Tritium Technology Research Articles

Recent developments from the European tritium handling programme

Several important projects of tokamaks have been undertaken worldwide and some of them inside the European Community, such as the Joint European Torus (JET) which has already furnished a large scientific contribution and the Next European Torus (NET) which can be considered as under project. These machines will be finally fuelled with deuterium-tritium mixtures (DT) and, then, the whole systems will include Tritium Plants (TP). Taking into account the risks inherent to tritium and the TP reliability to ensure, tritium handling has to be developed. In this paper, all main recent developments from the European Tritium Handling Programme are overviewed after having recalled the major problems to overcome in relation with the whole process. Firstly, the JET Active Gas Handling System, the conception of which can be considered as original as regards other “classical” systems, will be discussed because being basically interesting and potentially transposable. The other developments, mainly within the framework of the NET Tritium Technology Programme are after indicated. They concern DT (re) processing methods (DT purification and storage, tritium recovery from tritiated water and/or tritiated effluents...) as well as useful basic studies (aging studies of tritides, analysis of tritiated media...) in connexion with various technologies developed in several Laboratories of the Community. Globally all the results obtained in the last four years undoubtedly represent a considerable effort and constitute now and henceforth a valuable informations source. Finally conclusions are drawn as regards the future developments to achieve in relation with the existing results and the final objectives so that to optimize the European Tritium Handling Programme.

Application of SAES and HWT gas purifiers forthe removal of impurities from helium-hydrogen gas mixtures

An experimental programme is described which is performed in the framework of a tritium technology task for the Next European Torus (NET). The aim is to investigate the applicability of commercial gas purifiers for the removal of gaseous impurities such as N 2, CO, and CO 2 from a helium-hydrogen or a helium-tritium gas stream. In addition, it is intended to recover chemically bound hydrogen isotopes by decomposition of hydrocarbons, ammonia and water. Experimental results obtained in the PEGASUS facility are reported for the sorption of CH 4 by three non-uranium getters as a function of temperature and gas flow rate, the enhancement of CH 4 sorption efficiency by continuous removal of hydrogen, and cosorption behaviour of CO, N 2 and CH 4 in the presence of hydrogen. Gettering of CO and N 2 was found to be much more effective than that of CH 4. At temperatures above 200°C, an increasing tendency for the reaction 2H 2+CO→CH 4+1/2O 2 was found to occur during the passage of the gas through the getter. This reaction can be suppressed by using two getters in series: the first is operated at about 250°C to remove O 2, H 2O, CO and CO 2, and the second at 400°C or more for the removal of residual amounts of these species as well as for the removal of N 2 and CH 4.

The Chalk River Tritium Extraction Plant

The Chalk River Tritium Extraction Plant for removal of tritium from heavy water is described. Tritium is present in the heavy water from research reactors in the form of DTO at a concentration in the range of 1–35 Ci/kg. It is removed by a combination of catalytic exchange to transfer the tritium from DTO to DT, followed by cryogenic distillation to separate and concentrate the tritium to T 2. The tritium product is reacted with titanium and packaged for transportation and storage as titanium tritide. The plant processes heavy water at a rate of 25 kg/h and removes 80% of the tritium and 90% of the protium per pass. Catalytic exchange is carried out in the liquid phase using a proprietary wetproofed catalyst. The plant serves two roles in the Canadian fusion program: (1) it produces pure tritium for use in fusion research and development, and (2) it demonstrates on an industrial scale many of the tritium technologies that are common to the tritium systems in fusion reactors.

Fusion Nuclear Technology Research and Development in Japan

This paper discusses how basic research in various fields of fusion reactor technology has been undertaken since the mid-1970s in Japan. The major topics in the early years were materials research, superconducting magnet technology, and tritium science and technology. Current fusion nuclear technology research and development activities in Japan come under two administrative frameworks: the Science and Technology Agency (STA) and the universities under the control of the Ministry of Education, Science, and Culture. The long-term plan for fusion nuclear reactor technology development is discussed.

Developments of tritium technology for next-step fusion devices under JAERI-DOE (LANL) collaboration.

Since early 1980's, JAERI and DOE (LANL) have collaborated on developments of tritium technology for fusion reactors. About ten years collaboration, specially under the Annex III and IV agreements, has produced many remarkable outcomes on tritium fuel processing and tritium safety handling technologies. Under the Annex III agreement, the two “process-ready” components for tritium fuel processing such as a palladium diffuser and an electrolysis cell were tested. Through tritium tests, both components were verified to be feasible and attractive for various applications in fusion fuel processes. Based on the results obtained in the component tests, an integrated process bas been developed and designed by JAERI for full scale demonstration of the plasma exhaust processing. The process loop will be tested using tritium under the Annex IV agreement. In the first three years of joint operation of TSTA under Annex IV, a number of the integrated loop runs and non-loop tests of the subsystems were performed using 1...

Purification de l'hydrogene sous forme libre ou combinee dans un melange gazeux par reactions chimiques avec l'uranium

Within the framework of the european fusion program, we are dealing with the tritium technology aspect. Hydrogen, free or under a combined form within a H 2, N 2, NH 3, O 2, gaseous mixture, can be purified by chemical reactions with uranium metal. The resulting reactions consist in absorbing the impurities without holding back H 2. Working conditions have been defined according to two main goals: the formation of stable solid products, especially under hydrogenated atmosphere and the optimization of the material quantities to be used. Thermodynamical considerations have shown that the 950–1300 K temperature range should be suitable for this chemical process. Experiments performed with massive uranium set in a closed reactor at 973 K, have produced hydrogen according to the predicted reactions rates. But they have also pointed out the importance of interferences that might occur in the uranium-gas system, on the gases conversion rates. The comparison between the chemical kinetic ratings of the reactions of pure gases and the chemical kinetic ratings of the reactions of the same gases in mixture, has been set up. It proves that simultaneous reactions can modify the working conditions of the solid products formation, and particularly modify their structure. In this case, chemical kinetic ratings are increased up to their maximal value; that means surface phenomena are favoured as with uranium powder gases reactions.

Net and the European fusion technology programme

The NET programme consists of design activities and development of the related technologies. The design effort concentrates in general on those components whose feasibility has to be shown also via prototype testing. Two options for the basic machine have been considered and they are based on the same design solutions and technologies: NET I with 15 MA plasma current and inductively driven burn of about 300 s and NET II with 25 MA current and 700 s inductive burn. The Technology Programme, conducted by the Associated Laboratories, includes six main areas: superconducting magnet, plasma facing components, blanket, remote handling, tritium technology and safety. This programme is closely related to the NET design and aims at supporting, with experimental data, the assessment of the apparatus feasibility envisaged before entering into the engineering design in 1991.

Recent Results at the Tritium Systems Test Assembly

The Tritium Systems Test Assembly (TSTA) project at Los Alamos is charged with developing and demonstrating the tritium technology required for the magnetic fusion energy program. Since June 1987 the project has been operated by Los Alamos National Laboratory under the joint funding of the US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI). At TSTA the technology under development includes reactor exhaust gas processing technology and safety systems. This project is not currently investigating the processing of tritium extracted from a breeding blanket. The processing of tritium extracted from a breeding blanket is a possible upgrade to TSTA. This possible upgrade is the subject of an ongoing study among Los Alamos, JAERI and the Argonne National Laboratory (ANL). This installing such on upgraded process systems at TSTA in the 1990--1992 time frame. The exhaust gas processing system at TSTA is designed to demonstrate full-scale exhaust gas processing for a fusion machine of a size currently being designed for ITER. Of course, deuterium is also a major component of the exhaust gas. Current gas processing uses a total gas stream comprised of /approximately/65% deuterium, 35% tritium, 1-2% reactive impurities (currently nitrogen and methane) with the remainingmore » gas hydrogen and helium. The tritium inventory required to process through this loop at this rate is about 110 g. The main process loop at TSTA has been operating at this flow rate since July 1987. In addition to the operation of the main process loop, a number of nonloop experiments are underway at TSTA. This paper will update the status of both loop and nonloop studies at Los Alamos.« less

Overview of canadian fusion nuclear technology effort

The fusion nuclear technology program in Canada seeks to develop and apply specific areas of nuclear expertise to critical fusion issues. The primary focus is on tritium technology, remote handling and safety, with recent activities in nuclear balance of plant systems. The Canadian Fusion Fuels Technology Project (CFFTP) coordinates the effort by providing resources and direction to universities, national laboratories and industry who undertake fusion research and development and engineering. CFFTP has received funding from Atomic Energy of Canada, Ontario Hydro and the Province of Ontario. The total direct effort is about $10 million Canadian per annum, including contributions from subcontractors and cost sharing with external projects.

The role of a blanket tritium system on the fusion fuel cycle

Abstract The requirements of tritium technology are centered in three main areas, i.e., (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The gaseous tritium stream from the breeder tritium extraction system is significantly different from the plasma exhaust stream and, therefore, may have an important impact on the operation of the fuel processing system. For some blankets, such an aqueous solution blanket, the blanket tritium stream may dominate the fuel processing system in terms of component size and power consumption. The importance of the blanket interface to a fuel processing experiment, such as TSTA, has been identified. The initial work to define the blanket processing system, which is proposed to be added as part of TSTA, will be discussed here.

The european fusion nuclear technology effort

The role of fusion technology in the European fusion development strategy is outlined. The main thrust of the present fusion technology programme is responding to development needs of the Next European Torus. A smaller, but important and growing R&D effort is dealing with problems specific to the Demonstration, or Fusion Power, Reactor. The part of the programme falling under the somewhat arbitrarily defined category of “fusion nuclear technology” is reviewed and an outlook to future activities is given. The review includes tritium technology, blanket technology and breeder materials development, technology and materials for the protection of the first wall and of other plasma facing components, remote handling technology, and safety and environmental impact studies. A few reflections are offered on the future long-term developments in fusion technology.

Preface: Third Topical Meeting on Tritium Technology in Fission, Fusion and Isotopic Applications

"Preface: Third Topical Meeting on Tritium Technology in Fission, Fusion and Isotopic Applications." Fusion Technology, 14(2P2A), p. xiii

Tritium Technology Development in EEC Laboratories — Contributions to Design Goals for NET

An overview is given of the tritium technology activities carried out in the European national laboratories associated with the European Fusion Programme and in the European Joint Research Center. The relationship of these activities to the Next European Torus (NET) design priorities is discussed, and the current status of the research is summarised. Future developments, required for NET, which will be addressed in the definition of the next 5-year programme are also presented.

Overview of Canadian Fusion Fuels Technology Program on Safety

The Canadian Fusion Fuels Technology Project was formed in 1982 to undertake research and engineering in Tritium Technology and robotics for Fusion applications. The total current program is about $10 million per annum including contributions from subcontractors and cost sharing with external projects. The present paper gives an overview of the scope of the program and focuses on the safety related work.

Fusion Tritium Program in Japan

Nuclear Fusion Council, Atomic Energy Commission of Japan, has started to review the nuclear fusion R and D plan for the next stage, post JT-60. The council launched a subcommittee on fundamental issues in the nuclear fusion development in 1985, for review of the basic strategy of a development plan. The subcommittee presented an interim report in Feb. 1986 after 6 months discussion and the report was approved by the Nuclear Fusion Council. Two major R and D programs described in the interim report are the development of a Tokamak type large facility and the comprehensive development of the fusion reactor technology. The latter means to promote the reactor technologies which will be essential in the future to construct not only a D/T burning but also a DEMO reactor. The Nuclear Fusion Development Program in Japan is shown. The interim report recommended to organize two subcommittees to establish an integrated national R and D plan; one was for the design of the next step large facility and the other was for the R and program of the fusion technology. The subcommittee for the latter consisted of 7 working groups; one of them was organized for the tritium technology.

Tritium Technology Studies at the Tritium Systems Test Assembly

The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory has been in operation with tritium since June 1984. Presently there are some 30 g of tritium in the main process loop. This 30 g has been sufficient to do a number of experiments involving the cryogenic distillation isotope separation systems. In January 1986, two major experiments were conducted. During these experiments the fuel cleanup system was interfaced, through the transfer pumping system with the isotope separation system, thus permitting testing on the integrated fuel processing loop. This integration of these systems means that of the TSTA subsystem only the vacuum system remains to be integrated into the TSTA fuel processing loop. In the period of June 1984 through May 1986, the TSTA system had processed approximately 10/sup 8/ Ci of tritium. Total tritium emissions to the environment over this period have been less than 3 Ci as elemental tritium and 2 Ci as tritium oxide. Personnel exposures during this period have totaled less than 100 person-mRem. To date, the development of tritium technology at TSTA has proceeded in progressive and orderly steps. In two years of operation with tritium, no major design flaws have been uncovered.

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Tritium technology and fusion reactor fuel cycle.

Fuel cycle and tritium technology of fusion reactor are stated. Processing of plasma exhaust gas and tritium recovery from blanket are the two main fuel flows and their recent developments are explained. For self-sustaining of fuel cycle, high efficiency is essential in separation and recovery of tritium. Importance of tritium safe containment is also stressed.

Cat-D fueled reversed-field pinch reactor assessment

A quantitative comparison of the technology requirements, environmental and cost issues of DD. Compact Reversed-Field Pinch Reactor (CRFPR) relative to a DT/CRFPR has been performed. The first wall/blanket energy recovery cycle for the DD reactor is simpler and more efficient than the DT reactor. In other technology areas (such as magnets and vacuum systems) DD requirements are not significantly different than the DT reactor. Tritium technology for processing the plasma exhaust is required to DD reactors, but no tritium containment around the blanket or heat transport system is needed. Safety analysis shows similar consequences for the release of activated corrosion products or activated first wall/blanket structure. Consequences of all postulated DD accidents for tritium releases are significantly smaller than those from the DT reactor. Cost studies have been performed for a series of DD reactors and compared with the DT reactor.

Research and Development in Tritium Technology at the Institute of Radiochemistry, Nuclear Research Center Karlsruhe

The Tritium Technology Program at the Institute of Radiochemistry, Nuclear Research Center Karlsruhe has the following objectives: determination of physicochemical properties of solid breeding materials; tritium recovery from waste streams; development of tritium decontamination systems; and impurity monitoring in liquid metal blankets.