FEB-E氚循环系统计算机模拟

对聚变实验增殖堆工程概要设计(FEB-E)的氚燃料循环构造了一个动态子系统模型,研制了模拟氚燃料循环系统的计算机程序SWITRIMcode。计算了10个子系统中运行一年后的氚投料量和整个堆系统总的氚投料。     

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。     

FEB-E氚泄漏分析

由于氚本身非常昂贵而且氚的泄漏将造成环境污染,因此氚的控制对D-T加料的聚变堆来说是非常重要的问题。通常规定聚变堆环境周围总的氚泄漏率应<3.7×10~(11)Bq·d~(-1)。在聚变实验增殖堆FEB-E设计中采用高压氦气冷却,液态锂作氚增殖剂,因而氚与增殖材料之间具有强的化学亲合力。在正常工作条件下,     

基于积分分析方法对聚变堆启动氚量与所需氚增殖比的评估

托卡马克聚变堆的主要发展方式包括混合堆、纯聚变堆。关于托卡马克聚变堆氚自持的研究,国内外主要采用平均滞留时间方法进行研究,并且针对聚变功率较低的混合堆的氚自持研究较少。本工作采用更符合实际的积分分析方法分析了混合堆、纯聚变堆氚自持的启动氚量、氚增殖比(TBR)要求。研究结果表明：启动氚量、备用氚量与聚变功率具有线性关系,所需TBR与聚变功率呈反比例关系;混合堆聚变功率较低,所需TBR较高,工程实现所需TBR挑战较大,需要通过限制长期氚滞留量以降低所需TBR要求;纯聚变堆聚变功率高,所需TBR较低,工程实现所需TBR挑战较小,但备用氚需求达数十千克,应考虑氚系统的冗余设计或提高氚系统的可靠性、可维护性,以降低备用氚的使用规模;运行因子是聚变堆的一个重要设计指标,在此着重分析了运行因子对所需TBR的影响,并重新定义了一个聚变堆氚自持的关系式,以突出运行因子对氚自持的重要影响。     

磁镜聚变增殖堆概念设计初步报告

为满足下世纪上半叶核能迅速发展的需要,设计了为轻水堆提供充足核燃料的磁镜聚变增殖堆CHD。增殖堆能满足10个以上同等规模功率轻水堆的核燃料的需要,它可以在不需要进行再处理的情况下直接加浓燃料。为了抑制靠近等离子体区域的裂变,对压平的功率强度进行了计算。用这种办法,增强了直接加浓聚变增殖堆的燃料生产。为了减少MHD的压降,冷却剂LiPb轴向流入再生区。虽然在反应堆中氚的投料量很低,为了减少氚通过冷却剂管的渗透,必须研制特殊材料。由11个轻水堆电站和一个聚变增殖堆组成的系统的电成本为传统的轻水堆电站的1.05倍。     

磁镜聚变增殖堆概念设计初步报告

为满足下世纪上半叶核能迅速发展的需要,设计了为轻水堆提供充足核燃料的磁镜聚变增殖堆CHD。增殖堆能满足10个以上同等规模功率轻水堆的核燃料的需要,它可以在不需要进行再处理的情况下直接加浓燃料。为了抑制靠近等离子体区域的裂变,对压平的功率强度进行了计算。用这种办法,增强了直接加浓聚变增殖堆的燃料生产。为了减少MHD的压降,冷却剂LiPb轴向流入再生区。虽然在反应堆中氚的投料量很低,为了减少氚通过冷却剂管的渗透,必须研制特殊材料。由11个轻水堆电站和一个聚变增殖堆组成的系统的电成本为传统的轻水堆电站的1.05倍。     

流动锂液帘对堆芯等离子体影响的评估(英文)

运用零维模型评估了流动液态锂幕帘作为聚变实验增殖堆工程概要设计 (FEB-E) 第一壁对聚变等离子体的影响。得到了锂液帘工作温度对堆芯有效平均等离子体电荷?Zeff?,燃料稀释以及聚变功率之间的关系。表明在正常工作情况下,液态锂的蒸发对?Zeff?的影响不是很严重,但对燃料稀释和聚变功率的影响却较为敏感。作为一个例子,对较高功率密度的反剪切位形聚变实验增殖堆FEB-E设计方案 II,计算了液帘的流速与它表面最大温升的关系,结果表明:即便0.5m/s的低速流动液帘第一壁, 蒸发对聚变等离子体的影响也甚微。     

国外核聚变氚工艺研究简介

开发聚变能可从根本上满足人类对能源的需求。聚变氚工艺主要包括包层产氚工艺和废燃料再处理工艺,它是混合堆和聚变堆的关键技术之一。经过二十多年努力,核聚变研究取得很大进展,聚变氚工艺研究日益得到重视。这里对国外聚变氚工艺研究情况作一简要介绍。     

中国氦冷固态增殖剂实验包层系统氚输运初步分析

氚输运分析是开展中国氦冷固态增殖剂实验包层系统安全分析及未来聚变堆氚自持运行的重要研究内容之一。基于氚输运理论和固态增殖剂包层系统设计,利用FDS凤麟核能团队开发的聚变系统氚分析程序TAS,构建了固态增殖剂包层系统氚输运分析系统动力学模型。该模型氚输运结果与文献报道的吻合得很好,误差小于6%,验证了模型的正确性。针对中国氦冷固态增殖剂实验包层系统氚输运问题进行了两种计算方法(稳态、脉冲模式)的初步分析,获得了氚提取系统、氦气冷却系统回路氚分压,实验包层模块冷却流道、窗口室内氚提取系统和氦气冷却系统回路材料中氚滞留量,窗口室内氚提取系统和氦气冷却系统回路氚日渗透量等数据。最终对比结果显示,脉冲模式分析方法能够实时地跟踪源项的快速变化,更符合中国氦冷固态增殖剂实验包层系统实际运行情况。窗口室内氦气冷却系统回路材料中氚滞留量占到日产氚量的31.3%,因此需要在这些氚滞留损失严重的部位考虑适当的阻氚措施。     

氚增殖剂球床组件堆内辐照物理热工计算分析

为验证在中国先进研究堆(CARR)内进行国际热核聚变实验堆(ITER)氚增殖包层模块(TBM)辐照实验的可行性和安全性,进行了氚增殖剂球床组件堆内辐照物理及热工计算分析。氚增殖剂包层模块主要是固态氚增殖剂陶瓷球床。本文采用Monte Carlo粒子输运模拟程序对氚增殖剂球床进行堆内建模,计算球床的中子注量率、能量沉积和产额,得到不同功率下球床的中子注量率、发热功率和产氚速率以及球床组件引入反应堆的反应性。根据物理计算得到的组件各部件发热情况建立热工计算一维模型,通过更改反应堆功率得到满足实验要求的工况并采用三维程序进行验证。物理与热工计算分析的结果表明,在反应堆运行功率为20 MW的工况下球床组件各部件的温度均不超过限值。     

产氚聚变堆概念设计系统研究

提出了基于球环类型的先进氚生产堆概念设计,它是聚变能发展的中间应用。与传统托卡马克氚生产堆不同,设计中利用了球形环的先进等离子体物理性能和紧凑的结构特征,并尽量利用真空室内的空间安置氚生产包层以减少氚泄露而增加氚增殖率,达到年生产氚1000 g的目标,相应的堆利用因子为40％。在2D中子学计算的基础上提出了较为完整的初步概念设计。逐项进行了分析,同时对设计的风险、不确定性和后备方案也做了概括的解释。为下一步更详细、具体的概念设计提供了直接的依据和重要的参考价值。     

Overview of the tritium system of Ignitor

Among the recent design activities of the Ignitor program, the analysis of the tritium system has been carried out with the aim to describe the main equipments and the operations needed for supplying the deuterium–tritium mixtures and recovering the plasma exhaust.

In fact, the tritium system of Ignitor provides for injecting deuterium–tritium mixtures into the vacuum chamber in order to sustain the fusion reaction: furthermore, it generally manages and controls the tritium and the tritiated materials of the machine fuel cycle. Main functions consist of tritium storage and delivery, tritium injection, tritium recovery from plasma exhaust, treatment of the tritiated wastes, detritiation of the contaminated atmospheres, tritium analysis and accountability.

In this work an analysis of the designed tritium system of Ignitor is summarized.     

球环型产氚聚变堆概念设计研究

该研究基于球环类型的先进氚生产堆概念而设计,是聚变能发展的中间应用。与传统托卡马克氚生产堆不同,该设计利用球形环的先进等离子体物理性能和紧凑的结构特征,尽量利用真空室内的空间安置氚生产包层以减少氚泄露而增加氚增殖率,相应的堆利用因子为 40%。在二维中子学计算的基础上提出了较为完整的初步概念设计。在逐项分析的基础上对设计的风险、不确定性和后备方案也做了概括的解释。为下一步更详细具体的概念设计提供了直接的依据,具有重要的参考价值。     

Influence on moisture and hydrocarbons on conversion rate of tritium in catalytic reactors of fusion-DEMO detritiation system

Thoughtful consideration of abnormal events such as fire is required to design and qualify a detritiation system (DS) of a nuclear fusion facility. Since conversion of tritium to tritiated vapor over catalyst is the key process of the DS, it is indispensable to evaluate the effect of excess moisture and hydrocarbons produced by combustion of cables on tritium conversion rate considering fire events. We conducted demonstration tests on tritium conversion under the following representative conditions: (I) leakage of tritium, (II) leakage of tritium plus moisture, and (III) leakage of tritium plus hydrocarbons. Detritiation behavior in the simulated room was assessed, and the amount of catalyst to fulfill the requirement on tritium conversion rate was evaluated. The dominant parameters for detritiation are the concentration of hydrogen in air and catalyst temperature. The tritium in the simulated room was decreased for condition (I) following ventilation theory. An initial reduction in conversion rate was measured for condition (II). To recover the reduction smoothly, it is suggested to optimize the power of preheater. An increase in catalyst temperature by heat of reaction of hydrocarbon combustion was evaluated for condition (III). The heat balance of catalytic reactor is a point to be carefully investigated to avoid runaway of catalyst temperature.     

Tritium management and safety issues in ITER and DEMO breeding blankets

Safe, reliable and efficient tritium management in the breeder blanket faces unique technological challenges. Beside the tritium recovery efficiency in the tritium extraction and coolant purification systems, the tritium tracking accuracy between the inner and outer fuel cycle shall also be demonstrated. Furthermore, it is self-evident that safe handling and confinement of tritium need to be stringently assured to evolve fusion as a reliable technique. The present paper gives an overview of tritium management in breeder blankets. After a short introduction into the tritium fuel cycle and blanket basics, open tritium issues are discussed, thereby focusing on tritium extraction from blanket, coolant detritiation and tritium analytics and accountancy, necessary for accurate and reliable processing as well as for book-keeping.     

A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist.This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into the water coolant of a stratified blanket model, depending on the difference between the required tritium excess inventory and the measured tritium excess inventory. The compounds effectively reduce the amount of low energy neutrons available to react with lithium compounds, thus reducing the tritium breeding ratio. This controller reduces the amount of tritium being produced at the start of the reactor's lifetime and increases the rate of tritium production towards the end of its lifetime. Thus, a relatively stable tritium production level may be maintained, allowing the control system to minimize the stored tritium with obvious safety benefits. The FATI code (Fusion Activation and Transport Interface) will be used to perform the tritium breeding and controller calculations.     

Toward the ultimate goal of tritium self-sufficiency: Technical issues and requirements imposed on ARIES advanced power plants

Due to the lack of external tritium sources, all fusion power plants must demonstrate a closed tritium fuel cycle. The tritium breeding ratio (TBR) must exceed unity by a certain margin. The key question is: how large is this margin and how high should the calculated TBR be? The TBR requirement is design and breeder-dependent and evolves with time. At present, the ARIES requirement is 1.1 for the calculated overall TBR of LiPb systems. The Net TBR during plant operation could be around 1.01. The difference accounts for deficiencies in the design elements (nuclear data evaluation, neutronics code validation, and 3D modeling tools). Such a low Net TBR of 1.01 is potentially achievable in advanced designs employing advanced physics and technology. A dedicated R&D effort will reduce the difference between the calculated TBR and Net TBR. A generic breeding issue encountered in all fusion designs is whether any fusion design will over-breed or under-breed during plant operation. To achieve the required Net TBR with sufficient precision, an online control of tritium breeding is highly recommended for all fusion designs. This can easily be achieved for liquid breeders through online adjustment of Li enrichment.     

Mechanical design of a PERMCAT reactor module

The PERMCAT is a membrane reactor proposed for processing fusion reactor plasma exhaust gas: tritium removal is obtained by isotopic swamping operating in counter-current mode. In this work, a membrane reactor using a permeator tube of length about 500 mm produced via diffusion welding of Pd-Ag thin foils is described. An appropriate mechanical design of the membrane module has been developed in order to avoid any significant compressive and bending stresses on the very long and thin wall permeator tube: two expanded bellows have been applied to the Pd-Ag tube, so that it has been pre-tensioned before operating.The elongation of the metal permeator under hydrogenation has been theoretically estimated and experimentally verified for properly designing the membrane reactor.     

Analysis of the SEAFP Reactor Fuel Cycle

The safety aspects of a fusion reactor fuel cycle, which handles substantial quantities of tritium, have been assessed in the framework of the European Programme on Safety and Environmental Assessment of Fusion Power Long Term (SEAL). This study focused on the assessment of the tritium inventory that could be released from interlinked systems in accidental scenarios. A systematic review of the fuel cycle systems was performed by focusing attention on the main interfaces and to the possible propagation of accident sequences through these interfaces. For the bounding accident sequences identified, deterministic analyses were carried out to determine the accident consequences. Both process source terms (PST) and environmental source terms (EST) were estimated. Simultaneous failure of the primary and secondary containment was considered to be beyond the design basis, nevertheless a preliminary analysis has been carried out; a bounding accident sequence related to a double failure, involving a hydrogen fire, has led to a tritium environmental release of 5.3 g and the wall mechanical load deriving from the maximum hypoth-esizable hydrogen detonation has been defined. Tritium releases into the secondary containment are treated by the appropriate detritiation and by the vent detritiation system. The related EST has been estimated based on an overall tritium cleanup efficiency of 99%, deliberately chosen low to cause the EST to be overestimated. The maximum tritium environmental release is less than 11 g and corresponds to an in-vessel LOCA. For accidents initiating in the fuel cycle only, the maximum tritium release is at most 3.1 g.     

Optimization for the tritium isotope separation factor and permeation by selecting temperature and thickness of the diffusion Pd-Ag alloy cathode

Large quantities of tritiated water will be produced in the controlled fusion reactors for power generation. To eliminate the concentrated tritiated water and for recycling tritium, an industrial electrolyzer was developed. The aim of this paper is to give the design of this electrolyzer and the results for optimization in diffusion and in isotopic exchange by selecting thickness of a thimble-shaped Pd-25%Ag cathode working at high temperature and current. In this process, the tritium recovery system is based on the principle of the tritium diffusion Pd-Ag cathode which produces very pure hydrogen isotopes from enriched tritiated water.