Single-loop nuclear power plant with a pressurized-water reactor

Trust Tsentroénergomontazh. Translated from Atomnaya Énergiya, Vol. 71, No. 5, pp. 458–460, November, 1991.     

Knowledge-directed characterization of nuclear power plant reactor core parameters

A major life-limiting factor of the UK's Advanced Gas-Cooled Reactors (AGRs) is the condition of the graphite core. Installation of new measurement equipment is difficult and expensive, therefore maximizing the information gained from existing equipment is highly desirable. The main approach to determining the health of an AGR core is through periodic inspections undertaken during planned outages. However, there is the desire to supplement this inspection activity through the analysis of data gathered as part of routine plant operation. One such source of data is measurements taken during refueling and this paper describes knowledge-directed characterization of this refueling data, both spatially across the reactor core and temporally across the operational lifetime of the core. Characterization provides information relating to the current condition of the reactor core and allows suspected ageing trends to be visualized and confirmed. A standard approach for characterizing reactor core data is presented and applied to a variety of different reactor core parameters. The benefit of this approach is that it allows engineers to distill large volumes of refueling data into a readily understandable format in a short period of time. It also allows hypothesized trends relating to the ageing process within the core to be tested and provides supporting evidence for these hypotheses. The trending data is also valuable as it can form the basis of a predictive model of ageing of the reactor core. The ageing process of nuclear graphite is understood from theoretical and experimental viewpoints and this empirical data, gathered from operating reactors, further supports this understanding. This paper represents the initial exploration of using refueling data to construct a predictive model of AGR reactor core ageing.     

Decontamination of the reactor coolant pump in Maanshan nuclear power plant

To reduce the radiation dose that accumulated on the reactor coolant pump, decontamination work was carried out at the Maanshan Nuclear Power Plant. A four-step alkaline permanganate (AP)-CanDecon process was applied to remove the activity on the turning vane diffuser and pump impeller. The first step consisted of 8 h of AP treatment and 7 h of decontamination. It was followed by 2.5 h of AP treatment and 5 h of decontamination. An average decontamination factor of 2.9 was obtained. To understand the corrosion of the decontaminating reagents on the materials, coupons were installed in the decontamination tank. These were as-received and sensitized 304SS, alloy 600, casting stainless steel (CF-8), stellite-6, and carbon steels (A508 and A533). The exposure rates (mR h⁻¹) of the carbon steels were approximately five times higher in magnitude than those of the other materials. The decontamination levels (dpm per 100 cm²) of the A508 and A533 carbon steels were 5432 and 3701 respectively, while most of the rest of the materials were below the low limit of detection. Apparently, the corrosion product on the materials was a major factor in sustaining the exposure rate and the contamination level. The corrosion rate of the materials was also examined and compared with published data. An examination of the surface morphology of the materials after decontamination showed intergranular attack on the 304 stainless steel.     

Planetary nuclear power plant based on a thermionic converter reactor

The concept for a sealed planetary nuclear power plant, based on a high-temperature version of a thermionic nuclear power system, is examined. A high-temperature radiator makes it possible for the system to operate at a high power level. A sealed chamber gives the optimal matching of the construction with the operating conditions on the surface of the Moon and planets in the solar system, 1 figure. S. P. Korolev Rocket-Space Corporation énergiya. Translated from Atomnaya énergiya, Vol. 89, No. 1, pp. 20–22, July, 2000.     

一种低温堆核电站的动态数学模型

针对研发的采用一体化布置、全功率自然循环的低温核反应堆电站,建立了一个可用于大功率运行范围控制系统仿真的动态数学模型.模型采用了六组缓发中子动态方程(考虑了慢化剂温度和燃料温度反应性负反馈)、集中参数的堆芯传热模型以及自然循环流动模型,重点考虑了主回路自然循环对堆芯内冷却剂和燃料棒之间的传热系数、主换热器换热系数、主回路时间常数的影响.仿真结果表明,模型能够正确反映低温堆核电站的主要动态特性,可用于电站控制系统仿真.     

Self-contained co-generation nuclear power plant with integral nuclear reactor designed for remote and difficult-to-access areas

The paper gives an overview of the conceptual design of a reactor with enhanced safety intended for small NPPs to provide power supply for difficult-to-access and remote areas. The basic design features and the configuration of an integral nuclear reactor and the plant as a whole, as well as the main technical data are presented.     

AP1000核电厂反应堆冷却剂系统泄漏分析

第三代AP1000非能动核电厂的主要特征是采用非能动安全原理,使核电厂的系统、设备、构筑物大幅度简化,安全性、可靠性、经济性大幅度提高,以满足美国先进轻水堆业主要求文件的基本要求。本文针对美国业主要求文件(Utility Requirements Document,URD)第三卷第五章《专设安全系统》中对非能动先进轻水堆核电厂反应堆冷却剂系统压力控制功能的要求:在很小的反应堆冷却剂系统(Reactor Coolant System,RCS)净泄漏率(不大于2.27 m3·h-1)条件下,具有足够的系统冷却剂装量及补水能力,以保证在8 h(28 800 s)内不会触发自动降压系统而进行计算分析,本分析采用安全分析报告小破口失水事故(Loss of coolant accident,LOCA)分析采用的NOTRUMP程序,分析结果表明AP1000核电厂可满足上述美国URD要求。