R.G. Clemmer

The trio experiment

The TRIO experiment is a test of in-situ tritium recovery and heat transfer performance of a miniaturized solid breeder blanket assembly. The assembly (capsule) was monitored for temperature and neutron flux profiles during irradiation and a sweep gas flowed through the capsule to an analytical train wherein the amounts of tritium in its various chemical forms were determined. The capsule was designed to operate at different temperatures and sweep gas conditions. At the end of the experiment the amount of tritium retained in the solid was at a concentration of less than 0.1 wppM. More than 99.9% of tritium generated during the experiment was successfully recovered. The results of the experiment showed that the tritium inventories at the beginning and at the end of the experiment follow a relationship which appears to be characteristic of intragranular diffusion.

Fusion reactor blanket/shield design study

A joint study of tokamak reactor first-wall/blanket/shield technology was conducted by Argonne National Laboratory (ANL) and McDonnell Douglas Astronautics Company (MDAC). The objectives of this program were the identification of key technological limitations for various tritium-breeding-blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium-breeding-blanket concepts were evaluated according to the proposed coolant. The ANL effort concentrated on evaluation of lithium- and water-cooled blanket designs while the MDAC effort focused on helium- and molten salt-cooled designs. A joint effort was undertaken to provide a consistent set of materials property data used for analysis of all blanket concepts. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first-wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented.

STARFIRE, a commercial tokamak power plant design

Abstract STARFIRE is a design for a conceptual commercial tokamak electrical power plant based on the deuterium/tritium/lithium fuel cycle. In addition to the goal of being technologically credible, the design incorporates safety and environmental considerations. STARFIRE is considered to be the tenth in a series of commercial fusion power plants. STARFIRE has a 7-m major radius reactor producing 1200 MW of net electrical power from 4000 MW of thermal power, with an average neutron wall load of 3.6 MW/m 2 . The aspect ratio is 3.6 and a D-shaped plasma with a height-to-width ratio of 1.6 and average toroidal beta of 0.067 is used. The maximum magnetic field is 11T. Availability goals have been set at 85% for the reactor and 75% for the complete plant including the reactor. The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum for impurity control, most superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield.

The requirements for processing tritium recovered from liquid lithium blankets: The blanket interface

We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs.

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.

Wildcat: A commercial deuterium-deuterium tokamak reactor

AbstractWILDCAT is a conceptual design of a catalyzed deuterium-deuterium tokamak commercial fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing deuterium-tritium (D-T) designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete conceptual design.

The breeding blanket interface (BBI): recent results for the solid breeder and the aqueous salt solution blanket concepts

The Tritium Systems Test Assembly (TSTA) at Los Alamos, New Mexico, is a full-scale facility dedicated to testing tritium processing for fusion reactors. Adding a breeder blanket interface (BBI) to the TSTA is being studied. The BBI is to test the processing required for the tritium output streams for the various fusion-reactor breeder blankets. In the current phase of the study, the characteristics of the output from the various blanket types are being evaluated. Defining the output stream with respect to H/T ratio, impurity content, and radionuclide content is emphasized. Assessments of the solid breeder blanket (ceramic, Li/sub 2/O) and the aqueous salt solution blanket are reported. For the aqueous lithium salt solution blanket, the chemical systems are very complex, involving a number of phenomena, including radiolysis, corrosion, and electrolysis. A system of about 0.01 ITER scale may be added to the TSTA. For the solid breeder blanket, no critical issues are identified. A system of about half-ITER scale, possibly even full scale, may be added to the TSTA.<<ETX>>

The Breeder Blanket Interface (BBI) to TSTA: Requirements for an Aqueous Lithium Salt Blanket

A breeder blanket interface for an aqueous lithium salt blanket is defined for TSTA. High calculated radiolysis rates result in a high overpressure in the blanket and the need for a depressurizer a...

STUDIES OF FIRST WALL/BLANKET/SHTELD SYSTEMS FOR FUSION REACTORS

Two programs are underway at Argonne National Laboratory to develop experimental information on engineering issues of first wall and blanket systems for fusion reactors. These include a broadly based First Wall/Blanket/Shield Engineering Technology Program and the TRIO Blanket Processing Program which is addressing tritium recovery from solid tritium breeder materials.