Prototype of a nuclear rocket motor - the IRGIT reactor

The results of development work and power startup of a bench prototype of the reactor for a nuclear rocket engine - the IRGIT reactor - are presented for the development of a nuclear rocket motor. Solutions to problems of constructing and organizing the working process in the nuclear powr motor, the composition of the core, and the special features of the physics and thermophysics of the reactor are presented. The solutions are implemented in the construction of the IRGIT reactor and have passed firing tests.Translated from Atomnaya Énergiya, Vol. 97, No. 3, pp. 173–177, September, 2004.     

中国核农学的发展历程与成就

文章概述了我国核农学的创建与发展历程以及核技术农业应用研究所取得的主要成果。我国核农学始于1956年,经过50多年的发展,现已形成科研协作网络、学术交流网络、国际交流网络。这三大网络全面推动了中国农业核技术应用的核农学的形成与发展,在辐射育种、航天育种、同位素示踪技术在农业中的应用、农产品辐照贮藏和保鲜加工、辐射害虫防治、低剂量辐射刺激增产等方面取得了显著成绩。此外,文章还就中国核农学今后的发展构想以及核农学优先发展的主要领域提出了构思和建议。     

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.     

超高通量快中子研究堆核燃料概念设计研究

提高中子注量率是高通量研究堆的发展趋势,能够大幅加速反应堆材料研发进程。但若提高中子注量率至1016 cm?2·s?1将导致功率密度峰值相较于现有研究堆高数倍,对反应堆和核燃料设计带来许多挑战。为此,本文从中子学、传热、燃料材料堆内行为等方面半定量分析了提高中子注量率对核燃料性能的影响,并提出应对超高通量和功率密度挑战的设计措施,为发展超高通量快中子研究堆燃料设计提供指导。      

The MARS nuclear reactor plant: A inherently safe, small/medium size multipurpose nuclear plant

The crucial aspects of nuclear safety: the need of a safe shutdown and of a reliable decay heat removal system have been the starting points in the development of a medium size, inherently safe, multipurpose “new” nuclear reactor: the MARS nuclear reactor.     

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。     

超高通量快中子研究堆需求分析

实现超高快中子通量是世界先进研究堆的重要发展方向,对于加快第四代先进核能系统燃料及材料创新发展具有重要意义。本文从先进核能堆内结构材料与核燃料的辐照考验、长反应链超钚元素生产等角度,初步分析了我国建设超高通量快中子研究堆的必要性。在此基础上,确定了超高通量快中子研究堆的堆芯最大中子注量率及其冷却剂,给出了反应堆主要参数及冷却剂流动方案。反应堆热功率为200 MW,冷却剂为铅铋合金,最大中子注量率大于1016 cm?2·s?1。      

核农学在农业生态环境研究中的地位和发展战略

阐述了核农学的概念、研究范畴、研究进展及几十年来所取得的成就,并就核农学与农业生态环境的关系进行了探讨,提出了核农学在农业生态环境研究中的学科前沿、主要研究领域和优先发展领域,并就加强和改进核农学研究提出了几点建议,为核农学在生态环境中的应用及发展战略提供了基本素材与思考。     

Choosing suitable type of research reactor for Mongolia

Mongolian government said that one of the sources for energy supply will be a nuclear power. To utilize nuclear energy, it is required that Mongolia has to have sufficient number of national nuclear professionals. It is totally impossible to send all nuclear professionals to study in abroad for education degrees and do provide them with on the job training abroad. One of the key approaches to educate and train nuclear professionals locally is the need for Mongolia to have our own research reactor. The chosen research reactor will be used in various studies and for educational and training purposes. It is significantly important to get public acceptance to implement the nuclear power program. We are considering that it is suitable to choose TRIGA reactor of 300-500 kW power and of 10¹²⁻¹³ n/cm² s neutron flux.     

超临界水冷堆国内外研发现状与趋势

从我国核能长期发展的需求来看,研发第4代新型核能系统将确保核能的长期稳定发展。作为6种第4代未来堆型中唯一的水冷堆,超临界水冷堆具有经济性、延续性及可持续性等诸多综合优势,是国家水冷堆核电技术路线进一步发展的必然选择,也是清洁能源科学和技术领域基础研究国际竞争与合作重要的前沿与热点之一。本文将分析超临界水冷堆的技术特性及它在我国核能长期发展战略中的地位,总结国内外超临界水冷堆的研究现状与发展趋势,提出中国超临界水冷堆的发展方向与路线图。     

基于CFD方法的核动力系统热工安全特性研究进展

西安交通大学核反应堆热工水力团队（XJTU-NuTheL）长期致力于计算流体动力学（CFD）方法的核动力系统高精度热工水力模型开发及应用方面的研究。近些年,团队在单相CFD工程应用、两相CFD模型开发、大涡模拟（LES）及直接数值模拟（DNS）高性能并行计算、跨尺度多物理场耦合等方面取得了系列研究成果。主要包括：构建了核反应堆压力容器、蒸汽发生器、非能动余热排出系统换热器等核动力系统关键设备的三维多孔介质热工水力计算模型,建立了复杂物理现象及运动瞬变工况下的两相CFD数学物理模型,开发了CFD程序与核反应堆系统程序、堆芯子通道程序之间的跨尺度耦合以及与中子物理、力学程序之间的多物理场耦合分析平台。本文将重点阐述XJTU-NuTheL基于CFD方法在核反应堆热工水力研究方面的最新成果及进展,并提出CFD方法在核反应堆工程领域应用的主要挑战及发展方向,旨在促进CFD方法更好地服务于核动力系统设计与运行安全分析。     

数值反应堆的研究现状与发展建议

数值反应堆技术是基于多物理紧耦合的高精度、高分辨率、高置信度的高保真数值模拟技术,其目的是实现核反应堆内物理现象的精确数值呈现和分析,大幅度提高核反应堆的设计能力和安全运行能力,用数值技术驱动核能技术的快速发展。本文总结了国内外数值反应堆技术的研究现状,提出了数值反应堆技术的发展建议。     

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.     

Present status of the High Temperature Engineering Test Reactor (HTTR)

In Japan, the research and development on the High Temperature Gas-cooled Reactors (HTGRs) had been carried out for more than fifteen years since 1969 as the multi-purpose Very High Temperature gas-cooled Reactor (VHTR) program for direct utilization of nuclear process heat such as nuclear steel making. Recently, reflecting the change of the social and energy situation and with less incentives for industries to introduce such in the near future, the JAERI changed the program to a more basic ‘HTTR program’ to establish and upgrade the HTGR technology basis.The HTTR is a test reactor with a thermal output of 30 MW and reactor outlet coolant temperature of 950°C, employing a pin-in-block type fuel block, and has the capability to demonstrate nuclear process heat utilization using an intermediate heat exchanger. Since 1986 a detailed design has been made, in which major systems and components are determined in line with the HTTR concept, paying essential considerations into the design for achieving the reactor outlet coolant temperature of 950°C. The safety review of the Government started in February 1989. By request of the Science and Technology Agency the Reactor Safety Research Association reviewed the safety evaluation guideline, general design criteria, design code and design guide for the graphite and the high-temperature structure of the HTTR.The installation permit of the HTTR was issued by the Government in November 1990.     

PERFORM 60 - Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and continuous progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the effects of irradiation on materials microstructure. A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project. These tools allow to simulate irradiation effects on the constitutive behaviour of the reactor pressure vessel low alloy steel, and also on its failure properties. Relying on the existing PERFECT Roadmap, the 4 years Collaborative Project PERFORM 60 has mainly for objective to develop multi-scale tools aimed at predicting the combined effects of irradiation and corrosion on internals (austenitic stainless steels) and also to improve existing ones on RPV (bainitic steels).PERFORM 60 is based on two technical sub-projects: (i) RPV and (ii) internals. In addition to these technical sub-projects, the Users’ Group and Training sub-project shall allow representatives of constructors, utilities, research organizations… from Europe, USA and Japan to receive the information and training to get their own appraisal on limits and potentialities of the developed tools. An important effort will also be made to teach young researchers in the field of materials’ degradation.PERFORM 60 has officially started on March 1st, 2009 with 20 European organizations and Universities involved in the nuclear field.     

PGV-1500 steam generator. New problems and solutions

The basic questions concerning the development of a steam generator for a nuclear power plant with a VVé R-1500 reactor are presented. The basic design requirements which follow for steam generators from experience in operating analogs at nuclear power plants and taking account of the requirements for a reactor system are presented. The special features inherent to horizontal-type steam generators, which have been mastered and are used in nuclear power plants in our country, are noted. The domestic and world operating experience is taken into account in the development of the design. It is concluded that the design of the PGV-1500 steam generator satisfies the requirements for the concept of a VVéR reactor facility for a 1500 MW(e) unit of a nuclear power plant and is competitive on the world market for power-generating equipment for nuclear power plants. __________ Translated from Atomnaya énergiya, Vol. 99, No. 6, pp. 416–425, December, 2005. An erratum to this article is availabel at .     

数值反应堆原型系统开发及示范应用研究进展

数值反应堆是当前核反应堆工程与技术领域国内外研发热点之一。本文介绍了国家重点研发计划“数值反应堆原型系统开发及示范应用”重点专项总体研发进展，包括数值反应堆原型系统CVR1.0、核心软件开发与验证、多物理耦合进展，以及数值反应堆原型系统初步示范应用。     

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.     

A concept of radiation safety for nuclear power units in space and its implementation in the “Kosmos-1900” satellite

Conclusion Experience in ensuring radiation safety for nuclear units in outer space has shown the adequacy of using two independent systems: a booster system and a system for dispersing the reactor based on using active means to break up its structure.The theoretical calculations and experimental research performed in the Soviet Union, the engineering and construction development, and testing have confirmed the effectiveness of the booster and reactor dispersion systems as well as their capability to operate with the necessary reliability during normal operating conditions and in accident situations aboard space equipment.This article is a variant of a report presented at the Sixth Symposium on Nuclear Power in Outer Space (Albuquerque, January 1989).Translated from Atomnaya Énergiya, Vol. 66, No. 6, pp. 380–383, June, 1989.