Transfer of tritium in concrete coated with hydrophobic paints

An experimental study on tritium (T) transfer in porous concrete for the tertiary T safety containment is performed to investigate (i) how fast HTO penetrates through concrete walls, (ii) how well concrete walls contaminated with water-soluble T are decontaminated by a solution-in-water technique, and (iii) how well hydrophobic paint coating works as a protecting film against HTO migrating through concrete walls. The experiment is comparatively carried out using disks of cement paste which W(water)/C(cement) weight ratio is 0.6:1 with or without hydrophobic paints, and mortar disks which W/C/sand ratio is 0.6:1:2 with or without the paints. The hydrophobic paints tested in the present study are an epoxy polymer resin paint and an acrylic-silicon polymer resin one. After T exposure during specified time under a constant HTO vapor pressure in an acrylic box, the amount of water-soluble HTO on/in the disks is determined using a technique of H₂O dissolution during specified time. The results obtained here are summarized as follows: (1) HTO penetration in porous concrete can be correlated in terms of the effective diffusivity. (2) Its value in porous cement without coating is 1.2 × 10⁻¹¹ m²/s at 25 °C. (3) HTO penetrates only through pores in cement, and there is no path for HTO transfer in non-porous sand. (4) Rates of sorption and dissolution of HTO in disks of cement and mortar coated with the epoxy resin paint are correlated in terms of the effective diffusivity through the paint film which value is D_T = 1.0 × 10⁻¹⁶ m²/s. The rate-determining step is diffusion through the paint. (5) The epoxy resin paint works more effectively as an anti-HTO diffusion coating. (6) Another acrylic-silicon resin paint does not work well as anti-HTO diffusion coating. This may be because the hydrophobic property of the silicon resin paint is deteriorated with elongating the contact time with H₂O vapor or liquid. (7) The HTO uptake inside the epoxy paint is greater than that of the silicon one. (8) The permeation reduction factor (PRF) of HTO for the epoxy paint at steady-state is expected large, if HTO vapor only contributes to diffusion. However, when concrete surfaces coated with the epoxy paint are under wet conditions, the PRF value becomes smaller. All those results can be used to estimate the effect of HTO soaking in concrete walls in case of accidental T release in a fusion reactor room and to decontaminate wastes of tritiated concrete.     

Penetration of tritiated water vapor through hydrophobic paints for concrete materials

Behavior of tritium transfer through hydrophobic paints of epoxy and acrylic-silicon resin was investigated experimentally. The amounts of tritium permeating through their paint membranes were measured under the HTO concentration condition of 2–96 Bq/cm³. Most of tritium permeated through the paints as a molecular form of HTO at room temperature. The rate of tritium permeating through the acrylic-silicon paint was correlated in terms of a linear sorption/release model, and that through the epoxy paint was controlled by a diffusion model. Although effective diffusivity estimated by a diffusion model was obtained 1.1 × 10⁻¹³–1.8 × 10⁻¹³ m²/s for epoxy membranes at the temperature of 21–26 °C, its value was found to be hundreds times larger than that for cement-paste coated with epoxy paint. Hence, resistance of tritium diffusion through interface between cement-paste and the epoxy paint was considered to be the most effective in the overall tritium transfer process. Clarification of tritium transfer behavior at the interface is important to understand the mechanism of tritium transfer in concrete walls coated with various paints.     

Study on behavior of tritium in concrete wall

It is required to understand the tritium behavior in concrete for establishment of tritium safety technology of a fusion reactor or a tritium handling facility because the concrete wall is used as the final containment to prevent tritium release to the environment. This paper discusses about the effect of adsorption and diffusion of water and isotope exchange reaction between physically adsorbed water and chemically adsorbed water or structural water. It is known in this study that a large amount of tritium can be trapped to the concrete wall because cement paste has the nature of porous hydrophilic material.     

Desorption Rate of Adsorbed Tritiated Water from Molecular Sieve 5A by Exchange with Environmental Water Vapor

The desorption rate of tritiated water from molecular sieve adsorbed HTO, by exchange with the environmental water vapor, was measured. The molecular sieve, packed in a column, was initially changed with tritiated water and then humidified Ar gas was made to flow through it and the tritium concentration of effluent gas was measured. The desorption rate of tritiated water increased linearly with the water vapor pressure in the gas at constant flow rate. In the case where both the flow rate and the vapor pressure were kept constant, the amount of tritium left adsorbed on the molecular sieve decreased exponentially with time. It should be noted that the desorption rate was rather rapid even at room temperature and nearly all the tritiated water adsorbed on the molecular sieve was recovered by the flowing humidified gas at room temperature within several hours.     

Adsorption and Desorption Behavior of Tritiated Water on the Piping Materials

The behavior of tritium on the surface of various piping materials must be investigated for establishment of the safety confinement technology of tritium or for development of the effective fuel handling technology in a D-T fusion reactor, because tritiated water or gaseous tritium is captured on the piping surface through adsorption or isotope exchange reaction. The present authors carried out the water adsorption and desorption experiments on 304 stainless steel, copper, and aluminum in the temperature range from 5 to 100°C and in the partial pressure range of water vapor between 11.8 and 198Pa using a breakthrough method and quantified the amount of water adsorbed and the overall mass transfer coefficients in adsorption and desorption of water. It was observed in this study that aluminum adsorbed more water than stainless steel or copper. It was also observed that the adsorption and desorption rates of water for three materials showed almost the same values. The breakthrough behavior of tritiated water in a 100 m pipe of stainless steel was also evaluated applying the results of this work. It is concluded that water adsorption and desorption reactions influence the behavior of tritiated water in the piping system under the condition where the partial pressure of tritiated water vapor is lower than several pascals.     

The Second International Symposium on Advanced Nuclear Energy Research

The diffusion of tritiated water vapor into concrete walls of a fusion reactor building is studied to evaluate the capability of the tritium containment of concrete materials.

First, depth profiles of tritiated water in concrete are calculated to evaluate the capacity of the tritium containment by the sound concrete without cracks, and a 0.5-m-thick concrete wall is sufficient to prevent tritiated water releases to the environment in a normal operation of a fusion reactor over 50 yr. Second, simulations of the cleanup of tritiated water in the concrete reactor hall atmosphere taking into account the soaking are performed. Concrete porosity should be decreased to shorten the cleanup time of the reactor hall atmosphere. Surface coatings on the concrete, which apparently decrease the surface porosity, are effective measures to prevent the diffusion of tritiated water vapor into concrete during accidental releases.     

Tritium sorption on concrete protective coatings and the impact on operational safety

Exposures of concrete and selected coating materials to tritiated atmospheres have shown that tritium sorption on these materials and subsequent desorption are important parameters in defining tritium sources within a tritium-handling facility. Exposure time, tritium concentration and humidity of the air atmosphere affected the amount of tritiated water vapor sorbed. Some of the selected coatings reduced the tritium sorbed to less than 1% of unprotected concrete samples.Work funded by AECL Research and the Canadian Fusion Fuels Technology Project (CFFTP).     

Effect of hydrocarbons on the efficiency of catalytic reactor of detritiation system in an event of fire

Detritiation system of a nuclear fusion plant is mandatory to be designed and qualified taking carefully into consideration all the possible extraordinary situations in addition to that in a normal condition. We focused on the change in the efficiency of tritium oxidation of a catalytic reactor in an event of fire where the air accompanied with hydrocarbons, water vapor, and tritium is fed into a catalytic reactor at the same time. Our test results on the effect of these gases on the efficiency of tritium oxidation of the catalytic reactor indicated; (1) tritiated hydrocarbon produces significantly by reaction between tritium and hydrocarbons in a catalytic reactor; (2) there is little possibility of degradation in the detritiation performance because the tritiated hydrocarbons produced in the catalyst reactor are combusted; (3) there is no possibility of uncontrollable rise in the temperature of the catalytic reactor by heat of reactions; and (4) saturated water vapor could temporarily poison the catalyst and degrades the detritiation performance. Our investigation indicated a saturated water vapor condition without hydrocarbons would be the dominant scenario to determine the amount of catalyst for the design of catalytic reactor of the detritiation system.     

Out-of-pile tritium release study on Li4SiO4 pebbles from TRINPC-I experiments

Out-of-pile tritium release examinations of irradiated Li₄SiO₄ pebbles were performed in TRINPC-I experiments for evaluating material performance and verifying the system design. To generate tritium the specimens were irradiated with neutrons. Li₄SiO₄ pebbles were made by a freeze-drying method. In the experiments, concentrations of tritium in the form of tritium gas (HT + T₂) and tritiated water (HTO + T₂O) in the outlet streams of a reactor tube were measured separately with an ionization chamber and a liquid scintillation radiometer. The results show that the percentage of tritium gas (HT + T₂) and tritiated water trapped by the breeder pebbles were about 72% and 19% of totally released tritium, respectively. Thus, more tritium was released in the form of tritium gas in this work. In addition to tritium trapped by the breeder pebbles, the amount of free tritium was also measured by breaking on-line a quartz capsule containing Li₄SiO₄ pebbles, the percentage of which was 9% of totally released tritium. The temperature peaks of tritium gas mainly appeared at about 477 °C and 654 °C, while the temperature peak of tritiated water appeared at about 402 °C, under which most of tritiated water released.     

Optimization for the tritium isotope separation factor and permeation by selecting temperature and thickness of the diffusion Pd-Ag alloy cathode

Large quantities of tritiated water will be produced in the controlled fusion reactors for power generation. To eliminate the concentrated tritiated water and for recycling tritium, an industrial electrolyzer was developed. The aim of this paper is to give the design of this electrolyzer and the results for optimization in diffusion and in isotopic exchange by selecting thickness of a thimble-shaped Pd-25%Ag cathode working at high temperature and current. In this process, the tritium recovery system is based on the principle of the tritium diffusion Pd-Ag cathode which produces very pure hydrogen isotopes from enriched tritiated water.     

Effect of sweep gas species on tritium release behavior from lithium titanate packed bed during 14 MeV neutron irradiation

In a fusion reactor, the prediction of tritium release behavior from breeder blanket is important to design the tritium recovery system, but the amount of tritium generated is necessary information to do that. Hence, tritium generation and recovery studies on lithium ceramics packed bed have been started by using fusion neutron source (FNS) in Japan Atomic Energy Agency (JAEA). Lithium titanate (Li₂TiO₃) was selected as tritium breeding material, and its packed bed was enclosed by the beryllium blocks, and was kept at certain temperature during fusion neutron irradiation. During irradiation, the packed bed was purged with the sweep gas continuously, and tritium released was trapped in each gas absorber selectively by chemical form. In this work, the effect of sweep gas species on tritium release behavior was investigated. In the case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in the case of sweep by helium without water vapor, tritium in gaseous form was released first, and release of tritium in water form was delayed from gaseous tritium and was gradually increased.     

Tritium permeation experiments using reduced activation ferritic/martensitic steel tube and erbium oxide coating

Low concentration tritium permeation experiments have been performed on uncoated F82H and Er₂O₃-coated tubular samples in the framework of the Japan-US TITAN collaborative program. Tritium permeability of the uncoated sample with 1.2 ppm tritium showed one order of magnitude lower than that with 100% deuterium. The permeability of the sample with 40 ppm tritium was more than twice higher than that of 1.2 ppm, indicating a surface contribution at the lower tritium concentration. The Er₂O₃-coated sample showed two orders of magnitude lower permeability than the uncoated sample, and lower permeability than that of the coated plate sample with 100% deuterium. It was also indicated that the memory effect of ion chambers in the primary and secondary circuits was caused by absorption of tritiated water vapor that was generated by isotope exchange reactions between tritium and surface water on the coating.     

核设施氚气态释放后植物中有机氚的研究进展

从核设施释放到大气中的氚主要以氚化水（HTO）和氚化氢（HT）两种形式存在,最终以HTO的形式进入植物体。植物体中的氚有两种化学形态：自由水氚（TFWT）和有机氚（OBT）,其中OBT又被细分为交换性OBT和非交换性OBT。与TFWT相比,OBT在植物体内有较长的滞留时间和较大的剂量转换因子,在氚的食入剂量中OBT占主要份额,因此有必要对植物中的OBT展开全面研究。本文就植物中OBT的定义、交换性OBT和非交换性OBT的确定、OBT的形成过程及其影响因子、OBT预测模型的研究进行综述,同时对今后植物中OBT应重点研究的内容进行了简单分析,以期为植物中OBT的研究提供一定的参考。为准确评价OBT造成的辐射剂量,今后对OBT的研究中应着重从测量、夜间形成机理和环境中的行为等方面进行。     

Diffusion analysis on the thermal release of tritium from a neutron absorbing material

Tritium released from neutron irradiated borosilicate glass was determined by a specially designed sampling system and a liquid scintillation counter at temperatures in the range of 200–700°C. It was found that the chemical form of tritium released was tritiated water (HTO, T₂O) for the most part. Tritium produced in the glass would react with oxygen to form OT and diffuse out by a similar mechanism as the molecular diffusion of water in glasses. The diffusion coefficient of tritiated water in borosilicate glass obtained is expressed by D (cm²/s) = 5.3 × 10⁻⁴ exp( −128 kJ/mol)/RT). It is concluded from the diffusion analysis that the greater part of tritium produced in a neutron absorber, which is made of borosilicate glass, would remain in the glass for a few years of irradiation.     

Tritium: its generation and pathways to the environment at CANDU 6 generating stations

Tritium production rates and its pathways in CANDU 6 generating stations are studied in this paper. Tritium is generated primarily by the D(n, γ)T reaction in the reactor core. The major sources of tritium are moderator and coolant heavy water which are exposed to high neutron flux levels during reactor operation. A small amount of tritiated heavy water escape from heavy water filled systems is inevitable. Most of the tritiated heavy water which escapes is recovered through well-equipped heavy water recovery and collection systems. The vaporized tritiated heavy water is controlled by the D₂0 vapour recovery and ventilation systems. Similarly, the liquid tritiated heavy water is collected by portable D₂0 recovery tools and by an active drainage system. The remaining small amount escapes as tritiated heavy water and is lost to the environment in airborne and/or waterborne form.     

Enrichment and Volume Reduction of Tritiated Water Using Electrolysis Cell Having Hydrogen Permeable Cathode

The electrolysis rate and the separation factor for hydrogen isotopes are measured using the electrolysis cell having the hydrogen permeable cathode. As the hydrogen gas without the vapor of electrolyte is obtained by this method, decrease of the apparent separation factor by mixing with vapor can be avoided. It is also observed in this study that enrichment and volume reduction of tritiated water using the bipolar electrode electrolysis cell is effective because it gives small loss of tritium from the cell during volume reduction. The separation factor obtained in this study indicates that attachment of two or three sub-cells is enough for volume reduction of tritiated water.     

Measurement of tritium concentration in water by imaging plate

Concentration of tritium in water (4–400 kBq cm^?3) was measured by exposing an imaging plate without protection layer (Fujifilm, BAS-IP TR) to vapor for 2–48 h. It was found that tritium gradually penetrated into Eu-doped BaFBr phosphor and induced sufficiently intense photostimulated luminescence (PSL) even at the concentration of 4 kBq cm^?3. The intensity of PSL was proportional to tritium concentration in water. In addition, tritium absorbed in phosphor was reversibly released by keeping IP in air, and IP was able to be used repeatedly if total duration of exposure was ca. 24 h or less. The contamination of IP with tritium was not serious. It was concluded that IP technique has potential to measure tritium concentration in water without direct handling of tritiated water and with a minimum amount of radioactive waste.     

氦-3回路中氚的辐射安全分析

氚安全是确保燃料元件堆内功率瞬态试验的关键因素之一。本文首先分析了氚的来源和危害,提出了氦-3回路氚的防护和去污措施,然后讨论了氚在正常运行和事故时释放到包容箱和工艺间的量和处理措施,最后评价了氦-3系统发生不同安全措施失效的事故情况下工作人员的氚内照射剂量。结果表明:系统正常运行时工作人员所受最大剂量为1. 27μSv/d,除了氚安全措施全部同时失效且HT短时间全部被氧化成HTO的极限事故以外,在一般事故情况下氚对工作人员产生的最大剂量小于10 m Sv。     

Tritium in the DIII-D carbon tiles

The amount of tritium in the carbon tiles used as a first wall in the DIII-D tokamak was measured recently when the tiles were removed and cleaned. The measurements were made as part of the task of developing the appropriate safety procedures for processing of the tiles. The surface tritium concentration on the carbon tiles was surveyed and the total tritium released from tiles samples was measured in test bakes. The total tritium in all the carbon tiles at the time the tiles were removed for cleaning is estimated to be 15 mCi and the fraction of tritium retained in the tiles from DIII-D operations has a lower bound of 10%. The tritium was found to be concentrated in a narrow surface layer on the plasma facing side of the tile, was fully released when baked to 1000°C, and was released in the form of tritiated gas (DT) as opposed to tritiated water (DTO) when baked.     

Penetration behavior of water solution containing radioactive species into dried concrete/mortar and epoxy resin materials

Penetration behavior of radionuclides such as ¹³⁷Cs into dried concrete material, dried mortar material and epoxy paint for a few dozen days was observed using a solution containing fission products extracted from irradiated fuels to obtain fundamental information on the radionuclide penetration rate and depth. Hardly any radionuclides could penetrate into the epoxy paint. The radionuclide solution penetrated into concrete and mortar materials to a depth of a few millimeters for a few dozen days. The penetration behavior observed near the surface of concrete and mortar materials was similar to the diffusion of nuclides in media such as water-saturated concrete, bentonite and cement materials.