R. Scott Willms

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.

Fuel cleanup systems for fusion fuel processing

Abstract The fuel cleanup unit (FCU) of a fusion fuel processing system receives gas from the torus evacuation system which is composed of hydrogen isotopes (deuterium and tritium) contaminated with a few to possibly around 20% of impurities (methane, helium, protium, water, nitrogen, carbon oxides, ammonia, etc.). Because the usual processing step subsequent to the FCU, the cryogenic isotope separation system (ISS), operates at about 25 K, the FCU must remove all impurities which freeze at this temperature to prevent ISS plugging. This includes all impurities other than helium and protium. This first function of the FCU is termed purification. The collected impurities contain a substantial quantity of tritium largely in the form of methane (and higher hydrocarbons), water and ammonia. This tritium must be recovered from the impurities before the remaining non-radioactive components may be discarded. This second function is referred to as tritium recovery. To accomplish these two functions a number of techniques have been proposed and investigated. This paper presents these techniques and identifies the merits and range of utility of each method.

Initial testing of a low pressure permeator for tritium processing

Early tritium processing system operations led to relatively large environmental releases of tritium. This led to the inclusion of large oxidation/ adsorption systems between tritium systems and the facility stack so that tritium would be collected as water. While this has been an improvement, it has resulted in large amounts of tritiated water waste. Also it has been recognized for some time that there are cases where inert gases such as Ar, He and N2 are mixed with small amounts of tritium. If this gas is processed with oxidation/adsorption, the tritium, which is in the useful T2 form, is converted to the useless T2O form. Thus, there is a need for a system, which removes tritium from inert gases with high conversion so that the inert gas is ready for release to the environment, and that recovers tritium as T2. A system, which could meet this requirement, is one based on getters. The hydrogen isotopes are collected as metal hydride and the inert passes through. However, such systems are susceptible to rapid consumption of the metal if air or O2 is present.

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.

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.

Mathematical comparison of three tritium system effluent HTO cleanup systems

ABSTRACT It is important that air emissions from tritium systems be kept as low as reasonably achievable. Thus, over the years a number of gas detritiation systems have been developed. Recently there has been interest in lower-cost, simpler systems which do not convert HT to the much more hazardous HTO form. Examples of such systems are 1) a bubbler/dehumidifier, 2) a bubbler/collector, and 3) an adsorber/collector. A computer model of each configuration was written and run. Each system’s performance, including tritium buildup in liquid water, and tritium exhausted to the environment, are presented and compared.

Dynamic simulation of a proposed ITER tritium processing system

ABSTRACTDynamically simulating the fuel cycle in a fusion reactor is crucial to developing a better understanding of the safe and reliable operation of this complex system. In this work, we propose...

SURFACE CHARACTERIZATION OF TFTR FIRST WALL GRAPHITE TILES USED DURING DT OPERATIONS

ABSTRACT Surface characterization studies were performed on graphite tiles used as first wall materials during DT operation of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. These ex situ analysis studies revealed a number of interesting and unexpected features. In this work we examined the spatial and (where possible) the depth distribution of impurity species deposited onto the plasma facing surfaces using X-ray Photo-electron Spectroscopy (XPS) and Secondary Ion Mass Spectrometry (SIMS). This work determined that beyond the predominant species of carbon and oxygen, common impurities included silicon, boron, lithium and sulfur. Oxygen content in the plasma facing tile surfaces ranged from 20 to 50 atomic percent [excluding H-isotopes], clearly indicating an extensive zone of oxidized carbon. By contrast, carbon tile surfaces not exposed to the plasma have surface oxygen contents ranging from 2 to 6 atomic percent. Analytical measurements of the secondary impurities (B, Li, Si, S) levels were on the order of 1–4 atomic percent, (boron and lithium were injected for wall conditioning in TFTR.) The core level binding energies of these impurity species were consistent with the presence of common oxides or hydroxides (e.g., BxOy, Li2O, LiOH, Silicates). XPS measurements performed in concert with depth profiling indicated that the tile oxidized zone was significantly deeper than 1 micrometer into the (averaged) surface. Surface analytical results clearly indicate that plasma operations clearly redeposit injected impurities (Li, B) and the depth profiles and distributions of the hydrogen isotopes may be impactedand/or influenced by this deposition process. An attempt at determining hydrogen isotope concentration distributions was made using positive ion SIMS. Specific regions of some surfaces clearly indicated the presence of m/z=3 (HD, T) and m/z=15 (CH3, CHD, CT). Preliminary data examination using positive ion SIMS examination indicates that these mass markers are substantially higher in the near surface region when compared with spectra recorded deeper in the surface region. The deuterium and tritium concentrations were; however, sufficiently low or compromised bycommon isobaric interferencesthat accurate isotopic distributions using SIMS were not possible. These findings are in agreement with results reported by others. [Morimoto et al, Sun et al, reference 3 Haasz et al]

US tritium activities

ABSTRACT Since the last international tritium conference in 1995 the US continues its active interest in better understanding tritium and in using it safely and efficiently. US governmental tritium interests center around five major activities: 1) inertial confinement fusion, 2) fusion energy sciences (both magnetic and inertial), 3) tritium facility decontamination and decommissioning, 4) tritium production and 5) national defense applications. While the US interests have, roughly speaking, stayed the same, there have been significant changes in the US tritium community. There have been shifts in program emphases and changes in US tritium facilities with certain facilities either shutdown or being shutdown, and with new facilities under construction. This paper will provide an-overview of the US tritium activities and associated facilities.

Tritium Loss in Molten Flibe Systems

Abstract An emerging issue relative to beryllium technology in fusion involves tritium interactions with molten beryllium-bearing salts. Innovative designs for fusion reactors, both magnetic and inertially confined, feature the molten salt mixture 2LiF•BeF2, commonly called Flibe, as a tritium breeder and coolant. Tritium is bred in the Flibe as neutrons from the plasma are absorbed by Li atoms, which then transmute to tritium and helium. Transmutation of tritium from Be also occurs. Among the issues to be resolved for such coolant systems is the potential loss of tritium from the Flibe coolant to the walls of the system, particularly through heat exchanger tubes, and from there into secondary coolants or working fluids and the environment. Effectively removing tritium from Flibe in clean-up units is also important. In quiescent or low Reynolds number flow, tritium movement through Flibe is governed by diffusion. For Flibe in turbulent flow, as in heat exchanger tubes, transport is by turbulent mixing, and the same flow conditions and structural design features that maximize heat transfer to the heat exchanger walls will enhance the transport of tritium to those same surfaces. Analyses have been performed to estimate the fractional loss of tritium through heat exchanger tubes and release rates from Flibe droplets in vacuum disengagers in molten Flibe systems. The calculations suggest unacceptably large losses of tritium through heat exchanger tubes. The gravity of the implications of these estimates calls for experimental verification to determine if tritium losses through molten Flibe heat exchangers or other Flibe systems can really be so high and whether vacuum disengagers will really work. There is also a need for better information on evolution of tritium from Flibe droplets in a vacuum. Several experiments are presently being planned to address these issues and are discussed. These include experiments to induce tritium in Flibe using spontaneous fission neutrons, experiments in flowing loops to evaluated tritium losses through heat exchanger walls, and exploration of schemes for tritium extraction from molten Flibe.