Form Of Tritium Research Articles

Study of the behavior of tritiated water vapor in concrete materials

A fusion reactor requires high levels of safety and public acceptability, so safe confinement of tritium is a key issues. In order to understand tritium behavior in the concrete reactor components, sorption and desorption experiments were carried out as a function of the exposure time for sorption and the kind of purge gas for desorption. The samples used were cement paste, mortar and concrete. These samples were exposed to 740–1110Bq/cm3 of HTO vapor in 1kPa of H2O vapor at room temperature. Desorption experiments were carried out using N2 purging gas with water vapor by the thermal desorption method. The concrete materials became almost saturated with HTO within about 1–6months. Though the chemical form of tritium desorbed from the sample was mostly HTO, HT and other compounds were also detected above 673K.

Tissue-specific incorporation and genotoxicity of different forms of tritium in the marine mussel, Mytilus edulis

Marine mussels (Mytilus edulis) were exposed to seawater spiked with tritiated water (HTO) at a dose rate of 122 and 79 μGy h−1 for 7 and 14 days, respectively, and tritiated glycine (T-Gly) at a dose rate of 4.9 μGy h−1 over 7 days. This was followed by depuration in clean seawater for 21 days. Tissues (foot, gills, digestive gland, mantle, adductor muscle and byssus) and DNA extracts from tissues were analysed for their tritium activity concentrations. All tissues demonstrated bio-accumulation of tritium from HTO and T-Gly. Tritium from T-Gly showed increased incorporation into DNA compared to HTO. About 90% of the initial activity from HTO was depurated within one day, whereas T-Gly was depurated relatively slowly, indicating that tritium may be bound with different affinities in tissues. Both forms of tritium caused a significant induction of micronuclei in the haemocytes of mussels. Our findings identify significant differential impacts on Mytilus edulis of the two chemical forms of tritium and emphasise the need for a separate classification and control of releases of tritiated compounds, to adequately protect the marine ecosystem.

Influence of water concentrations to chemical forms of tritium generated in mercury through a nuclear reaction

In order to assess internal doses from tritium at the JSNS (Japan Spallation Neutron Source), a correlation between the chemical forms of tritium generated in mercury and the amount of water in it was studied by simulating the generation of tritium in the mercury target used for a neutron source. The ratio of [(3)H]H(2)O to all of generated tritium increased with increase in the water concentration. An empirical equation to predict the content of [(3)H]H(2)O from the water concentration is proposed.

Function of Water Molecule for Tritium Behavior on the Water-Metal Boundary

In order to investigate the function of water molecule for tritium transport behavior on the water-metal boundary, a series of experiments of tritium permeation into humid atmosphere was performed through pure iron piping with different surfaces of oxide etc., which contained about 1 kPa of pure tritium gas at 423 K. Chemical forms of tritium permeated into water were monitored continuously under purging outer jacket by <1000ppm of water vapor in Ar. Observation of metal surfaces was also carried out by Secondary Electron Microscope (SEM) and X-Ray Diffraction (XRD) analysis. The results were compared with those permeated into pressurized liquid water at 423 K.The actual tritium permeation rate into Ar with <1000 ppm of water vapor was not clearly changed that into liquid water. In the vapor atmosphere, a magnetite layer did not grow on the surface clearly, and tritium permeation rate and chemical species (∼100% of HTO) through pure iron piping with mechanically polished surface were not changed drastically comparing with data with a magnetite surface. On the other hand, hydrogen gas (HT) fraction of tritium permeated into the outer jacket increased drastically in case of a gold plating surface.

Tritium release from beryllium articles for use in fusion devices

Results obtained on radiation and magnetic field (MF) effects on tritium release at annealing of the beryllium pebbles from the EXOTIC-8-3/13 irradiation are presented in this study and compared with those for other irradiated beryllium materials. Abundance ratios of chemical forms of tritium in the EXOTIC-8-3/13 beryllium pebbles were determined: T 2 – 65%, T 0 – 23%, T + – 12%. A complete detritiation of these pebbles was achieved at 1123 K for 240 min; MF of 2.35 T had no appreciable effect on the tritium release. At 991 K for 240 min, the degree of detritiation was 96.6% without MF; MF of 2.35 T decreased it to 86.7%. At 940 K for 47 min, the degree of detritiation was 60%, 5 MeV fast-electron radiation of 14 MGy/h increased it to 76%, but the simultaneous action of the fast-electron radiation and MF of 1.7 T increased it to 88%.

Tritium sorption and desorption from JET beryllium tiles under temperature, electron radiation and magnetic field

Tritium release at annealing of samples cut from beryllium tiles exposed to D + D, D + T plasma in the Joint European Torus (JET) was investigated under 5 MeV fast-electron radiation at the dose rate 14 MGy h −1 and in a magnetic field of 1.7 T separately and simultaneously in order to evaluate possible effects of these factors. Abundances of chemical forms of tritium—molecular T 2 (44%), atomic T 0 (42%), and ionic T + (14%) and their distribution were determined in the plasma-exposed beryllium samples with lyomethods. Fast-electron radiation considerably increased the fractional tritium release at annealing by a factor of approximately 5. The magnetic field increased the fractional tritium release at annealing both without and under electron radiation by approximately 25%. Sorption of tritium in samples cut from an unused beryllium tile was investigated at about 1 Pa of T 2 at 773 K for 3 h both with and without simultaneous electron radiation and/or magnetic field. The amount of the tritium sorbed was found to be 3.3–6.2 kBq cm −2. Most of the tritium sorbed was localized in subsurface layers up to a depth of 80 μm of the beryllium samples.

The Effect of Water on Tritium Release Behavior from Solid Breeder Candidates

The authors have made a tritium release model to represent the release behavior of bred tritium from solid breeder materials using a series of studies.It has been observed that a large amount of adsorbed water and water produced by water formation reaction are released to the purge gas even though dry purge gas with hydrogen is introduced to solid breeder materials. According to our tritium release model, the presence of water in the purge gas and surface water on the material has a large effect on the tritium release behavior. In this study, the authors quantified the amount of adsorbed water and the capacity of the water formation reaction for various solid breeder materials (Li2TiO3, Li4SiO4, Li2ZrO3, LiAlO2). The effect of surface water on the chemical form of tritium released from the LiAlO2 blanket is also discussed in this study.

Observation of Tritium Distribution in Iron Oxide with Tritium Micro Autoradiography

To clarify the tritium permeation behavior, tritium distribution in iron oxidized in high temperature water was observed with tritium micro autoradiography. It was found that tritium was distributed homogeneously in the iron metal. However the oxide surface (magnetite) was found to contain a very low concentration of tritium. The inner layer of oxide could strongly effect the tritium permeation. From a comparison with the permeation experiment that had been reported in Ref. 1, it was suggested that tritium would mainly diffuse other path except the oxide lattice. According to the chemical form of tritium, which was released from iron surface into water, two assumptions were suggested. One is based on the different combination of tritium on the water-surface interface. The other is based on the oxidation mechanism.

Biological hazard issues from potential releases of tritiated dust from ITER

Tritiated dust will be generated during the operation of ITER. Several dedicated studies, involving in vitro and in vivo experiments on metal tritide and carbon tritide dust, have raised some concerns about the protection guidelines for workers exposed to tritiated dust, because they are based on the radiotoxicity of tritium in form of tritium gas (HT), tritiated water (HTO) or organically bound tritium (OBT). While the behaviour of HT, HTO and OBT in the human body is well understood, the same is not fully true for tritiated dust, because of the size distributions of the particles and the variety of base materials. The in vivo and in vitro studies on tritiated dust have shown the dependence of the tritium clearance and retention in the human body on the physico-chemical parameters (mainly the particle size and density). Some recent studies on JET dust relative to graphite and CFC dust have dealt with physico-chemical characterisation and with in vitro tritium dissolution studies. Some uncertainties still remain and further testing is necessary (in vitro and in vivo), restricting if possible the investigation to relevant parameters. In the working areas of ITER the radioactive dust concentration in air will be kept at a very low level. However, in the event of an accident, the atmosphere inside the buildings, including operative areas as well, might get contaminated with small amounts of tritiated dust and hence inhalation risks cannot be excluded. The present paper summarizes the results from previous investigations on the subject which are relevant in deriving sound dose conversion factors for tritiated dust. At the same time, it proposes a dose conversion factor for dust which is conservative in order to envelope all ITER relevant materials.

Tritium release from lithium silicate pebbles produced from lithium hydroxide

Out-of-pile tritium release experiments were performed using slightly overstoichiometric lithium orthosilicate pebbles irradiated in a research reactor. In the experiments, various sweep gases such as pure nitrogen gas, hydrogen/nitrogen gas and water vapor/nitrogen gas are used to investigate the effect of difference in sweep gas composition on tritium release form the breeder material. The experimental results indicate that almost all tritium was released as tritiated water vapor regardless of difference in gas composition of the sweep gas. Most of the chemical form of tritium released was tritiated water vapor even when the hydrogen/nitrogen gas was used for the sweep gas. It was also found that a considerably smaller amount of tritium was released when a pure nitrogen gas was used as the sweep gas. Comparison of the experimental results reveals that the tritium release rate is substantially enhanced if the sweep gas contains water vapor.

The Background Levels of Atomospheric Tritium Concentration in the Environment at NIFS Toki Site of Japan

Atmospheric tritium concentrations at the National Institute for Fusion Science (NIFS) Toki site of Japan, where the Large Helical Device (LHD) has been operating, were measured considering future deuterium plasma experiments and environmental safety. The major chemical forms of atmospheric tritium are water (HTO), hydrogen (HT) and methane (CH3T). Average tritium concentrations of HTO, HT and CH3T observed from January 2003 to March 2006 were 9.0 mBq/m3, 9.0 mBq/m3 and 2.0 mBq/m3, respectively. To examine about the systematic error of the air sampling device, we cross-checked with the sampling device of Kumamoto University. The values obtained with both devices were almost consistent. The HTO concentration principally depends on humidity in air. The HTO concentration in the collected water and the HT concentration tend to show seasonal variation. The atmospheric tritium levels measured at Toki were consistent comparing with another environmental values measured in Japan. The present atmospheric tritium monitoring would be useful for safety consideration.

Tritium Speciation in Nuclear Reactor Bioshield Concrete and its Impact on Accurate Analysis

Tritium ((3)H) is produced in nuclear reactors via several neutron-induced reactions [(2)H(n, gamma)(3)H, (6)Li(n, alpha)(3)H, (10)B(n, 2alpha)(3)H, (14)N(n, (3)H)(12)C, and ternary fission (fission yield <0.01%)]. Typically, (3)H is present as tritiated water (HTO) and can become adsorbed into structural concrete from the surface inward where it will be held in a weakly bound form. However, a systematic analysis of a sequence of subsamples taken from a reactor bioshield using combustion and liquid scintillation analysis has identified two forms of (3)H, one weakly bound and one strongly bound. The strongly bound tritium, which originates from neutron capture on trace lithium ((6)Li) within mineral phases, requires temperatures in excess of 350 degrees C to achieve quantitative recovery. The weakly bound form of tritium can be liberated at significantly lower temperatures (100 degrees C) as HTO and is associated with dehydration of hydrous mineral components. Without an appreciation that two forms of tritium can exist in reactor bioshields, the (3)H content of samples may be severely underestimated using conventional analytical approaches. These findings exemplify the need to develop robust radioactive waste characterization procedures in support of nuclear decommissioning programs.

Doses resulting from the inhalation of ITER generated tritiated dust

Dust particles generated in the ITER Tokamak may include some tritium. The coefficients derived to evaluate the doses resulting from the inhalation of the usual forms of tritium (gaseous, tritiated water, organic form) are not adequate for tritiated dust. This paper is a short literature study about the assessment of tritiated dust dose coefficients and their significance in worker, population and environmental protection for ITER

Tritium Release Behavior from Steels Irradiated by High Energy Protons

Thermal desorption behavior of tritium has been investigated for SS316 and F82H irradiated by 580MeV proton (SINQ-target3) up to 5.0 ∼5.9 dpa and 6.3∼9.1 dpa, respectively, in order to understand tritium transport in the irradiated materials. While the tritium release has only one peak at 670 K from irradiated SS316, that has two peaks at 510 K and 670 K from irradiated F82H. Those results indicate that only one kind of trap site exists in the SS316, and at least two kinds of trap site exist in F82H. As the results of tritium transport analysis of tritium release behavior, it was found that the trap site at 670 K for SS316 and F82H could be controlled by the same trap mechanism. As to the chemical form of tritium released from the steels, 1/2 and 1/3 of tritium was release as water vapor form from SS316 and F82H, respectively. It could be attributed to the growth of surface oxide on the metal surfaces during the TDS.

Tritium Behavior on the Water-Metal Boundary for the Permeation into Cooling Water Through Metal Piping

In order to accumulate data on tritium transferred to cooling water of a fusion reactor, a series of experiments of tritium permeation into water jacket pressurized to 0.8MPa by He gas was performed through pure iron piping, which contained about 1 kPa of pure tritium gas at 423 K. Chemical forms of tritium permeated into water were monitored periodically under continuous purging water jacket by He. Observation of metal surface was also carried out periodically by SEM and XRD analysis.The actual tritium permeation rate was about 1/5 level of the calculated value. Even if surface oxide layer (magnetite, porous & fine layers) grew in the water boundary, tritium permeation rate to water was not changed drastically. On the other hand, hydrogen gas (HT) fraction of tritium permeated in water jacket decreased drastically with oxide layer growth. Furthermore, permeated species and amounts were not affected clearly by the dissolved hydrogen in water by purging 1% H2 in He.

Magnetic Field Effects on Tritium Release from Neutron-Irradiated Beryllium Pebbles

The effects of temperature, magnetic field (MF), and ionizing radiation on the release of tritium from the Be pebbles irradiated in the BERYLLIUM experiment in 1994 in Petten, The Netherlands (irradiation neutron fluence 1.24 × 1025 m-2, irradiation temperature 780 K, and 3H content 7 appm) were investigated in this study. Simultaneous action of these factors corresponds to the real operating conditions of the blanket of a fusion reactor. The total amount of tritium in a separate pebble, the chemical forms of localized tritium (T0, T2, and T+), and the tritium distribution in the pebble volume were determined by a lyomethod (dissolution). Thermoannealing experiments were performed at a constant temperature of 1123 K for 2 h under the following conditions: separately both in MF (1.7 T) and under fast electron radiation (E = 5 MeV; P = 14 MGy·h-1) as well as under the action of all three factors. Tritium in the Be pebbles is localized for the most part as T2 (85 to 94%). The abundances of T+ (4 to 5%) and T0 (5 to 10%) are little. The tritium distribution in a pebble is not uniform; most of the tritium is localized in the inner part of a pebble. An MF of 1.7 T decreases slightly the fractional release of tritium under the given conditions of thermoannealing (from 30 to 25%), the fast electron radiation increases it (from 30 to 40%), but the simultaneous action of the MF and radiation increases it (from 30 to 54%). The effects observed are explained that the MF and radiation affect the concentration of main diffusing particles T0 in a beryllium matrix.

Chemical form of released tritium from solid breeder materials under the various purge gas conditions

Understanding of the release behavior of bred tritium from solid breeder materials is necessary to design tritium recovery system from blanket of a fusion reactor because permeation loss of bred tritium in the piping system or type of tritium recovery system depends on the tritium release behavior. It has been reported by the present authors that behavior of tritium release from solid breeder grain is consisted of diffusion in grain, tritium transfer at surface layer and surface reactions on grain surface such as adsorption or isotope exchange reactions. Chemical form of released tritium from Li 4SiO 4 (from FzK), LiAlO 2 (from JAERI), Li 2TiO 3 (from CEA) and Li 2ZrO 3 (from MAPI) under various purge gas condition is discussed in this study by using the data obtained from the out-of-pile tritium release experiment in JAEA and fitting results estimated by the tritium release model formed by the present authors. And then the tritium release behavior and chemical form of tritium in the test blanket module with solid breeder under the ITER condition is also discussed based on the estimation obtained using the tritium release model.

Ratio of tritiated water and hydrogen generated in mercury through a nuclear reaction

Tritium generated in a mercury target is a source of potential exposure of personnel at high-energy accelerator facilities. Knowledge of the chemical form of tritium is necessary to estimate the internal doses. We studied the tritium generation upon thermal neutron irradiation of a mercury target modified into liquid lithium amalgam to examine the ratio of tritiated water ([ 3H]H 2O) and tritiated hydrogen ([ 3H]H 2). The ratio between [ 3H]H 2O and [ 3H]H 2 generated in lithium amalgam was 4:6 under these experimental conditions.

Radiation quality of tritium

Tritium occurs from both natural and manufactured processes. In the environment, tritium can exist in the form of tritiated water (HTO) and in an organic form known as organically-bound tritium (OBT). Although, the concentrations of environmental OBT are relatively low, there is concern that current risk factors may underestimate the risk from OBT. Because tritium poses an internal hazard at cellular levels, microdosimetric techniques provide suitable tools for the study of radiation quality of tritium. In this study, microdosimetric simulations are performed for tritium uniformly distributed in a medium, and for tritium bound to biologically critical sites of dimensions from 10 nm to 2 microm. Results of local energy density are different for these two cases in microscopic regions. Based on the spatial distribution of energy deposition, dose mean lineal energies are calculated for tritium in forms of HTO and OBT. The dose mean lineal energies of OBT are about a factor of 1.7 higher than those of HTO in a wide range of target dimensions of biological interest. The results are consistent with radiobiological findings that OBT is about twice as effective as that of HTO.

Behaviour of three chemical forms of tritium in the environment after release from inertial fusion reactors

The increasing demand of results on the environmental effects of the tritium emission is due to the closer construction of experimental fusion facilities arriving to ignition and gain, and experimental reactor (ITER). In the current design studies, it is indicated that the elementary tritium (HT) overcomes in some cases in one order of magnitude the effects of potential releases of tritiated water vapour (HTO). To fully simulate the behaviour of these chemical forms, diffusion and deposition processes in the soil and vegetables need to be considered. In addition, this work also incorporates the penetration in the underground, re-emission and later conversion to organic tritium (OBT). he whole study drives to the conclusion that the behaviour of the tritium should be simulated using two well-differentiated studies: deterministic and probabilistic. Deterministic calculations are based on fixed meteorological data given a priori. The probabilistic study is based on measured real meteorological analysis every hour, and the probability that individuals can present dose for internal irradiation is considered. Both options have been considered for an specific environment of Mediterranean location of the system. Then the elementary tritium has been deposited in the soil, it can be oxidized by microbial action of the enzymes of the soil, and the resulting tritium form (in its oxidize form) goes back to the atmosphere. This process of re-emission is shown to be very important in this work since it has been typically considered that the inhaled tritium is only, HTO, when, in fact part of that account is due to the HT converted to HTO and re-emitted to the atmosphere. All our calculations have parallel also been performed for HTO, and they demonstrate that the HT contributes very significantly to the dose for inhalation through the re-emission processes. A final aspect of this work is the dosimetric analysis of the contamination through all ways: inhalation, re-emission and ingestion. Early and chronic doses have been evaluated.