Radiation Damage Study on Various Structural Refractory Alloys of a Multi-Purpose Reactor

Radiation damage properties of structural materials play a key role in design of a fusion–fission (hybrid) reactor. Refractory alloys offer a significant advantage of high neutron wall load capability under fusion neutron environment. In this study, main radiation damage parameters (displacement per atom (DPA) and helium production) on three different refractory alloys, namely W-5Re, TZM (Mo alloy) and Nb–1Zr used as structural material in a hybrid reactor were found. Neutron transport calculations were conducted with the aid of SCALE4.3 System by solving the Boltzmann transport equation with code XSDRNPM. The lowest radiation damage values were obtained for W-5Re alloy. Moreover, all investigated materials will require to be replaced frequently due to their radiation damage values during reactor life (~ 30 years).     

An Acoustically Driven Magnetized Target Fusion Reactor

We propose a new acoustic compression scheme for a MTF power plant. A strong acoustic wave is produced by piston impacts. The wave focuses in liquid PbLi to compress a pre-formed FRC plasma. Simulations indicate the possibility of building an economical 60 MWe power plant. A proof-of-principle experiment produces a small D-D fusion yield of 2000 neutrons per shot.     

聚变堆安全特性评价研究

确保核安全是未来聚变堆设计、建造和运行过程中必须坚持的最高原则,是聚变堆获得建造和运行许可的前提条件,也是聚变能得以吸引公众的主要理由之一。聚变堆具有高能中子、大量放射性氚、复杂结构、极端服役环境等特点,具有独特的潜在安全问题,因而必须开展针对性研究。本文将从聚变中子与放射性源项、热流体与能量传输、氚安全与环境影响、可靠性与风险管理、安全理念与公众接受度五个方面分别总结其安全特性,系统梳理其关键技术挑战,为建立聚变安全评价体系提供技术支持,进而服务于未来聚变堆的设计与建造。     

Design Comparison of Coated Structures for High Heat Flux Components in Fusion Reactor

High heat flux components in fusion reactors are examined from a viewpoint of structural design. Maximum admissible steady heat flux, which can be absorbed and removed by coated structures, was determined by one-dimensional numerical analysis basing on ASME code Sec. HI. Comparison of the current candidate materials is made in order to make heat flux as high as possible. The following conclusions were obtained. (1) Be-Cu structure can be used in order to remove high heat flux beyond 10 MW/m², however, the strong chemical activity of Be gives rise to a problem. (2) SiC-HT-9 and SiC-V alloy structures are promising in case of high heat flux less than 2 MW/m². (3) From thermal stress analysis tungsten and graphite are excellent coating materials and are available for many structural materials. The study of thermal shock resistance and thermal fatigue of these materials is now a problem to be solved.     

Modification of PROMETHEUS Reactor as a Fusion Breeder and Fission Product Transmuter

This study presents the analyses of the fissile breeding and long-lived fission product (LLFP) transmutation potentials of PROMETHEUS reactor. For this purpose, a fissile breeding zone (FBZ) fueled with the ceramic uranium mono-carbide (UC) and a LLFP transmutation zone (TZ) containing the ⁹⁹TC and ¹²⁹I and ¹³⁵Cs isotopes are separately placed into the breeder zone of PROMETHEUS-H design. The neutronic calculations are performed by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. A range of analyses are examined to determine the effects of the FF, the fraction of ⁶Li in lithium (Li) and the theoretical density (TD) of Li₂O in the tritium breeder zone (TBZ) on the neutronic parameters. It is observed that the numerical results obtained from both codes are consistent with each other. It is carried out that the profiles of fission power density (FPD) are flattened individually for each FF (from 3 to 10%). Only, in the cases of FF ≥ 8%, the system is self sufficient from the point of view of tritium generation. The results bring out that the modified PROMETHEUS fusion reactor has capabilities of effective fissile breeding and LLFP transmutation, as well as the energy generation.     

Utilization of Ceramic Uranium Fuels in ARIES-RS Fusion Reactor

This study presents the neutronic performance of the ARIES-RS fusion reactor design using different natural ceramic uranium fuels, namely UO₂, UN or U₃Si₂, dispersed in graphite matrix. These fissionable fuels inserted as micro spheres into the first range quadratic channels at the immediate neighborhood of the first wall in the inboard blanket to amplify fusion power and breed fissile fuel. Neutron transport calculations were performed with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S₈–P₃ approximation. Among the investigated fuels, UN showed the best neutronic performance while UO₂ and U₃Si₂ had similar performances. Numerical results pointed out that inserting fissionable fuel zone even with a small thickness (10 cm) in a pure fusion reactor increased fusion power from 2170 MW to 4500, 5250 and 4150 MW depending on the fuel type. Furthermore significant amount of fissile fuel was produced to be charged to light water reactors.     

Neutron Irradiation Effects on Superconducting Nb-Ti Wires for Fusion Reactor

A systematic study has been carried out on Nb-Ti wire of two different kinds of metallurgical structure, to examine the effects of neutron irradiation on the critical density J_c. The samples used were Nb-47.6 a/0 Ti (sample #A) and Nb-59.8a/0 Ti (sample #B), which were aged at 380°C for 0~10⁴min and irradiated to 1.3×10¹⁸n/cm² (E _n0.1 MeV). The sample temperature during the irradiation is believed to have been below 70°C. The values of J_c of both #A and #B aged up to 50 min were found to increase with irradiation. But when aged beyond 100 min, #B had its value of J_c lowered by the irradiation. The presence of Ti enriched precipitates such as α and ω phases in the samples was surmised from the behavior shown by the critical temperature T_c. The T_c of f #A and #B changed little by irradiation when aged not longer than 100 min, but with aging beyond 500 min, #B showed a decrease in its value of T_c. This decrease indicates that the Ti concentration in the matrix may have increased through radiation-induced breakup of the above-mentioned precipitates, which, in turn, would have brought about the reduction observed in J_c upon irradiation. It is concluded that superconducting Nb-Ti wire with J_c endanced by precipitation does not appear very resistant to neutron irradiation. This underlines the importance of the choice of superconducting materials to be used in fusion reactor magnets.     

聚变堆氚的环境安全评估

对国家863项目聚变实验增殖堆工程概要设计(FEB-E)进行了氚环境安全问题评估。FEB-E是采用液态锂作为包层氚增殖剂，每个包层模块各区之间用隔板隔开．中间通高压氦气冷却、包层第一壁和偏滤器也用氦气冷却。运用自行研制的SWITRIM程序和Sieverts’定律研究了正常工作状态下和事故状态下可能造成氚的环境污染水平。研究表明．正常工作状态下包层液态锂中的氚分压在10^-6～10^-8pa。造成氚环境污染的主要危险来自氚循环回路中的偏滤器子系统的抽出气体泄漏。因此，提高堆芯等离子体燃耗和真空系统设计性能是重要的。     

水冷动力堆用锆合金的疲劳

锆合金是水冷动力堆核燃料元件的包壳材料和堆芯的其它结构材料，在反应堆运行时，堆功率的波动和水冷却介质的流动使燃料组件及其它构件发生循环变形，在极端情况下出现破损。本文概述了堆内锆合金包壳循环变形的特点，并综述了锆合金的循环变形行为，循环变形下的组织结构演化，疲劳裂纹的扩展以及影响疲劳寿命的因素，在此基础上，针对高性能燃料元件的发展趋势，指出了有待进一步研究解决的问题。     

Utilization of Heavy Metal Molten Salts in the ARIES-RS Fusion Reactor

ARIES-RS is one of the major magnetic fusion energy reactor designs that uses a blanket having vanadium alloy structure cooled by lithium [1, 2]. It is a deuterium–tritium (DT) fusion driven reactor, having a fusion power of 2170 MW [1, 2]. This study presents the neutronic analysis of the ARIES-RS fusion reactor using heavy metal molten salts in which Li₂BeF₄ as the main constituent was mixed with increased mole fractions of heavy metal salt (ThF₄ or UF₄) starting by 2 mol.% up to 12 mol.%. Neutron transport calculations were carried out with the help of the SCALE 4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S ₈–P ₃ approximation. According to the numerical results, tritium self-sufficiency was attained for the coolants, Flibe with 2% UF₄ or ThF₄ and 4% UF₄. In addition, higher energy multiplication values were found for the salt with UF₄ compared to that with ThF₄. Furthermore, significant amount of high quality nuclear fuel was produced to be used in external reactors.     

聚变堆螺旋流道液态第一壁流动稳定性分析

聚变堆包层第一壁材料所面临高能粒子辐照、电磁辐射、高热负荷、复杂的机械负荷和相应的物理化学腐蚀制约其服役性能和使用寿命,是聚变能发展的瓶颈问题。液态第一壁由于液态工质自身的特点可以承受更高的热负载、中子壁负载以及更高的出口温度,且由于液态工质的不断更新不存在中子辐照损伤问题,在未来聚变堆应用中很具有吸引力。但由于液态金属在聚变堆强磁场作用下流动形成磁流体(Magnetohydrodynamic MHD)效应,维持液态第一壁在复杂的几何结构和苛刻的工作条件的稳定流动性是现有液态壁研究的难点问题。本文针对自由表面液态金属流动时产生的MHD特性,提出了螺旋流道液态壁流动方案,通过在真空室背壁上设置沿磁场方向的螺旋型流道,使流道内液态金属沿磁场运动,进而减少切割磁场产生的MHD效应。并参考典型聚变堆FDS-Ⅱ,建立了外包层三维模型与真实磁场位型,对方案进行MHD分析与优化,分析结果表明该方案可以在真空室表面形成完整、稳定的液态金属包裹,验证了该方案在磁场作用下液态第一壁流动稳定性与初步可行性。     

Suitability of Small Scale Linear Systems for a Fission–Fusion Reactor, Breeder, and Waste Transmutation

To date the magnetic fusion effort has been almost entirely devoted to only one application, that being a multi gigawatt central station electric plant. Given the already well established fission based industry, the likelihood that fusion will have any impact on curbing the current carbon-based energy demand in the 21st century is slim. This paper advocates that the first and primary use of fusion neutrons should be as the driver for a sub-critical fission nuclear energy system—a fission–fusion hybrid reactor. This system can also be utilized to transmute long-lived radioactive wastes, and breed fissile nuclear fuel for several additional fission reactors. A small-scale fusion system based on a reciprocating fusion cycle employing the magneto-kinetic compression of the Field Reversed Configuration (FRC) is particularly well suited for this application. The characteristics of this fusion neutron driver and the potential for transmutation of long-lived nuclear wastes and breeding of fissile nuclear fuel in a blanket are presented.     

Insertion of lead lithium eutectic mixture in RELAP/SCDAPSIM Mod 4.0 for Fusion Reactor Systems

RELAP/SCDAPSIM Mod 4.0 code was developed by Innovative System Software (ISS) for the analysis of nuclear power plants (NPPs) cooled by light water and heavy water. Later on the code was expanded to analyze the NPPs cooled by liquid metal, in this sequence: lead bismuth eutectic mixture, liquid sodium and lead lithium eutectic mixture (LLE) are inserted in the code. This paper focuses on the insertion of liquid LLE as a coolant for NPPs in the RELAP/SCDAPSIM Mod 4.0 code. Evaluation of the code was made for a simple pipe problem connected with heat structures having liquid LLE as a coolant in it. The code is predicting well all the thermodynamic and transport properties of LLE.     

Fusion Power—A Chemical Engineering View of the Integrated Enterprise

The purpose of this article was to achieve the beginning of an understanding of the integrated fusion enterprise from raw materials through power generation to decommissioning and waste disposal. The particular view point is that of a technically trained person who is only casually acquainted with the field. Emphasis is given to the chemical engineering aspects of controlled fusion power. It is concluded that there are indeed many areas in which the discipline of chemical engineering may contribute to the fusion effort. These areas include separation technology by physical and chemical means, heat and mass transfer in a packed bed blanket, tritium removal from molten coolants, distillation technology for isotope separation, and preparation of deuterium and lithium feed materials.     

Effect of Lithium Enrichment on the Tritium Breeding Characteristics of Various Breeders in a Fusion Driven Hybrid Reactor

Selection of lithium containing materials is very important in the design of a deuterium–tritium (DT) fusion driven hybrid reactor in order to supply its tritium self-sufficiency. Tritium, an artificial isotope of hydrogen, can be produced in the blanket by using the neutron capture reactions of lithium in the coolants and/or blanket materials which consist of lithium. This study presents the effect of lithium-6 enrichment in the coolant of the reactor on the tritium breeding of the hybrid blanket. Various liquid–solid breeder couples were investigated to determine the effective breeders. Numerical results pointed out that the tritium production increased with increasing lithium-6 enrichment for all cases.     

网格权窗减方差技术及其在聚变堆屏蔽分析中应用研究

在聚变堆辐射屏蔽计算中,如何有效解决深穿透问题是近年来国际聚变辐射安全领域关注的焦点之一。针对该问题,本文研究了直角坐标系与圆柱坐标系下基于网格的权窗减方差技术。本文基于超级蒙卡核模拟软件系统SuperMC实现了该方法,并选取减方差技巧的基准例题进行测试与分析,初步得出"粗划真空或密度很小的区域、细分密度大的区域"的网格划分规律,能有效提高网格权窗计算效率。基于该规律对聚变屏蔽基准问题进行对比分析,新的网格划分与原始网格划分的计算效率相比,FOM因子提高了1.92倍。减方差技巧的基准例题和聚变屏蔽基准问题计算中,SuperMC通量计算结果与MCNP相比偏差均在0.5%以下,证明了本文中方法的正确性。     

聚变数据库系统FusionDB研发与应用

聚变研究和设计是一项需国内外广泛合作的系统工程,积累和共享数据是当前重要任务。为了更好地整合聚变数据,FDS团队设计和研发了集聚变数据和数据处理与分析软件于一体的聚变数据库系统FusionDB,系统涵盖了聚变堆设计与安全分析关键数据,是国际上首个包括核数据、材料数据、部件数据、聚变物理实验数据以及核计算仿真和可靠性与概率安全分析等功能的综合型聚变数据库系统。FusionDB已应用于国际热核聚变实验堆ITER、中国科学院FDS系列聚变堆概念设计与研究中。     

Investigation on the Neutronic Performance of a Fusion Reactor Using Flibe with Heavy Metal Fluorides

Using liquid wall between the plasma and solid first wall in a fusion reactor allows to use high neutron wall loads and could eliminate frequent replacement of the first wall structure during reactor’s lifetime. Liquid wall should have a certain effective or optimum thickness to extend solid first wall lifetime to reactor’s lifetime and supply sufficient tritium for deuterium–tritium (DT) fusion driver. This study presents the effect of thickness of flowing liquid wall containing 90 mol % Flibe+10 mol % UF₄ or ThF₄ on the neutronic performance of a magnetic fusion reactor design called APEX. Neutron transport calculations were carried out with the aid of code Scale4.3. Numerical results brought out that optimum liquid wall thickness of ∼38 cm was found for the blankets using Flibe+10% UF₄ whereas, 56 cm for that with Flibe+10% ThF₄. Significant amount of high quality fissile fuel was produced by using heavy metal salt.     

Thermohydraulic and mechanical design assessment of tube-panel first wall for Tokamak Fusion Power Reactor

A first wall structural concept cooled by high temperature and pressurized water has been proposed for the Tokamak Fusion Power Reactor (SPTR-P). Among a number of candidate design concepts, a tube-panel structure was selected for the first wall design. Stainless steel serves as the first wall structural material. The first wall is separated from the blanket wall and has a circular cross-section coolant channel since this shape is the most desirable for resisting the mechanical load due to the pressurized cooling water. Feasibility of the thermohydraulic and mechanical design has been established by analyses under steady-state operating conditions. The effect of the heat load during plasma disruptions on the thermomechanical characteristics of the first wall has been clarified. The mechanical strength of the first wall of power reactor is inadequate to withstand the thermal load expected during plasma disruption in an experimental reactor.On leave from Kawasaki Heavy Industries Ltd.     

脉冲堆燃料的安全特性及其在小型动力堆中的应用

本文介绍了脉冲堆U-ZrH_(1.6)燃料的安全特性及具在国外小型动力堆中的有关应用情况。最终指出,U-ZrH_(1.6)是一种完美的燃料概念,除脉冲堆外,在小型动力堆中有着广阔的应用前景。