Deuterium Tritium Research Articles

Novel high temperature tritium blanket designs for confined spaces in spherical tokamak fusion reactors

Tritium self-sufficiency is one of the fundamental challenges for commercially viable deuterium–tritium nuclear fusion power stations. The combination of key high temperature radiation shielding materials that possess dense, high neutron absorption cross-section, and moderation properties, and tritium breeding materials could involve interesting design spaces for the central column challenge in spherical tokamaks. Potential tungsten alloys can be used for two functions: radiation shielding and structural material, providing a new design space window for spherical tokamak central column breeding space. In this paper, we present two novel high temperature concepts for the inboard side of the breeder blanket in a confined space, such as a spherical tokamak. A tungsten–rhenium–hafnium-carbide lithium-based design was found to offer the best TBR given a parameter optimisation based on shielding and thermal requirements. A silicon-carbide lead-lithium breeder design was also investigated. The highest TBR was found to be 0.135 in a 3D neutronics calculation with a W-24.5Re-2HfC (structural and shielding, wt%), Li (90% Li-6 enriched breeder), and tungsten pentaboride (W2B5) (shielding) option. Although this TBR is lower than unity, it will contribute to the reactor’s global TBR.

The fuel supply quandary of fusion power reactors

ABSTRACT Concern about adequate tritium resources to start up deuterium-tritium (DT) fusion reactors in the near term is misplaced: Proposed demonstration reactors and pilot plants will not realistically be buildable for a great many decades. More disturbing, tritium fuel self-sufficiency of eventual DT reactors may be unattainable, because it is unlikely that tritium-breeding blankets can completely replenish burned-up fuel as well as unburned tritium that is lost in numerous reactor subsystems. The concept of doubling time is unworkable. The only viable fuel solution for sustained operation demands deuterium-based reactors where the tritium is concurrently generated and burned up in the fusioning plasma.

Study of the Pure Deuterium Fuel Burning Rate in Z-Pinch Devices with Magneto-Inertial Confinement

The burning rate of pure deuterium (D-D) fuel in Z-pinch devices with magneto-inertial confinement was studied in this paper. The system of particle and energy balance equations for D-D fuel burning with a mixed D-T-3He fusion cycle (D-D, D-T, and D-3He reactions) was solved numerically, taking into account the densities of all reacted and produced ions (protons, deuterium, tritium, helium-3, and alpha-particles). The obtained results indicate that effective D-D fusion in Z-pinch devices can be successfully achieved under conditions of a hot, dense plasma with an initial temperature of 31 keV or higher. The initial ion density of deuterium and electron density were equal due to quasi-neutrality condition of the plasma, with both reaching 1024 m−3. Although the obtained results show that the burning rate of D-D fuel is approximately 2.3 times slower and its power density notably lower than that of D-T fuel, pure deuterium plasma can be considered as a promising alternative to well-studied deuterium–tritium plasma, with potential future applications in magneto-inertial fusion (MIF) facilities.

The consequences of tritium mix for simulated ion cyclotron emission spectra from deuterium-tritium plasmas

Abstract Measurements of ion cyclotron emission (ICE) are obtained from most large magnetically confined fusion (MCF) plasma experiments, and may be used in future to quantify properties of the fusion-born alpha-particle population in deuterium-tritium (DT) plasmas in ITER. ICE is driven by spatially localised, strongly non-Maxwellian, minority energetic ion populations which relax collectively under the magnetoacoustic cyclotron instability (MCI). ICE spectral peaks are typically observed at, near, or separated by, integer harmonics of the energetic ion cyclotron frequency. Here, for the first time, we study how simulated ICE spectra from DT plasmas vary with tritium concentration and compare with an observed JET DT ICE spectrum. We incorporate a population of thermal tritons, in addition to thermal deuterons and an energetic minority population of fusion-born alpha-particles, in simulations with the kinetic particle-in-cell (PIC) code EPOCH. This code has previously been used extensively for interpretation of ICE observations in terms of the self-consistent gyro-resolved collective Maxwell-Lorentz dynamics of tens of millions of simulation particles. Our simulation parameters are relevant to the ICE-generative outer midplane edge region of JET DT plasma 26148, which included 11% tritium. Quantifying the variation of simulated ICE power spectra with tritium concentration reveals that our simulation with 11% tritium concentration most accurately represents the observed ICE spectrum from this plasma. This outcome is encouraging for the diagnostic application of ICE to fusion plasmas containing two thermal ion species, including future DT plasmas.

Simultaneous enhancement of tritium burn efficiency and fusion power with low-tritium spin-polarized fuel

Abstract This study demonstrates that using spin-polarized deuterium-tritium (D-T) fuel with more deuterium than tritium can increase tritium burn efficiency (TBE) by at least an order of magnitude without compromising fusion power output, compared to unpolarized fuel. Although previous studies show that a low tritium fraction can enhance TBE, this strategy resulted in reduced fusion power density. The surprising improvement in TBE at fixed power reported here is due to the TBE increasing nonlinearly with decreasing tritium fraction but the fusion power density increasing roughly linearly with D-T cross section. A study is performed for an ARC-like tokamak producing 481 MW of fusion power with unpolarized 53:47 D-T fuel, finding the minimum startup tritium inventory ($I_{\mathrm{startup,min}}$) is 0.69 kg. By spin-polarizing half of the fuel and using a 60:40 D-T mix, $I_{\mathrm{startup,min}}$ is reduced to 0.08 kg, and fully spin-polarizing the fuel with a 63:37 D-T mix further reduces $I_{\mathrm{startup,min}}$ to 0.03 kg. Some ARC-like scenarios {are predicted to} achieve plasma ignition with relatively modest spin polarization. These findings indicate that, with advancements in helium divertor pumping efficiency, TBE values of approximately 10-40\% could be achieved using low-tritium-fraction and spin-polarized fuel with minimal power loss. This would dramatically lower tritium startup inventory requirements and reduce the amount of on-site tritium. More generally than just for spin-polarized fuels, {increased} plasma performance can be used to increase TBE. This strongly motivates the development of spin-polarized fuels and low-tritium-fraction operation for burning plasmas.

Neutron imaging of the deuterium-tritium tamping gas volume in an inertial confinement fusion hohlraum.

To benchmark the accuracy of the models and improve the predictive capability of future experiments, the National Ignition Facility requires measurements of the physical conditions inside inertial confinement fusion hohlraums. The ion temperature and bulk motion velocity of the gas-filled regions of the hohlraum can be obtained by replacing the helium tamping gas in the hohlraum with deuterium-tritium (DT) gas and measuring the Doppler broadening and Doppler shift of the neutron spectrum produced by nuclear reactions in the hohlraum. To understand the spatial distribution of the neutron production inside the hohlraum, we have developed a new penumbral neutron imager with a 12mm diameter field of view using a simple tungsten alloy spindle. We performed the first experiment using this imager on a DT gas-filled hohlraum and successfully obtained the spatial distribution of neutron production in the hohlraum plasma. We will report on the design of the spindle, characterization of the detectors, and methodology of the image reconstruction.

Conceptual design study of the tritium breeding unit for verifying the long-term performance of the breeding blanket

Fusion reactors produce high-energy neutrons through the Deuterium-Tritium (DT) reaction. The long-term performance verification and structural integrity assessment of the breeding blanket under continuous plasma operation are essential for the development of Demonstration fusion power reactor (DEMO) breeding blankets. The Korea Institute of Fusion Energy (KFE) has been planning the construction of a test facility called the Integrated Breeding Test Facility (IBTF) that uses a linear accelerator as the neutron source. A solid beryllium target has been considered to generate fusion-like neutrons. A conceptual design of the Tritium Breeding Unit (TBU), which is a key component of the breeding blanket, has been carried out. This study evaluated tritium production under fusion-like conditions using factors such as the position and thickness of the graphite block, the thickness of the first wall, and the radial length of the breeding zone. The results showed that increasing graphite block thickness and optimizing the dimensions of the TBU enhanced tritium production. However, these changes affected the structural integrity of the TBU. The side wall of the TBU experienced membrane plus bending stress, which exceeded the Level C and D acceptance criteria under the assumption of In-Box LOCA. Measures were investigated to ensure structural integrity while improving tritium production of the TBU.

Three-dimensional reconstruction of laser-direct-drive inertial confinement fusion hot-spot plasma from x-ray diagnostics on the OMEGA laser facility (invited).

A deep-learning convolutional neural network (CNN) is used to infer, from x-ray images along multiple lines of sight, the low-mode shape of the hot-spot emission of deuterium-tritium (DT) laser-direct-drive cryogenic implosions on OMEGA. The motivation of this approach is to develop a physics-informed 3-D reconstruction technique that can be performed within minutes to facilitate the use of the results to inform changes to the initial target and laser conditions for the subsequent implosion. The CNN is trained on a 3D radiation-hydrodynamic simulation database to relate 2D x-ray images to 3D emissivity at stagnation. The CNN accounts for the lack of an absolute spatial reference and the different bands of photon energies in the x-ray images. While previous work [O. M. Mannion etal., Phys. Plasmas 28, 042701 (2021) and A. Lees etal., Phys. Rev. Lett. 127, 105001 (2021)] studied the effect of mode-1 asymmetries on implosion performance using nuclear diagnostics, this work focuses on the effect of mode 2 inferred from x-ray diagnostics on implosion performance. A current analysis of 19 DT cryogenic implosions indicates there is an upper limit of ∼20% reduction in the neutron yield caused by an ℓ = 2 amplitude for ℓ2/ℓ0 ≤ 0.32. These conclusions are supported by 2D simulations.

Impurity behaviour in JET high-current baseline scenario for Deuterium, Tritium and Deuterium-Tritium plasmas

To support future ITER operation, experimental campaigns at the Joint European Torus (JET) with an ITER-like wall (tungsten divertor and beryllium main chamber) in pure deuterium (D), tritium (T) and Deuterium-Tritium (D-T) were performed. One of the most important challenges in recent years was the development of two main scenarios that investigated different approaches to achieve the high fusion power as well as good plasma confinement (Garzotti et al., 2023). The first one, so-called baseline scenario is relying on high plasma current (Ip≈3.5 MA), normalized beta βN < 2 and safety factor q95 ≈ 3 (Garzotti et al., 2023). On the other hand, the second one, so-called Hybrid scenario is operating at lower plasma current (flat-top Ip ≤ 2.6 MA) and density with respect to the baseline, higher normalized beta βN > 2 and safety factor q95 ≈ 4.8 (Hobirk et al., 2023).In this paper we focus on the impurity behaviour analysis for the baseline discharges at Ip = 3.5 MA and BT = 3.3 T with D, T and DT plasmas, in which the gas and power waveform were optimized to achieve the best possible performance. In particular, we study the impact of total heating power (Ptot + Palpha), flat-top gas flow and ELM (edge localized modes) frequency on mid-Z (Nickel (Ni), Copper (Cu)) and high-Z (Tungsten (W)) impurities. In addition, we compared the two best performing pulses of the baseline scenario (Ip = 3.5MA, BT = 3.3 T and Pin ≈ 35 MW) in D and DT in order to identify the causes responsible for the increase in radiation during the DT pulse, which led to an early plasma termination. All presented results rely on the data collected by the VUV as well as the bolometry system. Detailed analysis indicates that in the baseline scenario, higher radiation, which is most likely due to the tungsten (W), is observed for T and DT plasmas in comparison to D. Moreover, for the two best performing baseline pulses, tomographic reconstructions show that the radiated power density is mainly emitted from the low field side (LFS) of the plasma and W does not accumulate in the plasma center (Telesca et al., 2024).

Laser-assisted deuterium-tritium fusion: A quantum dynamical model

Deuterium-tritium (D-T) fusion is a key to generating safe, clean, and limitless energy on Earth in future fusion power plants. Its understanding at low collision energies is incomplete, as D-T fusion is a quantum tunneling process affected by resonances whose origin is linked to properties of not fully understood nuclear forces. Simplified quantum dynamical calculations of laser-assisted D-T fusion are presented, suggesting that laser-nucleus interaction can enhance the average D-T fusion probability by 7–70% at deep subbarrier energies using laser fields of intensity 1027–1029 Wcm−2 and a photon energy of 1 eV. Published by the American Physical Society 2024

The core–edge integrated neon-seeded scenario in deuterium–tritium at JET

Abstract This paper reports the first experiment carried out in deuterium–tritium addressing the integration of a radiative divertor for heat-load control with good confinement. Neon seeding was carried out for the first time in a D–T plasma as part of the second D–T campaign of JET with its Be/W wall environment. The technical difficulties linked to the re-ionisation heat load are reported in T and D–T. This paper compares the impact of neon seeding on D–T plasmas and their D counterpart on the divertor detachment, localisation of the radiation, scrape-off profiles, pedestal structure, edge localised modes and global confinement.

Diagnosing up-scattered deuterium-tritium fusion neutrons produced in burning plasmas at the National Ignition Facility (invited).

In the push to higher performance fusion plasmas, two critical quantities to diagnose are α-heat deposition that can improve and impurities mixed into the plasma that can limit performance. In high-density, highly collisional inertial confinement fusion burning plasmas, there is a significant probability that deuterium-tritium (DT) fusion products, 14.1MeV neutrons and 3.5MeV α-particles, will collide with and deposit energy onto ("up-scatter") surrounding deuterium and tritium fuel ions. These up-scattered D and T ions can then undergo fusion while in-flight and produce an up-scattered neutron (15-30MeV). These reaction-in-flight (RIF) neutrons can then be uniquely identified in the measured neutron energy spectrum. The magnitude, shape, and relative size of this spectral feature can inform models of stopping-power in the DT plasma and hence is directly proportional to α-heat deposition. In addition, the RIF spectrum can be related to mix into the burning fuel, particularly relevant for high-Z shell and other emerging National Ignition Facility platforms. The neutron time-of-flight diagnostic upgrades needed to obtain this small signal, ∼10-5 times the primary DT neutron peak, will be discussed. Results from several gain > 1 implosions will be shown and compared to previous RIF spectra. Finally, comparisons of experimental data to a simplified computational model will be made.

Optimization of Tokamak Blanket Design to Minimize Neutron Flux on Magnets

Fusion energy, a promising solution to global energy challenges, replicates the same processes that give our sun its energy, notably through deuterium-tritium (D-T) fusion reactions that produce significant energy. This study explores the mechanics and challenges of various fusion reactions, emphasizing the pivotal role of lithium breeder blankets in producing tritium, essential for sustaining D-T fusion in tokamak reactors. The helium-cooled lithium lead (LiPb) blanket design was simulated to optimize tritium breeding and neutron flux management. Results indicated a tritium breeding ratio (TBR) of 1.15, surpassing the self-sufficiency target of 1.1, with further improvements through increased lithium content and blanket thickness. Effective neutron shielding ensured safe operational limits for reactor components. These findings demonstrate the feasibility of achieving self-sustaining fusion reactions, essential for the viability of fusion power as a sustainable energy source. Future research will focus on advanced materials, refined simulations, and enhanced cooling technologies to further optimize fusion reactor designs.

JET machine operations in T&D-T

JET, the world’s largest operating tokamak with unique Be/W wall and tritium handling capability, completed a Deuterium-Tritium (D-T) campaign in 2021 (Maggi et al 29th Fusion Energy Conf.) following a decade of preparatory experiments, dedicated enhancements, technical rehearsals and training (Horton et al 2016 Fusion Eng. Des. 109–111 925). Operation with tritium raises significant technical, safety and scientific challenges not encountered in standard protium or deuterium operation. This contribution describes the tritium operational requirements, pulses and technical preparations, new operating procedures, lessons learned and details on the achieved operational availability and performance. The preparation and execution of the recent JET tritium experiments benefitted from the previous experience in 1991 (Preliminary Tritium Experiment), 1997 (DTE1 campaign) and 2003 (Trace Tritium Campaigns) and consisted of the following five phases: technical rehearsals and scenario preparation, tritium commissioning, 100% tritium campaign, D-T campaign (DTE2), tritium clean-up. Following the clean-up JET resumed normal operation and is currently undertaking a further D-T campaign (DTE3).

The thermodynamic properties of ZrCo hydrogen isotopes storage: The effect of Ta doping

ZrCo alloy exhibits distinct advantages in the storage of hydrogen and its isotopes. In this study, the phonon dispersion relation of ZrCo and ZrCoMx (M = H, D, T) is calculated. Subsequently, the thermodynamic properties of ZrCo and ZrCoMx compounds are investigated. Furthermore, an analysis is conducted on the temperature-dependent variations in the thermodynamic properties during hydrogen (deuterium, tritium) absorption reactions. Additionally, the impact of Ta doping (Zr0·875Ta0·125Co) on the thermodynamics of hydrogen storage (deuterium tritium) in ZrCo is discussed. The results indicate that H in the octahedral void leads to a higher thermodynamic value and a more stable system compared to H in the tetrahedral interstice. When ZrCo alloy store deuterium and tritium, there is an increase in ZPE, △H and △G of the system. At 300 K, the △H of ZrCo reacting with hydrogen, deuterium and tritium to form ZrCoM3 (M = H, D and T) are 85.94 kJ/mol, 90.09 kJ/mol and 92.16 kJ/mol, respectively. The ZPE and the △H of reaction are reduced upon Ta doping, thereby facilitating hydrogen (deuterium and tritium) circulating storage capabilities of alloy.

Optimization and experimental validation of deuterium-tritium neutron monoenergetics for diamond detector testing

Diamond detectors are excellent tools for measuring deuterium-tritium (D-T) neutron energy spectra. It has great potential for application in the field of magnetic confinement fusion. However, the current experimental values of the energy resolution of diamond detectors are much lower than the theoretical estimates. The purpose of this study is to find a method for experimental optimization. For this purpose, the D-T neutron energy distribution of the accelerator is theoretically analyzed and simulated. A method to increase the deuteron energy to obtain a better monoenergetic neutron source is proposed. This strategy was validated experimentally. The energy resolution of diamond for D-T neutrons achieves 1.21%, a Full Width at Half Maximum of 93 keV at 7.65 MeV. It is the highest resolution recorded to date. These results demonstrate the feasibility of the optimization method.

Demonstration of aneutronic p-11B reaction in a magnetic confinement device

Aneutronic fusion using commonly available fuel such as hydrogen and boron 11 (11B) is one of the most attractive potential energy sources. On the other hand, it requires 30 times higher temperature than deuterium–tritium fusion in a thermonuclear fusion reactor condition. Development of techniques to realize its potential for the experimental capability to produce proton-boron 11 (p-11B) fusion in the magnetically confined fusion device using neutral beam injection is desired. Here we report clear experimental exploration and measurements of p-11B fusion reactions supported by intense hydrogen beams and impurity powder dropper installed in the magnetic confinement plasma Large Helical Device. We measured a significant amount of fusion alpha particle emission using a custom designed alpha particle detector based on a passivated implanted planar silicon detector. Intense negative-ion-based hydrogen beam injectors created a large population of up to 160 keV energetic protons to react with the boron-injected plasma. The p-11B alpha particles having MeV energy were measured with the alpha particle detector which gave a fusion rate in a good agreement with the global p-11B alpha emission rate calculated based on classical confinement of energetic proton, using experimentally obtained plasma parameters.

Depth-resolved deuterium retention analysis in displacement-damaged tungsten using laser-induced breakdown spectroscopy

Fuel retention in plasma-facing components (PFCs) is a critical issue in future nuclear fusion reactors operating with Deuterium-Tritium (DT) regarding nuclear safety and fulfillment of the T cycle. However, during DT plasma operation, highly energetic neutrons will induce damage in the lattice of W PFCs causing enhanced fuel retention in defects or traps. Laser-Induced Breakdown Spectroscopy (LIBS) is a potential tool to monitor the T-content in situ in PFCs of future nuclear fusion devices. This article presents an ex situ study on pre-damaged W material after D plasma exposure to qualify the method and mimic conditions expected in a reactor. ITER grade W samples were displacement-damaged by 10.8 MeV W ions to a damage dose of 0.23 dpa and exposed to low temperature deuterium plasma at low energy in PlaQ. The resulting deuterium concentration was analyzed by using 3He Nuclear Reaction Analysis (depth resolution of ≈150 nm) as a well-established method, and LIBS (picosecond laser pulses, depth resolution of 15 nm). The sample with the highest deuterium concentration showed a deuterium-rich zone up to a depth of 1.13 μm using both techniques. This is close to the expected W ion-induced damage depth of ≈1 μm. The results imply that LIBS as an in situ technique for tritium monitoring could be a viable option for a reactor.

Baseline design of laser fusion research reactor with MW class laser facility

We propose a sub-ignition/burning reactor which is named the Laser-fusion Subcritical Power Reactor Engineering Method (L-Supreme). The reliabilities of L-Supreme in a MW class laser facility are assessed with respect to the following points: a reactor core, a target chamber, a target delivery system, an Exhaust Detritiation System (EDS), and neutron shielding. The Japan Establishment for Power-laser Community Harvest (J-EPoCH) would be applied as a MW class laser facility. A non-cryogenic glass balloon target filled with gaseous deuterium-tritium (DT) is contained in a target capsule. A chain-type magazine system might be used for a mass supply of the target capsules. Each target capsule is delivered to the center of a reactor core at 1 Hz. A batch of 10 000 laser shots would realize 0.22 MJ fusion power. The amount of tritium per batch is 1.51 × 1012 Bq. During laser experiments, unburned tritium is evacuated and transferred into an Exhaust Detritiation System (EDS). An evacuation rate of more than 0.1 m3 s−1 is required in order to recover less than 5000 Bq m−3 of the threshold of tritium concentration within 1 h. For safety, emergency situations such as tritium leakage in facilities are examined. The EDS works by internal circulation processes. Assuming leakage of tritium for a batch, an air circulation flow rate of 4100 Nm3 h−1 is required in an experimental hall for recovering less than 5000 Bq m−3 within 48 h. A primary and secondary neutron shield concept are proposed and would provide full neutron shielding. We conclude that it is possible to construct the L-Supreme system by marshalling current technologies.

Approach to startup inventory for viable commercial fusion power plant

With the increasing efforts to commercialize fusion power, private and government organizations are investing heavily in the development of technology to support a viable fusion power plant. Deuterium-Tritium (DT) fueled reactors are more prevalent than other proposed designs, requiring tritium processing and handling technology for safe operations and self-sufficiency. Further, each fusion power plant will need a specific-to-design startup inventory of tritium to begin operations. This startup inventory is required prior to breeding and is the minimum tritium inventory required to fill each processing component in the fuel cycle, to offset radioactive decay losses, and to avoid a zero-fuel situation for continuous operation. We present an approach to calculate the startup tritium inventory for a 500 MWth reactor, with considerations for reserve inventory for maintenance and commissioning. A baseline startup inventory was calculated to be approximately 327 gs. This value was obtained using modest assumptions about the technology and operating parameters of a fusion power plant. The required operating reserve inventory or the inventory necessary to keep a fusion power plant operational using only direct internal recycling for 24 h for the same plant design is approximately 642 gs. The approach and findings of this paper will enable fusion energy stakeholders to better utilize the existing scarce global tritium supply.