Tritium Handling Facility Research Articles

Analysis of the integral zeolite molecular sieve process for helium CPS application

Zeolite Molecular Sieves (ZMSs) are commonly adopted in tritium handling facilities for impurity removal from gaseous streams, particularly for tritiated water trapping. Several activities have been conducted to characterize the adsorption and desorption behaviour of the sieving materials. Instead, less attention has been put on the analysis of the integral process comprising the adsorption of the tritiated humidity, the regeneration of the molecular sieve and the recovery of the desorbed water. Focusing on the application for the EU-DEMO helium Coolant Purification System (CPS), this work presents a process simulator relying on a MATLAB dynamic model of the ZMS bed and on a RELAP5/MOD3.3 model for the regeneration loop. The dynamic model can simulate both the adsorption and regeneration modes of the ZMS by providing the temperature and the adsorbate concentration profile along the column, while the thermal-hydraulic model is used to assess the conditions of the regeneration loop. By coupling the outcomes of the two models it is possible to establish the effective regeneration efficiency of the process and the amount of the tritiated water recovered. For the case of helium CPS, the analysis demonstrates the feasibility of the proposed operative scheme and the capabilities of the regeneration procedure to ensure a 74.4 % regeneration efficiency of the ZMS bed.

Disposal procedure for contaminated surface of tritium handling facility in the decommissioning operation

For the decommissioning operation of tritium handling facility, surveillance of contamination with tritium on floor, wall and roof was performed. Significant amount of tritium contamination was detected in approximately 86% of total surface measurement points and maximally 70mm of scraping was required to eliminate tritium contamination. In some samples for measurement of tritium depth profile, the highest tritium concentration was observed at the inside but not the surface. This measurement result of depth profile is also discussed with our previous calculation results considering diffusion, adsorption, desorption of tritium in the concrete and isotope exchange reaction between tritiated water vapor in the concrete bulk and atmospheric water vapor.

Experimental investigation on feasible bioreactor using mechanism of hydrogen oxidation of natural soil for detritiation system

A passive reactor for tritium oxidation at room temperature has been widely studied in nuclear engineering especially for a detritiation system (DS) of a tritium process facility taking possible extraordinary situation severely into consideration. We have focused on bacterial oxidation of tritium by hydrogen-oxidizing bacteria in natural soil to realize the passive oxidation reactor. The purpose of this study was to examine the feasibility of a bioreactor with hydrogen-oxidizing bacteria in soil from a point of view of engineering. The efficiency of the bioreactor was evaluated by kinetics. The bioreactor packed with natural soil shows a relative high conversion rate of tritium under the saturated moisture condition at room temperature, which is obviously superior to that of a Pt/Al2O3 catalyst generally used for tritium oxidation in the existing tritium handling facilities. The order of reaction for tritium oxidation with soil was the pseudo-first order as assessed with Michaelis-Menten kinetics model. Our engineering suggestion to increase the reaction rate is the intentional addition of hydrogen at a small concentration in the feed gas on condition that the oxidation of tritium with soil is expressed by the Michaelis-Menten kinetics model.

A New Method to Measure Small Volumes in Tritium Handling Facilities, Using p-V Measurements

In a tritium handling facility it is important to know the volume of the parts of the system for accurate accountancy. The usual methods to determine volumes have several disadvantages concerning contamination issues and accuracy.In this work a novel method is used to determine volumes. The method does not require any liquid or precisely known reference volume. It is simple and the uncertainty of the measurement is lower than 1% for volumes from 10 cm3 to 10 dm3.In this paper the principle of the volume determination device (VDD) designed and built at the Tritium Laboratory Karlsruhe (TLK) is presented, together with a detailed uncertainty calculation. Also the first measurements with the VDD are discussed in detail.

Development of tritiated vapor absorbent applicable to the atmospheric detritiation system in a nuclear facility

The combination of hydrogen oxidation reactor packed noble metal catalysts and water vapor absorber has been applied to the atmospheric detritiation system of the tritium handling facility. Commercial synthetic zeolite such as molecular sieve 5A has been used as an adsorbent of ADS absorber. In the case of application of molecular sieve 5A to the ADS absorber of a large-scale tritium handling facility such as a future fusion plant, an absorber becomes huge due mainly to the difficulty in dehydration from molecular sieve 5A. Hence, application of CaY Faujasite-type zeolite with a high framework silica-to-alumina ratio to the adsorbent for atmospheric detritiation system was investigated. It was clear that the dehydration behavior at room temperature was significantly improved using CaY zeolite. In contrast, detritiation factor for CaY zeolite with a high framework silica-to-alumina ratio depended strongly on the space velocity through the absorber. To apply CaY zeolite with a high framework SiO(2)/Al(2)O(3) ratio to the ADS absorbent, the space velocity less than 250h(-1) was recommended to maintain the detritiation factor more than 1000. The steep increase in water adsorption at the relative pressure lower than 0.05 is a feature of synthetic zeolite with calcium cation. However, such an increase was not observed in water adsorption on CaY zeolite with a framework SiO(2)/Al(2)O(3) ratio more than 7.0. Consequently, the CaY zeolite with the framework SiO(2)/Al(2)O(3) ratio of 5.0 is a promising candidate as absorbent of ADS absorber.

Evaluation of tritium behavior in concrete

Concrete walls used in a fusion reactor or a tritium handling facility have the important role of separating the atmosphere in a facility from environmental air, since substantial tritium leakage is possible from a tritium system to the atmosphere of a tritium handling facility. Therefore, it is necessary to understand the transfer behavior of tritium through concrete. The experiments of this study were performed to observe the change of the tritium distribution with time in columns made of cement paste, mortar or concrete. It was found that exposure of the surface of concrete materials to HTO vapor for two months produced tritium permeation to the depth of 30 mm. Estimation of tritium penetration behavior was also performed in this study by applying adsorption, desorption and isotope exchange reactions and the estimated tritium profile was in good agreement with observed values.

Study on Permeation Behavior of Gaseous Tritium through Concrete Walls

Concrete walls play the role not only of the structural material but also of the final barrier of a multiconfinement system of tritium in a fusion reactor or a tritium-handling facility. Therefore, it is required that the behavior of tritium in the concrete materials be clarified to certify the radiation safety of a fusion reactor. The diffusion coefficient of hydrogen in cement paste is obtained by using the permeation experiment in this study, and it is found that the diffusion coefficient of hydrogen in the cement paste is only one order magnitude smaller than the diffusion coefficient of hydrogen in air. Calculation using the diffusion coefficient obtained in this study indicates that the gaseous tritium, HT or T2, can permeate rather rapidly to the outside through the concrete wall of a tritium-handling facility. This calculation implies that installation of a tritium recovery system with proper decontamination performance is required to minimize the tritium transfer to the outer environment.

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.

Study of Tritium Behavior in Cement Paste

The concrete materials are used as the partition wall of the tritium handling facilities. It is important to grasp the tritium behavior in the concrete wall for radiation safety. It is considered in this study that the surface water on the concrete materials consists of physically adsorbed water, chemically adsorbed water and structural water as in the case of porous adsorption materials. The adsorption capacity due to physically and chemically adsorption isotherms observed in this study shows that the amount of water adsorption on the cement paste is a quarter of the amount adsorbed onto the surface of activated alumina or molecular sieves 5A (MS-5A). It shows that concrete is easily contaminated with tritiated water.

Mass Transfer Behavior of Tritium from Air to Water through the Water Surface

It is anticipated that a certain amount of tritiated water exists in the atmosphere of tritium handling facilities, and it is recognized that the hazardous potential of tritiated water is rather high. Then, it is important to grasp the behavior of tritiated water for preserving of the radiation safety. The mass transfer behavior of tritium from air to water through the water surface was discussed in this study. The evaporation rate of water and the condensation rate of water were experimentally examined from measurement of change of the weight of distilled water. The tritium transfer rate from the tritiated water in air to the distilled water was also experimentally examined by using a liquid scintillation counter. Experimental results about change of tritium level in a small beaker placed in the atmosphere with tritiated water showed that diffusion of tritium in water and gas flow in the atmosphere gives considerable effect on tritium transfer. The estimation method of the tritium transfer made in this study was applied to explain the data at The Japan Atomic Power Company second power station at Tsuruga and good agreement was obtained.

Tritium Application: Self-Luminous Glass Tube (SLGT)

To manufacture SLGTs (Self-Luminous Glass Tubes), 4 core technologies are needed: coating technology, tritium injection technology, laser sealing/cutting technology and tritium handling technology. The inside of the glass tubes is coated with greenish ZnS phosphor particles with sizes varying from 4~5 [µm], and Cu, and Al as an activator and a co-dopant, respectively. We also found that it would be possible to produce a phosphor coated glass tube for the SLGT using the well established cold cathode fluorescent lamp (CCFL) bulb manufacturing technology. The conceptual design of the main process loop (PL) is almost done. A delicate technique will be needed for the sealing/cutting of the glass tubes. Instead of the existing torch technology, a new technology using a pulse-type laser is under investigation. The design basis of the tritium handling facilities is to minimize the operator's exposure to tritium uptake and the emission of tritium to the environment. To fulfill the requirements, major tritium handling components are located in the secondary containment such as the glove boxes (GBs) and/or the fume hoods. The tritium recovery system (TRS) is connected to a GB and PL to minimize the release of tritium as well as to remove the moisture and oxygen in the GB.

Manufacturing process of self-luminous glass tube (SLGT) utilizing tritium gas: Design of tritium handling facilities

Tritium, the essential material of self-luminous glass tube (SLGT), is a Β-ray emitting radioactive hydrogen isotope that requires a special handling facility. The design basis of a tritium handling facility is to minimize the operator’s exposure by tritium uptake and the emission of tritium to the environment. To fulfill the requirements, major tritium handling components are located in the secondary containment such as glove boxes (GBs) and/or fume hoods. Besides a tritium recovery system (TRS) and a tritium monitoring system are included in this facilities. To prevent both tritium release out to the room and air in-leakage into the GB, the GB is designed to maintain equal or slightly lower pressure than room atmosphere. TRS is connected to the GB and process loop (PL) to minimize the release of tritium as well as to remove moisture and oxygen in the glove box. TRS is composed of a molecular sieve adsorption bed (MS bed), a nickel catalyst bed, a metal getter, tritium monitors, and a circulation pump. TRS components are regenerated if needed and tritium is recovered for reuse.

Tritium Accounting by In-Situ Calorimetry of the JET Uranium Containers

ABSTRACTStorage of tritium in the form of metal hydride is a common technique in tritium handling facilities and is generally acknowledged as the only option for the storage of large tritium inventories in future fusion reactor applications. Since accounting for large inventories by the conventional TPVC (Temperature, Pressure, Volume, Concentration) is very cumbersome, it is highly desirable to perform accounting directly by the application of calorimetric methods, for example based on monitoring of temperature rise in the tritium storage container caused by heat of the tritium decay (1.95W/mol.T2). Following an earlier evaluation1 of the JET tritium storage containers by electrical simulation of heat of the tritium decay the viability of the method was proven by adiabatic calorimetry with known tritium inventories up to ≈5900TBq.

Levels of tritium concentration in the environmental samples around JAERI TOKAI

By the operation of research reactors, tritium-handling facilities, nuclear power plants, and a reprocessing facility around JAERI TOKAI, tritium is released into the environment in compliance with the regulatory standards.

The Design of the Gas Clean up System for the New Tritium Facility at AWE Aldermaston

A Gas Clean Up System (GCUS) has been designed, built and installed for the new tritium handling facility at Atomic Weapons Establishment (AWE), Aldermaston. The system is in the process of being commissioned and once this is complete and the facility is operational, it will be used for the removal of hydrogen isotopes from gaseous waste arisings within the facility and concentrating them in waste packages. The system also provides a depression for the sources of these waste arisings, particularly the inert gas gloveboxes, as part of the contamination containment within the facility. This paper describes the details of various sub-systems within the GCUS and their engineering, construction, installation and testing.

Separation of Hydrogen Isotopes from Nitrogen with Polyimide Membrane

Gas separation tests were performed with nitrogen containing small concentrations of deuterium or tritium on a small-scale polyimide membrane module purchased from Ube Industries, similar to that developed at the Japan Atomic Energy Research Institute (JAERI). Experimental results showed that this separation process could be potentially applied to tritium removal systems used in tritium handling facilities. It would allow to reduce significantly the gas volume to be treated by the conventional oxidation-adsorption process, and, therefore, the number or size of associated equipments. Hazards arising from handling of highly toxic tritiated water vapor due to conversion of elemental tritium, would hence be lowered.

Operating Experience with Tritium Accounting and Analysis Systems in Our Tritium-Handling Facilities

An integral part of the operation of a tritium-handling facility is the need to maintain a tritium accounting system which meets both operational and regulatory requirements. The Darlington Tritium Removal Facility processes heavy water to remove tritium for storage or sale. The Chalk River Tritium Laboratory is a general-purpose experimental facility, handling various quantities and specific activities of tritium in separate systems, including re-packaging of tritium from large metal-getter storage beds to smaller aliquots. Tritium accounting procedures, analytical techniques and operational experience gained to date in the operation of these tritium-handling facilities are described.

Tritiated Waste Management - Tritiated Oil Packaging and Decontamination

Tritium Handling facilities often have oil lubricated vacuum pumps in active portions of tritium systems. The interaction of tritium with the oil can lead to the formation of various classes of tritiated species, both volatile and non-volatile. These tritiated species can cause problemats in terms of handling, dosimetry and ultimate waste disposal. Knowledge of these species and their relative concentrations will allow the safe and cost effective use of oil-filled pumps in tritium environments. In order to understand more accurately the specific dosimetry hazards, improve handling procedures and minimize waste volumes, the distribution of tritium amongst the several classes of tritiated species was determined. The change in tritium content and distribution over time was also determined.

Percutaneous absorption of tritium-gas-contaminated pump oil.

One of the radiological problems encountered in tritium handling facilities is the hazards associated with tritium's ability to label and degrade organic materials. Experiments in which male hairless rats have been contaminated with tritium-gas-contaminated pump oil have demonstrated that tritium deposited on the skin provides an input of organically bound tritium and tritiated water in the body. The accumulation of organically bound tritium at the point of contact in the skin and in various tissues influenced tritium excretion in urine and feces. The retention of tritium in the body showed that tritium was mainly metabolized and assimilated as organically bound tritium. The distribution of tritiated water was rapid and uniform in the whole-body. Analyses of tritium excreted in animal urine and feces showed that a significant level of organically bound tritium was excreted shortly after exposure. The highest concentration of tritium activity was measured in the exposed area of the skin. An increased level of tritium accumulation in the liver and kidneys was seen. Dose calculations showed that the exposed skin had the highest dose, and the skin dose was primarily due to the retention of organically bound tritium at the point of contact. The interpretation of these data has indicated that the retention of short-term organically bound tritium in the skin may be a dominant factor for dosimetry purposes.

Lessons Learned during Startup of the Replacement Tritium Facility at the Savannah River Site

Tritium processing operations have been performed at the Savannah River Site since 1955. It was determined in the early 1980`s that new tritium facilities were required in order to meet increasingly stringent safety and environmental requirements while maintaining productivity. Construction of a new tritium handling facility, the Replacement Tritium Facility (RTF), began in 1987. The RTF incorporates the latest technology and design philosophy to mitigate accident consequences, enhance operational safety, minimize tritium losses to the environment, and ensure material safeguards and security. The RTF is currently completing startup testing. Tritium handling and processing operations are common to both national defense and fusion technology programs. Disseminating the lessons learned during the startup of a new tritium handling facility is important to the success of future tritium handling missions. 4 refs.