Yasunori Iwai

Analysis of hydrogen isotopes with a micro gas chromatograph

Abstract In the fuel cycle system of fusion reactors, analysis of hydrogen isotopes is very important from the view point of system control. The gas chromatograph (GC) with cryogenic separation column (cryogenic GC) is one of the most extensively used methods for the analysis of hydrogen isotopes. The micro GC with cryogenic column is expected to improve analysis time, that is a major disadvantage of conventional GC. The present authors have modified the micro GC to use its separation column at cryogenic temperature for H2, HD and D2 mixture analysis. Obtained retention time of H2, HD and D2 was about 85, 100 and 130 s, respectively. Peak resolution between H2 and HD, these are nearest each other, was about 1.0. These result suggests that the column developed in this work attained the practical level for the separation of hydrogen isotopes without tritium. Present detection limit of hydrogen isotopes was about 100–200 p.p.m., and it can be improved further by adjustment of separation column.

Tritium behavior in the Caisson, a simulated fusion reactor room

Abstract In order to confirm tritium confinement ability in the deuterium–tritium (DT) fusion reactor, intentional tritium release experiments have been started in a specially fabricated test stand called ‘Caisson’, at Tritium Process Laboratory in Japan Atomic Energy Research Institute. The Caisson is a stainless steel leak–tight vessel of 12 m3, simulating a reactor room or a tritium handling room. In the first stage experiments, about 260 MBq of pure tritium was put into the Caisson under simulated constant ventilation of four times air exchanges per h. The tritium mixing and migration in the Caisson was investigated with tritium contamination measurement and detritiation behavior measurement. The experimental tritium migration and removal behavior was almost perfectly reproduced and could almost be simulated by a three-dimensional flow analysis code.

Simulation study of intentional tritium release experiments in the caisson assembly for tritium safety at the TPL/JAERI

Abstract At the Tritium Process Laboratory (TPL) in Japan Atomic Energy Research Institute (JAERI), Caisson assembly for tritium safety study (CATS) with 12 m 3 of large airtight vessel (Caisson) was fabricated for confirmation and enhancement of fusion reactor safety to estimate the tritium behavior in the case, where the tritium leak accident should happen. One of the principal objectives of the present studies is the establishment of simulation method to predict the tritium behavior after the tritium leak accident should happen in a ventilated room. As for the understanding of initial tritium behavior until the tritium concentration become steady, the precise estimation of local flow rate in a room and time-dependent release behavior from the leak point are essential to predict the tritium behavior by simulation code. The three-dimensional eddy flow model considering, tritium-related phenomena was adopted to estimate the local flow rate in the 50 m 3 /h ventilated Caisson. The time-dependent tritium release behavior from the sample container was calculated by residence time distribution function. The calculated tritium concentrations were in good agreement with the experimental observations. The primary removal tritium behavior was also investigated by another code. Tritium gas concentrations decreased logarithmically to the time by ventilation. These observations were understandable by the reason that the flow in the ventilated Caisson was regarded as the perfectly mixing flow. The concentrations of tritiated water measured, and indications of tritium concentration by tritium monitors became gradually flat. This phenomena called ‘tritium soaking effect’ was found to be reasonably explained by considering the contribution of the exhaustion velocity by ventilation system, and the adsorption and desorption reaction rate of tritiated water on the wall material which is SUS 304. The calculated tritium concentrations were in good agreement with the experimental observations. The effectiveness of these codes was thus proved.

Simulation of Tritium Behavior after Intended Tritium Release in Ventilated Room

At the Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI), Caisson Assembly for Tritium Safety study (CATS) with 12 m3 of large airtight vessel (Caisson) was fabricated for confirmation and enhancement of fusion reactor safety to estimate tritium behavior in the case where a tritium leak event should happen. One of the principal objectives of the present studies is the establishment of simulation method to predict the tritium behavior after the tritium leak event should happen in a ventilated room. The RNG model was found to be valid for eddy flow calculation in the 50m3/h ventilated Caisson with acceptable engineering precision. The calculated initial and removal tritium concentration histories after intended tritium release were consistent with the experimental observations in the 50 m3/h ventilated Caisson. It is found that the flow near a wall plays an important role for the tritium transport in the ventilated room. On the other hand, tritium behavior intentionally released in the 3,000 m3 of tritium handling room was investigated experimentally under a US-Japan collaboration. The tritium concentration history calculated with the same method was consistent with the experimental observations, which proves that the present developed method can be applied to the actual scale of tritium handling room.

Results of experimental study on detritiation of atmosphere in large space

In order to obtain data on tritium removal from the atmosphere of a room, which is needed for designing a reliable effective atmosphere detritiation system in a fusion reactor and for detailed analysis on its safety, intentional tritium release experiments have been carried out in a controlled space called Caisson under various atmosphere conditions at the Tritium Process Laboratory (TPL) in Japan Atomic Energy Research Institute (JAERI). In case there is no tritiated water vapor in the released tritium gas, tritium was ideally removed by constant ventilation in spite of atmospheric conditions and residence time. On the other hand, when tritiated water vapor existed in the released tritium, residual contamination on the wall of the Caisson was detected and it was found that it depended on the initial humidity in the Caisson. This tritium removal behavior was successfully analyzed by considering the adsorption and desorption reaction rates of tritiated water on the wall by the constant ventilation.

A Steady-state Simulation Model of Gas Separation System by Hollow-filament Type Membrane Module

A steady-state simulation model of the gas separation system using a hollow-filament type membrane has been proposed. The mass transfer coefficients in the non-porous thin layer, in the porous support layer of the membrane and in the boundary layer of the membrane surface are estimated in the model. The four types of flow patterns: cross flow, mixing flow, concurrent flow and counter current flow, are also considered in the model. The mass transfer through the non-porous thin layer of the membrane controls the overall mass transfer by ~99%. The experimental observations of TPL (Tritium Process Laboratory in JAERI) for N2–H2 and Air—H2 systems agreed with the calculated results of the cross flow under a set of typical conditions (disposal volume of 2.78×10−3 Nm3/s, feed-side pressure of 3.44×105Pa, and permeated-side pressure of 1.07×104 Pa). The validity of the simulation method was thus proved. For Air-H2-H2O system also, the recovery ratios calculated for H2 are in good agreement with the experimental o...

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

Abstract 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 m3 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.

Tritium decontamination of TFTR carbon tiles employing ultra violet light

Tritium decontamination on the surface of Tokamak Fusion Test Reactor (TFTR) bumper limiter tiles used during the Deuterium-Deuterium (D-D) phase of TFTR operations was investigated employing an ultra violet light source with a mean wavelength of 172 nm and a maximum radiant intensity of 50 mW/cm 2 . The partial pressures of H 2 , HD, C and CO 2 during the UV exposure were enhanced more than twice, compared to the partial pressures before UV exposure. In comparison, the amount of O 2 decreased during the UV exposure and the production of a small amount of O 3 was observed when the UV light was turned on. Unlike the decontamination method of baking in air or oxygen, the UV exposure removed hydrogen isotopes from the tile to vacuum predominantly in forms of gases of hydrogen isotopes. The tritium surface contamination on the tile in the area exposed to the UV light was reduced after the UV exposure. The results show that the UV light with a wavelength of 172 nm can remove hydrogen isotopes from carbon-based tiles at the very surface.

A design study of water detritation and hydrogen isotope separation systems for ITER

A preliminary design study of the combined electrolysis catalytic exchange (CECE) process by using existing plant technology at the Japanese heavy water reactor (Fugen) was carried out for International Thermonuclear Experimental Reactor (ITER). Because of the large separation factor, the required column height is substantially reduced compared with a vacuum water distillation (WD) column. A preliminary design study of the hydrogen isotope separation system (ISS) was carried out for the ITER fuel cycle based on a reduced hydrogen gas flow rate from the CECE plant, and a new ISS cascade configuration was proposed. Major process parameters of the CECE and the ISS were compared with the ITER-FDR process systems.

Application of Pressure Swing Adsorption to Water Detritiation Process

Pressure swing adsorption (PSA) has been studied as a new water processing technique, detritiation of tritiated water, for a future fusion power plant. It will be a new tritium removal method having an additional function of isotope separation and quick dehydration, and it is expected to become the first stage of the system processing a large amount of tritiated water generated in a fusion plant. A series of the adsorption and dehydration experiments was carried out for a typical adsorbent of NaX zeolite as fundamental investigation to realize HTO/H2O separation system by PSA. It was clearly observed that break through time differs in H2O and HTO concerning the isotope separation function of NaX zeolite. It is certain that NaX zeolite can separate into the tritium concentrated water and the tritium reduced water by this difference of the break through time. The quick dehydration is attained by decompression and purge gas flowing. It was observed that a part of amount of the water released by the decompression was transferred by the purge gas, and the rest of water was adsorbed on the adsorbent again. After the re-adsorption phenomenon, the rest of water was gradually released by the diffusion. It is demonstrated that enlargement of pressure difference between adsorption and dehydration is effective to obtain high dehydration ratio. Furthermore, it was also verified that enough vapor removal capacity of purge gas is quite necessary to obtain high dehydration ratio.