Overview of Safety and Environmental Issues for Inertial Fusion Energy

This paper summarizes safety and environmental issues of Inertial Fusion Energy (IFE): inventories, effluents, maintenance, accident safety, waste management, and recycling. The fusion confinement approach among inertial and magnetic options affects how the fusion reaction is maintained and which materials surround the reaction chamber. The target fill technology has a major impact on the target factory tritium inventory. IFE fusion reaction chambers usually employ some means to protect the first structural wall from fusion pulses. This protective fluid or granular bed also moderates and absorbs most neutrons before they reach the first structural wall. Although the protective fluid activates, most candidate fluids have low activation hazard. Hands-on maintenance seems practical for the driver, target factory, and secondary coolant systems; remote maintenance is likely required for the reaction chamber, primary coolant, and vacuum exhaust cleanup systems. The driver and fuel target facility are well separated from the main reaction chamber.     

核设施安全监管与辐射环境安全监管

从历史沿革、管理对象、技术基础、区域覆盖和监管手段等5个方面比较了核设施安全监管和辐射环境安全监管的区别,分析了二者在根本目标一致性、顶层管理统一性和职能分工互补性上的联系,阐释两种监管共存的意义。     

核设施退役的环境安全

核设施退役的环境安全是退役的最终目标,是制约整个退役活动的关键,已成为世界各国公众关注的问题。我国在核设施退役活动中,在方案设计中采取了有效的安全措施,且退役活动严格遵照预定的安全措施实施,因而退役活动对环境安全的影响完全控制在预定值以下,对公众及其后代是安全的。也提到了可能对环境安全造成影响的核设施退役活动。本文对我国已实施的几项核设施退役活动对环境安全影响进行了初步探讨,目的是为了引起有关人士的共鸣,达到深入讨论,逐步统一认识,推进我国核设施退役工作的进展。     

Fusion safety and environmental R&D tasks in Japan

This paper aims at listing and evaluating the status of all the research and development (R&D) tasks necessary for the construction of a safe and environmentally benign fusion experimental reactor. At this time, it is not possible to define precisely the R&D tasks necessary for the licensing approval and those that are useful in improving safety but not necessarily required for licensing because the licensing procedure itself is still being discussed. Among the R&D tasks, the most important are considered to be those related to tritium safety, namely, those effective in reducing the uncertainty in tritium inventory in the plasma facing components and blanket, uncertainty in tritium permeation and leakage, and those to clarify tritium behavior in the containment and in the environment. The R&D tasks with second priority are judged to be those related to mobilization of the activation products such as activated erosion dust or the corrosion products. The volatilization of structural metal caused by the oxidation at high temperature seems to be highly unlikely but some experiments are needed to assure that this is the case.     

我国核安全法规概述

概述了我国核安全法规的范围及体系结构,从国家法律、国务院行政法规、部门规章、指导性文件四个层面介绍了我国现行核安全法规体系的内容,以及目前开展的核安全法规编写和修订工作.     

Progress of IRSN R&D on ITER Safety Assessment

The French “Institut de Radioprotection et de S?reté Nucléaire” (IRSN), in support to the French “Autorité de S?reté Nucléaire”, is analysing the safety of ITER fusion installation on the basis of the ITER operator’s safety file. IRSN set up a multi-year R&D program in 2007 to support this safety assessment process. Priority has been given to four technical issues and the main outcomes of the work done in 2010 and 2011 are summarized in this paper: for simulation of accident scenarios in the vacuum vessel, adaptation of the ASTEC system code; for risk of explosion of gas-dust mixtures in the vacuum vessel, adaptation of the TONUS-CFD code for gas distribution, development of DUST code for dust transport, and preparation of IRSN experiments on gas inerting, dust mobilization, and hydrogen-dust mixtures explosion; for evaluation of the efficiency of the detritiation systems, thermo-chemical calculations of tritium speciation during transport in the gas phase and preparation of future experiments to evaluate the most influent factors on detritiation; for material neutron activation, adaptation of the VESTA Monte Carlo depletion code. The first results of these tasks have been used in 2011 for the analysis of the ITER safety file. In the near future, this R&D global programme may be reoriented to account for the feedback of the latter analysis or for new knowledge.     

An Overview of Tokamak Alternatives in the US Fusion Program with the Aim of Fostering Concept Innovation

The US fusion program has operated for just over 50 years, during which time the tokamak has emerged as the most promising vehicle for a burning plasma experiment. However, many other concepts have been built and investigated as alternatives (and possible improvements) to the tokamak, perhaps to make energy from fusion an economic reality sooner. This Paper is an overview of the conventional alternatives to the tokamak and a set of those that are not so conventional with the aim of fostering concept innovation. Usually the devices are grouped into magnetic, inertial, electrostatic, or other categories, with sub-categories. Here, the groupings of conventional- and non-conventional-alternatives are used too. The conventional alternatives are those devices that have been adopted as serious alternatives, and for which many references are immediately available (e.g. rfp, mirror, stellarator, spheromak, laser ICF, etc). The non-conventional alternatives comprise approaches that are not being currently investigated or are worth consideration. In this grouping lie ideas like impact fusion, muon catalyzed fusion, and many historical ones (like the Elmo Bumpy Torus). Several examples of the physics of non-conventional alternatives are presented in summary form as examples of skunkworks in the hope that others will take up the challenge of concept innovation.     

Fusion Safety Research and Development in JAERI

This paper describes the current status and future plans of the fusion safety research and development regarding to the developments of the dust removal system and safety analysis code and the thermofluid experiments in the Japan Atomic Energy Research Institute (JAERI) for a fusion experimental reactor. The containment of the radioactive material is the key to achieve fusion safety. In the event of accidents, the source terms need to be evaluated with sufficient accuracy. Therefore, in JAERI, the dust characterization have been investigated and the dust removal system using electric force has been developed and tested. A safety analysis code including both thermal and plasma transient analyses under the various event sequences has been developed. Moreover, the preliminary experiments of thermofluid transients in the vacuum vessel such as Ingress of Coolant Event (ICE) and Loss of Vacuum Event (LOVA) have been started and the experimental results using preliminary LOVA/ICE apparatus during 1995–1996 are summarized in this paper.     

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。     

Fusion Safety Studies in Russia in 1996

The paper seeks to provide a summary report of observations and results of some Russian fusion safety studies performed in 1996. Release of tritium and helium from neutron irradiated beryllium at relatively high neutron fluences has a burst nature. With the growth of the beryllium temperature-increase rate to 90 K/s, the temperature of tritium burst release decreases from 800 to 450–500°C and for helium decreases from 1200 to 500°C. Characterization of carbon and tungsten dust produced in experiments simulating plasma disruptions revealed that dust particle distribution of sizes for graphites and carbon fiber composites has a bimodal nature with maxima in the range of 0.01–0.03 and 2–4 m for composite UAM and in the range of 0.14–0.18 and 2–4 m for graphite MPG-8. Chemical reactivity of beryllium with air was studied as well. A mathematical model for beryllium weight gain under its chemical interaction with air at temperatures of 700–800°C as a function of beryllium porosity, temperature, and interaction duration was developed.     

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。     

Characteristics and Contributions of the Three Major United States Toroidal Magnetic Fusion Facilities

This is Volume 1 of the report of a panel established by the U.S. Department of Energy Fusion Energy Sciences Advisory Committee (FESAC) and submitted in July 2005. The panel was charged to answer the following questions: What are the unique and complementary characteristics of each of the major U.S. fusion facilities? How do the characteristics of each of the three U.S. fusion facilities make the U.S. toroidal research program unique as a whole in the international program? How well do we cooperate with the international community in coordinating research on our major facilities and how have we exploited the special features of U.S. facilities in contributing to international fusion research, in general, and to the ITER design specifically? How do these three facilities contribute to fusion science and the vitality of the U.S. Fusion program? What research opportunities would be lost by shutting down one of the major facilities?     

核动力装置稳压器安全阀可靠性建模

基于对核动力装置稳压器安全阀系统功能及任务剖面的分析,建立了基本可靠性框图和任务可靠性框图,并采用普通慨率法得到了相应的可靠性数学模型.该可靠性模型为下一步进行故障模式、影响及危害性分析(FMECA)和可靠性分配提供依据.     

Environmental and Safety Aspects of Using Tritium in Fusion

Today most current fusion research and development activity is based on the expectation that the D–T reaction will be used for the first generation of fusion reactors. This mixture is the premier candidate for a fusion fuel on account of its outstanding energy gain. Fusion reactors will produce neither the problematic emissions now experienced from fossil-fuel-burning power plants nor the long-lived fission products and transuranic elements resulting from fission reactors. Even though, tritium is a radioactive isotope of hydrogen with 12.32 years half-life that exposes beta radiation. Although its specific activity is relatively weak but because of its gaseous state, it can leak easily from its container and contaminate its surrounding. In order to improving safety and reliability of fusion reactors, research groups jointly investigate radiation hazards resulting from the release of tritium and activation products during normal operations as well as accidental conditions. In this paper, some of the most significant safety and environmental aspects of tritium is briefly discussed.     

Environmental Radiation Safety Assessment in the Control Area at Nuclear Facilities

Atomic Energy - Procedural recommendations are made for a radiation safety assessment of the environment as well as the results of their approval based on monitoring of the radiation situation for...     

Overview of the Preliminary Safety Analysis of the National Ignition Facility

The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, New Mexico, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 individual laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF has been classified as a low hazard, radiological facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis report be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A Preliminary Safety Analysis Report (PSAR) has been approved, which documents and evaluates the safety issues associated with the construction, operation, and decommissioning of the NIF.     

Fusion Breeding: An Old,New Strategic Opportunity for Fusion

This article reviews the case for fusion breeding that the author has made over more than the last 15 years. It shows that even stipulating maximum success or ITER or NIF, pure fusion is nowhere near ready to provide economical power. Fusion breeding, using a device like one or both of these machines just might. A new direction for the American MFE base program is proposed. Fusion breeding could be a short cut which could cut out decades and decades from the development of pure fusion. Fusion breeding allows a midcentury power source, one which is carbon free, sustainable, economical, environmentally sound and have litte or no proliferation risk.     

国内外聚变核安全监管与许可初步分析研究

中国政府高度重视聚变发展,安全是聚变能发展的生命线,而核安全监管和相关许可制度是确保聚变能安全发展的必要手段。聚变堆具有其独特的安全特性,无法完全照搬目前基于裂变堆建立起来的法律法规等监管制度。本文初步梳理了国际(含ITER、国际原子能机构、国际能源署、欧盟、美国、韩国等)关于聚变核安全监管和许可的研究进展和相关经验,总结了我国目前在聚变核安全监管与许可方面的现状与存在的问题,为我国聚变核安全监管提出了发展建议。     

Requested Information Provided by the Three Major United States Toroidal Magnetic Fusion Facilities: Report of the 2005 FESAC Facilities Panel, Vol. 2

This is Volume 2 of a report of a panel established by the U.S. Department of Energy Fusion Energy Sciences Advisory Committee (FESAC) charged to review the three major U.S. fusion facilities. The Panel requested input from each of the three major U.S. toroidal magnetic fusion facilities. The request included an invitation to each facility program director to provide a document that addressed in detail the panel charge. This paper consists of the three documents that were received in response to that request.