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.     

Overview of R&D activities on tritium processing and handling technology in JAEA

In JAEA, the tritium processing and handling technologies have been studied at TPL (Tritium Process Laboratory). The main R&D activities are: the tritium processing technology for the blanket recovery systems; the basic tritium behavior in confinement materials; and detritiation and decontamination. The R&D activities on tritium processing and handling technologies for a demonstration reactor (DEMO) are also planned to be carried out in the broader approach (BA) program by JAEA with Japanese universities. The ceramic proton conductor has been studied as a possible tritium processing method for the blanket system. The BIXS method has also been studied as a monitoring of tritium in the blanket system. The hydrogen transfer behavior from water to metal has been studied as a function of temperature. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). A new hydrophobic catalyst has been developed for the conversion of tritium to water. The catalyst could convert tritium to water at room temperature. A new Nafion membrane has also been developed by gamma ray irradiation to get the strong durability for tritium.     

Tritium Safety-Related Studies at TPL of JAERI

Activities regarding tritium safety technology in the Tritium Process Laboratory (TPL) at Tokai Establishment of Japan Atomic Energy Research Institute are reviewed. Research and development of a new tritium removal system is being carried out by using a gas separation membrane which enable to make the ITER atmosphere detritiation system more compact and cost-effective. Techniques of gas flowing calorimetry and laser Raman spectroscopy are applied to develop new tritium accountancy methods. Studies of tritium-material interaction, such as plasma material interactions, radiochemical reaction of tritium in gas phase, radiolysis of tritiated water, and waste processing are being carried out under ITER/EDA and U.S.-Japan collaboration. Tritium release experiments for research of tritium behavior in confinements and environment and demonstration of safety related components are planned.     

Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor(CFETR) operating on a Deuterium-Tritium(D-T) fuel cycle. It is necessary to study the tritium breeding ratio(TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder(WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket,the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code(MCNP) and the fusion activation code FISPACT-2007.The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation.In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW.     

Requirements for helium cooled pebble bed blanket and R&D activities

This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R&D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM.The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R&D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R&D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.     

Japanese contribution to the DEMO-R&D program under the Broader Approach activities

Several technical R&D activities mainly related to the blanket materials are newly launched as a part of the Broader Approach (BA) activities, which was initiated by the EU and Japan. According to the common interests for these parties in DEMO, R&Ds on reduced activation ferritic/martensitic (RAFM) steels as structural material, SiC_f/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology are carried out through the BA DEMO R&D program, in order to establish the technical bases on the blanket materials and the tritium technology required for DEMO design. This paper describes overall schedule of those R&D activities and recent progress in Japan carried out by JAEA as the domestic implementing agency on BA, collaborating with Japanese universities and other research institutes.     

Tritium issues to be solved for establishment of a fusion reactor

In order to establish a D–T fusion reactor as an energy source, economical conversion of fusion energy to electricity and/or heat, attaining enough margins in tritium breeding, and insuring tritium safety must be simultaneously achieved. Scientists and researchers working on Tritium in Japan are now tackling with T related problems. Their research subjects can be categorized into two, i.e. researches on “Science and technology” to establish safe and economic Tritium fuel cycle for fusion reactors and “Tritium safety”. Many researchers from various universities, and institutes such as NIFS, JAEA and IEA (Inst. Environmental Science) in Japan are involved in various research programs. In this report, after brief introduction on Tritium related researches in Japan, important T issues to be solved for establishment of a fusion reactor will be summarized considering the handling of large amount of tritium, i.e. fuelling, D–T burning, T inventory, exhausting, refinement, confinement, permeation, leakage, contamination, regulation and tritium accountancy.     

Preliminary tritium safety analysis on China DFLL-TBM for ITER

The dual-functional lithium-lead test blanket module (DFLL-TBM) system was proposed to be tested in ITER. A tritium permeation model of the entire DFLL-TBM system was developed, and the tritium permeation and inventory in DFLL-TBM system were done based on the model during normal operation. Three classes of off-normal situations had been preliminarily analyzed, i.e. in-vessel TBM coolant leaks, in-TBM breeder box coolant leaks and ex-vessel TBM ancillary coolant leaks. The results showed that some issues required significant R&D effort to guarantee the tritium release to the environment below the allowable level, such as the tritium extraction from LiPb and helium coolant and very efficient detritiation system. And more analyses would be carried in the future to further assess the safety of DFLL-TBM.     

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.     

New progress on design and R&D for solid breeder test blanket module in China

ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R&D activities for each TBM module with the auxiliary system are introduced.The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li₄SiO₄ pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R&D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled.     

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.     

Design activities on helical DEMO reactor FFHR-d1

Based on high-density and high-temperature plasma experiments in the large helical device (LHD), conceptual design studies of the LHD-type helical DEMO reactor FFHR-d1 have been conducted by integrating wide-ranged R&D activities on core plasmas and reactor technologies through cooperative researches under the fusion engineering research project, which has been launched newly in NIFS. Current activities for the FFHR-d1 in this project are presented on design window analyses with designs on core plasma, neutronics for liquid blankets, continuous helical magnets, pellet fueling, tritium systems and plasma heating devices.     

A tritium permeation model for conceptual fusion reactor designs

A transient tritium permeation model is developed based on a simplified conceptual DT-fueled fusion reactor design. The major design features described in the model are a solid breeder blanket, a low pressure purge gas in the blanket, and a high pressure helium primary coolant. Tritium inventory in the breeder is considered to be due to diffusive hold-up and solubility effects. It is assumed that diffusive hold-up is the dominant factor in order to separate the solution for the breeder tritium concentration. The model was applied to the STARFIRE-Interim Reference Design, whose system parameters yielded a breeder tritium inventory on the order of grams, based on an average pellet radius of 10^?3 cm. The breeder pellets reach their steady-state tritium content in approximately 1.4×10⁴ s from system start-up, assuming continuous full power operation. Both the steady-state breeder tritium concentration and the time to reach that steady-state are proportional to the pellet radius squared. Other candidate solid breeders were considered, and their effect on the blanket tritium inventory was noted. The addition of oxygen to the primary coolant loop was required in order to keep the tritium losses through the heat exchanger to within the design goal of 0.1 Ci/day.     

聚变堆氚分析程序TAS1.0的设计与开发

氚是聚变堆的重要燃料之一,对聚变堆氚系统进行分析从而实行有效的氚控制是聚变研究的重要内容之一.在中国系列液态金属锂铅包层聚变堆概念设计研究基础上,利用现代软件工程方法及面向对象技术设计思想,发展了聚变堆氚分析程序TAS1.0,可用于聚变堆氚自持分析、氚燃料管理及氚安全性分析与研究,并可为聚变堆包层及燃料循环系统设计与分析提供技术支持.通过一系列的测试校验,表明了该程序的正确性与有效性.本文主要介绍该程序的系统设计、技术特点与程序测试.     

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.     

聚变能源中的氚化学与氚工艺研究进展及展望

核聚变被认为是人类社会未来的理想能源，对社会、经济的可持续发展具有重要的战略意义。氘氚聚变反应具有反应截面大、反应速率高、点火温度低及释放能量大等优点，是目前聚变研究的主要目标，而高效的氘氚燃料循环工艺与技术是实现聚变能源商业应用的基础。本文主要介绍氘氚燃料循环所涉及的等离子体排灰气中氚的快速回收、氚的增殖与提取、大规模氢同位素分离、氚测量等相关氚化学与氚工艺的研究进展及展望，以期对未来聚变能源氚工厂相关技术的研究提供借鉴和参考。     

Isotope Exchange Capacity on Li4SiO4 and Comparison of Tritium Inventory in Various Solid Breeder Blankets

It has been pointed out by the present authors that it is essential to understand such mass transfer steps as diffusion of tritium in the grain of a breeder material, absorption of water vapor into bulk of the grain, adsorption of water on surface of the grain, and exchange capacity of tritium to be trapped to surface of the grain together with two types of isotope exchange reactions for evaluation of the tritium inventory in a solid breeder blanket under various conditions. The isotope exchange capacity on the Li₄SiO₄ surface is experimentally obtained in this study. Most of the properties required for evaluation of the tritium inventory for various blanket materials have been already quantified by the present authors. Then it has become possible to compare the tritium inventory in solid breeder blankets packed with either Li₂O, LiAlO₂, Li₂ZrO₃, Li₂TiO₃ or Li₄SiO₄ using the calculation model previously presented by the present authors.     

氚化学与氚分析进展与展望

氚是聚变堆的重要燃料，氚的问题是制约聚变能源发展的瓶颈问题之一。氚化学中的科学技术问题是解决涉氚工艺的基础，氚分析测量是氚操作中获取氚信息的主要途径。本文综述了近年来氚化学与氚分析方面的研究进展，从氚与材料相互作用、涉氚材料中的氦行为、氚的氢同位素效应、氚的辐射生物效应，以及氚的分析测量方法五个方面对研究进展进行了分析，并对聚变堆中所面临的挑战进行了展望。     

Source Term and Mobilization Assessment in NSSR-1 of ITER

We describe the radioactive sources in the International Thermonuclear Experimental Reactor (ITER). The most important sources are co-deposited tritium, tritiated water, tokamak dust, and corrosion products. The co-deposited tritium is limited to 1 kg-T; the total on-site tritium inventory in the Basic Performance Phase (BPP) is 4 kg-T. Tritiated water concentrations are kept below 0.2 g-T/m³ in the divertor; other coolant loops have lower tritium concentrations. The in-vessel dust inventory is up to 100 kg-W, 100 kg-Be, and 200 kg-C. The activated corrosion product inventory is kept below 10 kg per loop.     

Efficiency and availability driven R&D issues for DEMO

DEMO is the main step foreseen after ITER to demonstrate the technological and commercial viability of a fusion power plant. DEMO R&D requirements are usually identified on the basis of the functions expected from each individual system. An approach based on the analysis of overall plant functional requirements sheds new light on R&D needs. The analysis presented here focuses on two overall functional requirements, efficiency and availability. The results of this analysis are presented here putting emphasis on systems not sufficiently considered up to now, e.g. the heating and current drive systems, while more commonly addressed systems such as tritium breeding blankets are not discussed in detail. It is also concluded that an overall functional analysis should be adopted very early in the DEMO conceptual design studies in order to provide a fully integrated approach, which is an absolute requirement to ensure that the ambitious goals of this device will be ultimately met.