用强迫对流方案,对比分析几种不同进气孔的设计,发现形成较优流场的规律,根据规律制定了优化方案,该方案可快速带走旋转电机的热量,避免其过热烧毁。经优化改进的反应堆压力容器主螺栓孔抛光机已成功应用于工程实践。