John R. Bartlit

Experimental Results from Hydrogen/Deuterium Distillations at the Tritium Systems Test Assembly

Fundamental operating parameters and column performance have been measured on the interlinked, four-column cryogenic distillation system at the Tritium Systems Test Assembly. Using mixtures of hydrogen and deuterium, data have been gathered and are given here on liquid holdup in the column packing and liquid level in reboilers (factors important to isotope inventory), pressure drop in the column packing, and product purity. Both Raman spectroscopy and gas chromatography were used for product analysis.

Single column and two-column H-D-T distillation experiments at TSTA

Cryogenic distillation experiments were performed 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 conditions. The liquid holdup in the packed section was about 10--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 under the condition that the vapor velocity was almost constant. For the two-column experiment, dynamic behavior of the cascade was observed.

Experience of TSTA Milestone Runs with 100 Grams-Level of Tritium

The first loop operation tests of the Tritium Systems Test Assembly (TSTA) with 100 grams-level of tritium were performed at the Los Alamos National Laboratory (LANL) in June and July, 1987. The July run was resumption of the June run, which was halted because of a loss of cryogenic refrigerant in the hydrogen isotope separation system.

On-line Tritium Process Gas Analysis by Laser Raman Spectroscopy at TSTA

Laser Raman spectroscopy has been applied to the on-line analysis of the operation of the cryogenic Isotope Separation System (ISS) at the Tritium Systems Test Assembly (TSTA). A flow-through cell was employed to permit near real-time observation of the dynamic response of the 3-column ISS. Accurate analysis of hydrogen isotopic mixtures may be made in less than 2 minutes. Full response to a change in the sampling point is achieved in approximately one minute. In this paper, response measurements are shown as well as static column profiles and dynamic response to induced parameter changes. Cross check of analysis was performed with radio-gas chromatography.

Combined gettering and molten salt process for tritium recovery from lithium

A new tritium recovery concept from lithium has been developed as part of the US/Japan collaboration on Reversed-Field Pinch Reactor Design Studies. This concept combines the gettering process at the front end to recover tritium from the coolant and a molten salt recovery process to extract tritium for fuel processing. A secondary lithium is used to regenerate the tritium from the gettering bed which, in the process, increases the tritium concentration by a factor of about 20. This way, the required size of the molten salt process becomes very small. A potential problem is the possible poisoning of the gettering bed by the salt dissolved in lithium.

The tritium systems test assembly at Los Alamos National Laboratory

Abstract The Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory is a facility for the integrated testing, in full scale, of the processes and safety systems required for the reprocessing and recycling of plasma exhaust gas from a tokamak reactor. The facility can reprocess and reuse tritium at the rate of 1000 g per day, with a tritium inventory of 130 g. TSTA has been operated since 1984 with increasing amounts of tritium and an increasing degree of system integration without significant releases to the environment or doses to workers. Full loop integration was achieved in 1988. Since 1987, the TSTA has been jointly funded and jointly operated by the U.S. Department of Energy and the Japan Atomic Energy Research Institute. Objectives include: the demonstration of long-term reliability and safety, the qualification of equipment for tritium service, and the verification of a facility that can be copied for future fusion machines, such as the International Tokamak Experimental Reactor (ITER). Future plans include the installation and integrated testing of a new impurity removal system of Japanese design, testing the effect on the plasma exhaust gas reprocessing systems of off-normal conditions in the torus, and the development and testing of technology for processing tritium contained in a breeding blanket product stream.

Introduction and synopsis of the TITAN reversed-field-pinch fusion-reactor study

Abstract The TITAN reversed-field-pinch (RFP) fusion-reactor study has two objectives: to determine the technical feasibility and key developmental issues for an RFP fusion reactor operating at high power density: and to determine the potential economic (cost of electricity), operational (maintenance and availability), safety and environmental features of high mass-power-density fusion-reactor systems. Mass power density (MPD) is defined as the ratio of net electric output to the mass of the fusion power core (FPC). The FPC includes the plasma chamber, first wall, blanket, shield, magnets, and related structure. Two different detailed designs TITAN-I and TITAN-II, have been produced to demonstrate the possibility of multiple engineering-design approaches to high-MPD reactors. TITAN-I is a self-cooled lithium design with a vanadium-alloy structure. TITAN-II is a self-cooled aqueous loop-in-pool design with 9-C ferritic steel as the structural material. Both designs use RFP plasmas operating with essentially the same parameters. Both conceptual reactors are based on the DT fuel cycle, have a net electric output of about 1000 MWe, are compact, and have a high MPD of 800 kWe per tonne of FPC. The inherent physical characteristics of the RFP confinement concept make possible compact fusion reactors with such a high MPD. The TITAN designs would meet the U.S. criteria for the near-surface disposal of radioactive waste (Class C, IOCFR61) and would achieve a high Level of Safety Assurance with respect to FPC damage by decay afterheat and radioactivity release caused by accidents. Very importantly, a “single-piece” FPC maintenance procedure has been worked out and appears feasible for both designs. Parametric system studies have been used to find cost-optimized designs. to determine the parametric design window associated with each approach, and to assess the sensitivity of the designs to a wide range of physics and engineering requirements and assumptions. The design window for such compact RFP reactors would include machines with neutron wall loadings in the range of 10–20 MW/m 2 with a shallow minimum COE at about 18 MW/m 2 . Even though operation at the lower end of the this range of wall loading (10–12 MW/m 2 ) is possible, and may be preferable, the TITAN study adopted the design point at the upper end (18 MW/m 2 ) in order to quantify and assess the technical feasibility and physics limits for such high-MPD reactors. From this work, key physics and engineering issues central to achieving reactors with the features of TITAN-I and TITAN-II have emerged.

Overview of the TITAN-I fusion-power core

The TITAN reactor is a compact (major radius of 3.9 m and plasma minor radius of 0.6 m), high neutron wall loading (~18 MW/m 2 ) fusion energy system based on the reversed-field pinch (RFP) confinement concept. The reactor thermal power is 2918 MWt resulting in net electric output of 960 MWe and a mass power density of 700 kWe/tonne. The TITAN-I fusion power core (FPC) is a lithium, self-cooled design with vanadium alloy (V-3Ti-1Si) structural material. The surface heat flux incident on the first wall is ~4.5 MW/m 2 . The magnetic field topology of the RFP is favorable for liquid metal cooling. In the TITAN-I design, the first wall and blanket consist of single pass, poloidal flow loops aligned with the dominant poloidal magnetic field. A unique feature of the TITAN-I design is the use of the integrated-blanket-coil (IBC) concept. With the IBC concept the poloidal flow lithium circuit is also the electrical conductor of the toroidal-field and divertor coils. Three dimensional neutronics analysis yields a tritium breeding ratio of 1.18 and a molten salt extraction technique is employed for the tritium extraction system. Almost every FPC component would qualify for Class C waste disposal. The compactness of the design allows the use of single-piece maintenance of the FPC. This maintenance procedure is expected to increase the plant availability. The entire FPC operates inside a vacuum tank, which is surrounded by an atmosphere of inert argon gas to impede the flow of air in the system in case of an accident. The top-side coolant supply and return virtually eliminate the possibility of a complete LOCA occurring in the FPC. The peak temperature during a LOFA is 991 °C.

The data collection system for failure/maintenance at the Tritium Systems Test Assembly

A data collection system for obtaining information which can be used to help determine the reliability and vailability of future fusion power plants has been installed at the Los Alamos National Laboratorys Tritium Systems Test Assembly (TSTA). Failure and maintenance data on components of TSTAs tritium systems have been collected since 1984. The focus of the data collection has been TSTAs Tritium Waste Tratment System (TWT), which has maintained high availability since it became operation in 1982. Data collection is still in progress and a total of 291 failure reports are in the data collection system at this time, 47 of which are from the TWT. 6 refs., 2 figs., 2 tabs.

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.