Deliberately small reactors and the second nuclear era

Smaller sized nuclear reactors were instrumental during the pioneering days of commercial nuclear power to facilitate the development and demonstration of early reactor technologies and to establish operational experience for the fledgling nuclear power industry. As the U.S. embarks on its “second nuclear era,” the question becomes: Will smaller sized plants have a significant role in meeting the nation's needs for electricity and other energy demands? A brief review of our nuclear history is presented relative to plant size considerations, followed by a review of several commonly cited benefits of small reactors. Several “deliberately small” designs currently being developed in the U.S. are briefly described, as well as some of the technical and institutional challenges faced by these designs. Deliberately small reactors offer substantial benefits in safety, security, operational flexibilities and economics, and they are well positioned to figure prominently in the second nuclear era.     

商用压水堆核燃料研发进展与应用展望

压水堆(PWR)是目前核电厂反应堆的主力堆型,而核燃料是反应堆的能量源泉和放射性裂变物质的主要来源,关乎核电厂的经济性和安全性。本文对当前国际上面向商用PWR应用研发的掺杂UO₂燃料、高铀密度燃料、微封装燃料和金属燃料的性能特点、技术状态及前景进行了归纳和评价。在掺杂UO₂燃料中,大晶粒燃料具有较高的技术成熟度,将在PWR实现大规模商用;高铀密度燃料和金属燃料在高温水腐蚀氧化问题以及事故下的行为仍待研究解决;具有极致安全的微封装燃料更适合特殊用途的小型反应堆。应协同开展先进燃料组件设计、建立设计准则以及研发高保真的性能分析技术等,以充分发挥新型燃料的可靠性及高燃耗优势。     

Evaluation of implementation of thorium fuel cycle with LWR and MSR

In order to construct a sustainable society, it is necessary to consider fairness beyond generations and between countries. It is expected that Asian countries continue growing their economy and will result consuming more energy. More CO₂ emission is not acceptable.Nuclear power has many advantages for reducing CO₂ emission. However, it still has concerns of nuclear proliferation, radioactive waste and safety. It is necessary to overcome these concerns if nuclear power is expanded to Asian countries. Thorium utilization as nuclear fuel will be an opening key of these difficulties because thorium produces less plutonium, less radioactive waste. Safety will also be enhanced. The use of molten-salt reactor (MSR) triggered by plutonium supply from ordinary light water reactor (LWR) with uranium fuel will allow implementation of thorium fuel cycle with electricity capacity of about 446 GWe around at 2050.The other important sector in a view of sustainability is transportation. Transportation is essential for economy growth. Therefore it is inevitable to reduce CO₂ emission from transportation sector. Electric vehicle (EV) will be used as a major mobility instead of gasoline engine cars. Rare-earth materials such as neodymium and dysprosium are necessary for producing EV. These materials are expected to be mined from Asian countries. It is often obtained with thorium as by-product. Thorium has not been used as nuclear fuel because it is not good for nuclear weapon and it does not have fissionable isotopes. Recent global trend of nuclear disarmament and accumulation of plutonium from uranium fuel cycle can support starting the use of thorium.Thorium utilization will help both to provide clean energy and to produce rare-earth for clean vehicle. These will create new industries in developing Asian countries. An international collaborative framework can be established by supplying resource from developing countries and supplying technology from developed countries. “THE Bank (THorium Energy Bank)” is proposed here as one part of such a framework.     

Requirements for 21st centry nuclear power plants

The development of energy production in the 21st century will be subject to more uniform per capita and regional consumption. Among the competing sources of energy, the positive qualities of nuclear power-unlimited fuel resources, high energy intensiveness, and ecological compatibility with the possibility of the wastes being highly concentrated—predetermine the development of large-scale nuclear power. The conditions for the development of such nuclear power are its ecological effectiveness and safety (of the reactors and the fuel cycle with the production of wastes), nuclear fuel breeding with adequate characteristics, and guarantees of nonproliferation of fissioning materials. Continuity in the development of nuclear power dictates the requirements for reactor systems in the near and distant future. The acceptable level of safety is closely related to the scales of nuclear power and the applications of nuclear energy sources. However, progress in decreasing the potential danger of reactors and decreasing the cost of protective systems is unavoidable. In choosing new directions, it is important to demonstrate the new qualities in the solution of the problems facing nuclear power in the future. An adequate diversity of reactor technologies could exist in the future. The requirements that will face nuclear power plants in the future stages of development and the expected stages of this development are discussed. The jourmal variant of this report at the 10th annual conference of the Nuclear Society “From the first nuclear power plant in the world to power engineering of the twenty-first century” (June 28–July 2, 1999, Obninsk) Russian Science Center “Kurchatov Institute”. Translated from Atomnaya énergiya, Vol. 88, No. 1, pp. 3–14, January, 2000.     

Low-capacity nuclear power: questions and answers

Low- and medium-capacity nuclear power plants for distributed power generation will play a large role at the present stage. The need for such plants in our country solely for supplying electricity to the northern regions will be at least 20 GW(e). One of the most promising developments of nuclear energy sources for autonomous heat and power generation is the Uniterm low-capacity plant with a water-cooled and -moderated reactor facility, where the rich experience in developing and operating domestic propulsion nuclear power facilities is used. It possesses unique user properties and meets the present-day requirements for safety, reliability, and ecological cleanliness.     

Current status and technical description of Chinese 2 × 250 MWth HTR-PM demonstration plant

After the nuclear accidents of Three Mile Island and Chernobyl the world nuclear community made great efforts to increase research on nuclear reactors and to develop advanced nuclear power plants with much improved safety features. Following the successful construction and a most gratifying operation of the 10 MW_th high-temperature gas-cooled test reactor (HTR-10), the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University has developed and designed an HTR demonstration plant, called the HTR-PM (high-temperature-reactor pebble-bed module). The design, having jointly been carried out with industry partners from China and in collaboration of experts worldwide, closely follows the design principles of the HTR-10.Due to intensive engineering and R&D efforts since 2001, the basic design of the HTR-PM has been finished while all main technical features have been fixed. A Preliminary Safety Analysis Report (PSAR) has been compiled.The HTR-PM plant will consist of two nuclear steam supply system (NSSS), so called modules, each one comprising of a single zone 250 MW_th pebble-bed modular reactor and a steam generator. The two NSSS modules feed one steam turbine and generate an electric power of 210 MW.A pilot fuel production line will be built to fabricate 300,000 pebble fuel elements per year. This line is closely based on the technology of the HTR-10 fuel production line.The main goals of the project are two-fold. Firstly, the economic competitiveness of commercial HTR-PM plants shall be demonstrated. Secondly, it shall be shown that HTR-PM plants do not need accident management procedures and will not require any need for offsite emergency measures.According to the current schedule of the project the completion date of the demonstration plant will be around 2013. The reactor site has been evaluated and approved; the procurement of long-lead components has already been started.After the successful operation of the demonstration plant, commercial HTR-PM plants are expected to be built at the same site. These plants will comprise many NSSS modules and, correspondingly, a larger turbine.     

Status and safety aspects of the 300 MW(e) GCFR demonstration plant

The Gulf General Atomic concept of the gas-cooled fast breeder reactor (GCFR) utilizes development that has already taken place on the high temperature gas-cooled reactor (HTGR), principally in plant systems and components, and it will benefit directly from the development work carried out to support the LMFBR — primarily the nuclear fuel development. Recent progress in the development work and engineering design for the 300 MW(e) GCFR demonstration plant is highlighted and the safety aspects of this plant are discussed.     

The advanced high-temperature reactor: High-temperature fuel, liquid salt coolant, liquid-metal-reactor plant

The Advanced High-Temperature Reactor is a new reactor concept that combines four existing technologies in a new way: (1) coated-particle graphite-matrix nuclear fuels (traditionally used for helium-cooled reactors), (2) Brayton power cycles, (3) passive safety systems and plant designs from liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants with boiling points far above the maximum coolant temperature. The new combination of technologies enables the design of a large [2400- to 4000-MW(t)] high-temperature reactor, with reactor-coolant exit temperatures between 700 and 1000°C (depending upon goals) and passive safety systems for economic production of electricity or hydrogen. The AHTR [2400-MW(t)] capital costs have been estimated to be 49 to 61% per kilowatt (electric) relative to modular gas-cooled [600-MW(t)] and modular liquid-metal-cooled reactors [1000-MW(t)], assuming a single AHTR and multiple modular units with the same total electrical output. Because of the similar fuel, core design, and power cycles, about 70% of the required research is shared with that for high-temperature gas-cooled reactors.     

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.     

A concept of the multipurpose liquid metallic-fueled fast reactor system (MPFR)

Against fossil fuels, the nuclear energy is the only alternative energy source in the next century. Such energy source as the future nuclear power plant is expected to meet the following requirements. First, high temperature output for the multiple energy conversion capability as the electricity generation and the production of alternative fuels (hydrogen), which can be used widely in transportation systems. Second, the capability for siting close to the energy consumption area without onsite refueling. Third, the capability for nuclear fuel breeding and incineration of long-lived fission products, and fourth, the harmonization between active and passive safety features. This paper describes the basic concept of the Multipurpose liquid metallic-fueled Fast Reactor system (MPFR), which satisfies all mentioned requirements with introducing the U-Pu-x (x: Mn, Fe, Co) liquid metallic alloys for the fuel. We can obtain such characteristics as high operational temperature of the reactor (between 550 °C and 1200 °C) and elongation of the core operational lifetime by the inherent fission product separation in the liquid fuel by using these alloys. The enhanced self-controllability is achieved by the thermal expansion of liquid fuel; and the re-criticality phenomenon at the core compaction events can be eliminated by discharging of the liquid fuel from the core.     

The European pressurized water reactor Result of the French–German cooperation of experienced NPP suppliers and operators

In 1989 Framatome and Siemens, the two most experienced European nuclear power plant suppliers, decided to join the efforts for the development of a new reactor type for the next generation in their equally owned subsidiary Nuclear Power International (NPI). In 1992 Electricité de France and the major German utilities operating nuclear power plants merged their own development programs with that of Nuclear Power International and initiated the European Pressurized Water Reactor (EPR) project. In order to reach the two major targets of the project, the licensability in both countries, France and Germany, and the competitiveness of nuclear energy with other alternative energy sources, the design basis which had differently developed in the two countries needed to be harmonized. In parallel, the licensing authorities of both countries extended their existing cooperation in the field of a safety survey of existing nuclear power plants to the definition of safety criteria for the next generation of nuclear power plants. Through this cooperation the licensability of EPR in France and Germany will be assured. Continuously performed cost analysis show in addition that also the second target of the project, the competitiveness with alternative primary energy sources, can be achieved. Thanks to the fruitful cooperation between all parties involved, satisfactory results have been achieved not by a simple superposition of existing design features but through a careful evaluation and combination of the best available alternatives. At the end of 1997 the basic design results were compiled in a final report. Subsequently an optimization phase was launched that further improves the competitiveness of the power generation costs.     

压水堆燃料辐照性能多物理耦合并行分析程序开发

核燃料元件是反应堆的核心部件,其性能影响反应堆的安全性与经济性,利用燃料元件性能分析程序开展燃料堆内稳态辐照性能分析对于燃料设计及安全评价具有重要意义。通过开发燃料温度分布、变形计算、裂变气体释放及内压等模型,结合燃料元件热工-力学多物理耦合计算分析耦合方案,基于先进并行计算方法构建了高性能并行化燃料性能分析程序Athena。利用典型商用压水堆核电站数据及同类程序计算结果进行了程序初步验证,结果表明Athena程序计算结果合理可靠。通过定义堆芯功率及热工水力边界条件,程序能够并行开展压水堆全堆芯燃料辐照性能分析,提高燃料辐照性能分析效率,是数值反应堆原型系统（CVR1.0）的重要组成。     

Similarity of high-temperature gas-cooled reactor technologies and designs in Russia and USA

Historical information concerning the development of high-temperature gas-cooled reactors in the USA and Russia is presented. The reactor facilities MHTGR (USA), VG-400 (Russia), VGM (Russia), GT-MGR (Russia, USA), and at the Fort St. Vrain nuclear power plant (USA) are described. The US programs for developing innovative high-temperature nuclear reactor technologies are examined. It is shown that the Russian and US technological developments for the fuel, reactor system, energy conversion system, and fission-product transport are similar.     

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation’s energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.     

摩洛哥坦坦地区核能海水淡化示范项目

摩洛哥王国准备采用我国开发的１０ＭＷ核供热堆作为热源，与高温多效蒸馏工艺相耦合，在坦坦地区建造核能海水淡化示范厂，日产８０００ｍ＾３淡水，可行性研究结果表明：该示范厂设计方案不存在技术障碍，其淡水生产成本和该地区相同规模的石化燃料淡化厂相当。     

Role of Nuclear Power in World Energy Production in the 21st Century

World growth of energy consumption in the 21st century is unavoidable. The most intense growth will occur in the developing countries. Of course, fossil fuels will continue to serve as the main source of energy, but the problems due to the growth of the raw materials base will also increase. The growth of nuclear power production will facilitate a successful solution to these problems. Various estimates show that by mid-century nuclear power will grow by a factor of four or five from the present level. Under such conditions, the solution of certain internal problems of nuclear power, such as preventing proliferation, handling spent fuel and radioactive wastes, and developing a reliable intrinsic raw materials base based on a closed fuel cycle and innovative nuclear technologies, will be exceedingly important for large-scale nuclear power to be successful. In this article, an assessment of the possible role of nuclear power in providing for the stable advancement of human civilization on global and regional scales is made on the basis of predictions of the growth of power production. __________ Translated from Atomnaya Energiya, Vol. 99, No. 5, pp. 323–336, November, 2005.     

Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads.The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany.In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented.Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of reactor core melting numerically is demonstrated that the molten core debris would be retained inside the reactor vessel due to the effective passive external water cooling of the vessel, keeping it completely intact.A short construction period of just 48 months from first concrete to provisional take over, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. Realistic average availability for a plant lifetime of 60 years and 12 months cycle is 94.5%. Systems and plant design were reviewed by expert groups of European utilities. With the SWR-1000, AREVA NP has developed a design concept for a BWR plant that is now ready for commercial deployment and which fully meets the most stringent international requirements in terms of nuclear safety and nuclear regulatory.     

Update of the French R&D strategy on gas-cooled reactors

Gas-cooled reactors take up a strong second role in France's R&D strategy on future nuclear energy systems as priority was given in 2005 to fast neutron reactors with multiple-recycle for their potential to optimally use uranium resource and minimize the long term burden of radioactive waste. Owing to the European past experience on sodium-cooled fast reactors (SFRs), this reactor type was logically selected as reference for a new generation fast neutron reactor intended to be tested as a prototype in the 2020s and be ready for industrial deployment around 2040. At the same time, the potential merits of a gas fast reactor (GFR) with ceramic clad fuel for a safe management of cooling accident are acknowledged for the potential of this reactor type to resolve critical issues of liquid fast reactors (safety, operability and reparability). A pre-feasibility report on a first concept of GFR was issued in 2007 that summed-up results of a 5-year international R&D effort on GFR fuel technology, reactor design and operating transient analyses. This report established a global confidence in the feasibility of this concept and its potential for attractive performances. Furthermore, it suggested directions of R&D to generate by 2012 an updated concept with improved performances and taking better benefit from GFR specific technologies.A second activity on gas-cooled reactors originates from the current interest of CEA's industrial partner AREVA in high or very high temperature reactors (V/HTR) for supplying hydrogen, synthetic hydrocarbon fuels and process heat for the industry. This activity currently encompasses R&D on V/HTR key technologies such as particle fuel fabrication, high temperature compact heat exchangers and coupling technologies to various power conversion systems. R&D on V/HTR and GFR are synergistic in various respects. The GFR can be viewed as a more sustainable version of the VHTR and synergies exist in research on heat resisting materials, helium system technology and power conversion systems. Both reactors require active research in materials and spur developments of new metallic alloys and ceramics applicable to other advanced nuclear systems.     

安全经济高效的先进能源

具有第四代安全经济特性的核电应该是人们期待的先进的清洁低碳能源。高温气冷堆是当今研发的第四代核电堆型之一,但现有的设计还存在需要排除的严重的安全隐患。堆芯不熔化,不等于说不会有严重事故发生。需要吸取国外球床高温堆和柱状高温堆两种实验堆型运行的经验教训、扩展安全观念和应对安全低概率事件,确保反应堆不出现后果极其严重的放射性释放事故。当热电转换系统采用与燃气蒸汽联合循环耦合应用的技术以后,会发挥高温堆所长,更大地提升转换效率,形成一种高安全低投资和高效率的双燃料清洁能源,可用于大堆或小堆的应用环境,可满足电力系统基本负荷和调锋负荷的需要。在工程设计上采取一系列改进和创新措施,包括釆用规则床模块化及地下反应堆设计以后,可在提高反应堆核心部位安全防卫能力的同时,防范低概率事件,成为一种新的安全经济高效的先进能源。     

A fusion–fission hybrid reactor with water-cooled pressure tube blanket for energy production

A fusion–fission hybrid reactor is proposed to achieve the energy gain of 3000 MW thermal power with self-sustaining tritium. The hybrid reactor is designed based on the plasma conditions and configurations of ITER, as well as the well-developed pressurized light water cooling technologies. For the sake of safety, the pressure tube bundles are employed to protect the first wall from the high pressure of coolant. The spent nuclear fuel discharged from 33GWD/tU Light Water Reactors (LWRs) and natural uranium oxide are taken as driver fuel for energy multiplication. According to thermo-mechanics calculation results, the first wall of 20 mm is safe. The radiation damage analysis indicates that the first wall has a lifetime of more than five years. Neutronics calculations show that the proposed hybrid reactor has high energy multiplication factor, tritium breeding ratio and power density; the fuel cannot reach the level of plutonium required for a nuclear weapon. Thermal-hydraulic analysis indicates that the temperatures of the fuel zone are well below the limited values and a large safety margin is provided.