[en] Integrated and closed tritium fuel loop is inevitable for fusion devices. One of the major accomplishment of the technology development for the fusion fuel loop is series of the experimental campaign of the Tritium Systems Test Assembly (TSTA) under the collaboration between US Department of Energy and the Japan Atomic Energy Research Institute (JAERI). The fuel processing loop for simulated plasma exhaust for refueling for continuous operation of fusion reactor has been successfully demonstrated. On the other hand, a number of design studies including ITER CDA was made for future fusion machines most are based on conservative design and tend to have complicated system and large tritium inventory, while flexibility of the system operation is limited. The present design study is intended to reveal some technical issues to be considered for practical fusion fuel loop in near future. Fusion devices will be operated in pulsed mode at least in early phase of tests, and increases tritium inventory and duty factor drastically. Operation scenarios for initial tokamak operation, short and long fusion reaction, and various cleaning modes should be discussed. The fuel cycle must process various species of gases and broad range of hydrogen isotopic contents in both pulse and stead state operation. It should be operated with very small inventory when the entire systems has small amount of initial tritium

[en] The Tritium System Test Assembly (TSTA) project at Los Alamos National Laboratory is charged with developing and demonstrating the fusion fuel processing requirement for the magnetic fusion energy program. The key fusion fuel streams are the plasma exhaust and the blanket stream. At present, the technology under development at TSTA includes only the plasma exhaust. The blanket is a logical upgrade to TSTA. This possible upgrade is the subject of an ongoing study among LANL, JAERI, and Argonne National Laboratory (ANL). This program, called The Breeding Blanket Interface (BBI), is being investigated with the intention of installing such a processing system at TSTA in the 1992--1993 time frame. The BBI program was initiated in July 1987. The conditions of the tritium product streams of various blankets were determined. Pre-conceptual designs of the BBI systems were developed for two blanket concepts: Aqueous Lithium Salt Solution Blanket and Solid Breeder Blanket. Preliminary estimates of costing and the schedule of the TSTA/BBI program have been developed. This paper covers the schedule and planning of the TSTA/BBI program. The goal of the program is to establish the engineering data basis of the fuel cycle for an ITER type machine. The operation of the TSTA/BBI hardware is expected to cover the period of 1993 to 1996. 9 refs., 2 figs., 3 tabs

[en] In 1987, the Japan Atomic Energy Research Institute (JAERI) and the US Department of Energy (DOE) signed a collaborative agreement (Annex IV) for the joint funding and operation of the Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory (LANL) for a five year period ending June, 1992. After this initial five year collaboration, the Annex IV agreement was extended for another two year period ending June, 1994. During the first five years, a number of the integrated process loop tests and non-integrated loop testing of the subsystems of TSTA were conducted. During integrated loop testing the vacuum system, fuel cleanup systems, isotope separation system, transfer pumping system and gas analysis system, are interconnected and tested using 100 g-inventories of tritium to demonstrate steady-state operation of a tritium fuel processing cycle for a fusion reactor. These tests have resulted in a number of significant accomplishments and an experience data base on research, development and operation of the fuel processing system. One of the most significant accomplishments during the initial five year period, was the continuous operation of the fuel processing loop for 25 days. During this 25-day extended operation, both the JAERI fuel cleanup system (J-FCU) and the original TSTA fuel cleanup system were operated under similar conditions of flow, pressure, and impurity content of the DT gas. Both fuel cleanup systems were demonstrated to provide adequate impurity removal for plasma exhaust gas processing. The isotope separation system was operated continuously, producing pure tritium while rejecting protium as an impurity

[en] Operating procedures are important for the safe and efficient operation of the Tritium Systems Test Assembly (TSTA). TSTA has been operating for four years with tritium in a safe and efficient manner. The inventory of tritium in the process loop is 100 grams and several milestone runs have been completed. This paper describes the methods used to operate TSTA. 3 refs., 1 fig

[en] The Fusion Safety Program (FSP) at the Idaho National Engineering Laboratory (INEL) conducts safety research in materials, chemical reactions, safety analysis, risk assessment, and in component research and development to support existing magnetic fusion experiments and also to promote safety in the design of future experiments. One of the areas of safety research is applying probabilistic risk assessment (PRA) methods to fusion experiments. To apply PRA, we need a fusion-relevant radiological dose code and a component failure rate data base. This paper describes the FSP effort to develop a failure rate data base for fusion-specific components

[en] Cryogenic distillation experiments were peformed at TSTA with H-D-T system by using a single column and a two-column cascade. In the single column experiment, fundamental engineering data such as the liquid holdup and the HETP were measured under a variety of operational condtions. The liquid holdup in the packed section was about 10 /approximately/ 15% of its superficial volume. The HETP values were from 4 to 6 cm, and increased slightly with the vapor velocity. The reflux ratio had no effect on the HETP. For the wo-colunn experiemnt, dynamic behavior of the cascade was observed. 8 refs., 7 figs., 2 tabs

[en] Separation experiments were performed with Isotope Separation System (ISS) at the TSTA from 1988 through 1989. These experiments included two column experiments with D-T mixture and four column experiments with D-T mixture and impurities (He and H₂). The objectives of these experiments were to obtain fundamental data of cryogenic distillation columns under wider range of operational condition than previous work, to observe the influence of He on the operation of cryogenic distillation columns, and to obtain fundamental data of cryogenics distillation columns under the influence of He. 2 refs., 7 figs., 2 tabs

[en] In March of 1988 full operation of the 4-column isotope separation system (ISS) was realized in runs that approximated the design load of tritium. Previous operations had been fraught with operating difficulties principally due to external systems. This report will examine the recent highly successful 6-day period of operation. During this time the system was cooled from room temperature, loaded with hydrogen isotopes including 109 grams of tritium, integrated with the transfer pumping, impurity injection, and impurity removal systems, as well as the remote computer control system. At the end of the operation 12 grams of tritium having a measured purity of 99.987% (remainder deuterium) were offloaded from the system. Observed profiles in the columns in general agree with computer models. A Height Equivalent to a Theoretical Plate (HETP) of 5.0 cm is confirmed. 3 refs., 5 figs

[en] The simulation of the units for hydrogen isotope separation by cryogenic distillation with packed columns can be carried out with a very efficient tool, based on a non equilibrium model [1]. This new approach enables a better representation of the physical phenomena involved in the columns to be obtained and consequently, better accuracy for the tritium inventory in the whole process. At the TSTA facility, one of main systems is the Isotope Separation System (ISS), fitted with an on-line laser Raman spectroscopy system: it allows to obtain very accurate composition profiles in the columns, rapidly and safely. Using experimental data, provided by TSTA, CEA, associated with PROSIM S.A., have carried out calculations on column 1. This study clearly demonstrates: the high efficiency of the experimental device; the good accuracy of the results in comparison with experimental data (composition profiles); and the efficacy of the method for design activities. 7 refs., 3 figs., 3 tabs

[en] The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory is a unique facility that can simulate and test reactor-scale fusion fuel processing systems and is operated jointly by the Japan Atomic Energy Research Institute (JAERI) and the US Department of Energy. A series of experimental campaigns of the TSTA loop were performed with approximately 100 grams of tritium in order to simulate and study fusion fuel processing at steady-state since 1984. Continuous processing of a simulated plasma exhaust (DT and H mixture, He, CH₄ and N₂) was successfully demonstrated for up to 25 days. However, besides prolonged steady state continuous operation, actual fusion machines will likely require various fuel processing modes. For instance, the International Thermonuclear Experimental Reactor (ITER) will operate as a pulsed machine. Also even for steady-state machines, transient operations for ramp-up, shut down, discharge cleaning or conditioning operation must be handled by the fuel system. After the successful completion of the original five year period of TSTA joint operation, an extended collaboration agreement was initiated in 1992 to conduct further experiments. TSTA has begun integrated fuel cycle tests which process gas streams varying in flow rate and composition