A reactor system for radiation research and materials testing

Translated from Atomnaya Énergiya, Vol. 68, No. 6, pp. 438–441, June, 1990.     

Reliability assessment of emergency exhaust system in a pool-type research reactor

The reliability of an extract system in a swimming-pool-type research reactor has been assessed. A global fault-tree analysis technique has been utilized. The basic event reliability data is based on both generic and reactor specific informations.The unavailability of the extract system is quantified in terms of the unavailability of the various functional requirements of the system. The unavailability is expressed as the probability of failure on demand. The computer system unavailability is determined from the minimal cutsets of the system. It is found that only three events have a major contribution to the top event, i.e., failures of compressed air supply, electric power supply and solenoid valve. A sensitivity analysis is performed to show the effects of variations in the data values of the dominant cutsets. An uncertainty analysis was also performed on the fault tree. The evaluations show that the reactor extract system lacks diversity and redundancy in most of its components. It is tolerant of most minor degradations, as these are taken care of by the operating policies and procedures. However, it can not tolerate common cause failures, e.g., simultaneous compressed air and electric power supply failure. Based upon the results obtained some recommendations are made.     

The design and experimental validation of an emergency core cooling system for a pool type research reactor

This paper presents the design of the Emergency Core Cooling System (ECCS) for the IEA-R1m pool type research reactor. This system with passive features, uses sprays installed above the core. The experimental program performed to define system parameters and to demonstrate to the licensing authorities, that the fuel elements limiting temperature is not exceeded, is also presented. Flow distribution experiments using a core mock-up in full-scale were performed to define the spray header geometry and spray nozzles specifications as well as the system total flow rate. Another set of experiments using electrically heated plates simulating heat fluxes corresponding to the decay heat curve after full power operation at 5 MW was conducted to measure the temperature distribution at the most critical position. The observed water flow pattern through the plates has a very peculiar behavior resulting in a temperature distribution which was modelled by a 2D energy equation numerical solution. In all tested conditions the measured temperatures were shown to be below the limiting value.     

Second trip system for NRU research reactor

For the past four decades, the NRU research reactor has played an important role at the Chalk River Laboratories, Atomic Energy of Canada Limited, serving as one of its major research and isotope production facilities. To ensure that it continues as an effective facility, compliant with the current safety standards, a comprehensive upgrade program is underway. Adding a second trip system (STS) is part of this upgrade program, aiming at improving the effectiveness and reliability of the overall shutdown function. This document describes the main features and basic principles of the STS.The STS is an independent, seismically qualified trip system, that guarantees reactor shutdown even if the existing trip system fails. It is designed based on 2 out of 3 general coincidence logic, with minimal interferences and changes to the existing system. In addition to the manual trip in the main control room, a remote manual trip is provided in the new Qualified Emergency Response Centre, which is also seismically qualified and always accessible. Thus, for any reason, if the main control room becomes uninhabitable, the reactor still can be manually shut down from this centre.     

反应堆中信号系统

介绍了反应堆中信号系统的设计思想、工作原理、系统构成以及在反应堆上的推广应用。     

The physics of rewetting in water reactor emergency core cooling

Surface rewetting is essential for the re-establishment of normal and safe temperature levels following dryout in rod clusters or boiler tubes, or following postulated loss-of-coolant accidents in water reactors. Rewetting experiments have been performed with tubes and rods with a wide range of materials and experimental conditions (surface temperatures 300–800°C, constant water flows 0.1–30 g s⁻¹). The physical processes involved in the rewetting of high temperature surfaces have been shown to be identical for both falling water films and bottom flooding. The variation of rewetting velocity with mass flow has been determined, and shown to be independent of hydraulic diameter over the range 0.2–6 mm of practical interest. Data have also been obtained on the mass ‘carryover’ fraction. Theoretical solutions for the rewetting velocities have been obtained by analysis of thermal conduction in the surface. At low mass flows, effectively one-dimensional (axial) conduction cools the surface ahead of the rewetting front, and gives agreement with experiment. At higher mass flows the rewetting velocity is substantially independent of surface thickness and conductivity. The present data and the available world data for rewetting are shown to be in agreement with the theory.     

大型物体辐射成像系统中散射问题的研究

在大型物体辐射成像系统中，成像质量是衡量成像系统的决定性指标。提出采用品质因子Q做为指标来描述成像系统，并结合蒙特卡罗模拟方法研究了在常见的X射线成像系统中发生的散射干扰对成像质量的影响．     

HAZOP-study on heavy water research reactor primary cooling system

By knowledge-based Hazard and Operability (HAZOP) technique, equipment malfunction and deficiencies in the primary cooling system of the generic heavy water research reactor are studied. This technique is used to identify the representative accident scenarios. The related Process Flow Drawing (PFD) is prepared as our study database for this plant. Since this facility is in the design stage, applying the results of HAZOP-study to PFD improves the safety of the plant.     

Optimal power control system of a research nuclear reactor

The power control system is a key control system for a nuclear reactor, which directly concerns the safe operation of a nuclear reactor. Much attention is paid to the power control system performance of nuclear reactor in engineering. The designers put a high value upon design of an optimal power control system. In this paper, a design method is applied to the design of power control system. According to the optimal control theory, an objective function, quadratic performance index with weight factors is proposed. Then, the objective function is transformed into frequency domain form by use of Paserval's theorem. In frequency domain, an optimal transfer function can be obtained at the lowest value of objective function. The system with optimal transfer function has an optimal performance. The transfer function of the power control system is derived from a typical research nuclear reactor. Using the state feedback theory, the transfer function is synthesized to the optimal transfer function. The simulative results with the optimal controller and with a conventional controller show that the performance of the optimal power control system is largely improved on dynamic characters. The method applied here not only can be used for research nuclear reactor but also can be easily extended to pressurized water reactor power plant and other fields.     

核应急辐射监测系统作业效能评估

为使核应急辐射监测系统性能指标考核评估结果能全面、准确地反映系统在实际作业场所中的作业能力,依据核应急辐射监测作业保障特点,构建了核应急辐射监测系统作业指标体系,并在此基础上提出采用"记号"的表达方式改进美国工业界武器系统效能咨询委员会提出的用于评估可靠性、可用性和能力的系统效能评估模型(Availability-Dependability-Capability-Model,ADC模型),使其评估结果不仅能定量地反映核应急辐射监测系统在作业过程中不同设备完成同一任务的总体效能,以及不同设备对总体效能的贡献,而且还能定量地反映核应急辐射监测系统中不同设备完成不同任务的作业效能。     

FRM-II: The new German research reactor

A new German high-flux research reactor is presently being built in Garching by the Technical University of Munich. The new reactor, called FRM-II, shall replace the existing ‘Forschungsreaktor München' FRM which has been operating very successfully for about 40 years now. The new reactor has been optimized primarily with respect to beam tube applications of slow neutrons, but will also allow to irradiate samples with thermal neutrons. Therefore, the FRM-II has been designed to provide a high flux of thermal neutrons in a large volume outside of the reactor core, where the neutron spectrum can be locally modified by using special spectrum shifters. The goal was to further obtain this high flux at a reactor power being as low as possible since this represents the best choice because of the lowest background radiation for the experiments, the lowest nuclear risk potential, the lowest costs and superior inherent safety features. In April, 1996, the project obtained the first partial nuclear licence which covered the general acceptance of the safety concept, the site opening and the construction of the reactor building. The final partial nuclear licence which allows nuclear start-up is expected for the year 2001.     

核与辐射应急准备分类及相应的管理要求

根据国际经验给出了基于危害评价的应急准备分类以及各类应急准备需要建立的基本能力,提出了应急管理行动、应急状态、应急响应的启动、缓解行动、应急评价和预测、防护行动和其他响应行动、公众应急通知、应急人员和应急援助人员的保护措施、医学响应行动、应急组织和人员配置、应急信息发布和公众沟通等11个主要要素对应5个应急准备类型的管理要求建议。     

The nuclear reactor circulation circuit as a radiation source

The solution of the problem of circulation circuits with a single radioisotope, which has been found earlier [1], is applied to the general case where several radioisotopes having radioactive progeny are formed in the substance to be activated. The problems of the absolute maximum circuit power and the consumption of neutrons per unit power for a number of elements which can be used as substances to be activated in the circuit are considered. From among them, the most promising are indium and its alloys.Special attention is paid to a circulation circuit where the substance to be activated contains a fissionable isotope (uranium circuit). It is shown that the specific power of such a circuit, all other conditions being equal, is considerably lower than the specific power of circuits with metallic indium or its alloys. As a particular case of a uranium circuit, the circulation from the reactor into the radiation unit,and the reverse,of fuel elements which have not burned up completely is considered. It is shown that, in this case, the power of the unit can be increased two- to fourfold in comparison with the power of a unit, which makes a single use of completely burned-up fuel elements.     

Monte Carlo simulation of the emergency shutdown system for the high-temperature pebble-bed nuclear reactor

A theoretical model of the emergency shutdown system for the high-temperature pebble-bed nuclear reactor is presented. The stochastic trajectory of an absorbent ball has been described as a fall with ‘absorption’ along the vertical axis and with planar diffusion in the horizontal directions. Small samples of the random packing of hard spheres and the absorbent balls' trajectories in them have been simulated by the direct Monte Carlo method. The values of model parameters have been estimated from the simulated distributions of the absorbent balls' displacement vectors. The estimated values agree qualitatively with experimental data.