Tritium Systems Test Assembly Research Articles

Tritium Activities in the United States

There have been many significant changes in the status of tritium activities in the US since the 4th Tritium Conference in October, 1991. The replacement Tritium Facility (RTF) at Savannah River Site and the Weapons Engineering Tritium Facility (WETF) at the Los Alamos National Laboratory are now operational with tritium. The Tokamak Fusion Test Reactor (TFTR) has initiated a highly successful experimental campaign studying DT plasmas, and has produced more than 10 Megawatts (MW) of fusion power in a D-T plasma. Sandia National Laboratory has ceased tritium operations at the Tritium Research Laboratory (TRL) and many of the activities previously performed there have been transferred to Los Alamos and Savannah River. The tritium laboratory at Lawrence Livermore National Laboratory has reduced the tritium inventory to <5 grams. The Tritium Systems Test Assembly (TSTA) at Los Alamos continues to be at the forefront of tritium technology and safety development for the fusion energy program.

A Self-Assaying Tritium Storage Bed for ITER

A safe metal-hydride ``self-assaying`` tritium storage bed, featuring accurate tritium assaying measurements is being designed, built, and tested at the Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory (LANL) for the International Thermonuclear Experimental Reactor (ITER) program. Time consuming inventory operations can be shortened by utilizing ``self-assaying`` tritium storage beds. Design considerations, calculations, problems, and construction details of the bed are presented. Sensitivity, predictability, and simplicity are optimized in this design by utilizing thermal radiation as the primary mode of heat transfer. Thermal analysis calculations have shown that the design may provide 15 times the ITER required sensitivity at full capacity of 150 gT.

Status and Future of the Tritium Plasma Experiment

The Tritium Plasma Experiment (TPE) has been recently upgraded and relocated at the Tritium System Test Assembly (TSTA) at Los Alamos National Laboratory. The first tritium plasma in the upgraded system was achieved on May 11, 1995. TPE is a unique facility devoted to experiments on the migration and retention of tritium in fusion reactor materials. This facility is now capable of delivering 100 to 200 eV tritons at a level of 1 A/cm2 to a 5 mm diameter sample, similar to that expected for the divertor of the International Thermonuclear Experimental Reactor (ITER). An aggressive research plan has been established, and experiments are expected to begin in June of 1995.

Recent Palladium Membrane Reactor Development at the Tritium Systems Test Assembly

The palladium membrane reactor (PMR) is proving to be a simple and effective means for recovering hydrogen isotopes from fusion fuel impurities such as methane and water. This device directly combines two techniques which have long been utilized for hydrogen processing, namely catalytic shift reactions and palladium/silver permeators. A proof-of-principle (PMR) has been constructed and tested at the Tritium Systems Test Assembly of Los Alamos National Laboratory. The first tests with this device showed that is was effective for the proposed purpose. Initial work concluded that a nickel catalyst was an appropriate choice for use in a PMR. More detailed testing of the PMR with such a catalyst was performed and reported in other works. It was shown that a nickel catalyst-packed PMR did, indeed, recover hydrogen from water and methane with efficiencies approaching 100% in a single processing pass. These experiments were conducted over an extended period of time and no failure or need for regeneration was encountered. These positive results have prompted further PMR development. Topics addressed include alternate PMR geometries and initial testing of the PMR with tritium. These are the subjects of this paper.

Operation of a simulated non-steady tokamak fuel loop using the tritium systems test assembly

Operation of a simulated non-steady tokamak fuel loop using the tritium systems test assembly

Performance of a palladium membrane reactor using an Ni catalyst for fusion fuel impurities processing

The palladium membrane reactor (PNM) provides a means to recover hydrogen isotopes from impurities expected to be present in fusion reactor exhaust. This recovery is based on reactions such as water-gas shift and steam reforming for which conversion is equilibrium limited. By including a selectively permeable membrane such as Pd/Ag in the catalyst bed, hydrogen isotopes can be removed from the reacting environment, thus promoting the reaction to complete conversion. Such a device has been built and operated at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL). For the reactions listed above, earlier study with this unit has shown that hydrogen single-pass recoveries approaching 100% can be achieved. It was also determined that a nickel catalyst is a feasible choice for use with a PMR appropriate for fusion fuel impurities processing. The purpose of this study was to systematically assess the performance of the PMR using a nickel catalyst over a range of temperatures, feed compositions and flowrates. Reactions which were studied are the water-gas shift reaction and steam reforming.

Tritium handling and processing experience at the Tritium Systems Test Assembly

Tritium handling and processing experience at the Tritium Systems Test Assembly

Tritium handling and processing experience at the Tritium Systems Test Assembly

Abstract In 1987 the Japan Atomic Energy Research Institute (JAERI) and the US Department of Energy (DOE) signed a collaborative agreement for the joint funding and operation of the Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory. This collaboration has continued through June 1994. During this seven year period the TSTA team has conducted many tritium experiments. These included a number of integrated process loop tests, as well as many off-line, stand-alone experiments to develop, test and validate new components and technologies for tritium service. The TSTA process loop has been operated in continuous mode for periods up to 25 consecutive days, and many loop experiments of shorter duration have been performed. In addition, a number of technical developments and component/system demonstrations have been accomplished. These tests have included development and evaluating of fuel cleanup systems and hydrogen isotope separation systems that are now available for consideration for ITER. This paper highlights some of the major accomplishments of this seven year collaboration.

Operation of a simulated non-steady tokamak fuel loop using the tritium systems test assembly

In order to develop a fuel system for a realistic fusion device in near future, a number of experimental campaigns of a simulated fusion fuel loop were performed under practical non-steady conditions at the Tritium Systems Test Assembly (TSTA). Some technical issues specific for non-steady fuel loop were identified and are being investigated further. The overall process loop was operated with non-steady inputs to better interface with pulsed tokamak operation, which requires a rather different and improved processing capability specific to each subsystem. The cryogenic distillation columns in the isotope separation are modified to provide side-stream recycle paths with isotopic equilibration function. This change improved separation characteristics with various feed compositions, and reduces the required number of columns for processing and resulted in a reduced tritium inventory in the isotope separation system (ISS). Another major technical development on the ISS is addition of a number of feed-back control loops that automatically operate the distillation columns stably under changing feed conditions. The plasma exhaust processing system composed of palladium diffuser, catalytic reactor, electrolysis cell and cold trap was operated mainly in the batch mode to handle a broader range of input flow rate and composition in various configurations to minimize tritium loss and inventory. The results demonstrated the overall capability and flexibility of the TSTA loop to serve as a fuel processing system under non-steady conditions; however, they imply that many technical issues arise in operating a practical fuel processing system. These may not be foreseen in the design stage and can only be determined during integrated tests under realistic operating conditions.

Tritium evacuation performance of a large oil-free reciprocating pump

Abstract A large oil-free reciprocating pump, with carbon polyimide composite piston rings and dynamic metal bellows seal, has been developed for tritium services at the Tritium Process Laboratory of the Japan Atomic Energy Research Institute. The design pumping speed is 54 m 3 h −1 for hydrogen gas at 5 Torr suction pressure and 875 Torr discharge pressure. Tritium evacuation performances were measured at 400–675 Torr discharge pressure at the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory under the US-Japan collaboration programme on fusion fuel technology. The pumping speed for tritium was greater than 120 m 3 h −1 at 4 Torr suction pressure and 500 Torr discharge pressure. The ultimate pressure with tritium was 0.8 Torr at 500 Torr discharge pressure. Performance data for tritium were slightly better than those for H 2 , D 2 , He and N 2 . The discharge pressure did not affect the evacuation performance. In order to check the long-term effect with tritium on pump, semicontinuous tritium pumping operations were performed for a total of 350 hours within 3 months. However, no performance changes were found. After stand-alone tests, the pump was successfully demonstrated as a DT fuel vacuum and transfer pump during a TSTA loop run.

Practical-scale tests of cryogenic molecular sieve for separating low-concentration hydrogen isotopes from helium

Abstract Earlier bench-scale work at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory examined a number of adsorbents for their suitability for separating low-concentration hydrogen (no tritium) from helium. One of the effective adsorbents was Linde 5A molecular sieve. Recently, experiments including tritium were conducted using practical-scale adsorbers. These tests used existing cryogenic molecular sieve beds (CMSBs) which each contain about 1.6 kg of Linde 5A molecular sieve. They are part of the TSTA integrated tritium processing system. Gas was fed to each CMSB at about 13 SLPM (standard liters per minute) with a nominal composition of 99% He, 0.98% H2 and 0.02% HT. In all cases, for an extended period of time, the beds allowed no detectable (via Raman spectroscopy) hydrogen isotopes to escape in the bed effluent. Thereafter, the hydrogen isotopes appeared in the bed exit with a relatively sharp breakthrough curve. It is concluded that cryogenic molecular sieve adsorption is a practical and effective means of separating low-concentration hydrogen isotopes from a helium carrier.

Performance of a palladium membrane reactor using an Ni catalyst for fusion fuel impurities processing

Abstract The palladium membrane reactor (PMR) provides a means to recover hydrogen isotopes from impurities expected to be present in fusion reactor exhaust. This recovery is based on reactions such as water-gas shift and steam reforming for which conversion is equilibrium limited. By including a selectively permeable membrane such as PdAg in the catalyst bed, hydrogen isotopes can be removed from the reacting environment, thus promoting the reaction to complete conversion. Such a device has been built and operated at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. This work was performed as part of the Annex IV collaboration between the US Department of Energy-TSTA and the Japan Atomic Energy Research Institute-Tritium Processing Laboratory. For the reactions listed above, earlier study with this unit has shown that hydrogen single-pass recoveries approaching 100% can be achieved. It was also determined that a nickel catalyst is a feasible choice for use with a PMR appropriate for fusion fuel impurities processing. The purpose of this study was to assess systematically the performance of the PMR using a nickel catalyst over a range of temperatures, feed compositions and flow rates. Reactions which were studied are the water-gas shift reaction and steam reforming.

Use of Magnesium for Recovering Hydrogen Isotopes from Tritiated Water

Reacting tritiated water with hot metal to recover the tritium from tritiated water has been practiced for considerable time. A metal frequently used for this purpose has been uranium. Recent work at the Tritium Systems Test Assembly at Los Alamos National Laboratory has focused on using magnesium for this purpose. This work was done as part of the Annex IV collaboration between the US Department of Energy/TSTA and the Japan Atomic Energy Research Institute/Tritium Processing Laboratory. Magnesium appears to have reactive properties that are as good as uranium and possibly better, and, of course, magnesium is easier to handle and less strictly controlled. Both bench-scale and practical-scale experiments were conducted with magnesium, including tests with tritiated water. Mg bed construction techniques and operating parameters were determined. Testing showed that the Mg packed bed was very effective for recovering hydrogen isotopes from water. However, when used for this purpose, either Mg or U is irreversibly consumed and must be disposed of as tritiated waste. It follows that this processing technique would be inappropriate for a large tritiated water processing operation. However, this technique may find utility for small-scale systems.

Failure rate data for glovebox components and cleanup systems at the Tritium Systems Test Assembly

Fusion facility safety and reliability/availability analyses require accurate component failure rate information to provide meaningful results. While fission reactor operating experience data may be adequate for some types of components, there are some data needs that are fusion-specific, such as tritium fueling and handling system information. This paper summarizes data analysis of tritium glovebox confinement systems and an air detritiation system from the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. These analyses benefit fusion design work by highlighting weak areas in designs to allow for modifications and upgrades, making future designs more robust. The TSTA results are generally smaller failure rates than the information from other industries, thus showing one of the benefits of gathering these fusion-specific data.

Using the tritium plasma experiment to evaluate ITER PFC safety

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capabilty of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 × 1023 ions/m2.s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures.

Molecular Sieve Regeneration System for Assaying HTO from Detritation Systems

A Molecular Sieve Regeneration System (MSRS) is being added to the existing Tritium Waste Treatment system (TWT) within the Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory. This system is an upgrade to the TWT to provide accurate measurements of the liquid waste generated from this system. Within the TWT, hydrogen isotopes are removed from the effluent gas stream by the catalytic conversion to water and the subsequent removal of water by molecular sieve trapping prior to the release to the environment. Within the TWT and similar systems, molecular sieve regeneration is required to rejuvenate the beds. The major difference of the MSRS and other regeneration systems is the capability of direct tritium assay of long-term storage waste containers. This is accomplished with loop-flow regeneration, water collection, and tritiated water assay by scintillation and calorimetric techniques. This paper describes the MSRS in detail and how it is interfaced with the Tritium Waste Treatment system.

Knowledge Collection and Operator Training Using an Artificial Intelligence System

AbstractA computer-based Expert Adviser has been developed to catalog technical expertise and dispense this knowledge to operators of a complex system of distillation columns at the Los Alamos National Laboratory (LANL). The Expert Adviser is also used to train personnel in operation of the LANL Tritium Systems Test Assembly (TSTA) Isotope Separations System (ISS). The methodology is applicable to a wide range of operations where documentation and transfer of knowledge is desirable.

Estimated Failure Rates for Some Selected Tritium System Components

Fusion facility safety and reliability/availability analyses require accurate component failure rate information to provide meaningful results. While fission reactor operating experience data may be adequate for some types of components, there are some data needs that are fusion-specific, such as tritium fueling and handling system information. This paper summarizes the results of a failure data analysis task on several components at Los Alamos National Laboratory’s Tritium Systems Test Assembly (TSTA). We took the failure reports and component operating information from the TSTA Failure/Maintenance Data Base. Calculated failure rates are on the same order of magnitude as screening fusion failure rate data, although a few failure rates were slightly higher. These new data should be of interest to fusion safety analysts and tritium systems designers.

Isotope Separation System Experiments at the TSTA

The recent results of the Isotope Separation System (ISS) operations at the Tritium Systems Test Assembly (TSTA) with 100 g of tritium indicate that the system generally satisfies design goals, while system stability problems remain to be solved. In this paper, the authors configure the ISS system for the three column mode. which is one of the promising cascade configurations in a fusion fuel cycle, to eliminate such instability and operated it for six days. Fluctuations in flows and liquid levels were improved. Column separation characteristics obtained were satisfactory and agreed with the numerical analysis. The amount of discharged tritium was an acceptable effluent level. This means that the existing ISS system can be used as a three column system and possibly be applied to numerous fuel concepts. Presently, a new laser Raman spectroscopic gas analyzer has been installed at the ISS. This on-line system enables studies of the ISS dynamic behavior for further stability and performance data.

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.