In the effort to deploy future fusion power plants and related technologies, educational and research programs for fusion enabling technologies are of high importance for pilot plants and production (energy) plants. This white paper attempts to discuss the current landscape of higher educational and research programs focused on fusion enabling technologies. The discussion is based on loose definitions of the potential largest areas of need and what is required to develop these programs from a federal/state perspective. It is noted that for educational programs to survive within the higher educational requirements, the programs will have to be intrinsically combined with funded research activities. This white paper is specifically focused on education and training programs for associates, bachelors, masters, and PhDs infusion enabling technologies. Historically there has been a disconnect between the plasma physics and fusion energy communities; thus future conversations around research and educational programs must balance the needs of both the historical plasma physics and fusion engineering communities from a place of casting a wider net. It is noted that plasma physics was addressed by other white papers at the Workforce Accelerator for Fusion Energy Development Conference held in Hampton, Virginia, USA, on 29-30 May, 2024. The challenges include (a) developing new or continued fusion enabling technology programs that will require funding over a long period of time and continuous support and patience for impact, (b) finding local faculty who are tied into fusion enabling technologies or can be recruited into them, and (c) generating interest into new institutions and communities. The primary issue is that educational and research program in fusion enabling technologies are limited in terms of the different technologies supported, as a limited number of universities and colleges exist. The overall recommendation is that if we are going to deploy more than one or more fusion pilot plants and commercial fusion power/production plants, we will need to initialize new programs and continue and expand existing programs. This will require the engagement of local faculty and staff from fusion community members (federal, national laboratories, universities, and companies), finding seed and long-term funding (research and educational), and inclusion in national activities.
Corrugated circular waveguides are essential radio-frequency (RF) components in electron cyclotron heating systems for magnetic confinement fusion devices such as tokamaks and stellarators. These waveguides enable low loss transmission of millimeter wave power over long distances (>10 m) between RF sources and in-vessel launching structures. Cold spray (CS) manufacturing, in which metallic microparticles are accelerated by a gas jet and impact a substrate to form a coating with high density and low oxidation, offers a novel fabrication pathway for these components. The method proposed in this paper deposits copper onto an externally corrugated cylindrical substrate followed by chemical removal of the substrate to produce an internally corrugated structure. As a first step to enabling fabrication by this method, spray experiments were conducted on flat corrugated substrates using a VRC Gen III high-pressure CS system controlled by a Fanuc M-710iC 6-axis robotic arm. Copper depositions onto these substrates, representative of similar to 100- to 200-GHz waveguides, are presented. Cross-sectional micrographs, material porosity, and surface roughness are reported for various spray parameters, postdeposition treatments, substrate materials, and corrugation aspect ratios and sizes.
The KArlsruhe TRitium Neutrino (KATRIN) experiment aims to determine the effective mass of the electron antineutrino by investigating the tritium beta-spectrum close to the kinematic end point. The Tritium Laboratory Karlsruhe hosts and operates the tritium parts of the experiment. A dedicated tritium loop system is tasked to provide the < 0.1% stabilized flow rate of tritium gas into the KATRIN source, maintaining a throughput of 40 gd(-1) and a tritium purity >95%, while at the same time acting as the interface to the established tritium handling infrastructure of the laboratory, reliably working for more than 3 decades. Since KATRIN's start of tritium operation in May 2018, more than 1350 operation days of the tritium "loop" system combined with KATRIN's tritium source have been achieved. This paper summarizes the tritium operation experience gained with special emphasis on the permeator performance, a key component enabling direct internal recycling.
The planned DEMOnstration fusion power plant is designed to operate in a pulse (BURN-DWELL) regime, using a balance of plant with an intermediate heat transfer system which, in some cases, utilizes helium and molten salt as heat transfer medium. For that gas-molten salt heat exchangers are one of the crucial components. Therefore, a W-type shell and tube heat exchanger with helium inside the tubes and molten HITEC salt outside the tubes is proposed. In order to optimize the heat exchanger design, an in-house numerical code and an evaluation function were developed. Searching the parameters space of more than 11 000 combinations (sets) with varying tube inner diameter, tube wall thickness, pitch, and number of tube heat exchangers with minimal values of the optimization coefficient were identified. The optimal heat exchanger design found has 35 000 tubes of 10 mm inner diameter, 12 mm outer diameter, and 21 mm pitch, and the equivalent size of the heat exchanger is circle divide 7.9 & times; 29.6 m with 0.3-MPa pressure loss in the primary heat transfer system helium and 3.6 MPa pressure loss in the intermediate heat transfer system molten HITEC salt, while transferring 262.8 MW of heat from helium to the HITEC molten salt.
This paper provides an overview of the tritium-related research activities conducted at the ENEA Brasimone Research Center, a leading facility in Italy dedicated to nuclear energy and fusion technologies. The Center's fusion research mainly deals with the development of the Breeding Blanket (BB) within the ITER and DEMO projects, especially of the water-cooled lithium lead BB and related materials.In this context, a key focus is the study of tritium behavior in liquid-metal environments, with a range of experimental devices and setups designed to study the fundamental tritium transport properties of structural and functional materials, tritium permeation in the gas phase and in contact with water, tritium extraction from liquid lithium-lead, and tritium recovery from coolants, particularly from helium. Protium or deuterium are used as substitutes for tritium in these studies.Additionally, the research at ENEA Brasimone includes the development and validation of tritium transport models, which are essential for design and safety analyses of future fusion reactors. This paper presents the experimental infrastructures and the main recent achievements, emphasizing how these experimental devices and modeling efforts contribute to improving the design and efficiency of tritium breeding systems, which are essential for the sustainable operation of fusion reactors.
The Hydrogen-3 Advanced Technology facility to be located on the United Kingdom Atomic Energy Authority Abingdon Culham Campus will build on its world-class tritium capability supporting future reactor fuel cycle development for projects such as the International Tokamak Experimental Reactor (ITER), Demonstration Reactor (DEMO), and Spherical Tokamak for Energy Production (STEP) by building an experimental tritium fuel loop at 1/10th scale of the proposed ITER fuel cycle. Fusion energy production requires the use of deuterium and tritium isotopes of hydrogen as its fuel sources. Quantitative analytical techniques capable of resolving hydrogen isotopologues are vital to tritium measurements, accountancy, and safety, as the radioactive nature of tritium requires close control of the location and quantities used within an experiment or facility. This work describes the use of micro gas chromatography to detect stable, nonradioactive, hydrogen isotopologues encountered within the tritium loop. Micro-GC uses sample sizes, on a microliter scale, through miniaturized sample injection. The separation of hydrogen isotopes was achieved by use of an external capillary cryo-column connected to the micro-GC, held in a Dewar of liquid nitrogen. The experimentation covers the use of micro-packed columns; capillary columns, including the effect of column length on peak resolution; and the Golay curve for neon carrier gas. It concludes that micro-packed columns are not capable of isotope separation in this setup and that the use of micro-GC is likely not suitable for large-scale plant-based analyses, and it gives the Golay curve for neon to determine the optimal carrier gas conditions.
Hydrogen isotope (H2, D2) separations in RHO-type zeolites having 0.36 & times; 0.36-nm pores were evaluated via temperature-programmed desorption and temperature-programmed adsorption. The synthesized Na,Cs-RHO zeolite was exposed to a mixed gas (H2/D2 = 50.7/49.3) at 196, 250, and 273 K, followed by desorption measurements starting at 77 K. A desorption peak was observed at approximately 190 K, and the D2/H2 selectivity was estimated to be approximately 1.4, which was independent of the exposure temperature. Furthermore, measurements confirmed adsorption of both H2 and D2 near 190 K. The adsorption of H2 and D2 at these high temperatures was also supported by sorption isotherm results.
The glove box detritiation system removes tritium from contaminated gas streams using a hydrogen-absorbing metal bed. Its absorption performance is governed by various operating parameters. Under extreme conditions, insufficient absorption capacity can lead to gas breakthrough and potential leakage, thereby reducing detritiation efficiency and compromising system safety. Conventional absorbing efficiency assessments are often inadequate because the actual temperature distribution within the bed is highly nonuniform, which causes nonnegligible variations in the removal efficiency.To address this, this study proposes an evaluation method based on multiple physical field coupling for a more precise assessment of the hydrogen concentration distribution. The absorption process within the metal bed is simulated using the finite element method, coupling absorption kinetics with the pressure-composition-temperature relationship. The results demonstrate that bed structure strongly affects the flow field, which subsequently alters the temperature distribution, and consequently, the absorption behavior. Finally, the hydrogen transport within the glove box is simulated by coupling the metal bed efficiency with the glove box flow field.
The Tritium Laboratory Karlsruhe (TLK) is a unique semitechnical facility dedicated to the development of tritium processing technologies encompassing analytical systems, processing methodologies, and experimental setups. Authorized to handle up to 40 g of tritium, TLK represents a central facility within the Helmholtz Association for tritium-related research and development. Since commencing operations in 1993, its proven confinement and processing philosophy has established an excellent safety record. This paper provides an overview of the current status of TLK's tritium infrastructure, summarizing operational experience, organizational frameworks, and recent technical progress. Developments addressing solutions to operational challenges, new facility interfaces, and the commissioning of a dedicated tritium isotope recovery plant are presented. An outlook on the planned tritium waste disposal facility underlines TLK's continuous commitment to innovation in tritium handling technologies.
Commissioned in 1983, the Joint European Torus (JET) has been the cornerstone of fusion research for four decades, achieving its most significant scientific milestones in its final operational years before ceasing operations in December 2023. Its transition to decommissioning represents a pivotal moment for Culham as it evolves toward a research-hub model for next-generation fusion reactors. JET will be the second major tritium-operated fusion facility to be decommissioned, following TFTR in 2002, whose tritium in-vessel cumulative inventory was significantly lower than JET (similar to 100 versus 1000 g). The JET Decommissioning and Repurposing Programme (JDR) must manage a substantial radioactive waste inventory, including approximately 9000 tonnes of low-level waste and around 250 tonnes of tritiated and activated intermediate-level waste (ILW), for which no long-term disposal route currently exists. As JET ILW is predominantly driven by tritium absorption, thermal detritiation offers a potential route for waste reclassification and disposal. This paper presents the expected waste inventory, the resulting JDR delivery strategy, and qualitative characterization results from ILW materials retrieved immediately after shutdown, which are expected to drive the overall waste management strategy.
Tritium plays a crucial role in nuclear fusion power plant designs, and adsorption beds are essential tools for managing tritiated water vapor. Series of tests were performed to investigate whether a saturated adsorption bed preferentially adsorbs heavy water vapor. The design of passive tritiated control systems could rely on adsorption beds preferentially trapping heavier isotopologues of water. This work investigates the displacement phenomenon and the effect of heavy water concentration on the performance of the bed. Significant displacement was observed when a humid stream of heavy water was diverted through a bed presaturated with light water, as indicated by changes in D2O and H2O partial pressures. After the capture of heavy water in the bed, the subsequent rise in D2O partial pressure depended on the heavy water humidity in the gas stream. Higher humidity led to faster and steeper mass transfer profiles in the adsorption bed, which were empirically fitted with sigmoid curves.
The Spherical Tokamak for Energy Production (STEP) will produce radioactive waste throughout its lifetime and decommissioning. In order to maximize the environmental sustainability of the program, it is essential to minimize the volume and activity of waste being sent to disposal by developing and optimizing new waste treatment techniques. The STEP program has been investigating the application of acids as a surface detritiation technique on 15g tritiated copper (Cu) and stainless steel 304 (SS304) samples originating from the Joint European Torus (JET). Hydrochloric acid, nitric acid (HNO3), sulfuric acid (H2SO4), and aqua regia have been assessed over an incremental 30 minute etch in order to determine the most effective acid for detritiating each material. Although HNO3 was found to be the most effective acid for the detritiation of Cu, the substantial mass loss observed suggests it may be too aggressive for a surface detritiation technique. SS304 experienced high detritiation from each acid, with H2SO4 and aqua regia achieving similar to 100% detritiation over the 30 minute period. The negligible mass loss observed throughout the SS304 etches may prove advantageous for a surface detritiation technique, potentially unlocking more favorable pathways within the waste hierarchy.
Currently, verification and validation (V&V) for fusion neutronics are largely performed manually: analysts convert benchmark models between code input formats and reformat the results for comparison, making the process time consuming, error prone, and hard to reproduce. Existing benchmark suites also rely almost exclusively on constructive solid geometry, thereby excluding from validation much of the practical neutronics workflow, including computer-aided design-based geometries, meshing, and transport on unstructured meshes. We propose a new approach to address these gaps, implemented in the open-source OpenMC-Fusion-Benchmarks (OFB) repository. The OFB provides a code-agnostic, standardized, and human-readable benchmark specification, with geometries supplied as Standard for the Exchange of Product (STEP) model data files and an automated meshing pipeline. The framework is designed to enable the near-fully automated V&V of neutronics codes and methods, from benchmark definition to simulation execution and results comparison. We illustrate the workflow on a representative benchmark and outline the remaining challenges and the future developments aimed at further automating and strengthening the validation of fusion neutronics tools.
Metal melting has been widely viewed as the most promising method for detritiating metallic waste, with its potential to release more tritium than other thermal methods. Furthermore, the homogenization of the metal postmelting is expected to distribute tritium throughout the ingot, reducing hotspots in the waste form. Removing tritium from wastes is important as it makes disposing or recycling of the metal far easier. Detritiation trials conducted with stainless steel showed promising indications, with detritiation efficiencies ranging from 68% to 97%. However, the low starting activity of the charge material (<100 Bq/g) introduced high statistical variability and potential cross-contamination effects, which impacted the reliability of some data. Despite these challenges, the observed tritium removal highlights the potential of metal melting as an effective method for treating fusion-related metallic wastes. Future studies with higher-activity samples and additional repetitions are planned to further investigate the impact of longer hold times and different crucible types on detritiation efficiency.
A study was conducted to develop neutron sources with magnetic system designs similar to those of two types of fusion devices: large helical device (LHD) and W7-X stellarators. The device dimensions, plasma confinement volumes, and plasma-confining magnetic field strengths were chosen to be close to those of the ITER tokomak experimental reactor. Calculations were performed using a spatiotemporal numerical code solving a system of diffusion and heat conductivity equations, the transport coefficients of which were determined by neoclassical transport processes. To ensure steady-state operation of the sources, continuous plasma heating was assumed, with the density maintained by fuel pellet injection. A total neutron flux of 5 & times; 10(19) s(-1) was demonstrated. The calculation results indicated that a neutron source using a magnetic system design similar to the W7-X stellarator is more productive.
Capillary porous systems (CPSs) containing liquid lithium are typically fabricated from tungsten fibers. These CPS units are often attached to a steel base using spot welding or simple clamps, methods that are inefficient for transferring high heat flux. In this work, the CPS was integrated into a base structure made of steel (AISI 316LN and AISI 420) and a nickel alloy (Ni34Fe). This study investigated the composite structures under static and cyclic thermal loads, measured their thermal diffusivity, and examined the effects of lithium corrosion. The composite with the Ni34Fe matrix demonstrated the highest resistance to static and cyclic thermal loading and exhibited thermal diffusivity values of up to 8.5 mm2/s. In contrast, composites with steel-based matrices failed after 100 cycles at 500 degrees C/50 degrees C, primarily due to the formation of brittle intermetallic phases. However, the main drawback of the Ni34Fe alloy was its susceptibility to grain boundary corrosion in liquid lithium at 500 degrees C.
There are numerous different technologies that can be used for water detritiation. There are two main approaches that are used for water detritiation, namely, volume reduction and tritium extraction. Among the different techniques, water distillation (WD) is a proven technology for water detritiation. WD, when used as a standalone technology for detritiation, is used for volume reduction. However, when coupled with thermal diffusion columns (TDCs) via a relatively small electrolyzer, it offers a low-cost viable solution for small-scale water detritiation and tritium extraction. This article presents a high-level layout, mathematical model, and simulation results of a water detritiation and tritium extraction plant. A base case simulation is performed with a 1 kg/h feed flow rate and a tritium concentration of 1 Ci/kg. The effects of different parameters, such as the bottom product flow rate, reflux ratio, and required height of the TDC column, are discussed.
Tritium, a radioactive and hazardous material, is one of the essential fuels for the deuterium-tritium (D-T) fusion reaction. In a fusion reactor, tritium is bred in the ceramic breeder material lithium meta-titanate (Li2TiO3) through interactions between lithium and neutrons originating from the plasma. In this study, the estimation of tritium permeation rates for a moderate-size fusion pilot plant has been carried out, as tritium can permeate from ceramic breeders to the coolant and subsequently through pipes and structural materials to the environment at elevated temperatures. In fusion reactors, reliable management of tritium is of fundamental importance from both safety and fusion fuel cycle perspectives. Therefore, tritium permeation rates from different loops to the environment, as well as its transfer between various loops, must be estimated to ensure accurate tritium accountancy in the fusion fuel cycle. A tritium transport model for the helium-cooled ceramic breeder blanket has been developed considering both steady-state diffusion-limited and steady-state surface-limited permeation regimes, and the results from these models have been compared. Tritium behavior under dynamic conditions has also been analyzed analytically for both diffusion-limited and surface-limited regimes. In addition, tritium permeation and retention within structural materials have been estimated for scenarios with and without trapping effects. A sensitivity analysis of the steady-state diffusion-limited tritium permeation model has also been performed by varying key parameters, including tritium generation rate, purge gas flow rate, and tritium extraction system efficiency, to evaluate the impact of each parameter on tritium permeation rates. Furthermore, tritium mass balance has been verified under steady-state conditions for both diffusion-limited and surface-limited permeation regimes, confirming the consistency and validity of the developed tritium transport model.
The dual-coolant lithium-lead and the water-cooled lithium-lead blankets driven by EUROfusion foresee PbLi as breeding material. Their efficient operation depends on correct definition of the hydrogen isotope transport parameters in liquid PbLi. However, there is large uncertainty in some of the experimental values published in the literature, which is unacceptable from the point of view of correct breeding blanket design. Furthermore, the exact lithium content and impurities that the PbLi alloy itself may contain represent another key aspect that needs to be addressed. The presented work summarizes experimental measurements made at the University of the Basque Country (EHU) on this matter during the last years, always in the framework of EUROfusion. The quality of PbLi samples was certified in terms of impurities and lithium content, and three different campaigns were carried out in close collaboration with CIEMAT. The samples have been tested using the absorption-desorption technique and introducing protium and deuterium alternatively. The results shown in the present work are focused on measurement of the solubility of protium and deuterium in PbLi expressed in terms of the Sievert's constant. The results obtained during the three different campaigns are shown and compared, combining measurements obtained by both absorption and desorption processes. These results are also compared with values previously published in the literature, and the isotope effect is also analyzed to check possible extrapolation of the results to the solubility of tritium in the PbLi alloy.
This paper presents a scientifically validated software architecture upgrade of the data acquisition and control system for a 100-kW, 45.6-MHz ion cyclotron resonance heating (ICRH) system at the Institute for Plasma Research. Existing control systems for high-power ICRH installations have predominantly relied on proprietary, platform-specific solutions that lack scalability and interoperability, which are critical deficiencies as fusion experiments grow in complexity. This work addresses these limitations by demonstrating the feasibility and performance advantage of an open-standard, EPICS-based architecture integrated with programmable logic controllers via MODBUS/TCP. The key scientific contributions include (i) a rigorous performance characterization of the EPICS-MODBUS/TCP interface latency under realistic operational loads; (ii) quantitative benchmarking of monitoring (100-ms update cycle), software-layer interlock response (<50 ms) and hardwired interlock circuit response (720 ns) against IEC 61511 safety integrity requirements; and (iii) empirical validation of error rates (<0.01%) and CPU utilization (15%-25%) during sustained 24-hour operation supported by real-time acquisition. Data acquisition is implemented on a USB-based multi data acquisition card to meet the1-kHz rate data storage requirement. The system has demonstrated successful power delivery exceeding 25 kW to a dummy load, with ongoing campaigns targeting full-power operation. The architecture's modular design reduces integration effort by an estimated 40% compared to the prior proprietary implementation and is directly applicable to other tokamak auxiliary heating systems globally.
