Tritium Behavior Research Articles

Introduction on tritium transport analysis model for HCCP breeding blanket system

In the context of implementing the He-cooled breeding blanket system for a fusion reactor, several crucial aspects need consideration. Among these, the development of a tritium transport model plays a pivotal role in ensuring safety and effective design. Given that tritium release into the environment from the breeding blanket system poses a radioactive risk, accurate calculation and prediction area essential to prevent potential incidents in a fusion reactor. To address this challenge, a collaborative effort between the Korea Institute of Fusion Energy (KFE) and the University of California, Los Angeles (UCLA) resulted in the creation of THETA-FR (Tritium/Hydrogen Enhanced dynamic Transport Analysis Tool for Fusion Reactor). THETA-FR is an integrated analysis tool that combines Matlab Simulink and COMSOL Multiphysics. Its purpose is to model and predict the dynamic transport phenomena of H isotopes (H/D/T) within the breeding blanket system. Specifically, THETA-FR focuses on transient tritium retention, permeation, and release amounts from the breeding blanket system into the environment. In this paper, we introduce the integrated system and components of THETA-FR, shedding light on various tritium behaviors within the HCCP breeding blanket FW/BU component. By leveraging THETA-FR, researchers and engineers can make informed decisions regarding tritium control and safety in fusion reactors.

Tritium Behavior from Vine to Wine

The hydrogen radioactive isotope, tritium, not only reflects seasonal variation in precipitation but also variations on a local level. To use this radioisotope in authentication and geographical indication procedures, the tritium levels in vines, grapes, and wine were measured in two forms: (1) tissue-free water tritium (TFWT), measured in the free water of the plant or the aqueous phase of the wine; (2) organically bound tritium (OBT), measured in the organic part of the plant or wine (as ethanol). This paper presents the tritium behavior in vines from May to October and in the wine produced from harvested grapes for both tritium forms (TFWT and OBT) in connection with environmental tritium concentrations in air and precipitations during the 2019–2023 vegetation periods near the Râmnicu Vâlcea city. If the tritium of tissue-free water was influenced by precipitation and air humidity during the harvest period, the organically bound tritium recorded a maximum level with a delay of 1–2 months compared with tritium’s seasonal variation in precipitation and air. The tritium values of early mean in precipitation and ethanol from wine were approximately an average value of 11 TU for all 5 years of observations.

Investigating advection–dispersion behavior for simulation of HTO and 238Pu transport in argillaceous shale with different varying degrees of weathering

Abstract Exploring the migration behavior of radioactive nuclides in the surrounding rock media of the disposal site is the basis for the safety evaluation of radioactive waste disposal sites. In this study, the column experiment was employed to evaluate the nuclide migration behavior in the surrounding rock medium of a near-surface disposal site in China and to investigate the advection–dispersion behavior of tritium (HTO) and plutonium-238 (238Pu) in highly weathered argillaceous shale with different degree of weathering. A reasonable numerical model was selected to fit the experimental breakthrough curves (BTCs) and obtain relevant migration parameters. The results showed that: (1) the breakthrough curves of HTO and 238Pu in fractured media with varying degrees of weathering exhibited clear “peak forward” and “tail dragging” phenomena; (2) the stream tube model can better fit the BTCs of the nuclides in the highly weathered fractures and obtain the average dispersion coefficient <D>, average distribution coefficient <K d>, and other parameters; (3) the two-region non-equilibrium model can better fit the BTCs of the nuclides in the penetrating fractures and weak weathering fractures and obtain the volumetric water content of the mobile liquid region (θ m), distribution coefficient (K d), and other parameters.

Tritium behavior in soil and mineral rock components used for plant cultivation

Immersion, percolation and tritium release experiments in peat and vermiculite soil samples were performed to analyze their behavior in this widely used medium for plant cultivation. Samples were immersed in tritiated water for 696 h and the isotope exchange capacity evaluated. A vertical flow regime was also considered with analysis for hydraulic conductivity to understand tritium mobility and therefore its availability. Peat soil showed a high tritium retention after percolation, but vermiculite seem to suppress its retention ability. The high moisture and organic content of peat enhanced its isotope exchange capacity. The falling head method was used to numerically evaluate the saturated hydraulic conductivity and outflow flux. Calculated isotope exchange capacity was 4.95×10-2 mol-T2O/g for peat and 3.38×10-2 mol-T2O/g for vermiculite. The tritium release experiment showed significant release of tritiated carbons in peat.

Interlaboratory comparison exercises for organically-bound tritium in the development of reference materials of environmental samples

The complex behavior of tritium and the probability of increasing tritium concentrations released in the environment were the promotors for the research and development of laboratory methods that enable to accurately determine the various forms of tritium including organically-bound tritium (OBT) for public and regulatory assurance. The measurement of tritium is a key step for dose and risk assessment. The Cernavoda Nuclear Power Plant (NPP) in Romania improved preparation methods and tested environmental matrices for OBT analysis through intercomparison exercises. This paper describes the international Organically-Bound Tritium (OBT) intercomparison exercise, organized by the Cernavoda Nuclear Power Plant (NPP) in 2019–2020, using fruit sample (quince) from Cernavoda town. Evaluation of the results from the participating laboratories was performed using both robust analysis (Algorithm A) method described in the ISO 13528:2015 standard and ANOVA method. The results obtained are encouraging as an increased number of participating laboratories did not change the observed dispersion of the results for activity concentration level around 50 Bq/L of combustion water. The stability of the remaining sample will be checked in time to investigate its use as a reference material for OBT analysis at the environmental levels.

Study on tritium permeation behavior in primary and second circuits of the hydrogen production system by methane steam reforming using HTGR

Hydrogen production by methane steam reforming (MSR) using the process heat from High-temperature gas-cooled reactor (HTGR) is one of the most promising technology for hydrogen production using nuclear energy in the near future. The permeation behavior of tritium produced by the reactor is extremely important when analyzing the safety of this hydrogen production system. There are two methods of connection between a nuclear reactor and a hydrogen production system. One is direct coupling that the steam reformer of the hydrogen production system is directly connected to the first circuit of the nuclear reactor. The other is indirect coupling via an intermediate heat exchanger (IHX) to connect nuclear reactor and hydrogen production system. Hence, this work proposed a model to analyze the difference in tritium permeation between these two coupling ways and to compare the tritium permeation behavior for different IHX alloys at various operating conditions. This model includes tritium's production rate, release fraction, and decay reduction in the system. The results show that the pre-exponential factor of the IHX alloy mainly limits the tritium permeation rate. In addition, the tritium permeation rate increases about two times when the temperature rises from 750 °C to 950 °C at a steady state. Further, the decrease of tritium permeation rate is seen from the calculation results for the system with IHX, e.g., for T = 750 °C, the mean permeation rate of tritium into the product gas is 9.31 × 10−12 mol/s at the whole year for indirect coupling with an IHX, compared to 3.86 × 10−12 mol/s for direct coupling, which reduced 58.54 % (This means that with IHX the permeation rate is lower than the system without IHX). These results will contribute to designing an effective and safe hydrogen production system by MSR using HTGR process heat.

The influence of tritium behaviour on spent fuel pool concrete

This paper investigates the effects caused by Tritium released from spent nuclear fuel (SNF) stored into a spent nuclear pool made of (reinforced) concrete. The release of tritium from fuel elements to coolant during normal LWRs operation is of 1 % while during 6–20 years of wet storage is assessed to 13–45 %. Permanent Tritium concentration in coolant is assessed to n·109 Bq·dm−3.The water contamination first and the filtration of the concrete pool walls then pose important safety problems to face from both radiological and structural point of view: during Tritium filtration, radiation affects significantly the structural and mechanical properties of concrete due to the breaking of atomic bonds.Under the effect of radiation concrete microstructure modifies; collision determines atom dislocation and so a lattice defect. Ionized rays may cause the decay of free or bonded water leading to the formation of H2 and O2 and produce explosive mixture.The damage from “soft” β-decay of Tritium inside the concrete could be harder due to short track of the particle and the resulting “volume” effect. Neither this effect nor to what extent Tritium movement in/through concrete material would take place over long-term are yet completely studied.Due to Helium (He) formation, when Tritium decays, the restructuring of water molecules with generation of reactive free radicals occurs. We have assessed these effects: the irradiation and accompanying volume expansion, formation of free radicals and corresponded chemical corrosion are not considerably influencing on the structure and sustainability of concrete biological shield. The most significant effect is the radiation breaking of van der Waals bonds that within 10 years can lead to the destruction of about 2 % of concrete.Concrete shield does not provide a complete barrier against Tritium release into the environment. The main threat of Tritium diffusion to coolant in SNF storage pools is caused by filtration of tritiated water through the concrete walls that may provide significant health hazards for employees of the nuclear industry. Results show that β-decay of tritiated water adsorbed in 1 m3 of concrete may destroy about of 0.05 mol (11.4 g) of cement stone within 10 years, which is less than 0.001%. Moreover, with 3. With reference to the 5th power unit of the Zaporizhzhia NPP, the volume of Tritiated water released from the pool is 0.67 m3 annually. Resultsshow that the tritium waste management concept needs to be revised.

Tritium Content in Komatsuna Cultivated in Tritium-Contaminated Peat Soil

Tritiated water from fusion power reactors will be the next major issue when fusion technology comes fully onstream. Effective radiation protection measures will be implemented when the scope of its behavior is well understood. To understand tritium behavior in the environment, komatsuna was cultivated in tritium-contaminated peat soil. It was indicated experimentally from water immersion experiments that the amount of tissue free water tritium in komatsuna depends on the tritium concentration in the soil and that the concentration in stems and leaves in komatsuna decreases as the tritium concentration in the soil decreases. The amounts of tritium retained in the roots were much less than that in the stems and leaves.

Structure and Kinetic Properties of a Molten FLiBe Mixture with Tritium

A study is performed of the self-diffusion of tritium and fluorine atoms, and the change in the structure of molten FLiBe upon raising the temperature of the system from 873 to 1073 K. The interaction between neutrons and both lithium and beryllium in molten-salt reactors (MSR) using FLiBe as a fuel salt results in the formation of large amounts of tritium. Tritium, which easily penetrates metallic structural materials at high temperatures, is a radionuclide hazard. Predictive models for the behavior of tritium in a molten fluoride salt must therefore be developed to solve the problem of MSR safety. The emergence of tritium in the system increases the average energy of interatomic bonds upon raising the temperature and concentration of tritium in the system. A rise in temperature is also accompanied by fluorine atoms creating a closer short-range order in the environment of tritium atoms. This is expressed in the formation of a high first peak of radial distribution function gT-F(r), an increase in the number of probable geometric neighbors, which is shown by Voronoi polyhedra, and fluorine atoms giving priority to fourth-order rotational symmetry in the environment of tritium atoms.

Ab initio study of hydrogen in titanium beryllides

Titanium beryllide Be12Ti is a candidate material for the neutron multiplier for the demonstration fusion reactor DEMO. Experimental studies show that under certain conditions, Be12Ti may contain inclusions of other phases such as Be2Ti, Be17Ti2. In this regard, it is important to study the behavior of tritium and its isotopes in the crystal lattices of these phases. All calculations are performed using ab initio methods. Solution energies of a hydrogen atom in all non-equivalent interstitial sites of the three studied titanium beryllides were found to be lower than that in pure beryllium. The formation energy of all types of vacancies in all studied beryllides is found to be higher than that in beryllium. The binding energies of a single hydrogen atom located both inside and outside the vacancies are calculated. Hydrogen inside monovacancy is more strongly bound as compared to that outside this vacancy. It turned out that in some cases hydrogen can be captured by vacancy being outside of it. The results obtained are essential for further study of interstitial diffusion of hydrogen and analysis of tritium retention in titanium beryllides.

Tritium Behavior in Water and Gas Produced by a Fully Tritium-Compatible Electrolyzer

The electrolysis process is essential in the water detritiation subsystem using the combined electrolytic catalytic exchange process. A special experimental program was designed to characterize a modified HOGEN H Series industrial electrolyzer. The tritium amount transferred to hydrogen gas, the water enrichment factor, and the number of hours necessary to attain a steady-state regime were parameters of primary interest in the experiments. To minimize the necessary time for a steady-state regime, the holdup of the water electrolyzer was chosen as the minimum value allowed for safe and constant parameter operation in all experiments. The stationary regime was attained after 120 h, with an enrichment factor near 5, and an amount of 18% to 19% of tritium transferred from tritium-enriched water to hydrogen gas. These parameters were obtained in all three experiments, and the modeling software of isotope separation by electrolysis confirmed the results.

Numerical simulation of tritium behavior under a postulated accident condition for CFETR TEP system

Numerical simulation of tritium behavior under a postulated accident condition for CFETR TEP system

Review of Recent Progress in Plasma-Facing Material Joints and Composites in the FRONTIER U.S.-Japan Collaboration

The plasma-facing components (PFCs) of future fusion reactors will have intricate structures and require multiple materials because no one material can simultaneously satisfy all the requirements of the component. The dissimilar material joints in PFCs must withstand extreme thermal and stress gradients under neutron irradiation. The Fusion Research Oriented to Neutron Irradiation and Tritium Behavior at Material Interfaces (FRONTIER) U.S.-Japan collaboration seeks to explore and explain the behavior of internal solid interfaces in PFCs under neutron irradiation. The first step of the collaboration was to identify the leading PFCs that should be studied further and prepare them for the next step, which will include neutron irradiation. Different strategies for material development are being pursued worldwide to produce robust PFCs. Here, an overview is presented of some of the most promising materials in the areas of copper alloys, tungsten-copper composites, tungsten-steel composites, additively manufactured tungsten, particle-reinforced tungsten, and tungsten and SiC fiber composites. Each material’s fabrication and benefits are described, and some discussion of remaining questions is given.

Advection–Dispersion Behavior for Simulation of H-3 and Pu-238 Transport in Undisturbed Argillaceous Shale of a Near-Surface Repository

In this study, a column experiment was employed to evaluate the nuclide migration behavior in the surrounding rock medium of a near-surface disposal site in China and to investigate the advection-dispersion behavior of tritium (H-3) and plutonium-238 (Pu-238) in highly weathered argillaceous shale. A reasonable numerical model was selected to fit the experimental breakthrough curves (BTCs) and to obtain the relevant migration parameters. The results show the following: (1) the internal structure of the highly weathered argillaceous shale exhibited heterogeneity, and the nuclide migration BTC showed characteristics of a "curve peak moving forward" and a "tail curve trailing"; (2) compared with other models, the stream tube mode could better fit the BTCs and obtain the average dispersion coefficient <D>, average distribution coefficient <Kd>, and other parameters; (3) compared to the results of the batch experiment, the distribution coefficient Kd obtained from the column experiment was smaller than that obtained from the batch experiment, which is speculated to be due to the influence of contact time and the contact area between the nuclide and the medium.

Simulation calculation of the influence of interstitial atoms on the desorption behavior of tritium in nuclear graphite

In order to explore the effect of interstitial atoms on tritium desorption behavior of nuclear graphite, the adsorption and desorption behaviors of hydrogen atoms on interstitial atoms were simulated by first-principles. The results indicated that the interstitial atoms can gradually adsorb 3 hydrogen atoms until the sp3 hybrid structure is formed. The CH3 structure with dangling bonds tends to desorb as ionic CH3, which will combine with free H or other CH3 to form CH4 or C2H6. The CH (or CH2) structure tends to polymerize with other CH (or CH2) structures to form C2H2 (or C2H4), which reaction energy barrier is slightly lower than that of the CH3 structure; Although the migration energy of the interstitial atoms with H adsorbed on the graphite flakes is not much different from that of the interstitial atoms without H, the energy barrier of their recombination with single vacancies is significantly greater than that of the interstitial atoms without H. Therefore, it can predict that the large number of interstitial atoms in decommissioned nuclear graphite will become important sites for chemisorption of tritium. In addition, the reaction energy barrier of the interstitial atoms possessed large adsorption energy with tritium gas desorption is significantly higher than that with carbon containing tritium products (including C2T2, C2T4, CT4 or C2T6). This work will provide technical guidance and support for the heat treatment of the adsorbed tritium in decommissioned nuclear graphite.

Numerical Simulation of Tritium Diffusion Behavior in FeCrAl Cladding

In this article, a simulation code for tritium diffusion behavior analysis in FeCrAl cladding is developed based on the COMSOL platform. The simulated results are in good agreement with experimental and theoretical results. The effects of different concentrations of tritium and temperature distributions on the diffusion behavior of tritium in FeCrAl cladding were further investigated. Finally, the optimal effect of different coating schemes on the tritium resistance of the FeCrAl cladding was investigated. The results show that higher temperatures lead to higher cladding diffusion coefficients, which will further lead to higher fluxes of tritium into and out of the cladding, which is found to further result in a higher tritium flux into and out of the cladding, as well as shorter tritium diffusion times. It is found that higher temperature will lead to more tritium flux into and out of the cladding and shorter time for the tritium flux to reach a steady state on the right side of the cladding. At the same time, the higher tritium partial pressure on the fuel side of the cladding will lead to a longer time for the tritium flux to reach a steady state on the water side of the cladding. The longer time to reach the steady state on the water side of the cladding increases the tritium flux into the cladding.

Levels and behavior of environmental tritium in East Asia

For a more systematic understanding of the levels of environmental tritium and its behavior in East Asia, a database on environmental tritium was established based on the literature published in the past 30 years. Subsequently, the levels and behavior of the environmental tritium were further studied by statistical analyses. The results indicate that the distribution of environmental tritium is inhomogeneous and complex. In areas without nuclear facilities, the level of environmental tritium has decreased to its background level, even though a certain number of atmospheric nuclear tests were performed before 1980. In general, the level of atmospheric tritium was marginally higher than the levels in precipitation and surface water; the levels in shallow groundwater and seawater were considerably lower. Furthermore, the levels of tritium in the atmosphere, precipitation, and inland surface water were strongly correlated with latitude and distance from the coastline. In soil and living organisms, the level of tissue-free water tritium (TFWT) was comparable to the tritium levels in local rainfall, whereas the persistence of organically bound tritium (OBT) in the majority of organisms resulted in an OBT/TFWT ratio greater than one. Conversely, extremely high levels of environmental tritium were observed near certain nuclear power plants and the Fukushima accident sites. These results highlight the requirement to know the tritium baseline level and its behavior in the environment beforehand to better assess the impact of tritium discharge. Further investigations of environmental tritium in East Asia using more efficient and adequate monitoring methods are also required.

Permeation behavior of gaseous tritium through the assembly composed of Zr and Al2O3 simulating Li rod

• An assembly simulating the Li rod structure for T production in HTGR was used. • The assembly was composed of inner Zr tube - outer Zr tube - Al 2 O 3 tube. • T permeation rate through the assembly from HT was much faster than that from HTO. • The rate-determining step was T permeation through the Al 2 O 3 tube. For starting up a fusion reactor stably, the tritium production by the nuclear reaction of neutrons and lithium in a high-temperature gas-cooled reactor has been studied. In order to suppress the migration of tritium to the whole reactor, it is important to design a Li rod with tritium confinement function and understand tritium behavior. A Li rod structure consisting of Zr - LiAlO 2 - Zr - Al 2 O 3 was proposed, and an assembly simulating proposed Li rod was fabricated. In the previous work using tritium gas containing mainly HTO rather than HT, the present authors succeeded in confining tritium in the assembly at 700 ℃ for 87 h. In this work, tritium gas containing mainly HT rather than HTO was supplied to the assembly, tritium permeation behavior was observed. Unlike the case of HTO, tritium permeated immediately after the introduction of HT. This result indicates that the chemical form of tritium significantly affects tritium confinement performance of the Li rod.

A numerical study on the effect of dust particles on tritium deposition on plasma-facing materials

Tritium deposition is a key issue for fusion devices. The existence of dust leads to increase of tritium deposition rates due to the large specific surface area (SSA) of dust grains. Experiments in JET-ILW have indicated that the deposition and retention behavior of hydrogen on dust grains is complicated. In present work, the deposition behavior of tritium on the dust is studied by a hybrid particle simulation method. Our results indicate that the large SSA of dust is not the only mechanism that increases tritium deposition. Dust size and the electrostatic sheath induced by the dust are also important factors that influence the tritium deposition. The retention rate enhanced by dust grains is also predicted based on our study.

Study on emergency ventilation optimization method for tritium leakage accident of high-pressure storage vessel

In the process of tritium storage, the high-pressure storage vessel may be damaged due to accidental conditions, such as impact or drop, which results in tritium leakage. The tritium gas will be released and diffused into the space immediately, which inevitably causes radioactive hazards to personnel and the environment. Ventilation is considered to be the best emergency treatment method for a tritium leakage accident. According to the characteristics of fixed tritium amount and rapid release of tritium in the tritium leakage accident of the high-pressure storage vessel, the numerical simulation of high-pressure released tritium behavior in a confined space is carried out based on Fluent and compared with CATS test data. The error is no more than 5%. The ventilation conditions are optimized according to different ventilation positions and different ventilation rates. It is founded that the tritium removal efficiency when the vent is set on the roof is 30% higher than that on the side wall. The tritium concentration in a confined space can be reduced to the background level after ventilation for 30 min at a rate of 0.5 m3 s−1. The optimal recovery time of the environment is given, which provides technical support for the emergency treatment of the tritium leakage accident of the high-pressure storage vessel.