Tritium Systems Test Assembly Research Articles

TSTA piping and flame arrestor operating experience data

The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium handling technology and experiment research at the Los Alamos National Laboratory. The facility was operated with tritium for its research and development program from 1984 to 2001, running a prototype fusion fuel processing loop with ∼100g of tritium as well as small experiments. There have been several operating experience reports written on this facility's operation and maintenance experience. This paper describes reliability analysis of two additional components from TSTA, small diameter copper gas piping that handled tritium in a nitrogen carrier gas, and the flame arrestor used in this piping system. The component failure rates for these components are discussed in this paper. Comparison data from other applications are also presented.

Operating experiences from existing fusion facilities in view of ITER safety and reliability

The objective of this ongoing activity is to develop a fusion specific component failure database useful to quantify probabilistic safety assessment, support design activity, quantify reliability, availability, maintainability, and inspectability (RAMI) analyses, and support any other uses where field experience can provide feedback to fusion facility design and operation. The data collection began in 1989. At first, it was to “harvest” already-published data from high-technology industries. Later, attention was also directed to collecting and analyzing operating experience data from existing tokamaks and other fusion experiments. Operating experience of European, American and Japanese facilities are being investigated. The main studies performed thus far have been on JET systems (such as Vacuum, Active Gas Handling (AGH), Neutral Beam Injectors, Power Supply and Ion Cyclotron Resonant Heating systems), ASDEX Electron Cyclotron Resonant Heating system, and the Tore Supra cryogenics, superconducting magnet and water cooling system of plasma facing component for Europe; DIII-D, TFTR, Tritium Systems Test Assembly and other facilities for the United States; and the Tritium Processing Laboratory for Japan. Additional systems will be investigated in the future. The paper is dedicated to give an overview on the overall activity, with indication of references, and a global overview on availability of different fusion devices.

Tritium Behavior Intentionally Released in the Room

To construct a fusion reactor with high safety and acceptability, it is necessary to establish and to ensure tritium safe handling technology. Tritium should be well-controlled not to be released to the environment excessively and to prevent workers from excess exposure. It is especially important to grasp tritium behavior in the final confinement area, such as the room and/or building. In order to obtain data for actual tritium behavior in a room and/or building, a series of intentional Tritium Release Experiments (TREs) were planned and carried out within a radiologically controlled area (main cell) at Tritium System Test Assembly (TSTA) in Los Alamos National Laboratory (LANL) under US-JAPAN collaboration program. These experiments were carried out three times. In these experiments, influence of a difference in the tritium release point and the amount of hydrogen isotope for the initial tritium behavior in the room were suggested. Tritium was released into the main cell at TSTA/LANL. The released tritium reached a uniform concentration about 30 ~ 40 minutes in all the experiments. The influence of the release point and the amount of hydrogen isotope were not found to be important in these experiments. The experimental results for the initial tritium behavior in the room were also simulated well by the modified three-dimensional eddy flow analysis code FLOW-3D.

Tritium Systems Test Assembly (TSTA) Stabilization

The Los Alamos National Laboratory (LANL) Tritium Systems Test Assembly (TSTA) project was begun in 1978 to develop, design, and demonstrate the technology and safe operation of selected tritium processing systems required for a fusion reactor. In 2001, the US Department of Energy (DOE) determined that TSTA’s mission was complete and that the facility should be stabilized.At the completion of the stabilization project in 2003, TSTA was categorized as a radiological facility. Before stabilization was complete, the tritium inventory at TSTA was grouped in the following categories: tritium gas mixed with hydrogen isotopes, tritiated water absorbed on molecular sieve, tritium held up as a hydride on various metals, and tritium held up in process components. For each of these, tritium content was characterized, a path for removal was determined, and the proper disposal package was developed. Hydrogen exchange, calorimetry, direct sampling, pressure/composition/temperature, radiological smear surveys, and controlled regeneration were used to determine the tritium inventory for each category of tritium.After removal, the tritium inventory was either (1) sent to other facilities for reuse processing or (2) buried at the LANL radioactive waste disposal site. One complete experimental system was packaged and transferred to another DOE site for future use. Special burial containers were designed and fabricated for the inventory buried at the LANL radioactive waste disposal site. The project was conducted with low tritium emission to the environment and negligible personnel exposure. After the tritium removal was complete, all remaining hardware and piping were opened and vented; the facility emission was below 1 Ci per day.

Simplified Estimation of Tritium Inventory in Stainless Steel

An important part of tritium facility waste management is estimating the residual tritium inventory in stainless steel. This was needed as part of the decontamination and decommissioning associated with the Tritium Systems Test Assembly at Los Alamos National Laboratory. In particular, the disposal path for three, large tanks would vary substantially depending on the tritium inventory in the stainless steel walls. For this purpose the time-dependant diffusion equation was solved using previously measured parameters. These results were compared to previous work that measured the tritium inventory in the stainless steel wall of a 50-L tritium container. Good agreement was observed. These results are reduced to a simple algebraic equation that can readily be used to estimate tritium inventories in room temperature stainless steel based on tritium partial pressure and exposure time. Results are available for both constant partial pressure exposures and for varying partial pressures. Movies of the time dependant results were prepared which are particularly helpful for interpreting results and drawing conclusions.

Comparison of Tritium Component Failure Rate Data

Published failure rate values from the US Tritium Systems Test Assembly, the Japanese Tritium Process Laboratory, the German Tritium Laboratory Karlsruhe, and the Joint European Torus Active Gas Handling System have been compared. This comparison is on a limited set of components, but there is a good variety of data sets in the comparison. The data compared reasonably well. The most reasonable failure rate values are recommended for use on next generation tritium handling system components, such as those in the tritium plant systems for the International Thermonuclear Experimental Reactor and the tritium fuel systems of inertial fusion facilities, such as the US National Ignition Facility. These data and the comparison results are also shared with the International Energy Agency cooperative task on fusion component failure rate data.

Development of Tritium Plant System for Fusion Reactors. Achievements in the 14-year US-Japan Collaboration.

Fuel processing technology and tritium safe-handling technology have been developed through US/DOE-JAERI collaboration from 1987 till 2001, and the technologies to construct the tritium plant system of ITER have been made currently available. This paper overviews the major achievements of this collaborative researches over fourteen years, which were performed mainly at the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory (LANL). The tritium plant system consists mainly of a fuel processing system, which includes a fuel cycle system and a blanket tritium recovery system, and a tritium confinement/removal system. The fuel cycle system recovers fuel from plasma exhaust gas and recycles it. In the collaboration, major key components and subsystems were developed, and the performance of the integrated system was successfully demonstrated over its one-month operation in which plasma exhaust model gas was processed at a processing rate of up to 1/6 level of the ITER. The technological basis of the fuel cycle system was thus established. Blanket tritium recovery technology was also successfully demonstrated using the TSTA system. Through the successful safeoperation of the TSTA, reliability of tritium confinement/removal system was verified basically. In addition, much data to confirm or enhance safety were accumulated by experiments such as intentional tritium release in a large room. Furthermore,distribution of tritium contamination in the vacuum vessel of the TFTR, a large tokamak of the Princeton Plasma Physics Laboratory (PPPL), was investigated in this work.

Behavior of tritium in the TSTA test cell combined with operation of the Experimental Tritium Cleanup (ETC) system

Tritium and deuterium are expected to be the fuel for the first fusion power reactors. Being radioactive, tritium is a health, safety and environment concern. Room air tritium clean systems can be used to handle tritium that has been lost to the room from primary or secondary containment. Such a system called the Experimental Tritium Cleanup (ETC) systems is installed at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. The ETC consists of (1) two compressors which draw air from the room, (2) a catalyst bed for conversion of tritium to tritiated water, and (3) molecular sieve beds for collection of the water. The exhaust from this system can be returned to the room or vented to the stack. As part of the US–Japan fusion collaboration, on two separate occasions, tritium was released into the 3000 m 3 TSTA test cell, and the ETC was used to handle these releases. Each release consisted of about one Curie of tritium. Tritium concentrations in the room were monitored at numerous locations. Also recorded were the HT and HTO concentrations at the inlet and outlet of the catalyst bed. Tritium surface concentrations in the test cell were measured before and at a series of times after the releases. Surfaces included normal test cell equipment as well as idealized test specimens. The results showed that the tritium became well-mixed in the test cell after about 45 min. When the ETC was turned on, the tritium in the TSTA test cell decreased exponentially as was expected. The test cell air tritium concentration was reduced to below one DAC (derived air concentration) in about 260 min. For the catalyst bed, at startup when the bed was at ambient temperature, there was little conversion of tritium to HTO. However, once the bed warmed to about 420 K, all of the tritium that entered the bed was converted to HTO. Immediately after the experiment, surfaces in the room initially showed moderately elevated tritium concentrations. However, with normal ventilation, these concentrations soon returned to routine levels. The data collected and reported here should be useful for planning for the operation of existing and future tritium facilities.

Numerical Estimation Method of the Hydrogen Isotope Inventory in the Hydrogen Isotope Separation System for Fusion Reactor

In the fuel cycle system of the ITER, a large fraction of tritium inventory is expected to be in the cryogenic distillation columns of the hydrogen isotope separation system (ISS). Therefore, the numerical estimation method of hydrogen isotopes inventory in the ISS with high precision is strongly required from safety point of view. Two series of experiments were performed to establish the numerical estimation method of the overall hydrogen isotope inventory in the ISS at steady state using ITER-scale large cryogenic distillation columns at the Tritium Systems Test Assembly in the Los Alamos National Laboratory under the US-Japan collaboration on tritium safety engineering. As a result of experiments, it was confirmed that the hydrogen isotope inventory in a cryogenic distillation column was estimated by the numerical estimation method proposed in this work with enough high precision from the engineering point of view, and it was proved that this method was applied for the ITER-scale cryogenic distillation columns. The precision of this estimation method was found to be almost independent on the composition profile in columns, and especially the liquid holdup ratio of deuterium to the volume of the column was less influential in the inner diameter of the packed section. In addition, the gaseous inventory in an ITER-scale cryogenic distillation column was found to have considerable impact on the total amount of holdup of the column.

Steady-State Computer Modeling of a Recent H-D-T Cryogenic Distillation Experiment at TSTA

ABSTRACTCryogenic distillation is the only technique with the capacity to handle the hydrogen isotope separation requirements of a fusion power plant. However, there are safety and cost considerations associated with the considerable tritium inventory that can accumulate in such an isotope separation system (ISS). The ISS must be able to reliably produce specified products while responding to varying input streams. To design an ISS that balances all of these considerations and operate it reliably, it is essential to have a computer model of the system. This allows for a better understanding of the system and the exploration of various parameter regions that would otherwise require very expensive experimentation.The value of such a model, however, is questionable until it is validated by comparison with actual experiments. Recently, as part of the Annex IV US/Japan collaboration, a series of tests were conducted on the ISS system at the Tritium Systems Test Assembly (TSTA) located at Los Alamos National Laboratory (LANL). This system has a fusion power plant-relevant capacity of 6 SLPM (standard liters per minute). These experiments employed light hydrogen (protium), deuterium and tritium. Conditions at five steady state conditions were measured. The measurements included concentration measurements at the column feed, top and bottom, and also at intermediate points.These measurements served as a benchmark for comparison to DYNSIM, the model that has been in use at LANL for many years.† This model was able to accurately predict the column concentration profile based on the measured pressure, temperature, reboiler heat, feed composition and flows for a set of significantly different operating conditions. These results impart confidence that the model is useful for future ISS design and for better understanding of existing system operations.

Comparisons of facility-specific and generic component failure rates for tritium-bearing components used in fusion research

This paper compares the component failure rates for continuous operation and function-on-demand components that confine tritium, and provides recommendations on values to use for fusion applications. Since a tritium release to the environment can create a significant public hazard, the safety and reliability of tritium systems is very important for fusion research experiments and future power plants. Probabilistic safety assessment techniques are used to evaluate the safety of tritium handling. A foundation for meaningful probabilistic safety assessment is the application of accurate component failure rate data. The component failure rate values discussed here have been calculated and analyzed from fusion research facilities, the US Tritium Systems Test Assembly and Japan's Tritium Process Laboratory. The generic failure rate data compared in this paper were taken from published sources in US and Canada.

Concentrating Low-Level Tritiated Water through Isotope Exchange

Trapping of tritium on polymers with specific functional groups was investigated as a means of treating waste streams containing low levels of tritium. Chemical exchange of tritium with hydrogen on the functional group was used as the mechanism for trapping. The polymers tested include Aurorez polybenzimidazole resin beads, Chelex 100 resin beads, Duolite GT-73, microcrystalline cellulose, and polyethylenimine. The tests were performed under simulated operating conditions on water obtained from the Radioactive Liquid Waste Treatment Facility at Los Alamos National Laboratory. Tritiated water from the Tritium Systems Test Assembly is discharged to this plant. Polyethylenimine is a water-soluble polymer that was tested using a stirred membrane cell with an ultrafiltration membrane. All of the polymers except polyethylenimine took up tritium from the water. Polybenzimidazole demonstrated the highest tritium uptake. The results are explained on the basis of the type of functional group, hydrogen bonding, and rigidity of the molecular structure of the polymer. The theoretical calculations indicate that significant isotope discrimination requires high-frequency modes with hydrogen bonding contribution and support the experimental findings. Modeling suggested trends that may lead to structures that are more efficient in trapping tritium.

Tritium Behavior Intentionally Released in the Radiological Controlled Room under the US-Japan Collaboration at TSTA/LANL

A series of planned tritium release experiments into the Tritium Systems Test Assembly (TSTA) tritium processing rooms (3000 m3), were performed under the US-Japan collaboration program at Los Alamos National Laboratory (LANL). These experiments were carried out to acquire data on the behavior of tritium in an actual tritium facility. These experiments were performed safely with no impact on the environment or worker safety using 37 GBq (1 Ci). The results showed that the migration of tritium within the facility was dominated by the residual airflow in the room and reached a uniform value in 30 ~ 40 minutes. After restoring the normal once through ventilation (5 air changes per hour), room tritium levels decreased to background in less than 1 hour. Residual surface contamination was detected (max. 1 Bq/cm2 level). Residual contamination levels were found to be in order of decreasing contamination: linoleum > epoxy coating > acrylic resin > butyl rubber > stainless steel, from soaking results. The surface contamination increase as a result of the tritium release, was reduced to background with the normal ventilation within a few days.

TPOP-II: Tritium Fueling at a Reactor Scale

As part of the International Thermonuclear Experimental Reactor (ITER) plasma fueling development program, Oak Ridge National Laboratory has fabricated a pellet injection system to test the mechanical and thermal properties of extruded tritium. This repeating single-stage, pneumatic, Tritium-Proof-of-Principle Phase II (TPOP-II) Pellet Injector has a piston-driven mechanical extruder and is designed to extrude and accelerate hydrogenic pellets sized for the ITER device. Tritium and deuterium-tritium (D-T) pellets have been produced in experiments at the Los Alamos National Laboratory Tritium Systems Test Assembly. About 38 g of tritium was used in the experiment. This paper presents results of the TPOP-II extrusion experiments. These extrusion experiments indicate that both T2 and D-T will require higher extrusion forces than D2 by about a factor of 2 and that the flow of the material may be characterized by static and dynamic shear strengths.

Tritium recovery from tritiated water with a two-stage palladium membrane reactor

A process to recover tritium from tritiated water has been successfully demonstrated at the Tritium Systems Test Assembly. A 2-stage palladium membrane reactor (PMR) is capable of recovering tritium from water without generating additional waste. A substantial amount of tritiated water will be generated in the International Thermonuclear Experimental Reactor both from torus exhaust and auxiliary operations. In addition, this process can be used to recover tritium from tritiated waste water being prepared for disposal in radioactive waste repositories. A large quantity of tritiated waste water exists world wide because the predominant method of cleaning up tritiated streams is to oxidize tritium to tritiated water. A series of cold (non-tritium) water processing experiments were run in preparation for the tritiated water processing tests. Optimum operating conditions were determined with these experiments. Tritium was recovered from a container of molecular sieve loaded with 2050 g (2550 std. l) of water and 4.5 g of tritium. The maximum water processing rate for the PMR system used was determined to be 0.5 slpm. A control method was developed to automatically liberate steam from the molecular sieve and add the desired amount of CO before injection into the PMR system. The maximum decontamination factor achieved in the 1st stage ranged from 100 to 260, depending on the inlet flow rate. Performance of the 2nd stage could not be measured because the outlet tritium concentration was below the background of the ion chamber used for analysis. Although the DF could not be measured, it is known that the DF was high because no tritium was detected, except during start-up, in the tritium waste treatment system that was downstream from the PMR system.

TFTR Tritium Operations Lessons Learned

The Tokamak Fusion Test Reactor which is the progenitor for full D-T operating tokamaks has successfully processed > 81 grams of tritium in a safe and efficient fashion. Many of the fundamental operational techniques associated with the safe movement of tritium through the TFTR facility were developed over the course of many years of DOE tritium facilities (LANL, LLNL, SRS, Mound). In the mid 1980`s The Tritium Systems Test Assembly (TSTA) at LANL began reporting operational techniques for the safe handling of tritium, and became a major conduit for the transfer of safe tritium handling technology from DOE weapons laboratories to non-weapon facilities. TFTR has built on many of the TSTA operational techniques and has had the opportunity of performing and enhancing these techniques at America`s first operational D-T fusion reactor. This paper will discuss negative pressure employing `elephant trunks` in the control and mitigation of tritium contamination at the TFTR facility, and the interaction between contaminated line operations and {Delta} pressure control. In addition the strategy employed in managing the movement of tritium through TFTR while maintaining an active tritium inventory of < 50,000 Ci will be discussed. 5 refs.

Inertial Confinement Fusion Target Filling at Los Alamos National Laboratory

Capability to fill inertial confinement fusion (ICF) targets with DT has recently been established at the Weapons Engineering Tritium Facility (WETF) at Los Alamos National Laboratory (LANL). The target filling system provides DT-filled glass targets for the U.S. National ICF Program. Tritium storage, purification, mixing, analysis, and high pressure capabilities at WETF are used to provide DT at pressures up to 400 atm to a target filling cell that can operate at temperatures to 400{degree}C. Isotopically pure tritium is obtained from the Tritium Systems Test Assembly at LANL, and typically has purities of 99% tritium or better. At WETF, a palladium-silver diffuser is used for removal of decay {sup 3}He from tritium prior to mixing with deuterium. After preparation, DT mixtures are stored in a passivated volume to minimize impurity accumulation from stainless steel. Analysis of tritium and DT mixtures is performed with a quadrupole mass spectrometer/beta scintillation detector system that utilizes an analytical technique previously developed at LANL to provide hydrogen isotope, helium, and impurity analysis. Glass targets are filled in aluminum eggcrates. The target filling cell has been designed to contain two eggcrates while maintaining isothermal conditions across the eggcrates during diffusion filling of targets. Results from amore » cryogenic condensation technique performed at Lawrence Livermore National Laboratory have confirmed the fill pressures. 3 refs., 5 figs., 1 tab.« less

Twelve Years of Tritium Operating Experience at the Tritium Systems Test Assembly

The Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory is a facility designed to develop and demonstrate, in full scale, selected tritium systems necessary for safe and efficien...

Studies of Tritiated Water Absorbed on Molecular Sieve

A systematic program to study the long term generation of gas which results when tritiated water is stored on molecular sieve is ongoing at the Tritium Systems Test Assembly (TSTA). The gas that is generated, as the tritium decays, may result in safety concerns because the pressure generated in the disposal container may lead to a failure of the container. The study addresses both the effects of helium evolution and the effects of radiolysis. The results of the study will aid in the development of packaging requirements for the final disposal container. 1 ref., 9 figs., 1 tab.

Demonstration of Regulatory Process Controls on the TSTA Cryogenic Distillation System

Due to the nature of its fusion reaction, most of the DT used to fuel the International Thermonuclear Experimental Reactor (ITER) will exit unreacted. This effluent will be purified by the fuel cleanup system so that the isotope separation system (ISS) will be fed only hydrogen isotopes and possibly some helium. The ISS will separate this feed into streams nominally composed of He/H2/HD, D2, DT, and T2. These products will be recycled through appropriate fueling systems back to the fusion reactor or returned to storage. The ISS will have to respond properly to feed conditions which change rapidly. Feed changes will propagate quickly through the entire cascade. It is apparent that an automatic control system will be required to respond to these changes and maintain product quality. It is convenient to divide such a distillation control system into `regulatory` controls and composition controls. The `regulatory` controls include liquid levels, flowrates, reboiler heats, and pressure. A system of regulatory controls has been devised and installed on the four-column cascade of cryogenic columns at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. This tritium-compatible ISS is designed for approximately 1/10 the ITER flowrate and is dedicated to fusion fuel studies. Details of the TSTA ISS have been published previously.