球形托卡马克堆嬗变中子学计算的比较研究

基于对球形托卡马克（ST）聚变堆的研究，提出了ST聚变－嬗变堆的设计概念。运用一维输运燃耗计算程序BISON3.0进行了优化设计，确定了适合于嬗变少额锕系MA核素的堆芯等离子体参数、包层结构及合适的换料周期。在一维计算的基础上，运用二维中子学程序TWODANT进行了二维中子输运计算；结合TWODANT给出的中子通量，运用一维放射性计算程序FDKR进行了燃耗计算，并给出了有关的计算结果。     

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根据HCSB-DEMO堆的设计要求,对不同尺寸的聚变堆能产生的聚变功率、中子壁负载和等离子体燃烧时间等进行计算与分析,给出了符合设计要求的堆芯参数。在所选定的堆芯参数条件下进行了零维功率平衡计算分析,给出了3组HCSB-DEMO堆的等离子体初步设计参数。     

非自持和非麦氏分布情况下托卡马克堆芯等离子体功率平衡

本文给出一组托卡马克堆芯等离子体功率平衡方程,它描述了在非自持和非麦氏分布情况下等离子体功率平衡的问题,计算结果表明。在现实可行的聚变堆工程和相应的等离子体参数情况下,堆芯中每秒可产生聚变中子PN大于10~19。     

ST嬗变堆中心柱的中子学及中子辐照效应研究

基于先进核数据库ENDF/B-VI和计算机程序对ST嬗变堆中心柱的中子学及辐照损伤的二维分析计算结果，分别对中心柱导体因中子辐照影响而引起电阻率、电阻、电流和欧姆电阻功率等的沿径向不均匀分布，以及辐照损伤对中心柱热工水力问题及更换寿命的影响进行了分析和计算。结果表明，中子辐照直接改变了中心柱导体材料的电阻率分布。热工-水力学分析和计算表明，电流不均匀分布可显著地延长中心柱的使用寿命，并估算出ST嬗变堆中心柱设计的更换寿命大约8年。     

快堆中子学计算程序NECP-SARAX的改进与验证

NECP-SARAX是西安交通大学NECP团队开发的用于快中子反应堆的中子学程序系统。为准确处理快中子反应堆中中等质量核素散射共振以及非弹性散射导致的复杂的中子慢化效应,SARAX程序最初采用连续能量的蒙特卡罗方法计算中子能谱从而获得堆芯计算使用的有效多群截面。由于蒙特卡罗程序计算效率低,且在低能量段统计偏差较大,提出采用基于点截面的超细群方法计算中子慢化能谱,避免了蒙特卡罗方法产生参数时存在的缺陷。堆芯计算采用多群中子输运,通过优化简化几何建模,改进了程序的实用性。采用多种微扰方法计算堆芯各种反应性系数,提出了基于中子输运微扰理论的虚拟密度方法以计算堆内组件变形导致的反应性变化。在进行堆芯瞬态计算时,采用了点堆和改进准静态两种方法,可用于一般快堆和快谱ADS的典型事故分析。OECD发布的一系列快堆基准题测试表明,SARAX程序在快堆计算中具有良好的精度,达到了与国外著名快堆程序相当的水平。有效增殖因子与连续能量的蒙卡计算结果相比偏差在300 pcm以内。同时,由于引入了虚拟密度理论和三维时空动力学模型,程序功能更加完善,可以更好地满足快堆工程设计的需求。     

聚变中子学的蒙特卡罗PVM并行模拟在ITER中子诊断设计中的应用

聚变能源很可能是人类文明得以维持发展的新型能源。未来的氘氚聚变堆的结构和工程设计很大程度上依赖于以聚变中子学为基础的计算。在过去的十余年中，很多的核数据库如FENDL和JENDL的检验工作围绕ITER设计而展开。聚变中子学计算包括中子和光子的输运计算。其计算目标是提供反应率和能谱等重要的信息。一维或二维的聚变中子学解析计算能提供一定精度的结果和高效率的优化设计，但对于一个三维的聚变托卡马克反应堆来说，只有蒙特卡罗方法能提供较精确的数值模拟结果。MCNP程序是由LANL实验室发展的用于中子和光子的蒙特卡罗计算的大型程序。PVM的并行计算环境能提高为MCNP程序的运行执行效率。     

次临界能源堆物理设计进展

聚变-裂变混合能源堆包括聚变中子源和次临界能源堆,主要目标是生产电能。回顾了国内外混合堆的发展历史,给出混合能源堆设计的边界条件和约束条件,说明次临界能源堆以铀锆合金为燃料、水为冷却剂的设计思想。利用输运燃耗耦合程序MCORGS计算了混合能源的燃耗,给出了中子有效增殖因数、能量放大倍数和氚增殖比等物理量随时间的变化。通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点。论述了混合堆的热工设计并进行了安全分析。对于燃耗数值模拟程序,通过多家对算,保证其计算结果的可信性。针对次临界能源堆的特点,利用贫铀球壳建立了贫铀聚乙烯装置和贫铀LiH装置,并且专门设计加工了天然铀装置,开展铀裂变率、造钚率、产氚率等中子学积分实验,验证了数值模拟的可靠性。     

常规导体托卡马克聚变核科学装置的堆芯参数研究

采用自主研发的零维系统程序ZERO-CODE，对基于铜线圈托卡马克概念的聚变装置开展了堆芯关键参数的设计。根据中子峰值壁面负载及相关物理、工程要求，确定了一组具有最优造价的堆芯参数。     

高温气冷堆堆芯中子时空动力学模拟计算

研究开发三维圆柱几何堆芯多群中子时空动力学改进准静态方法模拟计算程序.对给定的模块式高温气冷堆模型进行模拟计算.初始状态下,计算结果与中子扩散程序CITATION吻合很好.动态情况下,模拟堆芯反应性、堆内各能群中子平均注量率和堆芯相对功率等物理量随时间变化,计算结果与理论分析一致.     

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采用球形托卡马克(ST)等离子体位形,对氦冷嬗变包层、钠冷嬗变包层、氟锂铍(FLiBe)熔盐冷嬗变包层三种嬗变包层中子学方案进行了初步计算分析,并就各自的中子学特性进行了比较.结果表明,从嬗变长寿命放射性锕系核素237Np的角度考量,FLiBe冷却嬗变包层的嬗变性能最优.对氦冷嬗变包层的计算结果表明,通过改变初装料时237Np在次锕系元素中的成分比例,可使包层在比较长的运行时间(9.62年)内,保持稳定有效增殖系数、稳定功率、稳定产氚率.     

Activation Characteristics of Fuel Breeding Blanket Module in Fusion Driven Subcritical System

Shortage of energy resources and production of long-lived radioactivity wastes from fission reactors are among the main problems which will be faced in the world in the near future. The conceptual design of a fusion driven subcritical system (FDS) is underway in Institute of Plasma Physics, Chinese Academy of Sciences. There are alternative designs for multi-functional blanket modules of the FDS, such as fuel breeding blanket module (FBB) to produce fuels for fission reactors, tritium breeding blanket module to produce the fuel, i.e. tritium, for fusion reactor and waste transmutation blanket module to try to permanently dispose of long-lived radioactivity wastes from fission reactors, etc. Activation of the fuel breeding blanket of the fusion driven subcritical system (FDSFBB) by D-T fusion neutrons from the plasma and fission neutrons from the hybrid blanket are calculated and analysed under the neutron wall loading 0.5 MW/m2 and neutron fluence 15 MW.yr/m2. The neutron spectrum is calculated with the worldwide-used transport code MCNP/4C and activation calculations are carried out with the well known European inventory code FISPACT/99 with the latest released IAEA Fusion Evaluated Nuclear Data Library FENDL-2.0 and the ENDF/B-V uranium evaluated data. Induced radioactivities, dose rates and afterheats, etc, for different components of the FDS-FBB are compared and analysed.     

Axial Neutron Flux Evaluation in a Tokamak System: a Possible Transmutation Blanket Position for a Fusion–Fission Transmutation System

A sub-critical advanced reactor based on Tokamak technology with a D–T fusion neutron source is an innovative type of nuclear system. Due to the large number of neutrons produced by fusion reactions, such a system could be useful in the transmutation process of transuranic elements (Pu and minor actinides (MAs)). However, to enhance the MA transmutation efficiency, it is necessary to have a large neutron wall loading (high neutron fluence) with a broad energy spectrum in the fast neutron energy region. Therefore, it is necessary to know and define the neutron fluence along the radial axis and its characteristics. In this work, the neutron flux and the interaction frequency along the radial axis are evaluated for various materials used to build the first wall. W alloy, beryllium, and the combination of both were studied, and the regions more suitable to transmutation were determined. The results demonstrated that the best zone in which to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements of W alloy/W alloy and W alloy/beryllium would be able to meet the requirements of the high fluence and hard spectrum that are needed for transuranic transmutation. The system was simulated using the MCNP code, data from the ITER Final Design Report, 2001, and the Fusion Evaluated Nuclear Data Library/MC-2.1 nuclear data library.     

加速器驱动快/热耦合次临界系统的概念设计

对加速器驱动快／热耦合次临界系统进行了概念设计研究。在该系统中,内区的快包层和外区的热包层是相互独立的,快、热包层之间为空腔和B4C包层以实现单向耦合。快包层装以合金（MA＋Pu）Zr为燃料,热包层初始循环装以氧化物（Th＋Pu）O₂为燃料,平衡循环装以（Th＋^233 U＋Pu）O2为燃料。^99Tc,^129I和^135Cs分别以单质、NaI和CsCl的形式装入热包层。该系统具有较高的能量放大倍数、嬗变效率和燃料转换比:系统能量放大系数不低于320;锕系元素（MA）和裂变产物（FP）的嬗变支持比分别为1个和2个压水堆;热包层的燃料转换比为0．715。 Accelerator driven coupled fast/thermal subcritical system is conceptually designed. In the system, the inner/fast blanket and the outer/thermal blanket are separated each other by large vacuum and B4C coating for on edirection coupling. The metal type fuel （MA ＋ Pu）Zr is loaded into the fast blanket. The oxide type fuels （Th ＋ Pu） O2 and （Th ＋ ^233U ＋ Pu）O2 are loaded into the thermal blanket during the initial cycle and the equilibrium cycle, respectively. ^99Tc, ^129I and ^135Cs are loaded respectively in the form of pure technetium metal, sodium iodide and cesium chlorine into the thermal blanket. The system has good transmutation efficiency, high energy amplification factor and good fuel conversion ability: the energy amplification factor is above 320; the transmutation support ratios of MA and FP are about 1.0 and 2.0 PWRs respectively; the fuel conversion ratio in the thermal blanket is about 0. 715.     

Transmutation of 129I, 237Np, 238Pu, 239Pu, and 241Am using neutrons produced in target-blanket system ‘Energy plus Transmutation’ by relativistic protons

Target-blanket facility ‘Energy + Transmutation’ was irradiated by proton beam extracted from the Nuclotron Accelerator in Laboratory of High Energies of Joint Institute for Nuclear Research in Dubna, Russia. Neutrons generated by the spallation reactions of 0.7, 1.0, 1.5 and 2 GeV protons and lead target interact with subcritical uranium blanket. In the neutron field outside the blanket, radioactive iodine, neptunium, plutonium and americium samples were irradiated and transmutation reaction yields (residual nuclei production yields) have been determined using γ-spectroscopy. Neutron field's energy distribution has also been studied using a set of threshold detectors. Results of transmutation studies of ¹²⁹I, ²³⁷Np, ²³⁸Pu, ²³⁹Pu and ²⁴¹Am are presented.      

磁约束聚变堆次临界包层概念设计研究进展

为使磁约束聚变堆实现能量放大与氚自持,在其等离子体区周围设置次临界包层和产氚包层。采用天然铀合金燃料、轻水作冷却剂兼慢化剂,内嵌压力管式的次临界包层设计方案,通过对包层物理性能、结构概念设计、热工水力性能和安全分析,表明该方案可将聚变能量放大10倍以上,氚增殖比大于1.15,具有天然的临界安全性和良好余热安全性能。立足于近中期可利用的聚变技术,力争实现聚变能源的提前商用,为我国能源可持续发展提供一种有竞争力的技术选项。     

Fuel cycle for a fusion neutron source

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m³Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.     

美国激光惯性约束聚变能源研究综述

简要地介绍了美国激光惯性约束聚变能源( LIFE ) 的研究现状与发展前景。基于美国国家点火装置( NIF ) 的近期进展,美国利弗莫尔实验室提出了激光惯性约束聚变能源设想,并开始了分解研究。设想用新型二极管泵浦固体激光器产生1.4~2.0 MJ 的激光能量,靶丸聚变增益25~30,打靶频率10~15Hz,实现350~500 MW聚变功率,相当于聚变中子源强1.3×1020 ~1.8×1020 n/s。以此驱动次临界裂变包层,使能量再倍增4~10 倍,实现1 GW电功率的输出。采用创新设计的燃料元件,包层可达到90%以上的燃耗深度,形成一个安全、无碳、燃料资源丰富、核废料少、可持续发展的新型核能源系统。In this paper the present study situation and prospect of the American laser-based Inertial Confinement Fusion Energy ( LIFE ) are briefly introduced. It is based on recent progress of National Inertial Facility ( NIF ) and related research have begun. On the assumption of using laser energy of 1.4 to 2.0 MJ, the target fusion gain G=25~30, the repetition rate 10 to 15 Hz, the fusion power of 350 to 500 MW or neutron source power of 1.3×1020 to 1.8×1020 n/s could be achieved. For a sub-critical fission blanket driven by this fusion neutrons power, energy multiplication M of 4~10 and several GW of thermal power could be obtained. By novel design on fuel pins, burnup more than 90% would be achieved for heavy metals in the blanket. Inertial Confinement Fusion-fission energy is a promising concept, which characterized by inherent safety, richness in nuclear fuel resources, minimization of nuclear waste, non-CO2 emitting ,and it is a sustainable energy source.     

核用结构材料SIMP钢低活化特性的分析与比较

在中国聚变工程实验堆(CFETR)水冷陶瓷增殖(WCCB)包层的设计条件下,对水冷包层应用SIMP 钢,使用蒙特卡罗中子输运程序MCNP 与欧洲研制的材料活化计算程序FISPACT 耦合计算,分析了SIMP 钢结构材料的放射性比活度、衰变热、接触剂量率和辐照损伤等。通过与EUROFER-97、F82H 等多种低活化钢的对比发现,SIMP 钢在多个活化结果上达到国际上认可的低活化铁素体/马氏体钢(RAFM)的低活化特性。因此,SIMP 钢可作为未来聚变堆包层的候选结构材料。     

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Using a one-dimensional (1D) neutronics model, the neutronics performance in the China fusion engineering test reactor (CFETR) with latest design dimensions of vacuum vessel is calculated under the 2GW fusion power. The shielding effect of neutron reflecting material ZrH₂ on neutrons is calculated, and it is found that the 20cm reflector can shield 94.3% neutron fluence and 94.9% neutron nuclear heat. Meanwhile, the minimum shield blanket thickness corresponding to different neutron wall loads is calculated when CFETR is operated at 10FPY (full power year) and 20FPY. The results show that the minimum shield blanket thickness are 44cm, 53cm, and 65cm corresponding to the neutron wall loads with 1.0MW·m⁻², 1.5MW·m⁻², and 2.5MW·m⁻² respectively after the device is operated at 10 FPY; whereas the shielding blanket needs to be thicker in the radial direction to meet the neutron shielding requirements after the device is operated at 20FPY. The optimized size of the shielding blanket provides a significant reference for the design of CFETR advanced blanket.