Tritium Diffusion Research Articles

Physical property changes of neutron-irradiated aluminum nitride and their recovery behavior by annealing using a step-heating dilatometer

Aluminum nitride (AlN) is a candidate tritium permeation, electric insulation and corrosion barriers for several kinds of blankets such as molten salt–cooled (FLiBe) or liquid metal (Li-Pb or Li)-cooled blankets because of its desirably low dielectric constant and tritium diffusion. Commercially available AlN ceramic specimens were neutron-irradiated at two different fluences but the same irradiation temperature in the Japan Materials Testing Reactor. Specimen swelling was found to be slightly different for both conditions, with higher dose causing greater swelling. All irradiated specimens consisted from hexagonal AlN phase with α-Al2O3 phase occurring on the surface after long-time post-irradiation annealing in He atmosphere. The a- and c-axis experienced isotropic increase and degree of unit-cell volume change was almost the same with the macroscopic volume increase obtained from the length change. This result indicates uniform distribution of Frankel pairs. After step-wise thermal annealing by using a dilatometer up to 1673 K for 6 h at each step, the maximum recovery was found at 1573 K. Based on the recovery rates at each step by first-order analysis, macroscopic length recovery during annealing can be divided into three regions with different activation energies, low temperature (373–523 K) with ∼4.5 eV, intermediate temperature (523–873 K) with 0.5–1.0 eV, and high temperature (873–1273 K) with 2.0–2.5 eV. Over 1273 K, a slight increase of length was observed. It is thought that the expansion is due to oxidation.

Tritium transport model at the minimal functional unit level for HCLL and WCLL breeding blankets of DEMO

The Helium-Cooled Lithium-Lead (HCLL) and Water-Cooled Lithium-Lead (WCLL) Breeding Blankets are two of the four European blanket designs proposed for the DEMO reactor. A tritium transport model inside the blanket is fundamental to assess its preliminary design and safety features. Tritium transport and permeation are complex phenomena to be taken into account in the evaluation of the tritium balance, in order to guarantee tritium self-sufficiency and to characterise tritium concentrations, inventories and losses. In this context, the study has been performed at the minimal functional unit level of the outboard equatorial breeding blanket module, which is, during continuous operative conditions, one of the most loaded modules and this results in higher permeation phenomena. For these purposes, a 2D model for the breeder unit of HCLL and a 3D model for the single cell of WCLL breeding blanket concepts have been investigated. The models include advection-diffusion of tritium into the lead-lithium eutectic alloy, transfer of tritium from the liquid interface towards the steel, diffusion of tritium inside the steel, transfer of tritium from the steel towards the coolant, and advection-diffusion of diatomic tritium into the coolant. Thermal field, tritium generation rate profile, velocity field of lead-lithium and coolant have been also taken into account.

Tritium Diffusion Pathways in γ-LiAlO2 Pellets Used in TPBAR: A First-Principles Density Functional Theory Investigation

With superior thermophysical and thermochemical properties, γ-LiAlO2 has high compatibility with other blanket materials and is used in the form of an annular pellet in tritium-producing burnable absorber rods (TPBARs) to produce tritium by thermal neutron irradiation of 6Li. In radiation damaged γ-LiAlO2, different types of vacancies, defects of its constituent elements, and other trapping sites hinder the diffusion process of tritium. In this study, the first-principles density functional theory approach is used to study the diffusion mechanisms of tritium defect and its species, such as interstitial and substitutional tritium defects, oxygen–tritium vacancy defects, and interaction of tritium with oxygen vacancies in defective and nondefective γ-LiAlO2. The obtained results provide an understanding of how such defects hamper the diffusivity and solubility of tritium. By calculating several different diffusion pathways, our results show that the smallest activation energy barrier is 0.63 eV for substitu...

Effects of baking in deuterium atmosphere on tritium removal from tungsten

Tritium-containing polycrystalline tungsten, which was pre-irradiated with helium prior to tritium exposure, was isothermally heated under a vacuum or deuterium gas in order to investigate the effect of the deuterium on tritium removal from the tungsten. It is found that in the case of 3 h of baking at temperatures varying from 423 K and 573 K, tritium retention after baking in deuterium gas became smaller than that in a vacuum. At the baking temperature of 573 K, tritium removal was hastened in the case of baking in deuterium atmosphere. The results indicate that the baking in deuterium atmosphere is effective to remove tritium from tungsten. The acceleration of diffusion of tritium in bulk and recombination of tritium with deuterium at the surface might be possible mechanisms.

Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

Tritium transport in metal and molten salt fluids combined with diffusion through high-temperature structural materials is an important phenomenon in both magnetic confinement fusion (MCF) and molten salt reactor (MSR) applications. For MCF, tritium is desirable to capture for fusion fuel. For MSRs, uncaptured tritium potentially can be released to the environment. In either application, quantifying the time- and space-dependent tritium concentration in the working fluid(s) and structural components is necessary.Whereas capability exists specifically for calculating tritium transport in such systems (e.g., using TMAP for fusion reactors), it is desirable to unify the calculation of tritium transport with other system variables such as dynamic fluid and structure temperature combined with control systems such as those that might be found in a system code. Some capability for radioactive trace substance transport exists in thermal-hydraulic systems codes (e.g., RELAP5-3D); however, this capability is not coupled to species diffusion through solids. Combined calculations of tritium transport and thermal-hydraulic solution have been demonstrated with TRIDENT but only for a specific type of MSR.Researchers at Oak Ridge National Laboratory have developed a set of Modelica-based dynamic system modeling tools called TRANsient Simulation Framework Of Reconfigurable Models (TRANSFORM) that were used previously to model advanced fission reactors and associated systems. In this system, the augmented TRANSFORM library includes dynamically coupled fluid and solid trace substance transport and diffusion. Results from simulations are compared against analytical solutions for verification.

Results of the first Cool-down of the KATRIN Cryogenic Pumping Section

The Karlsruhe Tritium Neutrino (KATRIN) experiment uses the kinematics of tritium β-decay to determine the effective neutrino mass with a sensitivity of mν = 200meV/c2 (90% C.L.). In order to measure the decay electrons it is important to guide them adiabatically from the source to the spectrometer. In addition the diffusion of tritium into the spectrometers from the source has to be reduced by 14 orders of magnitude as tritium inside the spectrometers would induce additional background. For these two tasks the transport and pumping section were constructed. The last part of this section is the Cryogenic Pumping Section (CPS), which aims to reduce the residual gas flow by more than seven orders of magnitude. For this a cold argon frost surface, with a ≈ 2 m2) area that is maintained at 3 K, will be prepared to adsorb the incoming tritium molecules.Before the whole KATRIN setup will be connected together the performance of CPS will be tested on its own. This poster presents the measurement results of the first cool-down of the CPS.This work was supported by the GRK1694 and the Helmholtz Association.

Preparation of Li2TiO3 ceramic with nano-sized pores by ultrasonic-assisted solution combustion

Li2TiO3 is a candidate material for tritium breeding in the future nuclear fusion reactor. In this study, Li2TiO3 powder was synthesized by ultrasonic-assisted solution combustion synthesis (USCS) in a single step. The ultrasonic transducer with the power of 1000W was introduced in the synthesis process. The crystallite size of Li2TiO3 powder prepared by utilization of ultrasonic power is significantly decreased to ∼5.0nm, while the one obtained without ultrasonic power is 20.0nm. Li2TiO3 ceramic sintered from USCS powder at 800°C exhibits the small grain size of 330nm and the open pores size of 140nm. The crush load of the ceramic reaches 37.2N although the structure is porous. Compared with the ceramic prepared by solid-state reaction and conventional solution combustion synthesis, USCS sample has a higher conductivity of 2.0×10−6Sm−1 at room temperature, indicating the lower tritium diffusion barrier in the ceramic.

Radiation induced deuterium absorption dependence on irradiation temperature, dose rate, and gas pressure for SiC

During ITER and DEMO reactor operation Li-Pb blanket flow channel inserts made of SiC will be exposed to both radiation and tritium. Absorption, desorption, and tritium diffusion are expected to occur and will strongly depend on the irradiation conditions; temperature, and neutron and gamma fluxes. Previous results showed that marked deuterium absorption, associated with silicon deuterium bonding, occurs for SiC materials when subjected to a radiation field, and that this radiation enhanced absorption strongly depends on the ionizing radiation and also on displacement damage. Here the roles played by irradiation temperature, dose rate, dose, and deuterium gas pressure are addressed for reaction bonded SiC. The results show that radiation induced deuterium absorption depends linearly on total ionizing dose and deuterium gas pressure, but not on dose rate. Behaviour with irradiation temperature is more complex, and clear changes in the deuterium thermal desorption are observed to occur depending on irradiation temperature. SIMS results for high temperature loaded RB SiC are consistent with radiation enhanced diffusion.

An integrated model of tritium transport and corrosion in Fluoride Salt-Cooled High-Temperature Reactors (FHRs) – Part I: Theory and benchmarking

The Fluoride Salt-Cooled High-Temperature Reactor (FHR) is a pebble bed nuclear reactor concept cooled by a liquid fluoride salt known as “flibe” (7LiF-BeF2). A model of TRITium Diffusion EvolutioN and Transport (TRIDENT) was developed for use with FHRs and benchmarked with experimental data. TRIDENT is the first model to integrate the effects of tritium production in the salt via neutron transmutation, with the effects of the chemical redox potential, tritium mass transfer, tritium diffusion through pipe walls, tritium uptake by graphite, selective chromium attack by tritium fluoride, and corrosion product mass transfer. While data from a forced-convection polythermal loop of molten salt containing tritium did not exist for comparison, TRIDENT calculations were compared to data from static salt diffusion tests in flibe and flinak (0.465LiF-0.115NaF-0.42KF) salts. In each case, TRIDENT matched the transient and steady-state behavior of these tritium diffusion experiments. The corrosion model in TRIDENT was compared against the natural convection flow-loop experiments at the Oak Ridge National Laboratory (ORNL) from the 1960s and early 1970s which used Molten Salt Reactor Experiment (MSRE) fuel-salt containing UF4. Despite the lack of data required by TRIDENT for modeling the loops, some reasonable results were obtained. The TRIDENT corrosion rates follow the experimentally observed dependence on the square root of the product of the chromium solid-state diffusion coefficient with time. Additionally the TRIDENT model predicts mass transfer of corrosion products from the hot to the cold leg (as was observed in the experiments with salts containing UF4). In a separate paper the results of TRIDENT simulations in a prototypical FHR are presented.

Analysis on tritium permeation in spherical pressure vessel

Most of the tritium quantity stored in a pressure vessel decays into helium-3 and a tiny part permeates the inner surface of the vessel and enters the wall material with time. The tritium in the wall material not only diffuses in the wall and till permeates out of the vessel through the outer surface, but also decays into helium-3 which remains in the wall. The tritium permeating out of the vessel can contaminate environment and harm human being’s health and safety, therefore it is necessary to realize and control the permeation quantity. Directing at this situation that outer surface of the vessel is provided with general mass transfer boundary condition and the tritium inside the vessel is van der Waals gas, and taking simultaneously both decay and permeation of tritium within the vessel and decay and diffusion and permeation of tritium having permeated into the wall material into consideration, the theoretical model was developed and the theoretical formulas of total permeation quantity and surplus after decay of tritium were deduced. Through analytical calculation, the curves of total permeation quantity and surplus after decay of tritium versus time, total permeation quantity of tritium versus both convectivemass transfer coefficient of the outer surface and the wall thickness of the vessel were plotted and related laws were revealed. Corresponding to the theoretical calculation, a tritium stored experiment was performed, the sample of protecting Argon gas which was between the tritium stored vessel and an outer closed container was extracted, and the tritium content was measured by using the method of proportional counter. Then, through conversion, the transient surplus after decay of the tritium out of the vessel was obtained. The analytical calculations data were compared with experimented measurement ones, which indicates the reasonability of theoretical analysis. The obtained laws that tritium permeation quantity changes with such parameters as time etc. are useful for theoretical design of tritium storied vessels.

Hydrogen Diffusion and Trapping in Fluoride Salt Melts and Graphite Electrodes

Molten salt processes were developed decades ago as a promising technology for material processing and industrial process heat applications. This technology is still considered applicable for thermal energy storage and power production. Liquid salts can indeed be considered legitimate candidates for several heat transfer applications, thanks to favorable physical properties such as high boiling point, low vapor pressures, large specific heat, and large thermal conductivity. In particular, fluoride-based molten salts are considered as a coolant for several nuclear reactor advanced design owing to favorable neutronic cross sections and reduced corrosion of structural materials. FHR, Fluoride-salt cooled High-temperature Reactors, use flibe salt (2LiF-BeF2) to cool a graphite pebble-bed-fueled core. A detailed knowledge of the chemistry of fluoride melts is important for an efficient corrosion control and reliable FHR reactor design. The corrosion mechanisms in a molten salt system are different from aqueous systems; molten salt chemistry prevents the formation of a thermodynamically stable passivation layer on structural metals, therefore the corrosion control has to be achieved through the control of the redox potential of the fluoride salt. Solubilities and diffusivities of metals, rare earths, fission products and tritium in the fluoride melt and within graphite framework, are of direct interest for the FHR design community. Ionic coordination in the liquid and gas phase also needs to be fundamentally understood in order to address the above mentioned problems. This paper will provide a review of the electrochemical techniques and spectroscopy methods used in the past to study fluoride salt systems. Furthermore, state of the art of techniques used in other high temperature molten salt environments and low temperature ionic liquids will be reviewed. Tritium management is the main design challenge for the FHR reactors. Tritium is produced through transmutation of Lithium-6 in a neutron environment and might be released into the atmosphere through the heat exchanger walls. However, in FHR reactors, the graphite pebble bed and graphite reflector may be the ultimate sink for tritium. Goal of the ongoing research at UW Madison is to define tritium diffusion and absorption into graphite, using hydrogen as a surrogate. This paper will address theoretical electrochemical techniques used in aqueous system to study absorption of hydrogen into metals (Potentiostatic Double-Step [1]) and also experimental techniques (Electrochemical Impedance Spectroscopy [2]) applicable to the same issue. These techniques will be applied initially to graphite/water systems and later to the graphite/flibe system. The paper will present preliminary results and report the challenges pertaining to experimental setup with fluoride salt systems, due to the high melting temperatures, the highly corrosive nature of the salt, in addition tothe toxicity of Beryllium, main component of flibe salt. REFERENCES: [1] B. Pound, G. Wright, et al., ‘A potentiostatic double-step method for measuring hydrogen atom diffusion and trapping in metal electrodes—II. Experimental’, Acta Metall., vol. 35, no. I, pp. 263–270, (1987). [2] D. D. Macdonald and M.C.H. McKubre. “Impedance Measurements in Electrochemical Systems”. Chapter 2 in Modern Aspects of Electrochemistry. 14, 61(1982).

Design of a permeator against vacuum for tritium extraction from eutectic lithium-lead in a DCLL DEMO

One of the most important issues in future fusion power plants is the extraction of tritium generated in the breeders in order to achieve self-sufficiency. When the breeder is a liquid metal one of the most promising techniques is the Permeation Against Vacuum, whose principle is based on tritium diffusion through a permeable membrane in contact with the liquid metal carrier and its further extraction by a vacuum pump. A conceptual design of permeator has been developed, taking into account the features of a DEMO reactor with a Dual Coolant Lithium Lead (DCLL) breeder blanket. The study is based on the analysis of different membranes and geometries aiming at the overall efficiency (extraction capability) of the device, as well as its compatibility with the breeder material. The permeator is based on a rectangular section multi-channel distribution where the liquid metal channels and vacuum channels are alternated in order to maximize the contact area and therefore to promote tritium transport from the bulk to the walls. The resulting permeator design has an excellent estimated extraction efficiency, of 80%, in a relatively compact device.

Technical Means for Safe Management of Radioactive Waste Containing Tritium

The paper considers the possibility for separate safe storage of radioactive waste containing tritium using passive protection of concrete, clay, silica gel and additional engineering barrier in the form of container made of polymeric materials. It describes some physical properties of the container made of two polymeric materials for storage of tritium radionuclides separated from other radionuclides. The experts calculated the thickness of the container walls taking into account the ability of tritium diffusion as the main physical-chemical characteristics of tritium. The use of Teflon coating protects the plastic container in case of fire.

Trapping and thermal diffusion for energetic deuterium implanted into SiC

During ITER and DEMO reactor operation Li–Pb blanket flow channel inserts made from SiC will be exposed to both radiation and tritium. Absorption, desorption, and tritium diffusion are expected to occur and will strongly depend on the irradiation conditions; temperature, and neutron and gamma fluxes. Reaction bonded (RB) SiC samples were deuterium implanted at both room temperature and 450°C at different implantation energies and the corresponding TSD spectrum was obtained for each implantation energy. After implantation the samples were subjected to SIMS analysis. The TSD spectra obtained for all the samples implanted at different energies are very similar and characterized by a prominent deuterium desorption occurring at temperatures between 450 and 1000°C with a maximum that exhibits a clear trend to shift toward higher temperature as either implantation energy or implantation temperature increase. SIMS analysis before heating the deuterium implanted samples indicates that the implanted deuterium has a tendency to become bonded to Si rather than to C. The SIMS analysis shows that once heated up to 1000°C only part of the implanted deuterium was thermally released. The temperature shift observed when increasing the deuterium implantation energy and, hence, penetration, implies a deuterium diffusivity value at 700°C of about 8.5×10−17m2/s.

Determination of tritium generation and release parameters at lithium CPS under neutron irradiation

This paper describes the main parameters of tritium generation and release from lithium capillary-porous system (CPS) under neutron irradiation at the IVG.1M research reactor. The experiments were carried out using the method of mass-spectrometric registration of released gases and using a specially constructed ampoule device. Irradiation was carried out at different reactor thermal powers (1, 2 and 6 MW) and sample temperatures from 473 to 773K.In the experiments a very small tritium release was detected likely due to its high solubility in liquid lithium. It can be caused by formation of lithium tritide during tritium diffusion to the lithium surface.

Irradiation effects and hydrogen behavior in H2+ and He+ implanted γ-LiAlO2 single crystals

Gamma-phase lithium aluminate (γ-LiAlO2) is a breeder material for tritium, a necessary substance for strategic stockpile and fusion power systems. A fundamental study of structural evolution and tritium diffusion in γ-LiAlO2 under displacive irradiation is needed to fully assess the material performance. This study utilizes ion implantation of protium (surrogate for tritium) and helium in γ-LiAlO2 single crystals at elevated temperatures to emulate the irradiation effects. The results show that at 573 K there are two distinct disorder saturation stages to 1 dpa without full amorphization; overlapping implantation of H2+ and He+ ions suggests possible formation of gas bubbles. For irradiation to 1021 H+/m2 (0.36 dpa at peak) at 773 K, amorphization occurs at surface with H diffusion and dramatic Li loss; the microstructure contains bubbles and cubic LiAl5O8 precipitates with sizes up to 200 nm or larger. In addition, significant H diffusion and release are observed during thermal annealing.

High efficiency 4H-SiC betavoltaic power sources using tritium radioisotopes

Realization of an 18.6% efficient 4H-silicon carbide (4H-SiC) large area betavoltaic power source using the radioisotope tritium is reported. A 200 nm 4H-SiC P+N junction is used to collect high-energy electrons. The electron source is a titanium tritide (TiH3x) foil, or an integrated titanium tritide region formed by the diffusion of tritium into titanium. The specific activity of the source is directly measured. Dark current measured under short circuit conditions was less than 6.1 pA/cm2. Samples measured with an external tritium foil produced an open circuit voltage of 2.09 V, short circuit current of 75.47 nA/cm2, fill factor of 0.86, and power efficiency of 18.6%. Samples measured with an integrated source produced power efficiencies of 12%. Simulations were done to determine the beta spectrum (modified by self absorption) exiting the source and the electron hole pair generation function in the 4H-SiC. The electron-hole pair generation function in 4H-SiC was modeled as a Gaussian distribution, and a closed form solution of the continuity equation was used to analyze the cell performance. The effective surface recombination velocity in our samples was found to be 105–106 cm/s. Our analysis demonstrated that the surface recombination dominates the performance of a tritium betavoltaic device but that using a thin P+N junction structure can mitigate some of the negative effects.

Study on a method for loading a Li compound to produce tritium using high-temperature gas-cooled reactor

Tritium production using high-temperature gas-cooled reactors and its outflow from the region loading Li compound into the helium coolant are estimated when considering the suppression of tritium outflow. A Li rod containing a cylindrical Li compound placed in an Al2O3 cladding tube is assumed as a method for loading Li compound. A gas turbine high-temperature reactor of 300MW electrical nominal capacity (GTHTR300) with 600MW thermal output power is considered and modeled using the continuous-energy Monte Carlo transport code MVP-BURN, where burn-up simulations are carried out. Tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. A GTHTR300 can produce 400–600g of tritium over a 180-day operation using the chosen method of loading the Li compound while minimizing tritium outflow from the cladding tube. Optimizing tritium production while suppressing tritium outflow is discussed.

Tritium through-diffusion test in non steady state: Can the effective diffusion coefficient be determined?

The diffusion coefficient of tritium in its liquid form through cementitious materials is an indicator of durability in the domain of nuclear waste disposal. Because the cement-based materials designed for the waste storage have very high confinement properties, diffusion tests tend to last several months if not years in some cases. Therefore, basing the determination of the tritium diffusion coefficient on the experimental data measured during non steady state is tempting. The aim of this work is to demonstrate that the determination during the transient regime of the effective diffusion coefficient and the material porosity leads to non-negligible underestimation of these quantities. Moreover, it appears that, when using a numerical model for the determination, their validity can only be judged on the basis of more than one criterion, whether those criteria are obtained from the experimental data or from a numerical model.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.