Tritium Behavior Research Articles

Codeposition of deuterium with ITER materials

The levels of retention in codeposited layers of each of the three ITER materials (C, Be and W) are compared. Scaling laws, based on the conditions during the codeposition process (surface temperature, incident particle energy and ratio of the depositing fluxes), are presented to allow prediction of expected retention under ITER conditions. Retention in carbon codeposits scales inversely with incident particle energy, whereas in the metallic codeposits the retention level scales proportionally to increasing particle energy. The differing scaling of retention with incident particle energy provides insights into which material may impact the global retention in ITER depending on where it may form codeposits. In addition to the amount of retention, the release behaviour of tritium from codeposits will influence the tritium accumulation rate within ITER. The thermal release behaviour of T (or D) from codeposits can be used to evaluate the effectiveness of baking at different temperatures as a means of tritium removal. Finally, the desorption kinetics from Be and W codeposits are contrasted. In the case of W codeposits, the duration of the baking cycle is important in determining the removal efficiency, whereas with Be codeposited layers, the maximum achievable bake temperature plays the leading role in determining removal efficiency.

R&D on support to ITER safety assessment

After performing its first ITER safety assessment in 2002 on behalf of the French “Autorité de Sûreté Nucléaire (ASN)”, the French “Institut de Radioprotection et de Sûreté Nucléaire (IRSN)” is now analysing the new ITER Fusion facility safety file. The operator delivered this file to the ASN as part of its request for a creation decree, legally necessary before building works can begin on the site. The IRSN first task in following ITER throughout its lifetime is to study the safety approach adopted by the operator and the associated issues. Such a challenging new technology calls for further in-house expertise and so in parallel a R&D program has been set up to support this safety assessment process, now and in the next years. Its main objectives are to identify the key parameters for mastering some risks (that would have been insufficiently justified by the operator) and to perform some verifications with methods and codes independent from the operator's ones. Priority has been given to four technical issues (others could be investigated in the future, like the behaviour of activated corrosion products). The first issue concerns the simulation of accident sequences with the help of the ASTEC European system code, developed by IRSN (jointly with its German counterpart, the GRS) for severe accidents in Pressurised Water Reactors. A preliminary analysis showed that most of its physical models are already applicable, e.g., for thermal-hydraulics in accidents caused by water or air ingress into the vacuum vessel (VV) or dust transport. Work has started in 2008 on some model adaptations, for instance oxidation of VV first wall materials by steam or air, and on validation on the ITER-specific ICE and LOVA experiments. Other model improvements are planned in the next years, as feedback from the work done for the other technical issues and from the code validation. The second issue concerns the risk of gas explosion due to concentrations of hydrogen and carbon monoxide in the ITER main volumes (VV but also the neighbouring volumes), produced by wall materials oxidation. A consistent program of modelling in the Computational Fluid Dynamics (CFD) TONUS IRSN code and of experiments in the TOSQAN (IRSN) and ENACEFF (CNRS/Orléans, France) facilities has been defined: inertising as means of mitigation, flammability limits, flame acceleration and transition to detonation, gas combustion. The third issue concerns the risk of dust explosion due to concentrations in the ITER volumes after remobilisation of dust deposits on walls. A consistent program of modelling in the CAST3M code (CEA), in collaboration with the Carthagene University (Spain), and of experiments in the BISE and TOSQAN IRSN facilities has been defined. The concerned phenomena are dust spatial distribution, remobilisation and entrainment, and explosion. The studies will also address the characterisation of dust and the control and mitigation processes. The fourth issue concerns the tritium behaviour. The theoretical analysis addresses its retention in the VV first walls, its chemistry in the gas phase during transport in the cooling circuits, and in the liquid phases during trapping in the neighbouring volumes or buildings. The efficiency of the diverse processes foreseen for detritiation, either in the VV or in the other rooms, in relation with the ventilation systems, is being also investigated.

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

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

Fusion materials development program in the broader approach activities

Breeding blankets are the most important components in DEMO. The DEMO blanket has to withstand high neutron flux typically 15–30 dpa/year under continuous operation. Therefore integrated and effective development of blanket structural materials and breeding/multiplying materials is essential in the blanket development for DEMO. In parallel to the ITER program, broader approach (BA) activities are initiated by EU and Japan. Based on the common interest of each party towards DEMO, R&D on reduced activation ferritic martensitic (RAFM) steels as a DEMO blanket structural material, SiC f/SiC composites which have potential for use in DEMO blankets, advanced tritium breeders and neutron multiplier for DEMO blankets, and tritium technologies including tritium behavior studies in advanced materials for DEMO blanket applications will be carried out as a part of the BA activities.

Analysis of Tritium Behavior in Very High Temperature Gas-Cooled Reactor Coupled with Thermochemical Iodine-Sulfur Process for Hydrogen Production

The tritium concentration in the hydrogen product in Japan's future very high temperature gas-cooled reactor (VHTR) system coupled with a thermochemical water-splitting iodine-sulfur (IS) process (VHTRIS system), named GTHTR300C, was estimated by numerical analysis. The tritium concentration in the hydrogen product significantly depended on undetermined parameters, i.e., the permeabilities of a SO3 decomposer and a H2SO4vaporizer made of SiC. Thus, the estimated tritium concentration in the hydrogen product for the conservative analytical condition ranged from 3.4 × 10−3 Bq/cm3 at STP (38 Bq/g-H2) to 0.18 Bq/cm3 at STP (2,000 Bq/g-H2). By considering the tritium retained by core graphite and the reduction in permeation rate by an oxide film on the heat transfer tube of the IHX and the HI decomposer, the tritium concentration in the hydrogen product decreased to the range from 3.3 × 10−5 Bq/cm3 at STP (0.36 Bq/g-H2) to 5.6 × 10−3 Bq/cm3 at STP (63 Bq/g-H2), which were smaller than those for the conservative analytical condition by factors of about 3.2 × 10−2 and 9.6 × 10−3, respectively. The effectof the helium flow rate in the helium purification system on the tritium concentration in the hydrogen product was also evaluated.

Recent results of R&D activities on tritium technologies for ITER and fusion reactors at TPL of JAEA

At Tritium Process Laboratory (TPL) of Japan Atomic Energy Agency (JAEA), tritium technologies for a fusion reactor have been carried out up to date. The design studies of Air Detrtiation and Vent Detritiation System (ADS/VDS) of ITER have been carried out in JAEA as a contribution of Japan to ITER. For the tritium processing technologies, our efforts have been focused on the R&D of the tritium recovery system of ITER test blanket system. A ceramic proton conductor has been studied as an advanced blanket system. A series of fundamental studies on tritium safety technologies for ITER and for fusion DEMO plants has also been carried out. The main R&D activities in this field are the tritium behavior in a confinement and its barrier materials, monitoring, accountancy, detritiation and decontamination, etc. Especially, for the fundamental studies for a DEMO plants, a part of the studies will be carries out a new facility at Rokkasho in Aomori in the Broader Approach (BA) program in coming 10 years. For this purpose, a design study of the facility at Rokkasho has first been started.

Deuterium and Helium Release and Microstructure of Tungsten Deposition Layers Formed by RF Plasma Sputtering

It is important to evaluate tritium behavior in tungsten deposition layers considering a long-term plasma operation. In this study, tungsten deposition layers were formed by deuterium or helium RF plasma sputtering. The release behavior of deuterium or helium from the layers were observed by a thermal desorption method. When a tungsten deposition layer does not contain oxygen, the retained deuterium is mainly released as D2. When oxygen exists in the layer, the majority of deuterium is released as water vapor. Tungsten deposition layers have an amorphous structure and consist of fine grain with size of 2-3 nm. Numerous bubbles are observed in the layers. A formation of tungsten deposition layer in a fusion reactor may make tritium control more difficult.

Tritium Behavior Intentionally Released in the Room

To construct a fusion reactor with high safety and acceptability, it is necessary to establish and to ensure tritium safe handling technology. Tritium should be well-controlled not to be released to the environment excessively and to prevent workers from excess exposure. It is especially important to grasp tritium behavior in the final confinement area, such as the room and/or building. In order to obtain data for actual tritium behavior in a room and/or building, a series of intentional Tritium Release Experiments (TREs) were planned and carried out within a radiologically controlled area (main cell) at Tritium System Test Assembly (TSTA) in Los Alamos National Laboratory (LANL) under US-JAPAN collaboration program. These experiments were carried out three times. In these experiments, influence of a difference in the tritium release point and the amount of hydrogen isotope for the initial tritium behavior in the room were suggested. Tritium was released into the main cell at TSTA/LANL. The released tritium reached a uniform concentration about 30 ~ 40 minutes in all the experiments. The influence of the release point and the amount of hydrogen isotope were not found to be important in these experiments. The experimental results for the initial tritium behavior in the room were also simulated well by the modified three-dimensional eddy flow analysis code FLOW-3D.

Tritium Research Activities Under the Broader Approach Program In JAEA

The R&D for tritium technologies to a demonstration reactor (DEMO) plant are carried out in the Broader Approach (BA) program in Japan: 1) tritium accountancy technology; 2) basic tritium safety research; and 3) tritium durability test. A multi-purpose facility will be constructed at Rokkasho in Japan to carry out the above R&D. Beta and gamma radioisotopes as well as tritium (370 TBq/year) can be handled in the facility.At TPL (Tritium Process Laboratory) of JAEA, a series of R&D programs for tritium technologies have been carried out. The main R&D activities in this field are: tritium behavior in a confinement; monitoring; detritiation; and decontamination. In this paper, the results of recent activities at TPL of JAEA are also summarized from the viewpoint of the related R&D subjects under the BA program.

Concentration Profiles of Tritium Penetrated into Concrete

Concentration profiles of tritium in cement paste, mortar and concrete were measured after exposure to tritiated water vapor for a given time. Tritium penetrated a distance of about 5 cm from the exposed surface during an exposure of 6 months. The model of tritium behavior in concrete materials reported by the present authors was developed in this study with the consideration of the effects of sand and aggregate on both the diffusion coefficient of tritiated water vapor and the isotope exchange capacity. Predictive calculations based on the tritium transport model were also carried out in some situations of tritium leakage. The results of the calculations show that a large amount of tritium will be trapped in the concrete walls, and the trapped tritium will be gradually released back to the tritium handling room over the time of months to years even after the decontamination of the room is completed.

Hydrogen Release Rate of SUS316 in Ar Gas Flow and Vacuum

To understand the mechanism of hydrogen release from materials used in vacuum technology, the release behavior of tritium from stainless steel SUS316 previously loaded with this gas was investigated isothermally at 423 K either under a stream of argon or under vacuum conditions. Depth profiles of the heated specimens were determined by chemical etching. The experimental release rate and the tritium depth profile in the solid could be simulated with a model based on bulk diffusion. The diffusion coefficients used to fit the depth profiles of specimens heated under vacuum at 423 K were non-discernible from those obtained with specimens heated under a stream of Ar at the same temperature.

Analysis of Tritium Behavior in Very High Temperature Gas-Cooled Reactor Coupled with Thermochemical Iodine-Sulfur Process for Hydrogen Production

Analysis of Tritium Behavior in Very High Temperature Gas-Cooled Reactor Coupled with Thermochemical Iodine-Sulfur Process for Hydrogen Production

Studies on the behaviour of tritium in components and structure materials of tritium confinement and detritiation systems of ITER

Confinement and the removal of tritium are key subjects for the safety of ITER. The ITER buildings are confinement barriers of tritium. In a hot cell, tritium is often released as vapour and is in contact with the inner walls. The inner walls of the ITER tritium plant building will also be exposed to tritium in an accident. The tritium released in the buildings is removed by the atmosphere detritiation systems (ADS), where the tritium is oxidized by catalysts and is removed as water. A special gas of SF6 is used in ITER and is expected to be released in an accident such as a fire. Although the SF6 gas has potential as a catalyst poison, the performance of ADS with the existence of SF6 has not been confirmed as yet. Tritiated water is produced in the regeneration process of ADS and is subsequently processed by the ITER water detritiation system (WDS). One of the key components of the WDS is an electrolysis cell. To overcome the issues in a global tritium confinement, a series of experimental studies have been carried out as an ITER R&D task: (1) tritium behaviour in concrete; (2) the effect of SF6 on the performance of ADS and (3) tritium durability of the electrolysis cell of the ITER-WDS. (1) The tritiated water vapour penetrated up to 50 mm into the concrete from the surface in six months' exposure. The penetration rate of tritium in the concrete was thus appreciably first, the isotope exchange capacity of the cement paste plays an important role in tritium trapping and penetration into concrete materials when concrete is exposed to tritiated water vapour. It is required to evaluate the effect of coating on the penetration rate quantitatively from the actual tritium tests. (2) SF6 gas decreased the detritiation factor of ADS. Since the effect of SF6 depends closely on its concentration, the amount of SF6 released into the tritium handling area in an accident should be reduced by some ideas of arrangement of components in the buildings. (3) It was expected that the electrolysis cell of the ITER-WDS could endure 3 years' operation under the ITER design conditions. Measuring the concentration of the fluorine ions could be a promising technique for monitoring the damage to the electrolysis cell.

Numerical Study on Tritium Behavior by Using Isotope Exchange Reactions in Thermochemical Water-Splitting Iodine—Sulfur Process

One potential problem in the hydrogen production system coupled with the high-temperature gascooled reactor (HTGR) is transmission of tritium from the primary coolant to the product hydrogen by permeation through the heat transfer tubes. Tritium accumulation in the process chemicals in the components of a hydrogen plant, a thermochemical water-splitting iodine-sulfur (IS) process, will also be a critical issue in seeking to license the hydrogen plant as a non-nuclear plant in the future. A numerical analysis model for tritium behavior in the IS process was developed by considering the isotope exchange reactions between tritium and the hydrogen-containing process chemicals, i.e., H2O, H2SO4 and HI. The tritium activity concentration in the IS process coupled with the high-temperature engineering test reactor (HTTR), the HTTR-IS system, was preliminarily evaluated in regard to the effects of some indeterminate parameters, i.e., equilibrium constants of the isotope exchange reactions, permeability of tritium through heat transfer tubes, tritium and hydrogen concentrations in the secondary helium coolant, and the leak rate from the secondary coolant loop. The results describing how the tritium activity concentration changes with variations in these parameters and which component has the maximum tritium activity concentration in the IS process are described in this paper.

Study on Chemical Behavior of Energetic Deuterium Implanted into Li2TiO3

Deuterium ion implantation and subsequent X-ray Photoelectron Spectroscopy (XPS) and Thermal Desorption Spectroscopy (TDS) experiments were performed with varying implantation temperatures to reveal chemical behavior of tritium produced in Li2TiO3. These experimental results showed that there were four deuterium trapping states; two of which were interacted with and without oxygen near the surface, and the other two were interacted with E’-center and with oxygen with the formation of O-D bond in the bulk. These trapping states of deuterium in the bulk were almost the same as those of tritium generated in thermal neutron-irradiated Li2TiO3. The total amount of deuterium retention in the bulk was almost constant until O-D bonds formed in the bulk were decomposed, indicating that tritium trapping could proceed under hot atom chemical reactions. It was concluded that E’-center could trap the implanted deuterium more frequently than oxygen with the formation of O-D bonds in the bulk. Annihilations of them due to oxygen recovery could increase the retention of D with the formation of O-D bonds, resulting in the almost constant deuterium retention ratio up to its decomposition temperature of 573 K.

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.

Detrapping behavior of tritium trapped via hot atom chemical process in neutron-irradiated ternary lithium oxides

Tritium detrapping behavior in neutron-irradiated ternary lithium oxides was investigated by the comparison of the annihilation of irradiation defects with the tritium release. It was revealed that the annihilation of irradiation defects would consist of two processes; namely the fast and the slow ones. The slow annihilation process has correlation with the tritium release, indicating that E′-center or F +-center could act as tritium trapping site, and from its activation energy of each sample, the annihilation of E′-center and F +-center could be attributed to the recovery of oxygen via diffusion, triggering the tritium release. Meyer–Neldel plots of these results indicate that the slow annihilation process was governed by the formation entropy of a pair of vacancy and interstitial atom of oxygen. Therefore, the trapped tritium would be detrapped by oxygen recovery to E′-center or F +-center, and its kinetics would be determined by the population of oxygen vacancy under thermal equilibrium.

Recovery of Tritium from Flibe Blanket in Fusion Reactor

The present paper is to describe the behavior of tritium in Flibe as a self-cooled liquid blanket of a fusion reactor quantitatively. In order to avoid the generation of corrosive TF, Flibe is maintained under reduction atmosphere to transform TF to T2 to keep a faster reaction rate compared with a residence time in a self-cooled blanket. The most important point is to clarify whether or not the redox control of Flibe can be achieved by Be rods inserted in a blanket within a limited contact time. The dissolution rate of a Be rod and the TF reduction reaction rate of Be + 2TF = BeF2 + T2 in Flibe were experimentally determined under the JUPITER-II collaboration work. Close agreement was obtained between experiment and our simplified complete-mixing model. Especially, the reaction between Be and F− ion immediately after the contact was found to be limited by diffusion of F− ion. The behavior of tritium generated in a Flibe fuel cycle was simulated under a Flibe flow condition of FFHR-2.

Recent Activities on Tritium Technologies for ITER and Fusion Reactors at JAEA

The design studies of Atmosphere Detirtiation System (ADS) have been carried out in Japan Atomic Energy Agency (JAEA) as a contribution of Japan to ITER. The performance of ADS has also been investigated under accidental conditions such as fire and co-existing of a poison gas for catalyst like SF6. There is no degradation of Detritiation Factor (DF) under co-existing of CO or CO2 up to 20% as a simulated fire condition. However, only 0.1% of SF6 degrades the DF from more than 1000 to 50, following reduction of water by SF4 etc. (decomposition products of SF6) at 773K of catalyst bed.For the tritium processing technologies, our efforts have been focused on the R & D of the tritium recovery system of breeding blanket. In case of ITER Test Blanket Module, a cryogenic molecular sieve bed system was designed and demonstrated. Furthermore, electro-chemical pumping system using a proton conductor is also investigated to design more effective system. The durability of electrolysis cell for Water Detritiation System (WDS) has been investigated and it is expected that the cell can endure more than 3 years’ operation under the ITER WDS design condition.A series of fundamental studies on tritium safety technologies has been carried out as another major activity of JAEA for ITER and future fusion demo reactors. Tritium behavior in various confinement materials, tritium monitoring & accountancy, and detritiation were studied under collaboration programs with universities, using Caisson Assembly for Tritium Safety study.

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.