球床包层聚变—裂变混合堆热工安全分析

球床包层混合堆与板状元件包层混合堆相比较，前者在核燃料生产和安全方面可能具有更多的优越性。本应用ＴＨＥＲＭＩＸ程序和辅助程序对我国开发的托卡马克堆芯氮气冷却球床包层聚变－裂变合堆的包层进行了热工计算。计算中考虑了不同的燃料球材料及稳态，卸压和断流事故工况。计算结果表明，只要选用合适的燃料球材料和设置适当的控制保护系统，具有快速卸料罐的托卡马克堆芯氦气包层聚变－裂变混合堆的概念设计在安全上的可行的。     

PWR（U）乏燃料中超铀元素在混合堆快裂变包层内嬗变研究

从中子学角度对ＰＷＲ（Ｕ）乏燃料中的超铀元素（２３８Ｐｕ，２３９Ｐｕ，２４１Ｐｕ，２４１Ａｍ，２４３Ａｍ，２３７Ｎｐ，２４４Ｃｍ）在聚变－裂变混合堆快裂变包层内嬗变的可行性进了研究。利用一维中子输运和燃耗计算程序ＢＩＤＥＣＡＹ译不同燃料组分的四个快裂变包层进行分析计算。结果表明，在聚变－裂变混合堆快裂变包层内安全，高效地嬗变ＰＷＲ（Ｕ）乏燃料中的超铀元素是可能的。     

混合堆增殖乏燃料组件中子学特性初步研究

本文主要对聚变-裂变混合堆增殖乏燃料在压水堆组件中使用的可能性进行了初步研究。根据聚变 裂变混合堆增殖乏燃料的特点,给出了的聚变-裂变混合堆增殖乏燃料压水堆组件设计方案,分析组件的燃料温度系数、慢化剂温度系数等参数。结果表明：聚变 裂变混合堆乏燃料组件的特性与全铀组件的特性相似。在相同的易裂变同位素质量百分比情况下,本文给出的组件设计方案的功率不均匀系数更小。研究结果可为未来实现聚变 裂变混合堆和压水堆联合循环系统提供技术支持。     

聚变-裂变混合堆的氚工艺问题

文章描述了聚变堆和聚变-裂变混合堆的氚工艺问题。根据聚变堆和聚变-裂变混合堆的特点讨论了对包层氚增殖材料的要求,列举了几种可作氚增殖的合理材料特性。给出了几种从包层提取氚和从废聚变燃料中回收氚的方法。最后对混合堆的氚安全及防护问题进行了讨论。     

聚变—裂变混合堆安全性初探

对聚变－裂变混合堆的安全性进行了初步分析和探讨。主要利用改进后的混合堆放射性程序ＦＤＫＲ对混合堆产生的核废物及放射性进行计算，并将结果与压水堆、高温气冷堆和液态金属冷却快中子增殖堆进行了比较。结果表明，混合堆与裂变动力堆相比有较好的安全性。     

聚变－裂变混合堆安全性初探

对聚变－裂变混合堆的安全性进行了初步分析和探讨．主要利用改进后的混合堆放射性程序ＦＤＫＲ对混合堆产生的核废物及放射性进行计算，并将结果与压水堆、高温气冷堆和液态金属冷却快中子增殖堆进行了比较。结果表明，混合堆与裂变动力堆相比有较好的安全性。     

核能源前景浅见

文章展望了裂变堆、纯聚变堆和聚变-裂变混合堆的前景,分析了混合堆的低聚变条件和很高的能量与燃料增殖能力等重大优点。认为作为由裂变能源过渡到纯聚变能源的桥梁,聚变-裂变混合堆应成为未来核能源的方向之一。     

聚变—裂变混合堆及其在我国核能发展中的作用

本文概要介绍聚变和聚变-裂变混合堆基本原理及其作用。聚变-裂变混合堆可以为压水堆或快堆提供充足的核燃料。它和压水堆或快堆组成的系统具有经济可行性。在解决我国核能发展中燃料短缺问题和促进纯聚变能源的发展方面可望发挥重要的作用。     

托卡马克商用混合堆堆内燃料循环优化设计

简要介绍了聚变－裂变混合堆难内燃料循环研究的方法、程序和程序的改进，提出了适用于托卡马克商用混合堆ＴＣＢ设计的３种堆内燃料循环模型，研究了堆内燃料的装卸模式与增殖燃料２３９Ｐｕ生产量的关系，增殖燃料加浓度的选择，提出了抑制裂变直接加浓核燃料概念，并给出了有关的计算结果。结果表明，在ＴＣＢ设计中，采用抑制裂变直接加浓核燃料模式，可实现年产２３９Ｐｕ燃料２２００ｋｇ且加浓度大于３％。结果还表明，采用分区卸料方式，可有效地减小系统的功率摆动和裂变率，这对商用混合堆设计尤其重要。     

不产生长寿命高放废物的先进核能系统

阐述了核废物分离－－嬗变（Ｐ－Ｔ）处置和先进核能系统（ＡＮＥＳ）和重要性及其物理基础，讨论了对化学分离的要求和现状，并对裂变堆、聚变－裂变混合堆、加速器驱动次临界堆等核废嬗变炉为主的３类先进核能系统作了简要讨论。最后，对我国开展核能系统研究的发展战略提出了建议。     

Gas-cooled fusion-fission hybrid reactor systems

In this paper, the concept of the fusion-fission hybrid reactor is reviewed, and a system of classification for hybrid blanket designs is suggested. The advantages and disadvantages of gas cooling for hybrid reactor systems are discussed and the design implications of using gas cooling in a hybrid blanket are presented. Five of the more complete gas-cooled hybrid reactor conceptual design studies are discussed, and the fission-suppressed hybrid blanket concept is identified as offering potentially significant advantages in terms of inherent safety features and reduced technology development requirements compared to higher power fission blankets. It is concluded that helium is attractive as the coolant for hybrid reactor systems, and that technically viable reactor designs have been developed using helium cooling. The helium-cooled fission-suppressed hybrid blanket, based on thorium fuel for production of²³³U, is identified as being a particularly attractive candidate for further hybrid reactor development work.     

多用途小型堆ACPR100概念设计

中国广核集团提出了一种新的陆上多用途小型堆ACPR100,具有一体化设计、模块化布置、非能动安全、多用途等特点,目前已完成概念设计。本文主要介绍了ACPR100堆芯核设计、子通道热工水力分析、冷却剂系统分析、典型事故分析等研究成果。研究结果表明：ACPR100具备高安全性能、良好的冷却剂系统平衡及符合陆上小型堆用户需求的长周期换料等特点。     

Application of PSA Methodology to Design Improvement of JAERI Passive Safety Reactor (JPSR)

A probabilistic safety assessment (PSA) technique was applied to the design of JAERI Passive Safety Reactor (JPSR). A PSA was performed to clarify safety features and identify vulnerabilities of the original design. Based on the PSA results and considering thermal-hydraulic analyses and experiments, the JPSR design was improved to enhance plant safety. The improved design was re-evaluated with the PSA. Initiating events selected in this study were: large-break LOCA, medium- and small-break LOCAs, SGTR, main steam line break, loss of offsite power, loss of feed water, and other transients. Fault tree analyses were used to evaluate the system unavailabilities. The total core damage frequency due to internal events was estimated to be less than 10^?7/RY. The contribution of high frequency non-LOCA events could be significantly reduced by the design modification. The dominant initiating event was the small break LOCA and the dominant sequence was the failure of residual heat removal system. The present study indicated that the improved JPSR design has sufficient safety margin and the PSA methodology is very effective to improve reactor safety systems in a conceptual design phase.     

微压供热堆HAPPY200总体技术方案

以安全、经济、成熟的核能供热技术为目标,研发了微压供热堆HAPPY200。通过对HAPPY200的总体方案设计、系统关键参数、堆芯方案、热工水力设计、结构方案、主要工艺系统方案、设备方案以及安全评价等方面开展论证和分析,完成了整个核供热系统的概念设计。HAPPY200采用基于大容积水池的安全系统,实现了反应堆系统的非能动安全。HAPPY200的技术方案具有高度安全、系统简化、技术成熟、建造周期短、运行维护费用低及供热品质高等特点,具有广阔的市场前景和市场竞争力。目前已完成HAPPY200的概念设计并确定了示范堆的厂址,正在开展工程设计。     

European lead fast reactor—ELSY

The conceptual design of the European Lead Fast Reactor is being developed starting from September 2006, in the frame of the EU-FP6-ELSY project. The ELSY (European Lead-cooled System) reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, while fully complying with the Generation IV goal of sustainability and minor actinide (MA) burning capability. Sustainability was a leading criterion for option selection for core design, focusing on the demonstration of the potential to be self sustaining in plutonium and to burn its own generated MAs. To this end, different core configurations have been studied. Economics was a leading criterion for primary system design and plant layout. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Low capital cost and construction time are pursued through simplicity and compactness of the reactor building (reduced footprint and height). The reduced plant footprint is one of the benefits coming from the elimination of the Intermediate Cooling System, the low reactor building height is the result of the design approach which foresees the adoption of short-height components and two innovative Decay Heat Removal (DHR) systems. Among the critical issues, the impact of the large mass of lead has been carefully analyzed; it has been demonstrated that the high density of lead can be mitigated by compact solutions and adoption of seismic isolators. Safety has been one of the major focuses all over the ELSY development. In addition to the inherent safety advantages of lead coolant (high boiling point and no exothermic reactions with air or water) a high safety grade of the overall system has been reached. In fact the overall primary system has been conceived in order to minimize pressure drops and, as a consequence, to allow decay heat removal by natural circulation. Moreover two redundant, diverse and passive operated DHR systems have been developed and adopted. The paper presents the overall work performed so far.     

Proposal of Integrated Test Facility for in-Vessel Thermofluid Safety of Fusion Reactors

A vacuum vessel (VV) of a tokamak fusion reactor like the International Thermonuclear Experimental Reactor (ITER) consists the first confinement barrier that includes the largest amount of radioactive materials such as tritium and activation products. The ingress of coolant event (ICE) is a design basis event in the ITER where water is used as the coolant. The loss of vacuum event (LOVA) is also considered as an independent design basis event. Based on the results of ICE and LOVA preliminary experiments, an integrated in-vessel thermofluid test is being planned and conceptual design of the facility is in progress. The main objectives of the integrated test are to investigate the consequences of possible interaction of the ICE and the LOVA and to validate the analytical model of thermofluid events in the VV of the fusion reactor. This paper introduces a conceptual design of the integrated test facility and a testing plan.     

液态金属内单个气泡上升行为的MPS法数值模拟

液态金属冷却核反应堆采用气泡泵的概念设计来提升堆芯自然循环能力。液态金属内气液两相流动特征将直接影响核反应系统一回路的自然循环能力及堆芯安全。本研究通过采用移动粒子半隐式（MPS）方法,对液态金属中单个上升气泡的气泡动力学行为进行数值模拟。分析了铅铋合金中3种初始直径不同的单个氮气泡在上升过程中的气泡形状和速度的变化趋势;对比了初始直径相同的单个氮气泡在液钾、液钠、铅铋合金、钾钠合金和锂铅合金5种液态金属中的上升行为;同时将模拟得到的气泡形状与Grace经验关系图进行了对比,验证了MPS方法数值模拟结果的正确性。     

Safety evaluation of the inherent and passive safety features of the smart design

SMART (system-integrated modular advanced reactor) is a 330 MWt advanced integral PWR, which is under development at KAERI for seawater desalination and electricity generation. The conceptual design of the SMART desalination plant produces 40,000 m³/day of potable water and generates about 90 MW of electricity, which are assessed as sufficient for a population of about 100,000. The SMART enhances safety by adopting the inherent safety design features such as the elimination of large break loss of coolant accidents, substantially large negative moderator temperature coefficients, etc. In addition, the safety goals of the SMART are achieved through the adoption of passive engineered safety systems such as an emergency core cooling system, passive residual heat removal system, safeguard vessel, and reactor and containment overpressure protection systems. This paper describes the design concept of the major safety systems of the SMART and presents the results of the safety analyses using a MARS/SMR code for the major limiting accidents including transient behaviors due to desalination system disturbances. The analysis results employing conservative initial/boundary conditions and assumptions show that the safety systems of the SMART conceptual design adequately remove the core decay heat and mitigate the consequences of the limiting accidents, and thus secure the plant to a safe condition.     

The tokamak hybrid reactor

At a time when the potential benefits of various energy options are being seriously evaluated in many countries throughout the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the Commercial Tokamak Hybrid Reactor (CTHR) [1]. This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of ‘client’ Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode.