Tritium Release Behavior Research Articles

Tritium release from lithium orthosilicate pebbles deposited with palladium

Slightly overstoichiometric lithium orthosilicate pebbles are fabricated from lithium hydroxide and silica by a melting and spraying method in a semi-industrial scale facility. The authors performed out-of-pile annealing tests using the lithium orthosilicate pebbles irradiated in a research reactor. Moreover, the effect of the deposition of palladium in the lithium orthosilicate pebbles on the behavior of tritium release was investigated. The lithium orthosilicate pebbles were irradiated in a research reactor. In the out-of-pile annealing experiments, the temperature of lithium orthosilicate pebbles was raised from ambient temperature to 1173 K at a constant rate of 5 K/min under the stream of 0.1% hydrogen/nitrogen sweep gas. The experimental results indicate that almost all tritium was released as tritiated water vapor from the virgin lithium orthosilicate pebbles. It was also found that a lager amount of tritium was released as the molecular form (HT) from the lithium orthosilicate pebbles deposited with palladium.

Gas bubble network formation in irradiated beryllium pebbles monitored by X-ray microtomography

The effective and safe operation of helium cooled ceramic breeder blankets with beryllium as a neutron multiplier requires among others an efficient tritium release. A micrometric resolution computer aided microtomography (CMT) setup located at the European Synchrotron Radiation Facility made possible the 3D reconstruction of interconnected channel networks of helium bubbles in beryllium pebbles, thus enabling the identification of open porosities in the micrometer range. Beryllium pebbles of 2 mm diameter were neutron irradiated at 770 K to a fluence of 1.24 × 10 25 nm −2, resulting in 480 appm helium and 12 appm tritium. After annealing at 1500 K, CMT was performed on the pebbles at 4.9 and 1.4 μm spatial resolution, respectively, followed by the post-processing of the reconstructed pebble volumes. Besides a bimodal pore distribution with a smaller population around 10 μm diameter and a high density of partly interconnected pores around 40 μm diameter, a swelling of 17% was found. The spatial distribution of the void fraction network will be discussed together with implications on tritium release behaviour.

Effects of multi-ion irradiation on microstructural changes in lithium titanate

The irradiation behavior of Li 2TiO 3 under a fusion reactor environment was simulated by simultaneous irradiation of Li 2TiO 3 by the triple ion beams and the respective single ion beams of O 2+, He + and H +. The microstructural changes in Li 2TiO 3 caused by the irradiation were measured by FT-IR photoacoustic spectroscopy. The results suggest that the amount of TiO 2 formed is proportional to the dpa and that the method of irradiation does not affect the dependence of formation of TiO 2. On the other hand, the amount of defects and/or radiolytic products generated by irradiation, which is considered to trap hydrogen near the surface, is found to be affected by the method of irradiation. Such phenomena are believed to affect the tritium release behavior from Li 2TiO 3, and durability of Li 2TiO 3 and compatibility of Li 2TiO 3 with other components of the breeder blanket such as structural materials in the fusion reactor system under operation.

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.

Characteristics of Tritium Release Behavior from Solid Breeder Materials

A tritium release model has been developed by the present authors. The tritium release curves estimated by this tritium model give good agreement with experimental curves for Li4SiO4, Li2TiO3, Li2ZrO3 or LiAlO2 under various purge gas conditions in our out-of-pile bred tritium release.The characteristics of tritium release behavior from various solid breeder materials carried out by us and in EXOTIC experiments at Petten are discussed in this study.

Study on Correlation Between Tritium Release Behavior and Annihilation Behavior of Irradiation Defects in Neutron-Irradiated Li4SiO4

Tritium release from thermal neutron-irradiated Li4SiO4 is initiated with the annihilation of E’-centers by recovering O- with diffusion of O-. Electron Spin Resonance (ESR) shows that differences in the formation of irradiation damage between 14 MeV and thermal neutrons in Li4SiO4 result in different tritium release behaviors. The kinetics for the annihilation of irradiation defects has been determined. The contribution of elastic collisions by 14 MeV neutrons was much higher than that of thermal neutrons. Isothermal annealing experiments show that annihilation of irradiation defects consisted of two processes: namely, the fast and slow annihilation processes. Their activation energies were determined to be 0.13 and 0.39 eV, respectively. Comparing the experimental results for the thermal and 14 MeV neutronirradiated Li4SiO4 shows that the activation energies of the slow annihilation process were significantly different. These results relate to the density of irradiation defects, which in turn depend on the contribution of the recoil particles produced by nuclear reactions to form irradiation damaged sites.

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.

Tritium diffusivity in crystal grain of Li 2TiO 3 and tritium release behavior under several purge gas conditions

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. Tritium release curves estimated using the tritium release model gave good agreement with observed tritium release curves from Li 4SiO 4, Li 2ZrO 3 or LiAlO 2. Tritium release behavior from Li 2TiO 3 under humid purge gas, dry purge gas and dry purge gas with hydrogen conditions is discussed in this study, tritium release curves using the release model that we proposed previously give a good agreements with experimental tritium release curves. Tritium effective diffusivity in the crystal grain of Li 2TiO 3 is also estimated in this study using a curve-fitting method applied to the release curves obtained under the humid purge gas condition. It is discussed that change of color of Li 2TiO 3 surface under hydrogen purge gas condition is observed and this phenomenon might affect tritium release behavior from Li 2TiO 3.

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.

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.

Manufacturing methods and characterisation of titanium beryllides

Beryllium pebbles of about 1 mm diameter are presently the reference neutron multiplier in helium cooled pebble bed (HCPB) blanket designs, both in Japan and Europe. However, some concern exists with Be pebbles, mostly because of the uncertainty of T-release properties during DEMO reactor operation conditions. Since the recent past, beryllides like Be 12Ti are considered to be one of the most promising materials for advanced neutron multipliers due to their relatively high melting point, good chemical and corrosion stability and much better tritium release behaviour compared to pure Be. Development of scaleable manufacturing methods bypassing the intrinsic brittleness of beryllides is considered to be the most challenging. In this work, we present a detailed study of two different routes of titanium beryllide fabrication, including characterisation of obtained specimens by means of scanning electron microscopy (SEM), transmission electron microscopy (TEM), X-ray analysis and impurity analysis performed by plasma emission spectrometer.

In-pile tritium release behaviour of lithiummetatitanate produced by extrusion–spheroidisation–sintering process in EXOTIC-9/1 in the high flux reactor, Petten

The irradiation programme EXOTIC (extraction of tritium in ceramics) is carried out within the European framework for the development of the helium cooled pebble bed concept. The EXOTIC-9/1 is the latest experiment in the series of EXOTICs that are irradiated in the high flux reactor in Petten. Tritium release and inventory in lithium containing ceramic pebbles are key properties to be tested in a TBM. New production routes of pebbles are developed, leading to different thermomechanical and tritium release properties. The objective of the EXOTIC-9/1 is to study in-pile tritium release behaviour of the latest developed lithiummetatitanate pebbles (Li 2TiO 3). The pebbles are produced by a extrusion–spheroidisation–sintering process at CEA. The new pebbles differ with respect to porosity from the lithiummetatitanate ceramics tested in the previous EXOTIC 8 programme. The pebbles have diameter in the range from 0.6 to 0.8 mm. Irradiation of EXOTC-9/1 started at 24 March 2005, and will continue until the end of 2006, in total about 400 irradiation days. The temperature is varied between 340 and 580 °C. Begin of Life (BOL) tritium production rate is 0.56 mCi/min. Based upon the in-pile tritium release measurements and the analysis of the tritium residence time it can be concluded that tritium release in the new batch of the high density Li 2TiO 3 pebbles irradiated in EXOTIC 9/1 is rather slow compared to the ceramics irradiated in the EXOTIC 8 irradiation campaign. In this paper, the in-pile tritium behaviour will be reported during normal operation and during transients in temperature, purge gas chemistry and gasflow. The collected data is compared to tritium release data from ceramics irradiated in previous EXOTIC experiments with respect to tritium inventory, residence time and porosity.

Hot atom chemical behavior of tritium in neutron-irradiated Li2TiO3 and Li2ZrO3

The annihilation behavior of irradiation defects induced in neutron-irradiated Li2TiO3 and Li2ZrO3 were investigated with the tritium release behavior. It was revealed that the common characteristics in both samples were that the annihilation process of irradiation defects consisted of two first-order processes and E’-center could act as tritium trapping site, and otherwise was the way how the E’-center annihilated. The difference was suggested to attribute to the mobility of M as M4+ (or M3+, etc.).

Estimation of tritium release behavior from solid breeder materials under the condition of ITER test blanket module

A tritium release model from solid breeder materials has been developed. Tritium release curves estimated using this model gave good agreement with tritium release curves from various solid breeder materials obtained by out-pile experiments under various purge gas conditions in previous study. In this study, tritium release behavior from Li 2TiO 3 and LiAlO 2 are estimated for the neutron irradiation condition and operation sequence of the ITER test blanket module test program (neutron wall loading of 0.78 MW/m 2 and repetition of burn time for 400 s with a dwell time of 1400 s). The results are evaluated to discuss the tritium release and inventory performance of different breeder materials.

Measurement system for in-pile tritium monitoring from Li 2TiO 3 ceramics at WWRK reactor

A 220-day irradiation of Li 2TiO 3 ceramics with 96% enrichment of isotope 6Li was carried out at WWRK reactor. One of the study goals was to examine tritium release behavior during Li burn-up. To achieve this goal three types of ceramics samples were examined simultaneously using a system for in-pile tritium monitoring: one (pebbles) – under constant temperature of 650 °C, and two (pebbles and pellets) – within temperature change ranges from 500 to 900 °C. Flows of tritium release from ceramics during the various reactor campaigns, as well as tritium generation rates for each ampoule are presented in the paper. The main result is justification of burn-up in 6Li reaching up to 20% of Li 2TiO 3 ceramics.

Migration and release behavior of tritium in SS316 at ambient temperature

BIXS measurements indicate that immersion into water or chemical etching of SS316 contaminated with tritium at moderate temperatures causes an immediate reduction of the outermost surface concentration of tritium. The fraction of surface tritium removed by water, i.e. 30–50%, is small in comparison to the total tritium present in the specimen. Allowing a specimen to age whose surface and subsurface had been removed by etching up to a depth where the concentration of tritium is mostly constant revealed that within a few months a re-growth of tritium up to a saturation value higher than half of that originally present on the specimen takes place. Concurrently, a small but steady liberation of tritium at rates increasing from 0.1 to 0.3kBq/h was noticed.

Tritium Release Behavior from Steels Irradiated by High Energy Protons

Tritium Release Behavior from Steels Irradiated by High Energy Protons

Model calculation of tritium release behavior from lithium titanate

Among the various tritium transport processes in lithium ceramics, the importance and the detailed mechanism of surface reactions still remains to be elucidated. The dynamic adsorption and desorption (DAD) model has been developed to calculate the tritium release behavior from Li 2O surface. In the DAD model, the tritium release behavior is considered to be controlled by the surface coverage of adsorbed species, such as OH −, O 2−, H −, determined by the H 2 and H 2O concentrations in the gas phase and the temperature. In a previous paper, the tritium residence time on the Li 2O surface was calculated from the model, and the calculation results were compared with the experimental results. It was shown that the calculation results agreed well with the experimental results. In the present paper, a model for the surface behavior of tritium release from Li 2TiO 2 was constructed, according to the concept of the model for Li 2O. The calculation results are compared with the experimental results and the validity of the tritium release mechanism assumed in the model is discussed.

Tritium release from neutron-irradiated Li 2O: Transport in porous sintered pellets

The tritium release behavior from Li 2O sintered pellets (81–88% T.D.) is examined by isothermal heating tests. (1) For the 88% T.D. specimens, the fraction of residual tritium is found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (2) For the 81% T.D. specimens, which are annealed after irradiation at 630 K for 4 h, the fraction of residual tritium is also found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (3) For the 81% T.D. specimens as irradiated, the tritium release rate is found to follow the square-root law of the annealing time. The rate-determining process is controlled by Kohlrauch stretched exponential form. Tritium trapped in irradiation defects released with recovering the defects by isothermal heating.

Solid breeder test blanket module design and analysis

This paper presents the design and analysis for the US ITER solid breeder blanket test articles. Objectives of solid breeder blanket testing during the first phase of the ITER operation focus on exploration of fusion break-in phenomena and configuration scoping. Specific emphasis is placed on first wall structural response, evaluation of neutronic parameters, assessment of thermomechanical behavior and characterization of tritium release. The tests will be conducted with three unit cell arrays/sub-modules. The development approach includes: (1) design the unit cell/sub-module for low temperature operations and (2) refer to a reactor blanket design and use engineering scaling to reproduce key parameters under ITER wall loading conditions, so that phenomena under investigation can be measured at a reactor-like level.