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The Commission for 'Corresponding to Radiation Disaster of the Japanese Radiation Research Society' formulated a description of potential health effects triggered by tritium. This was in response to the issue of discharging water containing tritium filtered by the Advanced Liquid Processing System (ALPS), generated and stored in Fukushima Daiichi Nuclear Power Station after the accident. In this review article, the contents of the description, originally provided in Japanese, which gives clear and detailed explanation about potential health effects triggered by tritium based on reliable scientific evidence in an understandable way for the public, were summarized. Then, additional information about biochemical or environmental behavior of organically bound tritium (OBT) were summarized in order to help scientists who communicate with general public., (© The Author(s) 2021. Published by Oxford University Press on behalf of The Japanese Radiation Research Society and Japanese Society for Radiation Oncology.)

• T permeability was obtained in tungsten in W via a gas-driven permeation technique with a flowing H-T permeation system. • T permeability was measured in low T partial pressure of(5.8-6.8)x10-2 Pa over the temperature range from 573 to 873 K. • A scanning Auger microscope was used to characterize surface conditions of W before and after T permeation experiment. A flowing hydrogen-tritium permeation system was used to investigate tritium permeation behavior in polycrystalline tungsten. Tritium permeability was measured with T 2 partial pressure ranging from 5.8 × 10−2 Pa to 6.8 × 10−2 Pa over the temperature range from 573 to 873 K. The obtained tritium permeability in polycrystalline tungsten was P T [mol m−1s−1 Pa−0.5] = (2.13 × 10−9)exp(−93.32 [kJ mol−1]/ RT). This is one of the first tritium permeability data from any polycrystalline tungsten via a gas-driven permeation technique. The tritium permeability in polycrystalline tungsten was compared with the reported hydrogen isotope permeability in other polycrystalline tungsten. [ABSTRACT FROM AUTHOR]

Understanding of thermodynamic equilibria of multi-components hydrogen isotopes is required to accurately measure tritium permeability at the expected low tritium partial pressure and non-negligible high hydrogen partial pressure in a fusion blanket system. A gas-driven tritium permeation system that is capable of independently controlling hydrogen and tritium partial pressures was developed at Idaho National Laboratory to accurately measure low partial pressure tritium permeability. The thermodynamic equilibria for hydrogen (H) – tritium (T) permeation through metal are discussed to accurately measure tritium permeability, and the experimental conditions required for evaluating tritium permeability in H-T system are presented. [ABSTRACT FROM AUTHOR]

The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to evaluate in-vessel tritium inventory in the nuclear environment for fusion safety. The electrical upgrade were recently carried out to enhance operational safety and to improve plasma performance. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium and eliminating heat stress issue. In November 2015, the TPE successfully achieved first deuterium plasma via remote operation after a significant three-year upgrade. Simple linear scaling estimate showed that the TPE is expected to achieve Γ i max of >1.0 × 10 23 m −2 s −1 and q heat of >1 MW m −2 with new power supplies. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, FNSF, and DEMO for improving in-vessel tritium inventory assessment in fusion nuclear environment. [ABSTRACT FROM AUTHOR]

This paper describes a new experimental system designed to investigate tritium mass transfer properties in materials important to fusion technology. Experimental activities were carried out at the Safety and Tritium Applied Research (STAR) facility located at the Idaho National Laboratory (INL). The tritium permeation measurement system was developed as part of the Japan/US TITAN collaboration to investigate tritium mass transfer properties in liquid lead lithium eutectic (LLE) alloy. The experimental system is configured to measure tritium mass transfer properties at low tritium partial pressures. Initial tritium permeation scoping tests were conducted on a 1 mm thick α-Fe plate to determine operating parameters and to validate the experimental technique. A second series of permeation tests was then conducted with the α-Fe plate covered with an approximately 8.5 mm layer of liquid lead lithium eutectic alloy (α-Fe/LLE). We present preliminary tritium permeation data for α-Fe and α-Fe/LLE at temperatures between 400 and 600 °C and at tritium partial pressures between 1.7E − 03 and 2.5 Pa in helium. Preliminary results for the α-Fe plate and α-Fe/LLE indicate that the data spans a transition region between the diffusion-limited regime and the surface-limited regime. Additional data is required to determine the existence and range of a surface-limited regime. [ABSTRACT FROM AUTHOR]

Abstract: Recent progress in experiments, modeling and theory has increased confidence that ITER can achieve its goal of high Q (>10) in inductive operation. Further, experimental results in the ‘improved hybrid’ regime suggest a possibility of high Q (>10) operation in a long pulse (>1000s) with benign ELMs. However, considerable uncertainty still exists in the prediction of several key aspects; for example, tritium retention, disruption, impurity control, ELMs, SOL transport and dust. ITER requires flexibility of operation scenarios at least in the early operation phase to accommodate uncertainties in prediction, to explore wide operational spaces and to incorporate newly developed control schemes. Because of these uncertainties conservative assumptions are adopted in performance predictions, and step-wise implementation of reactor-relevant plasma-facing material, such as tungsten, are planned. [Copyright &y& Elsevier]

Abstract: A baseline strategy for dust and tritium-inventory control and recovery in ITER has been established and preparations are underway for its implementation. Limits on dust and tritium-inventory are an integral part of the ITER safety case and are fixed at 1kg for tritium, 1000kg for mobilisable dust and 11kg (beryllium)/76kg (tungsten) for dust on hot surfaces. Maximum average T-retention rates of ∼1g/shot are estimated for baseline inductive operation at QDT =10, suggesting that the in-vessel T-retention could reach the administrative limit of 640g in as little as ∼2months of operation. Baking is expected to remove a significant fraction of the T co-deposited on the divertor targets. Despite large uncertainties, dust quantities are expected to remain well below safety limits over the divertor cassette lifetime. In situ aspiration during divertor cassette exchange is foreseen as the main dust removal technique. [Copyright &y& Elsevier]

Abstract: We have applied a tritium imaging plate technique to measure the tritium distribution profile on surface and in bulk of various metal materials after exposure to a deuterium–tritium plasma in a linear plasma experimental apparatus. The experimental tritium concentration profiles in mm range are interpreted according to a simple hydrogen diffusion model in each metal. We found that a significant amount of tritium is localized in near surface regions and is clearly distinguishable from tritium diffused in the bulk. The amount of surface tritium is not likely correlated to bulk properties (diffusivity and solubility), but is related to trapping in surface defects or metal impurities such as oxide and carbide. [Copyright &y& Elsevier]

Abstract: The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials. [Copyright &y& Elsevier]

Abstract: The Tritium Plasma Experiment (TPE) has been used to investigate deuterium fuel retention behavior in tungsten and molybdenum-materials utilized for plasma-facing surfaces in some existing tokamak plasma devices and under consideration for future devices. Although several studies have been performed over the past several years on these metals, many issues remain unresolved, including for example blister formation mechanisms and correlation to surface conditions. In this study, we expose several metal samples to deuterium ion fluences up to 1026 ions/m2 and measure retention behavior with thermal desorption spectroscopy. Fractional retention of up to 2.0×10−5 is found for W at 600K, and Mo similarly retains deuterium at a fraction of 1.5×10−5 at 600K. Blistering was found for W samples exposed at temperatures above 453K, whereas blistering was not observed for Mo samples at any experiment temperature. [Copyright &y& Elsevier]

Abstract: The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C β-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073K over the duration of ∼3000h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation. [Copyright &y& Elsevier]