Tritium Control 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.

A novel solution for hydrogen monitoring in fusion processes: 3D printed BaCe0.6Zr0.3Y0.1O3-α sensors

The control of tritium (3H) in liquid-based tritium breeding blankets (TBBs) is one of the key points to assure their correct performance to prove the 3H sufficiency. For that, electrochemical sensors are a good option for tritium quantification since they can perform in-situ and online measurements.One of the multiple challenges for hydrogen detection is to find suitable materials that can operate at high temperatures and in harsh environments. In this context, perovskite-type ceramics, such as BaCe0.6Zr0.3Y0.1O3-α (BCZY), have elevated proton conductivity and exceptional stability even in reducing atmospheres and at elevated temperatures, which make them strong contenders for high-temperature hydrogen sensing applications.In the present study, we describe the development of a dual-mode hydrogen sensors based on 3D-printed BCZY, which allows a high degree of design flexibility. Crucible geometries were tested and characterized using X-ray diffraction and scanning electron microscopy. Electrochemical sensors were constructed and characterized at 400, 450 and 500 °C in both, amperometric and potentiometric configuration. The results demonstrated the capability of 3D-printed BCZY sensors for hydrogen detection in fusion reactors, offering a breakthrough solution for monitoring fusion processes.

Study on tritium transport characteristics in fluoride-salt-cooled high-temperature advanced reactor

Fluoride-salt-cooled high-temperature reactors (FHRs) are a promising reactor class that combines prominent attributes from the Gen-Ⅳ reactors, raising attractive interests of the international community. However, tritium control has been a major challenge in FHRs because of the neutron reactions of lithium. It's necessary to study the transport characteristics of tritium in the fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) and to provide theoretical support for appropriate tritium control measures and safe operation of the FHRs. In this paper, the inclusive transport characteristics of production, graphite adsorption, diffusion, dissolution, and permeation on tritium in the primary loop are analyzed. The mathematical and physical model of the tritium transport phenomenon is established, and an analysis model of tritium transport characteristics is also developed. Then, the tritium transport characteristics in the FuSTAR primary loop are analyzed. The results show that the tritium in the FuSTAR primary loop mainly exists in the form of Diatomic Tritium (T2), and it's found that Tritium Fluoride (TF) and T2 are adsorbed by graphite after being generated in the core, and this process reaches saturation in about 150 Effective Full Power Days (EFPDs). It also indicates that almost 48% of tritium penetrates the secondary loop through the heat exchanger, and only about 3.8% of tritium is adsorbed on the core graphite. TF makes the corrosion of the heat exchanger and downcomer more serious. Finally, the safety of tritium transport in FuSTAR is conservatively estimated, with a tritium emission dose of 0.1674 mSv/a. This study can provide a theoretical reference for the structural design, tritium recycling, and radiation protection of FHRs.

Analysis of Coolant Purification Strategies for Tritium Control in DEMO Water Primary Coolant

A major objective of the European fusion program is the design of the DEMOnstration power plant named DEMO. Up to now, most fusion experiments have been dedicated to a plasma physics investigation while, in DEMO-oriented activities, large attention is devoted also to other systems necessary to produce tritium and to convert the fusion power to electricity. The blanket region, responsible for tritium breeding, is characterized by high tritium concentrations, high temperature, and large heat transfer metallic surfaces in which tritium can permeate. Therefore, the problem of tritium permeation and the resulting tritium content in the primary coolant are of great relevance for DEMO. For the pre-conceptual design of the Water-Cooled Lead–Lithium variant, the tritium permeation rate from blanket into coolant was assessed and possible mitigation strategies were suggested. Starting from a review of the CANDU tritium experience, a preliminary assessment of the maximum tritium concentration target in the DEMO primary coolant was performed and different strategies (off-line, on-line, and hybrid) for the water coolant purification system coupled with the DEMO operating scenario were analyzed. The intent is to identify suitable solutions to reduce the tritium concentration inside the water coolant, having in mind the complexity of a water detritiation process.

Adsorption of hydrogen isotopes on graphene

We investigated the possibility of using graphene for control of hydrogen isotopes by exploring adsorption, reflection, and penetration of hydrogen isotopes on graphene using molecular dynamics. Reflection is the dominant interaction when hydrogen isotopes have low incident energy. Adsorption rates increase with increasing incident energy until 5 eV is reached. After 5 eV, adsorption rates decrease as incident energy increases. At incident energies greater than 5 eV, adsorption rates increase with the number of graphene layers. At low incident energies (<1 eV), no isotopic effects on interactions are observed since the predominant interaction is derived from the force of π electrons. Between 1 eV and 50 eV, heavier isotopes exhibit higher adsorption rates and lower reflection rates than lighter isotopes, due to the greater momentum of heavier isotopes. Adsorption rates are consistently higher when the incident angle of the impacting atoms is smaller between 0.5 eV and 5 eV. At higher energies (>5 eV), larger incident angles lead to higher reflection and lower penetration rates. At high incident energies (>5 eV), crumpled graphene has higher adsorption and lower penetration rates than wrinkled or unwrinkled graphene. The results obtained in this research study will be used to develop novel nanomaterials that can be employed for tritium control.

A simulation study of tritium removal from molten salt at high temperature with tritium permeation through metallic material

Tritium control was a challenge of the development of the Molten Salt Reactors. Researchers have proposed a method using the metallic permeator to remove tritium directly from molten salt, which was based on the property of tritium easily permeating through the metallic materials at high temperature. A shell and tube type of metallic permeator was simulated and analyzed for tritium removal efficiency and its influencing factors. Results showed the tritium removal efficiency of the permeator would be improved by increasing the proportion of tritium in T2 (or HT) form to total tritium in molten salt, the mass transfer coefficient of T2 (or HT) transferring from molten salt surface to metal surface, and the hydrogen permeability of the permeator. However, these effects were associated and the single factor might only work at certain conditions. And for now, the process of T2 (or HT) transferring from molten salt surface to metal surface is the rate-determining step in tritium removal process of the permeator, which indicated the limitation of tritium removal efficiency of the permeator was due to the low mass transfer coefficient rather than the low permeability of tritium permeating through the permeator metal. Thus, improving the mass transfer coefficient by the structure optimal design of the permeator is an available approach to get a high-efficiency of tritium removal.

Development of tritium dynamic transport analysis tool for tritium breeding blanket system using Modelica

Demonstration of the engineering feasibility of tritium breeding, including tritium generation, tritium extraction, tritium control and tritium safety is one of the main objectives of ITER Testing Blanket Module (TBM) Program. As one of TBM concepts, the tritium transport assessment of the China Helium Cooled Ceramic Breeder TBM and its ancillary systems (called test blanket system, TBS) is absolutely vital to understand and analyze the dynamic tritium behaviors in the system in detail according to the ITER operation scenario. Considering the requirements of the tritium transport analysis for the HCCB TBS, an integrated dynamic tritium transport analysis tool has been developed by the system simulation language “Modelica”. The multiple dynamic physics effects are considered in the tool, including the aspects impacting the tritium transport, simplified thermal-hydraulics considering only the linear loss and local loss in fluid, 1-d thermal conduction in solid, hydrogen isotope transport in fluid, 1-d tritium permeation in solid, tritium solution and recombination on solid surfaces, tritium desorption from ceramic breeder etc. A simplified model of the HCCB TBS is modeled to verify the functions and accuracy of this tool. The results, such as dynamic change of tritium concentration/partial pressure in fluids, dynamic change of tritium inventory in solids, tritium permeation fluxes etc., have been obtained, which are preliminary verified at various aspects by comparing with analytical and hand calculations.

Development of an on-line sensor for hydrogen isotopes monitoring in flowing lithium at DONES

• One challenging aspect on the development of DONES is the purity of the lithium. • An online sensor based on the PAV technique is proposed for H-isotopes monitoring. • The design consists of a capillary tube made of a Pd/Ag membrane spirally rolled. • An experimental facility is designed to demonstrate the feasibility of the monitor. • Theoretical simulations have been carried out using the EcosimPro platform. DONES (Demo Oriented Neutron Source) consists of an irradiation facility based in the nuclear stripping reactions that occur when one energetic deuteron beam impinges on a liquid lithium target flowing along an open channel. Since impurities enhance the erosion/corrosion effect, one of the challenging aspects on the development of the neutron source is the purity of the lithium. One group of impurities to be controlled are hydrogen isotopes in which the control of tritium as part of the produced impurities is mandatory due to its radiological risk. R&D activities are being performed for the development of different kinds of hydrogen sensors since there is no availability of commercial online monitors that could work under the lithium environment. As an extrapolation from other fusion applications, an online sensor based on the permeation against vacuum technique is proposed in this paper. The sensor design consists of a capillary tube made of a Pd/Ag membrane whose inner volume is subjected to vacuum promoting the H-isotopes permeation from the liquid. The vacuum line is connected to a mass spectrometer for the isotope detection. In addition, an experimental facility has been designed to experimentally demonstrate the feasibility of this monitor. It is made of a two-vessels system for lithium melting and H-isotopes detection. Furthermore, theoretical simulations have been carried out using the EcosimPro platform. Results show the expected behaviour of the sensor working under static conditions in the facility.

Deuterium removal from beryllium co-deposits by simulated strike-point sweeping

Abstract Beryllium (Be) co-deposition with hydrogen isotopes will probably be the main mechanism of tritium accumulation in ITER. To reduce the use of baking of the first wall and divertor, raising strike-points during a pure deuterium (D) discharge has been proposed to heat the co-deposits directly and enhance tritium desorption. The feasibility of such plasma-induced removal as a potential means of tritium control has been investigated. Be-D co-deposited layers with a thickness of around 3 μm were created at 428–458 K and exposed to D or H plasma at various temperatures (673–873 K) for various times (180–1200 s). The remaining D content was measured by thermal desorption spectroscopy and compared to the control samples. Plasma-induced desorption was found to be efficient in removing significant amounts of D out of the co-deposits, removing 85% of D from the layer after D plasma exposure at 773 K for 1200 s and 84% of D after exposure at 873 K for only 180 s.

Effects of helium implantation with heavy ion irradiation on deuterium permeation in yttrium oxide coating

In a fusion reactor, strict control of tritium is required from the viewpoints of sufficient fuel balance and radiological hazard. Fabrication of tritium-impermeable ceramic coatings is a promising solution, and ceramic coatings have been investigated for several decades. In recent years, several irradiation tests have been conducted for the ceramic coatings to understand irradiation effects on their characteristics and functions. However, there are few experimental reports on multi-irradiation tests of neutrons and alpha particles produced by nuclear reactions and radioactive decay of tritium. In this study, yttrium oxide coatings were fabricated by reactive and non-reactive radio frequency magnetron sputtering, and their microstructures and deuterium permeation behaviors were investigated after heavy ion irradiation and helium implantation tests simulating irradiation damage by neutrons and generated alpha particles. All irradiated samples showed the formation of an amorphous layer and a nanocrystalline layer or region. In particular, an iron-helium-simultaneous-irradiated sample contained voids including nanometer-size spheres, indicating the formation of helium bubbles. The irradiated samples showed grain growth and/or crystallization at a higher temperature than the unirradiated samples during deuterium permeation tests. That suggests that the voids and the helium bubbles prevented structural change at low temperatures. Moreover, the helium bubbles tended to be more stable than the voids. On the other hand, nickel-single and nickel-helium-sequential-irradiated coatings deposited by non-reactive sputtering showed few voids and helium bubbles, resulting in a different behavior of deuterium permeation.

Diffusion Characterization of Hydrogen Isotopes in Hastelloy N Alloy for the Application of Fluoride-Salt-Cooled High-Temperature Reactors (FHRs)

With the method of gas-driven permeation, a series of permeation experiments was carried out using Hastelloy N alloy membrane in an elevated temperature range of 400°C to 800°C with different hydrogen isotopes. A complete set of permeability, diffusivity, and Sieverts’ constant for hydrogen and deuterium in Hastelloy N alloy was successfully obtained. The isotope effect in the diffusion process was analyzed and compared with references. The ratios of diffusive transport parameters for hydrogen and deuterium were a permeability ratio of ФH/ФD = 1.32exp(0.34kJ/RT), a diffusivity ratio of DH/DD = 1.15exp(−0.41kJ/RT), and a Sieverts’ constant ratio of KS,H/KS,D = 1.16exp(0.21kJ/RT). The result that the permeation flux of deuterium was decreased after introducing hydrogen could be used to suppress the permeation of tritium in future tritium control of the Fluoride-salt-cooled High-temperature Reactor (FHR). Compared with NiO, the Cr2O3 formed in the surface oxidation layer of Hastelloy N alloy showed better hydrogen permeation barrier performance after baking above 700°C in air.

Transient analysis of tritium transport characteristics of thorium molten salt reactor with solid fuel

Tritium management is a vital factor that needs to be considered before a molten salt reactor (MSR) or fluoride-salt-cooled high-temperature reactor (FHR) is constructed. The analysis of tritium transport characteristics under steady and transient conditions is essential in the conceptual design of MSRs and FHRs. In this study, tritium transport characteristics analysis is conducted for a fluoride-salt-cooled pebble bed reactor, i.e., thorium molten salt reactor with solid fuel (TMSR-SF) under accident conditions. The tritium transport characteristics analysis (TAPAS) code is improved with a power model and heat-and-mass transfer model to adapt to TMSR-SF. The transient responses of the tritium transport characteristics of TMSR-SF under the unprotected reactivity insertion accident (URIA) and unprotected overcooling accident (UOC) are determined. It is found that the amount of tritium fluoride (TF) adsorbed on graphite varies, indicating the dynamic equilibrium between the TF adsorption and desorption on graphite. Due to the nuclear density variations of Li-6 and Be-9 in the fluoride salt under the accident conditions, the tritium production rate increases and decreases under the UOC and URIA conditions, respectively. Generally, under these two accident conditions, the tritium permeation rates are lower than that of the normal operation condition. In these cases, it is unnecessary to develop extra safety facilities for tritium control. This study contributes to the tritium management of TMSR-SF.

Evaluation of public dose from FHR tritium release with consideration of meteorological uncertainties

Tritium management is a potentially significant issue in fluoride-salt-cooled high-temperature reactors (FHRs), as these reactors can produce tritium at high rates. Potential impact of the tritium released into the environment needs to be investigated to help determine the maximum-allowable tritium-release rate from an FHR plant. In this study, a dose assessment on the public resulting from FHR tritium release was performed via computational modeling. Three potential locations for FHR construction, i.e., the Hanford site, Idaho Falls in Idaho, and Oak Ridge in Tennessee, were selected. Atmospheric tritium dispersion was modeled using computer code family GENII and a parametric study of key meteorological variables was carried out. An uncertainty analysis was performed to examine the reliability of the prediction of dose for the year 2020. It is discovered that conditions in favor of lower public dose level from FHR tritium release include low atmosphere temperature, high wind speed, high relative humidity, and high tritium release point. It is also discovered that for different geological locations, the dominance of meteorological parameters differs significantly. Among the three locations modeled, although the Hanford site might be the most suitable location for FHR construction in the past, in the near future, Oak Ridge would possess advantages in the dose assessment aspect over the other two. We assumed that the tritium release rate from an FHR plant is given at 18.5 TBq/day and compared the probability of the maximum individual dose exceeding the regulatory limit (0.4 mSv/y). According to the prediction of dose for the year 2020, this probability is extremely low. While for Idaho Falls, it is 91.62% and for the Hanford site, 44.27%. The results indicate that effective measures should be taken for tritium control in FHR.

Fusion Blankets and Fluoride-Salt-Cooled High-Temperature Reactors with Flibe Salt Coolant: Common Challenges, Tritium Control, and Opportunities for Synergistic Development Strategies Between Fission, Fusion, and Solar Salt Technologies

Recent developments in high-magnetic-field fusion systems have created large incentives to develop flibe (Li2BeF4) salt fusion blankets that have four functions: (1) convert the high energy of fusion neutrons into heat for the power system, (2) convert lithium into tritium—the fusion fuel, (3) shield the magnets against radiation, and (4) cool the first wall that separates the plasma from the salt blanket. Flibe is the same coolant proposed for fluoride-salt-cooled high-temperature reactors that use clean flibe coolant and graphite-matrix coated-particle fuel. Flibe is also the coolant proposed for some molten salt reactors (MSRs) where the fuel is dissolved in the coolant. The multiple applications for flibe as a coolant create large incentives for cooperative fusion-fission programs for development of the underlying science, design tools, technology (pumps, instrumentation, salt purification, materials, tritium removal, etc.), and supply chains. Other high-temperature molten salts are being developed for alternative MSR systems and for advanced Gen-III concentrated solar power (CSP) systems. The overlapping characteristics of flibe salt with these other salt systems create significant incentives for cooperative fusion-fission-solar programs in multiple areas. We describe the fission and fusion flibe-cooled systems, what has created this synergism, what is different and the same between fission and fusion in terms of using flibe, and the common challenges. We review (1) the characteristics of flibe salts, (2) the status of the technology, (3) the options for tritium capture and control in the salt, heat exchangers, and secondary heat transfer loops, and (4) the coupling to power cycles with heat storage. The technology overlap between flibe systems and other high-temperature MSR and CSP salt systems is described. This defines where there are opportunities for cooperative programs across fission, fusion, and CSP salt programs.

RIPER: An irradiation facility to test Radiation Induced Permeation and release of deuterium for fusion blanket materials

An efficient reduction of gaseous hydrogen isotope permeation through a metal wall is essential in several applications for the tritium inventory control of fusion energy devices. Tritium is also a safety and biological hazard and tritium losses by permeation through the materials is a critical point to be avoided to ensure safety in fusion energy. Ceramic coatings are candidates to reduce tritium permeation, into the breeder blankets of future devices, due to its low hydrogen permeation rate, excellent corrosion resistance and good mechanical properties. Furthermore, these functional materials of future fusion reactors are expected to be subjected to elevated temperature and high level of radiation fluxes and their impact on the tritium inventory has not been properly assessed mainly due to the lack of data available in the literature on this effect. Radiation induced changes in the composition and microstructure of the materials may alter tritium control in future devices. In this context, a new permeation facility “Radiation Induced Permeation” (RIPER) has been developed at CIEMAT in order to obtain relevant data on permeation of hydrogen isotopes in structural/functional materials under irradiation. This system allows one to differentiate and characterize gas concentration driven permeation, and permeation due to different parameters applied to the sample conditions like temperature, irradiation dose or electric field. The reliability of the system has been checked by means of permeation measurements for different materials in comparison with values found in literature and different measurements during irradiation have been performed. The experimental challenges of these types of measurements are reflected in this paper.

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.

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.

Mass transport analysis for tritium removal in FHRs

Tritium control and mitigation is one of the most significant issues in Fluoride Salt-cooled High-temperature Reactors (FHRs). To address this issue, a cross-flow tritium removal facility has been proposed to remove molecular tritium, T2, from the primary coolant. To model the performance of the design, a method is developed based on the logarithmic mean difference of the square root of the tritium partial pressures on the upstream side and the purging gas side. A computer code has been developed based on this global analysis method to study the effectiveness of tritium removal in such a cross-flow tritium removal facility. The code is benchmarked against experimental data from a hydrogen separation experiment in the literature and compared with the results from a code based on traditional finite volume method. Evaluation of the cross-flow tritium removal facility designed for FHR systems has also been performed. The results show how the key factors, i.e., the fluid flow rates and tritium inlet concentration affect the performance of the facility.

Numerical study of tritium transport characteristics in Thorium Molten Salt Reactor with Solid Fuel (TMSR-SF)

Abstract Tritium control is one of the most key issues for the development of Molten Salt Reactor (MSR). The tritium transport characteristics in 2MWth Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF) proposed by Shanghai INstitute of Applied Physics (SINAP) is studied in this paper. Three works are conducted: Firstly, an integrated numerical model is established, including tritium production, adsorption, diffusion, corrosion and permeation. Secondly, the Tritium trAnsPort chAracteristics analysiS code (TAPAS) for TMSR-SF is developed and benchmarked. The results prove the fidelity and accuracy of TAPAS. Finally, the distributions of the key variables including TF, T2 and Cr2+ in TMSR-SF system at both steady and transient state are obtained. In addition, some methods for tritium mitigation are evaluated by TAPAS. Numerical results show that tritium can be efficiently removed from the molten salt via graphite absorption. The amount of T2 at steady state is two orders of magnitude more than TF. Tritium permeation can be prevented by increasing 7Li enrichment and reducing redox potential. Tungsten or similar low-permeability material is a candidate for tritium barrier when covered on the surface of the heat exchanger wall. This work provides some valuable considerations for the tritium control in the MSR.

A volume-averaged model of nitrogen–hydrogen plasma chemistry to investigate ammonia production in a plasma-surface-interaction device

Nitrogen impurity seeding is a promising technique for increasing the radiative power dissipation rate in the edge plasma of a fusion device. It will be required in future fusion devices such as ITER to reduce the directed heat flux on the divertor strike-points to within erosion limits. However, chemical reactions between nitrogen and fuel isotopes may complicate tritium control measures by increasing in-vessel retention and impacting the gas-handling plant. To gain insight into the nitrogen–hydrogen plasma chemistry a volume-averaged (global) model is developed and compared with experimental measurements in the MAGnetised Plasma Interaction Experiment plasma device. A set of 702 reactions is compiled and used to model the population dynamics of 51 relevant neutral, ionic, electron, surface and metastable excited state species. Stable equilibrium values are compared to results from an experimental investigation in which a combination of mass spectrometry, Langmuir probe analysis and optical emission spectroscopy is used to determine neutral and positive-ionic trends under the same conditions. The dominant ammonia production mechanism is found to be the Langmuir–Hinshelwood reaction between adsorbed atomic hydrogen and above 25% hydrogen concentration. For lower hydrogen proportions the Eley–Rideal reaction between free atomic hydrogen and is found to dominate. The dominant loss mechanism (for all compositions) is found to be electron impact dissociation into neutral fragments.