Tritium Migration Research Articles

Influence of heating rate and storage condition on thermal desorption of deuterium in tungsten

The retention of hydrogen isotopes in tungsten has been a critical issue in future fusion facilities. Thermal desorption spectroscopy (TDS) is a common method for the studies of hydrogen isotope retention. To explore the effect of heating rate in TDS measurement, a series of deuterium (D) desorption experiments with different heating rates were conducted. The results were compared with TMAP7 (Tritium Migration and Analysis Program, version 7) simulations and Kissinger's theory. An empirical expression for the analysis of desorption spectra with different heating rates was derived. The expression was verified in another set of experiments on helium (He) desorption. Furthermore, the formation of molecules containing D atoms during TDS measurements was investigated. Specimens exposed to D plasma were stored in air and vacuum respectively for three months prior to desorption experiments. Signals of H, H2, HD, D2, HDO and D2O were detected. The reasons for the formation of these species were explored.

Toward a high-fidelity tritium transport modeling for retention and permeation experiments

Tritium Migration Analysis Program version 8 (TMAP8), the latest version of TMAP, was developed within the framework of the Multiphysics Object-Oriented Simulation Environment (MOOSE). Created at Idaho National Laboratory (INL), MOOSE is an open-source, dimension-agnostic, fully coupled, and fully implicit multiphysics platform featuring massively parallel computation capabilities. Using TMAP8, tritium transport in a divertor monoblock was analyzed to elucidate the effects of pulsed operation (up to fifty 1,600 s plasma discharge and cool-down cycles) on the tritium in-vessel inventory source term and ex-vessel release term (i.e., tritium retention and permeation) for safety analysis. With its built-in Message Passing Interface capability, TMAP8 can, in under 2 h, simulate tritium transport in three different layered materials (i.e., tungsten, copper, and copper-chromium-zirconium alloy) in 2D geometry, using a single device/computer with 10 cores. The MOOSE-based TMAP8 code can leverage other MOOSE tools developed under the Nuclear Energy Advanced Modeling and Simulation program to perform tritium and thermal transport in complex geometries and multiphysics environments. And via its massively parallel computation, MOOSE will enable the fusion pilot plant designers to conduct high-fidelity multiphysics modeling for the design of the divertor and blanket systems as well as for the safety analysis.

Deuterium retention in cyclic transient heat loaded tungsten with increasing cycle numbers

Surface damage and microscopic defect evolution of tungsten (W) armor under transient heat loads are key factors for fuel retention in fusion reactors. In this work, experiments were conducted to investigate the effects of cyclic thermal shocks on deuterium (D) retention and surface blistering in W. Thermal shock experiments were conducted on recrystallized W using an electron beam with a power density of 0.15 GW m−2 across 100–1500 cycles, followed by D plasma exposure with high-fluence (∼1 × 1026 D m−2). The results demonstrate that samples subjected to 500 and 1500 cycles exhibit a significant presence of sub-grains within 90 μm. Notably, the inhibition of blistering induced by thermal shock leads to a substantial reduction in D retention (5.45 × 1019 D m−2) at lower cycle numbers (100 cycles) compared to the reference sample (2.35 × 1020 D m−2) which was only exposed to D plasma. When cycle numbers increase to 500 and 1500, D retention reaches 1.98 × 1020 D m−2 and 4.56 × 1020 D m−2, respectively. Based on the tritium migration analysis program, we propose that total D retention is a consequence of the competition between defects reduced by thermal shock-induced suppression of blistering and defects generated by plastic deformation induced by thermal stress. D retention initially decreases with the increase in cycle numbers, followed by a subsequent rise, with the inflection point slightly higher than 500 cycles. Additionally, due to the extensive scope of thermal stress, an escalated exposure period will result in substantial D captured by heat-induced defects, consequently intensifying the D retention. Whether there exists an upper limit to D retention induced by the increasing thermal shock cycles necessitates further experimental analysis. Nonetheless, it is evident that thermal shock significantly contributes to D retention within a profoundly deep bulk region under high cycles.

The Impact of Tides and Monsoons on Tritium Migration and Diffusion in Coastal Harbours: A Simulation Study in Lianyungang Haizhou Bay, China

Many nuclear power plants have been built along China’s coasts, and the migration and diffusion of radioactive nuclides in coastal harbours is very concerning. In this study, considering the decay and free diffusion of radioactive nuclides, a local hydrodynamic model based on the FVCOM was built to investigate the migration and diffusion of the radioactive nuclide tritium in Haizhou Bay, China. This model was calibrated according to the observed tidal level and flow velocity and direction, which provide an accurate background. This study aimed to evaluate the impact of tides and monsoons on the migration path and concentration variations in tritium over time. The results demonstrated that the simulated flow field can reflect real-life receiving waters. The distribution of the tritium concentration is affected by the flow field, which is related to the tides. Moreover, more severe radioactive contamination was exhibited in winter than in summer because monsoons may have hindered the migration and diffusion of tritium within the harbour. Given the poor hydrodynamic conditions and slow water exchange in the open ocean in Haizhou Bay, the diffusion rate of radioactive nuclides outside the bay area was higher than that within it.

Multiphysics analysis of thermal mechanical behavior and tritium migration in FeCrAl and Cr-coated Zircaloy cladding under PWR normal operating and transient conditions

Both Cr-coated Zircaloy and the FeCrAl alloy have garnered attention as potential substitutes for PWR cladding materials, due to their impressive qualities, such as robust high-temperature oxidation and corrosion resistance. However, FeCrAl alloy is found to accelerate the tritium permeation throughout the cladding, this heightened tritium permeation prompts easier release from the cladding into the coolant. Consequently, the assessment of material tritium permeation becomes a crucial aspect in fuel performance analysis. In this work, the multiphysics models of Cr-coated Zircaloy and FeCrAl cladding are reviewed firstly. Then the multiphysics models are implemented in CAMPUS code and the validation of tritium migration model is presented. Subsequently, the fuel performance under normal and accident (LOCA and RIA) conditions and the tritium migration in three types of claddings are simulated and discussed. The simulation results show that the utilization of FeCrAl cladding, instead of Zircaloy, results in an overall reduction in fuel temperature under both normal and LOCA conditions. Furthermore, both the Cr-coated Zircaloy and FeCrAl cladding are found to effectively delay the cladding failure time under LOCA condition. However, the total strain of FeCrAl cladding is found to be significantly higher than that of the Cr-coated Zircaloy cladding under RIA condition. And the Cr-coated Zircaloy is found to have better tritium resistance effect than FeCrAl cladding, which greatly reduces the coolant contamination treatment. Therefore, based on the combined consideration of fuel performance and tritium resistance, Cr-coated Zircaloy emerges as the preferred candidate cladding over the FeCrAl cladding.

Computer simulation of tritium release behavior of Li4SiO4 ceramic breeder with an improved packed bed model

In order to have a deep understanding of the tritium migration and release mechanisms in ceramic breeders, computer simulation of the tritium release behavior of Li4SiO4 ceramic breeder was performed by using an improved tritium release model. The influences of various factors, such as surface adsorbed water, water vapor in the purge gas, hydrogen in the purge gas, grain size and tritium production amount, on the tritium release process were systematically investigated. The simulation results have shown that: 1) The surface adsorbed water and water vapor in the purge gas can remarkably facilitate tritium release as tritiated water (HTO) in low temperature region (typically <450 °C); 2) Adding hydrogen to the purge gas is effective to promote tritium release as tritium gas (HT), but the effectiveness is sensitive to surface adsorbed water and water vapor in the purge gas; 3) Large grain size and tritium production amount can relatively weaken the effects of surface adsorbed water and water vapor in the purge gas. In general, the simulation results were consistent with the observations in tritium release experiments, which were helpful to further elucidate the complex behavior of tritium release from Li4SiO4.

MELCOR-TMAP: The integration of MELCOR for fusion and TMAP4 for fusion reactor systems safety analysis and tritium inventory tracking

The design of fusion reactors requires the prediction of radioactive tritium migration during normal operating conditions and accident scenarios. Safety standards such as those from the U.S. Department of Energy (DOE) require that fusion reactor designs limit the release of hazardous material to the environment. In addition to safety requirements, the rarity of tritium supplies for burning D-T fusion plasma requires tritium breeding blankets. Computational modeling of migration of tritium and various reaction products throughout the fusion reactor serves to create designs that fulfill these requirements. MELCOR-TMAP, owned and developed at Idaho National Laboratory (INL), is an integrated system-level engineering code adapted for the analysis of fusion reactor systems. The history of the conception and evolution of the MELCOR-TMAP code, as well as its current capabilities, recent updates including a corrected composite material wall nodalization scheme, and plans for future development, are discussed.

Incorporation of tritium by pepper and eggplant cultures with short-term exposure to tritium oxide

Incorporation of tritium by pepper and eggplant cultures with short-term exposure to tritium oxide

Effect of neutron dose on the tritium release behavior of Li2TiO3–0.5Li4SiO4 biphasic ceramic

Neutron irradiation with different fluence and flux on the Li2TiO3-0.5Li4SiO4 biphasic ceramic powders was conducted. The tritium release behavior was studied by thermal desorption spectroscopy (TDS) with different heating rates and isothermal heating at various temperatures. It was found that the tritium TDS spectra were shifted toward the higher temperature side as the neutron fluence was reduced. The tritium migration in the biphasic ceramic was governed by diffusion and de-trapping processes with the activation energies of 0.29–0.50 eV and 1.00–1.13 eV for the neutron-irradiated samples with the fluence of 8.25 × 1015 and 3.96 × 1016 n cm−2. For isothermal heating, there is another tritium desorption peak when the sample was continually heated to higher temperatures after isothermally heating at 690 K and 800 K, indicating that tritium still remained in the ceramic under isothermal heating conditions. The intensity of the second peak was decreased at the higher isothermal heating temperatures. The diffusion model with detrapping reaction can fit the experimental data, suggesting that both the detrapping and diffusion processes should be considered to estimate tritium release characteristics for the tritium breeding biphasic ceramics.

Tritium Transport Modeling at the Pore Scale in Ceramic Breeder Materials Using TMAP8

Fusion reactors depend on the blanket material to breed and release tritium at the same rate or faster than it is consumed by the fusion reaction. Cellular ceramic breeders (CCBs) are dense materials that can maintain a high tritium breeding ratio while promoting tritium release because of highly connected pores. Assessing the tritium breeding capabilities of these materials requires a combination of extensive experimental and modeling efforts. In this work, we develop and calibrate a multiphysics model of tritium transport. This novel model accounts for ceramic and pore diffusion, trapping and detrapping, and several surface reactions at the pore surface. We perform a sensitivity analysis and calibrate the model by comparing its predictions against experimental measurements of deuterium absorption. The calibrated model is then used to model tritium absorption in samples with different pore microstructures to investigate the effect of pore interconnectivity on tritium absorption. The model is part of the development of the multiscale, multiphysics framework for tritium transport [i.e., the Tritium Migration Analysis Program (TMAP8)], which is itself built on top of the finite-element multiphysics framework multiphysics object-oriented simulation environment (MOOSE). This study demonstrates some of TMAP8’s capabilities and is the first step toward assessing the tritium breeding capabilities of ceramic breeder material designs.

Analysis on Emergency Treatment Effect of Tritium Leakage Accident in the Nuclear Reactor Under Different Ventilation Conditions

The high-pressure tritium leakage is a typical nuclear leakage accident, which can cause radioactive harm. The study aims at the tritium leakage problem in the high-pressure vessel of a nuclear reactor, installing a typical tritium leakage scenario and establishing the thermodynamic model of tritium leakage behavior which describes the transient process of tritium leakage. To analyze the effect of different ventilation scenes on tritium removal efficiency, the study sets up six different ventilation scenes and carries out the numerical simulation of tritium migration and mixing behavior based on Fluent. The result shows that different ventilation scenes will form different swirl distributions; the more complex the swirl distribution is, the faster the tritium concentration decreases. The shortest time required to decrease tritium concentration in space to the background level at the ventilation rate of 1 m3 s−1 is about 830 s and the longest time is about 1100 s. The overall tritium removal efficiency can be effectively improved when the swirl concentration distribution area is opposite to the exhaust outlet position. The setting method of ventilation is optimized, which provides technical support for the emergency treatment of tritium leakage accidents in the room.

A comparative study on the hydrogen dissolution and release behaviors in the zirconate proton conductors by TDS and TMAP4 analysis

Proton-conducting oxides are potential materials for electrochemical devices such as fuel cells, hydrogen pumps, hydrogen sensors, and the tritium purification and recovery system in nuclear fusion reactors. The hydrogen concentration in oxide materials is important, but its precise measurement is difficult. In this study, thermal desorption spectroscopy (TDS) was used to investigate hydrogen dissolution and release behavior in proton-conducting oxides, yttrium (Y), and cobalt (Co) doped barium-zirconates in the temperature range of 673-1273 K using deuterium (D2) and heavy water (D2O). Specimens were prepared with conventional powder metallurgy: the powder of three zirconates, BaZr0.9Y0.1O3-a (BZY), BaZr0.955Y0.03Co0.015O3-a (BZYC), and CaZr0.9In0.1O2.95 (CZI) was pressed into discs and fired in the air at 1873 K for 20 h. The densities of the sintered BZY, BZYC, and CZI specimens were 98 percent, 99.7 percent, and 99.5 percent of the theoretical densities respectively. XRD, SEM, and EDX were performed for the structural and morphological analysis of the sintered samples. From TDS measurement, a similar trend of temperature-dependent hydrogen solubility was obtained for all samples compared to the tritium imaging plate (TIP) methods literature data of HT- and DTO-exposed samples. To compare the experimental results of the deuterium desorption profile derived by TDS analysis, the simulation code of the tritium migration analysis program, version 4 (TMAP4) was employed.

Assessment of BISON capabilities for component-level prediction of tritium transport in fusion and fission applications

We assessed the hydrogen migration and redistribution model implemented in the BISON fuel performance code for component-level tritium transport applications in fusion and fission example scenarios. We developed BISON models of an ITER heat exchanger, light water reactor fuel cladding, and a fluoride salt cooled high temperature reactor heat exchanger. Reasonable agreement between reported values in reference studies and BISON predictions demonstrated the ability of the BISON models to predict tritium transport behavior through steel materials. Next, sensitivity and uncertainty analysis was used to understand key parameter sensitivities within the BISON model for evaluation of tritium migration and redistribution. . The sensitivity analysis showed that the diffusivity activation energy was the most important parameter and that the heat of transport was the least important parameter for tritium migration in steels. Overall, the capabilities of the BISON code for component level modeling of tritium transport are promising.

Comparative Studies of Hydrogen Dissolution and Release Behavior in Zirconate Oxides by TDS and TMAP4 Analysis

Proton-conducting oxides are potential materials for electrochemical devices such as fuel cells, hydrogen pumps, hydrogen sensors, and the tritium purification and recovery system in nuclear fusion reactors. The hydrogen concentration in oxide materials is important, but its precise measurement is difficult. In this study, thermal desorption spectroscopy (TDS) was used to investigate hydrogen dissolution and release behavior in proton-conducting oxides, yttrium (Y), and cobalt (Co) doped barium-zirconates in the temperature range of 673-1273 K using deuterium (D2) and heavy water (D2O). Specimens were prepared with conventional powder metallurgy: the powder of three zirconates, BaZr0.9Y0.1O3-a (BZY), BaZr0.955Y0.03Co0.015O3-a (BZYC), and CaZr0.9In0.1O2.95 (CZI) was pressed into discs and fired in the air at 1873 K for 20 h. The densities of the sintered BZY, BZYC, and CZI specimens were 98 percent, 99.7 percent, and 99.5 percent of the theoretical densities respectively. XRD, SEM, and EDX were performed for the structural and morphological analysis of the sintered samples. From TDS measurement, a similar trend of temperature-dependent hydrogen solubility was obtained for all samples compared to the tritium imaging plate (TIP) methods literature data of HT- and DTO-exposed samples. To compare the experimental results of the deuterium desorption profile derived by TDS analysis, the simulation code of the tritium migration analysis program, version 4 (TMAP4) was employed.

Phase composition control of biphasic tritium breeding ceramics and correlation behavior of lithium and tritium

• High purity of the Li2TiO3-xLi4SiO4 ( x = 0.5, 1) biphasic ceramic powders . • A dependence on the phase ratio for the high-temperature stability. • Favourable tritium diffusion coefficient for the biphasic Li2TiO3-xLi4SiO4. • Fast prediction of the tritium release property by the electrical characteristics. Biphasic tritium breeding ceramics were prepared by the microwave-induced solution combustion (MSCS) method in a single step. The phase ratio of lithium titanate and lithium silicate can be effectively controlled by changing the Si/Ti atom ratios in the aqueous precursor. The combustion process is completed within ∽100 s, and the nano-structure Li 2 TiO 3 -Li 4 SiO 4 ceramic power can be obtained directly after the combustion reaction. This preparation approach is promising for the scale production of tritium breeding ceramics with various chemical compositions. Besides, as Li 4 SiO 4 absorbs CO 2, which leads to the formation of Li 2 SiO 3 and Li 2 CO 3 , the powder needs further calcination treatment at 983 – 1023 K. For the ceramic bodies sintered at 1073 K and 1223 K, the Li 2 TiO 3 - x Li 4 SiO 4 ( x = 1) ceramic shows high phase stability and small grain size of ∽200 nm. However, for the case of Li 2 TiO 3 - x Li 4 SiO 4 ( x = 0.5) ceramic, the Li 4 SiO 4 phase is transformed into Li 2 TiSiO 5 and Li 2 SiO 3 with a lithium loss in sintering process. On the other hand, the evaluation of the tritium migration behavior using the electrochemical workstation is performed. The calculation results show that the lithium diffusion coefficient ( D Li ) can reflect the D tritium for the ceramics with various phase compositions. It is found that the D tritium is not proportional to the lithium density in the ceramic. The Li 2 TiO 3 - x Li 4 SiO 4 ( x = 1) ceramic has a favorable tritium migration behavior than mono-Li 4 SiO 4 ceramic.

Quantitative prediction of dynamic HTO migration behavior in the soil and non-negligible evapotranspiration effect

Tritium is mainly produced from nuclear facilities apart from nuclear tests. After being released to the environment, tritium would cause water & food contamination due to its radioactivity and mobility. This study investigated dynamic characteristics of tritiated water (HTO) migration in the soil and evapotranspiration effect based on realistic environmental conditions. The influences of soil types and time-varying environmental factors such as precipitation and evapotranspiration on tritium migration behaviors were specially discussed under normal continuous and accidental short-term release conditions. Radiation dose caused by dynamic tritium evapotranspiration was evaluated. The results show that tritium migration velocity in the soil is much higher than other particles such as cesium due to negligible adsorption of tritium by the soil. Tritium migration in the soil from up to down is attributed to precipitation. On the contrary, evapotranspiration factor would carry tritium movement along the opposite direction. A considerable fraction approximately 55% of tritium deposited in the soil would be reemitted into the air from bare soil and plant leaves due to evapotranspiration effect. Subsequently, the radiation dose caused by second plume due to evapotranspiration effect might be higher than the first plume due to direct release from the nuclear facility under routine discharge.

Preliminary Estimation of Tritium Migration for A-FNS Lithium Target System

In the Advanced Fusion Neutron Source (A-FNS), an accelerator-driven fusion-relevant neutron source that is planned for development in Japan, a few grams (3.5 g at full power operation) of tritium will be generated every year, mainly in the lithium target system. Since the generated tritium would migrate out of the lithium target system, it is necessary to estimate the tritium migration into and out of the lithium target system for the design of detritiation systems for the A-FNS. Therefore, a preliminary estimation is performed in the present study. As a result, it is found that almost all of the generated tritium in the lithium target system would be trapped in the impurity removal system, while less than 0.5% would migrate out. It is also indicated that the amount of tritium that would migrate out of the lithium target system would be able to be processed with the existing techniques so far.

Hydrogen isotope dissolution and release behavior in Y‐doped BaCeO3

AbstractOxides having a perovskite structure are known to exhibit good proton conductivity in a high‐temperature region, which makes them potential candidates in the tritium recovery and purification system of the electrochemical devices of fusion reactors such as hydrogen pumps, hydrogen sensors, and tritium monitoring systems. But the dissolution of hydrogen into the proton conductor and its release behavior is not well known yet. In this study, the dissolution and release behavior of hydrogen was investigated in BaCe0.9Y0.1O3−α proton conductor by exposing deuterium gas (D2) and heavy water vapor (D2O) at 600 or 700°C for 5 h. The dissolved deuterium amount in the sample was measured using a temperature‐programmed desorption analysis (TDS) method. The relationship between the amount of deuterium dissolved in the sample with the exposure pressure was also investigated. At room temperature (RT), the diffusivity of hydrogen (e.g., deuterium) was calculated from the change of the amount of residual hydrogen in the exposed sample due to the change of storage time. The Tritium Migration Analysis Program, Version 4 (TMAP4) simulation code was used to simulate and compare the experimental value of the obtained diffusivity. Although diffusivity at RT has not been reported in the past, the diffusivity is close to the extrapolation value of the literature data. From the comparison of this research results with literature, it may consider that the dominant diffusion mechanism does not change regardless of the temperature range.

Assessment of the tritium distribution in the vegetation cover in the areas of underground nuclear explosions at the Semipalatinsk test site

This paper provides results of assessment of the tritium distribution in the vegetation cover in the areas of underground nuclear explosions at the Semipalatinsk Test Site (STS). The research was conducted at the former “Degelen” test site along the streams that are one of the main channels of tritium migration from underground nuclear explosions epicenters. The dominant plant species Carex supine and Achnatherum splendens that belong to different ecological groups in relation to humidity were selected as representatives of the vegetation cover.The TFWT (tissue free water tritium) and OBT (organically bound tritium) activity concentrations in the vegetation cover were measurement. TFWT activity concentration in the samples of both plant species had high values with an average of up to 30 kBq kg−1. The OBT activity concentration was 1–2 orders of magnitude lower than the TFWT in all plant samples. The TFWT and OBT activity concentrations in vegetation samples are closely correlated (r = 0.75, p < 0.05). No statistically significant difference was found between the content of tritium in the samples of Carex supine and those of Achnatherum splendens taken at the same locations.OBT/HTO ratios for vegetation samples of both species were close to equilibrium ratio used in environmental transfer models. In some cases, OBT/HTO ratios were significantly lower than one, which indicates that simple environmental transfer models may not accurately predict the behavior of HTO and OBT in different environmental compartments.The average OBT/HTO ratio for soil samples (0.6 ± 0.1) close to the equilibrium value indicates the equilibrium condition at the research area.The obtained [OBT]plant/[OBT]soil ratios indicate that soil organic matter accumulates tritium from year to year. However, in some locations with high tritium contamination ratios [OBT]plant/[OBT]soil were more than one due to OBT activity in soils is almost the same as OBT activity in plants.It was found that the nature of the spatial distribution of tritium in the vegetation cover in the areas of underground nuclear explosions is complex, and obviously depends on the location of the tunnels in which nuclear tests were conducted, as well as on the peculiarities of the hydrological regime of underground and surface waters, which are the main channels of tritium migration in the research area. Thus, the vegetation cover reflects the spatial distribution of tritium contamination in the sites of underground nuclear explosions and can be used as an indicator of the radiation situation when monitoring radiation-hazardous areas.

Verification and validation of mHIT code over TMAP for hydrogen isotopes transport studies in fusion-relevant environments

The accurate prediction of tritium inventory and permeation fluxes in the breeding blanket of a D-T fusion reactor is a key aspect for future thermonuclear power plants licensing. Tritium permeation into structural materials could give rise to potential issues concerning the fuel self-sufficiency and can be lost into the environment with resulting radiological risks for the population. In the frame of hydrogen isotopes transport modelling, the Tritium Migration Analysis Program (TMAP) code is considered a referred code for the safety design of nuclear fusion power plants. It is mainly applied to Plasma Facing Components (PFCs), which are subjected to intense particles implantation and surface heat fluxes.Due to some code limitations and to the need of having a more easy-to-use computer code in a multiphysics framework, in the last years the commercial software COMSOL Multiphysics has been used to assess tritium transport studies as an application to ITER, DEMO and CFETR fusion reactors, but it was never validated against TMAP. Within this paper, a COMSOL-based code named mHIT (multi-trapping Hydrogen Isotopes Transport code) is presented. A set of verification and validation (V&V) problems were addressed, with the aim of substantiating the capabilities of the code in common fusion-relevant experimental set-ups and to include different physics according to the level of detail needed for a given application.