Tritium Behavior Research Articles

Tritium transport analysis for tokamak exhaust processing system of tritium plant

Abstract As an important part of tritium plant, tokamak exhaust processing (TEP) system is used for extracting hydrogen isotopes from exhaust gas of tokamak which contains high concentration of tritium and deuterium. A detailed tritium transportation analysis for TEP system can help to obtain more precise and accurate results of tritium behavior in the tritium plant. In this paper, a detailed model of TEP system is established for the particular tritium transportation simulation during its operation. Mean residence time method is mainly used to simulate tritium transport in TEP, and a new calculating method is developed as optimization for Pb-Ag membrane processing. The tritium inventories and tritium losses of components in TEP system are obtained. The influences of reflux ratio of one membrane permeator on the tritium transportation behavior are also analyzed.

Diffusion, permeation and solubility of hydrogen, deuterium and tritium in crystalline tungsten: First principles DFT simulations

Tungsten (W) has shown promise to be the most favored plasma facing material in nuclear fusion with deuterium (D) and tritium (T) as fuel. In order to identify a better material, atomistic understanding of the behavior of D/T with W is highly desirable. Here, we report the mechanistic pathways of diffusion of H, D and T in bcc W employing density functional theory (DFT) using PBE and PBE-D3 functional and harmonic transition state theory. The activation diffusion barrier was determined using nudge elastic band techniques. The zero point energy (ZPE) was incorporated by phonon calculations to include the isotope effects. The surface adsorption of H, D and T is predicted to be exothermic, whereas surface to sub-surface and bulk absorption is endothermic but the vacancy induced absorption in bulk W was predicted to be exothermic. H, D and T atoms were observed to be diffused preferably from tetrahedral site to the nearest tetrahedral site. The calculated diffusion coefficients, rate constants, permeability constants and solubility are found to be higher for H compared to its heavier isotopes D and T. The trends of diffusion, permeation and solubility for H, D and T are well in agreement with the experimental results. The diffusion values predicted for H isotopes in bcc W from PBE-D3 method is quite close to the experimental values compared to PBE method. But, the calculated and experimental ratio of diffusivity, permeability and solubility of any two H isotopes follow the similar trend and no considerable change is noticed between PBE and PBE-D3 methods. Further, the calculated diffusion coefficients are shown to be one order of magnitude lower than Cr and two orders than Fe and thus might be the basis for considering W as plasma facing materials. The present study might be useful in modeling the behavior of tritium in W metal for which data is either scattered or least available.

Expected Environmental Effects of Long-Term Tritium Supply—Lessons Learned

Fusion reactors are considered an almost unlimited source of power for the future, with tritium [deuterium-tritium (D-T) mixture] being the most convenient fuel for these facilities. Due to tritium-specific features and its self-radioactivity there are a number of issues when handling a large amount of tritium for long-term operations of fusion reactors. This paper presents research results of tritium behavior in the D-T mixture and in the environment during operations at tritium facilities. The results provide the basis for an assessment of the future environmental effects of long-term operation of a fusion reactor/tritium facility. The results of experimental and modeling studies include tritium reactions with carbon oxides under radiolysis, radiolysis of DTO, tritium dispersion in atmosphere under rough terrain, tritium oxide washout from atmosphere, tritium oxidation in soil with soil bacteria, tritium oxide migration in the unsaturated soil layer, and tritium uptake by plants, crops, and lichens. Both experimental and observational data have been used to develop models to predict tritium behavior in the environment under similar conditions. The models have been verified and validated.

Physical Chemistry and Technology of Alkaline Liquid-Metal Coolant: A Retrospective-Perspective Look

The results of studies on the physical chemistry, mass transfer, and technology of alkali liquid-metal coolant are reported. The state of the coolant is determined by the interaction coolant-impurities-structural materials- protective gas. The impurity sources and their intensity were determined: sodium and sodium-potassium alloy – oxygen, hydrogen (tritium), carbon, products of corrosion of structural materials, nitrogen, and the protective gas, lithium- nitrogen. The following data were obtained on the impurities in the coolant: form, equilibrium concentration, solubility, reaction kinetics, and mechanisms of heterogeneous and homogeneous mass transfer. It was shown that the required concentration of the impurities in sodium and the sodium-potassium alloy guaranteeing the design-basis parameters and a low rate of corrosion of the structural materials is achieved on purification by means of cold traps. Deeper purification of coolants in high temperature NPF intended for use in space is achieved by means of getters (hot traps). The behavior of tritium and hydrogen in the sodium loops in NPP with fast reactors was studied. A new combined system is proposed for purification from impurities in high-temperature NPF for hydrogen production at sodium temperature ~900°C.

Distribution of tritium concentration in the 0–25 cm surface soil of cultivated and uncultivated soil around the Qinshan nuclear power plant in China

In this study, tritiated water (HTO) and organically bound tritium (OBT) activity concentration at different depth soil layers (0–5 cm, 5–10 cm, 10–15 cm, 15–20 cm, and 20–25 cm) were measured in uncultivated and cultivated soil samples collected in the vicinity of the Qinshan nuclear power plant (QNPP) in July, September, and December 2018. The concentration difference, the spatial and temporal distribution, the seasonal variation, and the OBT/HTO ratios were investigated. The average ratios of HTO concentration between uncultivated and cultivated soil moisture were 1.20 ± 0.24, 1.39 ± 0.46 and 0.95 ± 0.14 in July, September and December, respectively, the corresponding values for OBT were 1.17 ± 0.31, 1.22 ± 0.49 and 1.08 ± 0.28. In generally, the highest HTO concentration in uncultivated soil was found for the topsoil (0–5 cm) in July and September and for the deeper soil layer (20–25 cm) in December, while for cultivated soil, the highest levels were found for the middle layer soil in July, for the topsoil (0–5 cm) in September, and for the deeper layer soil (20–25 cm) in December. Both soils, the vertical profile distribution of OBT concentration showed no consistent tendency, and there were no significant differences in the HTO and OBT concentrations between different soil layers, except for the highest concentration. Whether uncultivated soil or cultivated soil, HTO activity concentrations showed an apparent spatial distribution and seasonal variability, decreasing with the distance to the release sources and with sampling time, while OBT concentrations showed lower spatial and seasonal variability than HTO. In most cases, the OBT/HTO ratios were less than 1, with average values of 1.01 ± 0.48 and 1.06 ± 0.86 for cultivated soil and cultivated soil samples, respectively. The results of this work suggest that farming may affect tritium behavior in soil, while the spatial and temporal distribution of tritium is only slightly impacted.

Exhaust behavior of tritium from the large helical device in the first deuterium plasma experiment

ABSTRACT The tritium exhaust behavior from the Large Helical Device (LHD) was observed in the first deuterium plasma experimental campaign. Tritium in the exhaust gas was monitored at the conducted by use of the ionization chamber and water bubbler system with the discrimination of chemical forms. The observation results indicated that (i) tritium on the surface of the first wall and divertor tiles as plasma facing components was released by the hydrogen isotope exchange reaction of the glow discharge cleaning operation and the diffusion-limited process was suggested in the tritium release behavior from the bulk, (ii) the amount of tritium release from the LHD vacuum vessel was about one-third of the produced tritium and the mostly produced tritium was still retained at the end of plasma experimental campaign, (iii) the ratio of exhausted tritium from the LHD vacuum vessel was larger than that in the case of JT-60U based on the carbon materials as the plasma-facing components. It indicated that the tritium inventory would be reduced and controlled by the kind of plasma-facing materials.

Development of a Monitoring Technique for the Permeation Behavior of Tritium in Pure Nickel to Pure Water

A technique to monitor the permeation behavior of tritium in metals to pure water was successfully developed. A metal membrane separated two containers: one is for tritium loading as an upstream side, and the other is for tritium permeation release as a downstream side. Tritium was loaded by gas absorption at controlled temperatures of 303 K, 323 K, and 373 K and pressures of 4 and 8 kPa at the upstream side. Pure water in the downstream side was automatically and continuously circulated to a solid scintillation counting apparatus by which the tritium concentration in the pure water was directly measured for more than 100 h. When the present technique was applied, almost diffusional permeation behavior of tritium at the nickel-water interface was demonstrated.

Helium blanketing and gas circulation effect of hydriding reaction with ZrCo

The tritium storage bed receives and supplies tritium gas for fusion facilities. Inert gases, such as helium-3, which is produced by tritium beta decay, disturb the absorption behavior of tritium in the bed and this is called the Helium Blanketing Effect (HBE). This study focused on the HBE and gas circulation as a way of breaking the HBE for the hydriding reaction with ZrCo. The hydriding reaction with the application of 100 % hydrogen without helium-4 was completed within five minutes under 120 kPa initial hydrogen partial pressure and 20 °C reactor temperature. With 5% helium-4 in the hydrogen, it took more than 1000 min for the hydriding reaction to be completed under the same conditions. The HBE was broken with the application of a gas-circulation process. Even with 50 % helium-4 in the hydrogen, the hydriding reaction was completed in five minutes under the same conditions. In an experiment conducted at various flow rates of gas circulation, breaking the HBE was observed even at a small flow rate. The gas circulation process can efficiently break the HBE and separate the helium-3 produced in the tritium storage bed.

Tritium Research and Development Status at KAERI

Korea has 26 nuclear power plants (NPPs). Out of these 26 plants, 4 are Canada Deuterium Uranium (CANDU) reactors at the Wolsong nuclear power site. In CANDU reactors, deuterium oxide is used as a moderator/coolant, and tritium is produced whenever a deuterium oxide nucleus captures a neutron. The Wolsong Tritium Removal Facility was designed to remove tritium generated in CANDU reactors. We are introducing tritium environmental protection not only at the Wolsong NPP but also at the High-Flux Advanced Neutron Application Reactor (HANARO) and in high-temperature gas-cooled reactors (HTGRs). We present a tritium behavior analysis code and assess the concentration of tritium in combustible dry active waste. Advanced techniques are introduced to transfer tritium from tritiated water to the gaseous phase. In addition, research on the nuclear fusion tritium storage and delivery system, which is part of the fuel cycle, has been carried out. In this paper, we present the recent progress in the effort to develop tritium systems at the Korea Atomic Energy Research Institute.

Tritium Transport and Distribution in a High-Temperature Gas-Cooled Reactor

The High-Temperature Gas-cooled Reactor (HTGR) has been selected as one of the next-generation nuclear power plants because of its passive safety features. The Korea Atomic Energy Research Institute (KAERI) has been studying how to utilize HTGR efficiently and safely. The HTGR uses graphite as a moderator and helium as a coolant. Once tritium is produced, it is released into the coolant; once released from the core, tritium travels within the primary loop. Because the coolant is gas phase, it is easy to transport to other systems. While it circulates in the primary loop, tritium is involved in processes that include leakage, purification, and permeation. KAERI has been developing a tritium behavior analysis code named TRitium Overall Phenomena analYsis (TROPY) to analyze tritium transport and predict the amount of tritium in the loop in the HTGR core. In this paper, the functions of the TROPY code are introduced, and the amount of tritium in each loop and the amount released into the product hydrogen from the MHTGR 350-MW(thermal) core are explained.

An overlooked pool of hydrogen stored in humic matter revealed by isotopic exchange: implication for radioactive 3H contamination

Tritium (3H), the radioactive isotope of hydrogen, is a contaminant occurring in most nuclear liquid releases from nuclear industries. The behavior of tritium in the environment is an ongoing societal concern, notably the storage of tritium in organic compounds in waters, soils and sediments. In particular, the fate of tritium depends on the availability of hydrogen for isotopic exchanges in natural organic matter, especially in lignin-like molecules. Therefore, we studied here H exchangeability using natural lignite and leonardite humic matter samples. We compared results from gas–solid isotopic exchanges (αiso) with those from widely used methods of functional group characterization by means of deprotonation (αdepro = αcarboxyl +αphenol +αalcohol). Since standard methods for titrating total acidity and determining protonated carboxyl groups are limited by the unconstrained slow deprotonation of humic material, we have modified methods to maximize exchangeable hydrogen reactivity by extending the reaction time in reagent solutions out to 3 weeks, while total OH groups are investigated by the standard acetylation method for alcohol group determination. Results show that more than two times less hydrogen remains available for deprotonation than for isotopic exchanges with an average of αiso = 35.9% versus αdepro = 15.5%. This finding reveals the existence of a neglected pool of hydrogen likely to play a key role in processes based on hydrogen exchange, such as metal complexation by natural ligands. Results also show the reliability of the isotopic exchange method for hydrogen speciation in soil-type matrices.

Behaviour of Tritium and Tritiogenic Helium in Freshwater Lens Groundwater Systems: Insights from Langeoog Island, Germany

Tritium (3H) and its daughter product 3He have been widely used as tracers in hydrological studies, but quantitative analyses of their behaviour in freshwater lenses and the transition zone in coastal aquifers are presently lacking. In this paper, the fate of 3H and 3He in the freshwater lens and the transition zone as well as the saltwater wedge is studied using numerical variable-density flow and transport models of different degrees of complexity. The models are based on the conditions on the German island of Langeoog, which is uniquely suited for this purpose because of the high 3H concentration of the North Sea. It is found that most bomb-related tritiogenic 3He still resides in the freshwater lens, making it a useful tracer for young (<60 years) groundwater. Differences in dispersive transport between 3H and 3He can cause an apparent age bias on the order of 10 years. Under favourable conditions, 3H from seawater can penetrate deep into the offshore part of the aquifer and has potential to be used as a tracer to study saltwater circulation patterns. Our modelling suggests that the field-observed 3H in the transition zone does not originate from seawater but from freshwater affected by the bomb peak. Yet in models with a low (αL=0.5 m) dispersivity, no 3H was sequestered into the transition zone and the transition zone width was underestimated. Better results were obtained with αL=5 m, a value that is higher than in comparable modelling studies, which suggests that further work is needed to better understand the controls (tides, lithological heterogeneity, or transience of recharge and pumping) on transition zone mixing processes.

Study on impact of Cr and Mo on diffusion of H in 2.25Cr1Mo steel using first-principle calculations

Chrome-molybdenum steel (2.25Cr1Mo steel) is one of the main materials of a steam generator (SG) in high-temperature reactor-pebblebed modules (HTR-PM). It is essential to analyze the source term of tritium in this material, because the behavior of tritium in a SG is important for performing source term analysis in normal and accident conditions. In this article, the diffusion behavior of H atom in 2.25Cr1Mo steel was calculated to estimate the diffusivity of tritium using first-principles density functional theory. To develop and simplify the model of hydrogen diffusion in 2.25Cr1Mo steel, the impact of Cr and Mo on the diffusion of hydrogen in bcc-Fe were first calculated, all the possible diffusion paths were considered, and the minimum energy path was obtained. The diffusion activation energy and pre-exponential factor of the diffusion coefficient were obtained from Vienna Ab initio Simulation Package combined with the climbing image-nudged elastic band method. The results indicate that the minimum energy path for the impurity H atom is from one tetrahedral interstitial site to an adjacent tetrahedral interstitial site. The function of the diffusion coefficient of H in 2.25Cr1Mo steel with temperature T can be expressed as D=1.486×10−7×(−16.350kJ/molRT) (m2/s). The diffusion coefficient of our calculation and some of the previous experiments have an excellent quantitative agreement, which indicates the reliability of our crystalline model and the practicability of the present theoretical approach. More importantly, the computational results in this work can be treated as a good screening method to collect reasonable experimental data, which will provide a good reference for tritium source term evaluation in the SG of the HTR-PM.

Development of Field Estimation Technique and Improvement of Environmental Tritium Behavior Model

Development of Field Estimation Technique and Improvement of Environmental Tritium Behavior Model

Transient analysis of tritium transport characteristics in fluoride-salt-cooled high-temperature reactor

Tritium control is a challenging issue for the commercialization of Fluoride-salt-cooled High-temperature Reactor (FHR). In this paper, the tritium behaviors in a FHR system are briefly introduced. A tritium transport characteristics analysis code (the updated TAPAS code) is developed and validated to research the dynamic responses of tritium behaviors during reactor accidents. Two typical accidents for the FHR are simulated, including unprotected reactivity insertion accident (URIA) and unprotected overcooling (UOC) accident. Numerical results show that tritium transport characteristics are strongly coupled under transient conditions; they are quite different from those of normal operation, ranging from tritium production and speciation, adsorption by graphite, chromium corrosion and deposition to tritium permeation through structural materials. The results are discussed in detail in comparison with our previous study. Based on current calculation models, the T2 permeation rate decreases under the URIA and UOC, therefore, extra safety facilities for tritium control under these two transient conditions are not necessary. This study may provide a valuable reference for the tritium control in the FHR.

First Principles Study of Tritium Diffusion in Li₂TiO₃ Crystal with Lithium Vacancy.

Li2TiO3 is one of the most significant breeder materials and has potential applications in future fusion reactors. Defect models with three types of lithium vacancies were considered to study the diffusion behavior of tritium in Li2TiO3 by the density functional theory calculations. The possible tritium adsorption sites inside the lithium vacancy were examined and analyzed. The energy barrier of all diffusion paths between different adsorption sites was calculated and the minimum energy barrier is about 0.45 eV, which indicates that the tritium atom diffuses freely inside the lithium vacancy; when a tritium diffuses across the crystal in the typical three directions, our results reveal that the tritium atom prefers to move along the [010] direction. Furthermore, we found that the minimum energy barrier for the tritium atom to escape the trap of Li vacancy is 0.76 eV. After the tritium jumping out of the Li vacancy, the minimum energy barrier is 0.5 eV for the tritium atom diffusing in the crystal. Therefore, we predict that tritium can easily escape from the trap of the Li vacancy and then diffuse across the crystal. Such results are beneficial to the tritium release process in Li2TiO3 and could provide theoretical guidance for the future applications of the Li2TiO3 materials.

Tritium adsorption in the lithium vacancy of Li2ZrO3: A first principles study

Tritium adsorption in the irradiation defects of the Li2ZrO3 is a fundamental problem to understand the tritium behavior during the release process. A comprehensive computational study of tritium/helium adsorption in the lithium vacancy of bulk Li2ZrO3 is presented by the density functional theory calculations. The most stable tritium adsorption position has been found and it is determined by the neighboring lithium–oxygen interactions. The results reveal that the intrinsic defect is the lithium vacancy with one electron and it transforms to be the neutral state after a tritium atom is adsorbed. Moreover, helium is adsorbed almost in the center of lithium vacancy without bonding with surrounding oxygen atoms, which could diffuse easily in the bulk Li2ZrO3. Therefore, we predict that the intrinsic Li vacancy tends to adsorb a positive ion T[Formula: see text] other than a neutral T atom. Our results provide theoretical support to understand the T behavior in the Li2ZrO3 crystal.

Comparison of release behavior of water vapor and tritiated water vapor from natural soil by heating

In a water-cooled nuclear fusion power plant, a large amount of tritiated water will be handled. Assuming severe accidents including the release of tritiated water to the surrounding environment, it is important to understand the tritium behavior in natural soil. In this study, two samples of natural soil were collected from different places in the Hakozaki campus of Kyushu University and the release behavior of water vapor and that of tritium by heating to 1000°C were separately investigated and compared. Three release peaks of water vapor appeared at 120, 270 and 450°C for both samples. Also, three release peaks of tritium appeared at the approximately same temperature as water vapor release from both samples of soil immersed in tritiated water. The chemical form of released tritium was mainly tritiated water vapor. These results suggest that tritium sorbed in soil during immersion in tritiated water exists with water originally contained in the soil. The amount of tritium released from each sample soil by heating to 1000°C was estimated to be only about 15% of that expected to be sorbed in the soil. Tritium remaining in the soil after heating is presumed to be retained in structural water stably present in clay minerals constituting natural soil.

Thermal Release Behavior of Tritium from Tungsten After Implantation by Glow Discharge

Tritium release behavior in a tungsten sample after exposing to tritium ions with energy about 200 eV created by glow discharge has been studied by both β-ray–induced X-ray spectrometry (BIXS) and imaging plate (IP). The tungsten sample was heated stepwise in a vacuum vessel at temperatures from 400 to 1000 K in experiments, and results obtained from both BIXS and IP measurements showed that the amount of tritium absorbed on the sample surface decreased more than 97% after heating at 800 K. Both intensity and shape of the measured X-ray spectrum have been specified to estimate the change of the tritium depth profile after each heat treatment. Besides, the Monte Carlo Stopping and Range of Ions in Matter (SRIM) code has been introduced to calculate the initial tritium depth profile just after being irradiated by glow discharge. Analysis shows that tritium atoms locate around 3 nm in depth before annealing, and tritium distribution becomes uniform in the near-surface layers (around several nanometers) gradually after heat treatment. At about 800 K, the relative tritium concentration in the near-surface layers reaches its maximum value compared with tritium in the deeper part of the tungsten sample. Then more and more tritium diffuses deeper into the sample as the temperature increases.

Tritium Transport Analysis in a 2-MW Liquid-Fueled Molten Salt Experimental Reactor with the Code TMSR-TTAC

In the Thorium-Based Molten Salt Reactor (TMSR), tritium is produced at a high rate, which results in huge difficulties regarding tritium control. Tritium distributions in a 2-MW liquid-fueled molten salt experimental reactor (TMSR-LF1) were simulated with the TMSR–Tritium Transport Analysis Code (TTAC) (TMSR-TTAC) that was developed for analysis of tritium behaviors in the TMSR. The simulation for normal operation showed that about 60% of the tritium would permeate through the metal walls of the system, 25% of the tritium was removed by the purge gas system, and 15% of the tritium was absorbed on the core graphite. In addition, the effects on tritium distribution of the chemical-redox potential in fuel salt, the tritium permeation behavior through the metal walls, and various tritium removal methods in the TMSR-LF1 have also been simulated. The simulation results based on those conditions are analyzed in this paper to improve the knowledge of tritium behavior in the TMSR-LF1 and to provide reliable methods and strategies for tritium control in the TMSR system.