空间放射性同位素电池发展回顾和新世纪应用前景

迄今为止,美俄两国已向空间发射了80多台空间核电源(包括同位素电池和反应堆电源)。重点回顾了20世纪放射性同位素电池的研发历史和空间发射现状;概括介绍了目前放射性同位素温差发电器(RTG)业已达到的技术水平和提高热电转换效率的最近动向;综述了美国、俄罗斯和欧洲航天局在21世纪初期(20001/2015)使用RTG的空间和太阳系探索计划,展现了RTG的广阔应用前景。     

Chemistry and technology of Molten Salt Reactors - history and perspectives

Molten Salt Reactors represent one of promising future nuclear reactor concept included also in the Generation IV reactors family. This reactor type is distinguished by an extraordinarily close connection between the reactor physics and chemical technology, which is given by the specific features of the chemical form of fuel, representing by molten fluoride salt and circulating through the reactor core and also by the requirements of continuous ‘on-line’ reprocessing of the spent fuel. The history of Molten Salt Reactors reaches the period of fifties and sixties, when the first experimental Molten Salt Reactors were constructed and tested in ORNL (US). Several molten salt techniques dedicated to fresh molten salt fuel processing and spent fuel reprocessing were studied and developed in those days. Today, after nearly thirty years of discontinuance, a renewed interest in the Molten Salt Reactor technology is observed. Current experimental R&D activities in the area of Molten Salt Reactor technology are realized by a relatively small number of research institutions mainly in the EU, Russia and USA. The main effort is directed primarily to the development of separation processes suitable for the molten salt fuel processing and reprocessing technology. The techniques under development are molten salt/liquid metal extraction processes, electrochemical separation processes from the molten salt media, fused salt volatilization techniques and gas extraction from the molten salt medium.     

船用核动力的发展特点与启示

美国、前苏联/俄罗斯船用核动力技术长期保持世界领先,其发展经验和技术脉络具有极高的参考价值。本文通过对美国、前苏联/俄罗斯船用核动力发展的主要历程和技术进行分析研究,创新总结归纳出其反应堆系统基本型、通用试验平台、差异化配置等共同发展规律,并从管理模式、技术路线以及发展趋势等方面挖掘提炼出美国和前苏联/俄罗斯船用核动力技术遵循的一系列共性特点和差异化特征,可为船用核动力发展提供一定的参考和启示。      

模块式高温气冷堆核电站设计特点

模块式高温气冷堆具有安全、灵活、可靠、经济性好的优点,受到核技术先进国家的重视。本文着重介绍了美国新近推出的模块式高温气冷堆核电站的设计特点和安全特性。     

The role and place of gas-cooled reactors in the structure of nuclear power provision

The possibilities are considered of using high-temperature gas-cooled reactors to generate heat with various temperature potentials for supplying energy to industrial production, supplying heat, and generating electrical power. Brief information is given on the characteristics and design of high-temperature gas-cooled reactor schemes developed in Russia and also on the GT-MHR project being developed as part of international cooperation. An account is given of research on high-temperature gas-cooled reactors for the purposes of power technology and the generation of electrical power. The basic criteria and safety principles are described. An estimate is made of the possible fraction of high-temperature gas-cooled reactors in the structure of power provision in Russia, taking account of the specific consumption of natural uranium and the accumulation of radioactive waste if there is no reprocessing of the depleted fuel. OKBM. Translated from Atomnaya énergiya, Vol. 87, No. 2, pp. 87–91, August, 1999.     

数字化的核反应堆物理启动系统的研制和应用

介绍了数字化物理启动系统的构成和基本工作原理，及其在10MW高温气冷实验堆物理启动试验过程中的首次成功运用。实践证明：该系统不但运行可靠，计算迅速准确．减轻了人员劳动强度，且与同类模拟系统相比，具有实时监测显示、试验结果透明度高的特点。     

Heavy water reactors and nuclear power plants in the USSR and Russia: Past, present, and future

The history of the development of heavy-water nuclear reactors and the assoiated, installations in the USSR and Russia is presented. Research reactors constructed at the ITEP and under the scientific direction of the ITEP in other countires (Yugoslavia), industrial heavy-water nuclear reactors, and the Maket zero-power reactor are described. Heavy-water gas-cooled reactors for nuclear power plants are discussed in detail: the nuclear power plant with an A-1 reactor, constructed in Czechoslovakia, and the design of maximum-safety nuclear power plant. Electronuclear neutron generators and subcritical nuclear reactors and the possibility of using the for burning weapons plutonium are examined. The electronuclear neutron generator developed at the ITEP is described. State Science Center of the Russian Federation—Institute of Theoetical and Experimental Physics. Translated from Atomanaya énergiya, Vol. 86, No. 4, pp. 310–321, April, 1999.     

Renewed interest in lead cooled fast reactors

In the last few years a number of compact designs of lead-alloy cooled systems have been promoted. Moreover, in Russia a design effort was started earlier on the pure lead-cooled BREST reactor but this effort does not appear to be strongly funded any more. But now the lead cooled and compact STAR-LM reactor is promoted in the US and in the European Union there is some interest in a mediumsized lead-cooled fast reactor (LFR). It has brought some nuclear industries, a large utility, several research centers and universities together to ask the European Commission for a partial funding of design and safety efforts. A 600 MWe LFR design is proposed which would be useful for base load operation but as a fast system it could also be used for load following. Because of the possible plant simplifications and the use of pure lead, the economics of such a system should be good. Moreover, efficient fuel utilization, the burning of higher actinides and a closed fuel cycle make it a sustainable system. Whether, this larger system has the same inherent / passive safety characteristics as smaller LFRs needs to be examined. In this paper the passive emergency decay heat removal by reactor vessel aircooling of such a larger system is investigated. Moreover an inlet blockage in a subassembly of a low power density LMR is analyzed. Furthermore, the pros and cons of lead vs. lead/bismuth coolants are discussed.     

Safety concept and operational criteria of a nuclear process heat plant

A plant design of a high-temperature nuclear reactor (HTR) of 500 MW(th) coupled with two coal gasification lines has been developed (PNP-500). The general objective of the safety concept is to maintain the same high standard as that developed for conventional nuclear power plants. As burnable gases are transported within the primary circuit of the reactor, the containment building and in close proximity outside of the containment, novel safety considerations are arising. The related safety concepts and the results of experiments with exploding gases are discussed. Also specific operating criteria are discussed which are due to the coupling between a nuclear reactor and the two gasification lines.     

Seismic isolation development for the US advanced liquid-metal reactor program

GE Nuclear Energy, in association with a US Industrial Team and support from the US National Laboratories and Universities, is developing a modular liquid-metal reactor concept for the US Department of Energy (DOE). The objective of this development is to provide, by the turn of the century, a reactor concept with optimized passive safety features that is economically competitive with other domestic energy sources, licensable, and ready for commercial deployment. One of the unique features of the concept is the seismic isolation of the reactor modules which decouples the reactors and their safety systems from potentially damaging ground motions and significantly enhances the structural resistance to high energy, as well as long-duration earthquakes. Seismic isolation is accomplished with high-damping natural-rubber bearings. The reactors are located in individual silos below grade level and are supported by the isolator bearings at approximately their center of gravity.This application of seismic isolation is the first for a US nuclear power plant. A development program has been established to assure the full benefits from the utilization of this new approach and to provide adequate system characterization and qualification for licensing certification. The development program, which is supported by the US Department of Energy (DOE), Argonne National Laboratory (ANL), Energy Technology Engineering Center (ETEC), the University of California at Berkeley (UC-Berkeley), General Electric (GE), and Bechtel National, Inc. (BNI), is described in this paper and selected results are presented. The initial testing indicated excellent performance of high-damping natural-rubber bearings. The development of seismic isolation guidelines is in progress as a joint activity between ENEA of Italy and the GE Team.     

A nodal dynamic model for control system design and simulation of an MHTGR core

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high-temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high-temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power-distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.     

Abstracts on Transport of Radioactive Materials

Abstract

When compared to many other countries, the system of regulating nuclear transport in the USA and the organisational structure of that system often appears overly complex and confusing to some persons outside the USA (and sometimes even to persons in the USA). The objective of this paper is to provide a timely reference source for the readers of The International Journal of Radioactive Materials Transport, particularly those outside the USA, and hopefully to help to clarify some of that confusion. The sources of regulations and the structures of the primary regulatory organisations of the USA and their interfaces and responsibilities are described. The major features of the regulations of the principal regulators—The US Department of Transportation and the US Nuclear Regulatory Commission—are discussed. Although the nuclear transport regulations of the USA have, since 1968, been based essentially on the IAEA standards contained in Safety Series No 6, a few notable differences do exist between the application of certain of those standards in the USA regulations and the relevant Safety Series No 6 standard. The most notable of these differences and the rationale for the existence of the difference are discussed.     

美、俄两国核电与铀工业的现状及走向

介绍了美国和俄罗斯的核电、铀生产现状及未来走向。美国与俄罗斯是世界上最早拥有核电的国家。历经40多年的发展，美国的核发电量占全美总发电量的20％，俄罗斯则为11％(2001年）。美国核电用铀的6％出自本土；而俄罗斯的铀产量的近一半用来出口。美国的未来核电用铀在很大程度上将依从于国际市场和各级、各类铀的库存；在安民兴国的氛围中，俄罗斯依托其殷实的铀资源，确立了核发电宏图。可以认为，尽管美国与俄罗斯两国核电、铀工业的走向会对世界铀矿地质勘查和铀生产产生重要的影响，但全球铀工业在近10a里仍将呈现供、需基本平衡的态势。     

Project of the GT-MHR high-temperature helium reactor with gas turbine

In February 1995, MINATOM of Russia and General Atomics (USA) signed the Agreement for the development and design of the GT-MHR facility with a modular helium reactor and a gas turbine intended to be constructed in Russia. This Agreement was subsequently expanded by the participation of Framatom and Fuji Electric. The GT-MHR facility is designed for burning weapons-grade plutonium and utilization of the heat produced in the direct gas-turbine cycle with electricity production efficiency of about 50%. In future such facilities with uranium fuel will be proposed for use as commercial NPPs. A GT-MHR prototype and fuel production facility are intended to be constructed at the Siberian Chemical Combine in Seversk (Tomsk-7). In accordance with the Agreement, a conceptual design of the GT-MHR should be developed in September 1997. As a part of the conceptual design, a reactor module with a power conversion system is being designed and plutonium fuel is being developed.     

SSTAR: The US lead-cooled fast reactor (LFR)

It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV lead-cooled fast reactor system.     

Metallic fuels for advanced reactors

In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.     

Concept and salient design features of a high-temperature nuclear reactor with solid coolant

The salient features of using a solid substance to cool the core of a nuclear reactor and the associated advantages and limitations are examined. Conceptual proposals concerning the core design and the arrangement of the in-reactor space of a high-temperature nuclear reactor with a solid coolant are presented. Evaluated data and some results for a model reactor are presented. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 156–161, September, 2007.     

小型氟盐冷却高温堆耦合布雷顿循环系统分析与研究

为满足小型氟盐冷却高温堆（FHR）能量转换需求,开发与之匹配的高效、紧凑、无水冷却动力转换系统,本文对比了超临界二氧化碳（SCO₂）、空气、氩气（Ar）、氮气（N₂）、氙气（Xe）5种气体工质在不同布雷顿循环构型中的热电转换效率、?效率、?损失分布。研究发现,SCO₂布雷顿循环相比其它工质循环具有最高的热电转换效率和?效率,且结构更为紧凑,易于小型化和模块化,与小型氟盐冷却高温堆耦合更具优势;进而对SCO₂布雷顿循环进行构型优化,得出匹配小型氟盐冷却高温堆的最佳循环构型方式,构成固有安全模块化小型氟盐冷却高温堆热电转换系统,为西部能源利用提供新研究思路。      

State of the art and prospects of fast neutron reactors

On the basis of experience of fast reactor design, construction and operation gained in Russia, this paper outlines their state of the art. The high maturity and efficiency of this type of nuclear power development in Russia and the equalization of the economic characteristics of thermal and fast reactors is shown, as well as the expediency of improvement of nuclear power environmental characteristics owing to fast reactors incorporation.     

俄罗斯核应急管理体制和技术支持体系分析北大核心CSCD

俄罗斯是核能应用强国,在核能稳步发展的同时,逐步建立了体系完备、功能完整、运转高效的国家核应急系统,并实现了核应急管理体制与国家应急管理体制的有机结合,采用的是垂直管理模式,具有学习和借鉴的价值。本文首先介绍了俄罗斯核应急管理体制,然后分析了俄罗斯核应急技术支持体系,最后从应急管理机制、突发事件的预警与监测、信息与资源共享、技术体系能力建设等几个方面阐述了对我国核应急响应技术能力建设与发展的若干启示与思考。