Liquid Tritium Research Articles

Self-diffusion of liquid deuterium hydride and liquid tritium.

We present a quasi-elastic neutron scattering study of liquid deuterium hydride carried out using the Disk Chopper Spectrometer at the National Institute of Standards and Technology. Under saturated vapor pressure, the self-diffusion constant of deuterium hydride obeys an Arrhenius law D=D0exp-EA/kBT, where the prefactor D0 is given by D0=9.5±1.2Å2/ps and the activation energy is given by EA = 58 ± 2K. We apply the quantum law of corresponding states to the known diffusion constants of the hydrogen isotopologues. From this application, we estimate that D0≈9.1Å2/ps and EA ≈ 75K in liquid tritium. Young's theory of quantum-mechanical effects in van der Waals fluids [Young, Phys. Rev. A 23, 1498 (1981)] is shown to apply to the diffusion constants of the liquid hydrogens. Our results underscore the importance of nuclear quantum effects in shaping the properties and behavior of the hydrogen isotopologues.

Comparative Study on the Reduction of Tritium Breeding Ratio Caused by Inventory Changes of a Solid-State Tritium Breeding Blanket in a Fusion Demonstration Reactor Using MCNP and FISPACT-II

The self-sufficient supply of tritium in the deuterium-tritium nuclear fusion demonstration reactor (DEMO) is a fundamental design requirement. But, it is hindered by depletion of tritium breeding materials resulting in reduction of tritium breeding ratio (TBR) less than the initial design value especially in the solid-state tritium breeding blanket (TBB) of the DEMO. Unlike the liquid tritium breeding blanket of DEMO, compensation measures of the depleted breeding material in the solid-state TBB will be its substitution depending on the reduction rate of TBR. To estimate the replacement period of the solid-state TBB, it is required to estimate the reduction rate of TBR according to the operation conditions of the DEMO and the physical configuration of a solid-state TBB. In this study, the representative simulation codes, MCNP and FISPACT-II, are used for assessment of the reduction rate of TBR with the benchmarking model which is modified from the one poloidal segment of the TBB in the Korean-DEMO. After 3 full power-year operations with the neutron irradiation with the benchmarking model, the TBR simulated by MCNP is reduced to 96.84% of the initially calculated TBR, but the TBR calculated by FISPACT-II is reduced to 90.57% of the initially calculated TBR.

A New Approach to Nuclear Fusion Based on Muon Catalysis by Means of High-Energy Ball Lightning and an Accelerative Mechanism

A new method of realizing nuclear fusion reactions based on muon catalysis and the accelerative mechanism is proposed. High-energy ball lightning is periodically generated in a reactor chamber filled with deuterium gas and directed into a container containing liquid tritium. The entry of ball lightning into the tritium is accompanied by the generation of muons and mesomolecules due to a cascade process. Following the ball lightning, a high-energy plasma jet moves under the influence of traveling transverse electromagnetic waves. Deuterium ions and electrons of the jet, accelerated by intense transverse electromagnetic waves, interact with the tritium. Nuclear fusion reactions occur with the participation of muonic molecules at very low temperature. The developed method resolves all physical and technical problems that are inherent in existing traditional methods.

Fabrication and characterization of zirconium oxide coating on tubular steel by metal organic decomposition

In fusion reactor blankets, multifunctional ceramic coatings fabricated on structural materials have been investigated for tritium permeation reduction, electrical insulation, and corrosion protection. However, it has been challenging to manufacture a homogeneous coating on the complicated components in the blanket breeder, such as ducts. In terms of sustainable development, it is crucial to establish a ceramic coating technique on the inner walls of fuel pipes in liquid blanket systems. In this study, we applied the metal organic decomposition method to fabricate zirconium oxide coatings on 316L austenitic stainless steel tubes and characterized their properties as liquid tritium breeding pipes. Homogeneous zirconium oxide coatings with a thickness of 395 ± 12 nm have formed on the inner walls of SS316L tubes. The compatibility of the coated tubes with liquid lithium-lead was examined under static and flowing conditions at 550–600 °C for prolonged exposure of up to 2000 h. Under static conditions, ZrO2 coatings demonstrated exceptional compatibility with the liquid Li-Pb as they protected structural material from corrosion. At a flow rate of 0.16 m/s, the coatings were partly delaminated due to different thermal expansion coefficients of the coating and substrate and acceleration under a dynamic system. The coating thickness increased up to 650 ± 26 nm after exposure for 2000 h, which indicated ZrO2 coatings reacted with Li-Pb to form corrosion products. Besides, ZrO2 coatings achieved the insulating requirements of magnetohydrodynamic insulators as their electrical resistivities were on the order of 104–106 Ω m at the temperature range of 250–550 °C.

Development of 2.2 μm cavity ring-down spectrometer for tritiated water analysis

A rapid and simple tritium analysis method is required for tracer application and the quantitative evaluation of radioactive waste. In this study, we focused on cavity ring-down spectroscopy (CRDS), which is an ultra-sensitive laser absorption spectroscopy, and developed a spectrometer for tritium analysis. A current modulation-assisted acoustic optical modulator switching method was developed in the prototype setup containing a 2.2 μm diode laser for accessing the 2ν1 absorption band of tritiated water vapor. The benefit of this switching method was investigated using the Allan deviation and compared to conventional acoustic optical modulator-only and current-only switching methods. Using the prototype setup with the proposed switching method, CRDS of stable H2O vapor was demonstrated. The detection limit for liquid tritium water analysis was estimated to be 2 × 101 kBq/10 μl for ten-minute measurements.

An Innovative Approach to Prepare Liquid-Solid Dual-Phase Flowable Tritium Breeder with Low MHD Effect

In present paper, a novel flowable tritium breeder is prepared by mixing the Li2TiO3 micro-powders and liquid GaInSn alloy, where GaInSn alloy is used to simulate the fluid behaviors of lithium-based liquid tritium breeder, forming a type of composite characterized by liquid-solid dual phase. In detail, the effects of the volume fraction of ceramic micro-powders on viscosity and conductivity of the composite in magnetic field are the focus. The XRD results prove that the obtained Li2TiO3 micro-powders contained Li2TiO3 phase without impurities. The results shows that once the magnetic field intensity exceeds the critical value, the viscosity of liquid GaInSn metal becomes significantly greater than that of liquid-solid dual-phase composites. Furthermore, the addition of Li2TiO3 micro-powders could effectively reduce the magneto hydro dynamic (MHD) fluid effect, and the dual-phase composites exhibit comparatively lower flow resistance under the strong magnetic field. Moreover, the conductivity of the tritium breeder composites decreases rapidly with the addition of Li2TiO3 micro-powders. The MHD pressure-drop-increasing rate decreases with the increase of viscosity, which indicates that the addition of Li2TiO3 micro-powders effectively reduces the MHD effect. The conductivity of the composites increased slightly and then remained stable after static placing for several tens of minutes. The present investigation provides a novel insight into the fabrication strategy of tritium breeder materials with low MHD effect.

Solubility of Bi in Li–Pb Eutectic Alloy Between 508 and 623 K

Liquid lithium–lead (Li–Pb) eutectic alloy is one of the candidate liquid tritium breeding materials of fusion blanket and it contains bismuth (Bi) as an initial impurity or transmutation product from lead. The saturated solubility of Bi, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$S$ </tex-math></inline-formula> (wt.ppm), in Li–Pb was measured in the temperature range from 508 to 623 K. The measured results were plotted on the graph with <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$S$ </tex-math></inline-formula> (wt.ppm) versus 1/ <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$T$ </tex-math></inline-formula> . From these plots, a positive temperature dependence, where the solubility decreases as the temperature is lowered, is confirmed. The equation of the solubility of Bi in Li–Pb was obtained, too. In addition, the dissolution enthalpy of Bi in Li–Pb, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\Delta H_{\mathrm {sol}}$ </tex-math></inline-formula> , was quantified from this equation. The precipitation morphology of Bi in Li–Pb was also investigated. X-ray diffraction was used to analyze the precipitates collected from the surface of the liquid phase. As a result, the peak pattern of Li3Bi was found from the precipitate. Moreover, the amount of polonium produced when the cold trap is introduced to a Li–Pb loop of a fusion reactor was predicted from the solubility obtained in this study.

The synergy of heavy-ion irradiation and lithium-lead corrosion on deuterium permeation behavior of ceramic coating

Functional ceramic coatings have been investigated for several decades to mitigate tritium permeation through structural materials in a fusion reactor blanket. For the establishment of a liquid blanket system, the coatings require not only tritium permeation reduction but also the tolerance to irradiation damage induced by neutrons and corrosion by liquid tritium breeders. In this study, deuterium permeation measurements under exposure to liquid lithium–lead were performed for a heavy-ion-irradiated zirconium oxide coating sample to elucidate the synergy of irradiation and corrosion on hydrogen isotope permeation. The irradiated coating decreased the deuterium permeation flux under lithium–lead exposure by more than two orders of magnitude in comparison with the unirradiated one. Each activation energy of permeation and diffusion of the coating sample was larger than that of the unirradiated one at around 450 °C and decreased to the level of the unirradiated one at 600 °C. The results indicated that the voids formed by ion irradiation aggregated in grain boundaries of the coating, resulting in an increase in the energy barrier of diffusion at low temperatures and then a decrease due to the defect recovery at high temperatures. Besides, a corrosion product layer formed on the coating during lithium–lead exposure increased the energy barrier of solution. The synergy of irradiation and corrosion on permeation was developed at lower temperatures before the defect recovery.

Submicron-thick yttria-stabilized zirconia coating as an advanced tritium permeation barrier

Tritium permeation through structural materials is one of the critical issues in a fusion reactor from the viewpoints of an efficient fuel cycle and radiological safety. Ceramic coatings have been studied as tritium permeation barriers (TPBs) and shown sufficient reduction of hydrogen isotope permeation. In high thermal efficiency liquid blanket concepts, however, not only tritium permeation but also corrosion of the structural materials by liquid tritium breeders such as lithium–lead (Li–Pb) is a serious concern. This study focuses on the fabrication of two kinds of yttria-stabilized zirconia (YSZ) coatings by metal organic decomposition and characterization through several tests such as deuterium permeation and static Li–Pb exposure. The 4 mol%-Y2O3-containing YSZ coating of less than 100 nm in thickness showed three orders of magnitude lower deuterium permeation flux than uncoated substrate after crystallization and grain growth, and the high permeation reduction performance maintained through the thermal cycle tests. The 10 mol%-Y2O3-containing YSZ coating sample showed a high permeation flux due to the defects formed during the permeation measurement. Both coatings showed compatibility with Li–Pb. A phase transformation was confirmed in the 4 mol%-Y2O3-containing YSZ coating after bending tests, suggesting that the coating sample achieved a self-repairing property. These results indicate that the 4 mol%-Y2O3-containing YSZ coating would apply to the TPB.

Deuterium permeation through multi-layer ceramic coatings under liquid lithium-lead exposure condition

Corrosion of tritium permeation barrier ceramic coatings by liquid tritium breeders is an unavoidable concern in a high thermal efficiency fusion reactor. In this study, we have established a deuterium permeation measurement system under static Li-Pb exposure conditions to investigate the deuterium permeation behavior of multi-layer ceramic coatings under corrosive environments. The system successfully detected the coating degradation on the moment by measuring an irregular increase in deuterium permeation. The permeation behaviors showed that the coating degraded more seriously as the number of the interfaces between the different ceramics increased.

Assessment of tritium effluent from Prototype Generation IV Sodium-cooled Fast Reactor

A tritium transport model for a pool-type sodium cooled fast reactor was developed for estimation of tritium effluents from the Prototype Generation IV Sodium-cooled Fast Reactor (PGSFR). For this purpose, the transport model developed for Experimental Breeder Reactor II and Fast Flux Test Facility was updated and validated. The tritium production rate from ternary fission and activations in PGSFR were calculated by the Argonne Reactor Computation (ARC) code suite for fast reactor analysis, and due to the potential uncertainties in source terms and variations of cold trap performance and material permeability during reactor operation, the tritium distributions and tritium effluents were estimated for several bounding cases. The upper bounding gaseous and liquid tritium release rates from PGSFR are ~ 440.7 Ci/year and 14.7 Ci/year, respectively. The gaseous tritium effluent is about few orders of magnitude smaller than the regulatory constraint of the tritium concentration in the air. The peak tritium concentration in the steam-generator feedwater is ~ 1.1x10-3 μCi/g, which is required to be diluted prior to releasing as tritiated water. The accumulated tritium in the cold trap after five cycle operations is ~ 52,000 Ci. Because of significant conservatism in these upper bounding estimations, it is expected that the tritium effluent at the nominal case with plausible source terms would be about 2 orders of magnitude smaller than that of the upper bounding case.

Lithium-lead corrosion behavior of zirconium oxide coating after heavy-ion irradiation

For a strict control of tritium migration in fusion reactor fuel systems, tritium permeation barrier coatings have been developed for several decades. In liquid blanket concepts, corrosion of the coatings by liquid tritium breeders is a serious concern in addition to tritium permeation. Furthermore, the coatings are exposed to high-energy neutrons in an actual reactor, which would bring a synergy of irradiation and corrosion. In this study, zirconium oxide (ZrO2) coatings prepared by metal organic decomposition were exposed to static lithium-lead (Li-Pb) for 500 h at 500−600 ºC after iron-ion irradiation with the damage concentration of up to 16 displacement per atom at room temperature and 500 ºC. Corrosion products formed on all the exposed samples, and their amount and stability depended on the damage density and temperature during irradiation and Li-Pb exposure. Voids were formed by the irradiation and segregated in grain boundaries of the ZrO2 coating. The thicknesses of irradiated coatings decreased with damage concentration; therefore, Li-Pb corrosion of the coatings would be promoted by microstructure change caused by heavy-ion irradiation. These results suggest that the Li-Pb compatibility of the ZrO2 coatings would be degraded under irradiation environments.

Numerical study on tritium dispersion in coastal waters: The case of Hangzhou Bay, China

Identification of tritium dispersion characteristics is critically important to recognize the environmental impact due to discharge of radioactive liquid effluents from nuclear industry. For the first time, this study simulated the dispersion behavior in the coastal waters of tritium discharged from Qinshan nuclear power base, where the sole heavy water reactor is located and the largest tritium amount was discharged in China. Both the flow field and tritium concentration distribution in Hangzhou Bay were displayed under normal and accidental discharge conditions considering the realistic water dynamic characteristics. Effects of river and tidal flow on tritium dispersion, effects of tritium evaporation factor, as well as public radiation dose were specially discussed. The results show that tidal flow is the major factor in deciding tritium dispersion for the Hangzhou Bay case. River flow could also affect tritium dispersion, particularly in the southwest region of the discharge point. Under normal discharge condition, the maximum tritium concentration in the vicinity of Qinshan nuclear power base is at 104 Bq/m3 level, which is an order of magnitude higher than the natural background tritium. Under accidental discharge condition, maximum tritium concentration decreases along with an increasing tritium dispersion time. Results furthermore indicate that a considerable tritium evaporation fraction appears for the Hangzhou Bay case. In spite of this, radiation dose involving ingestion and inhalation pathways associated with liquid tritium discharge is still much lower than the regulatory dose limit. This study is expected to recall attentions to tritium dispersion characteristics apart from providing fundamental simulation data for tritium distribution in the waters of Hangzhou Bay.

Corrosion tests of multi-layer ceramic coatings in liquid lithium-lead

Tritium permeation through structural materials in fusion reactor blanket systems is a critical issue from the viewpoints of fuel loss and radiological safety. Ceramic coatings have been investigated to prevent tritium permeation; however, corrosion of the coatings by blanket materials, especially corrosive liquid tritium breeders such as lithium-lead is serious. In our previous study, the improvement of permeation reduction performance using an erbium oxide (Er2O3)-zirconium oxide (ZrO2) two-layer coating was confirmed, but it did not show substantial corrosion resistance. In this study, various multi-layer coatings were fabricated by metal organic decomposition and then exposed to lithium-lead under static conditions to investigate precise corrosion behaviors of multi-layer coatings. After lithium-lead exposure for 1000 h at 600 °C, the outermost Er2O3 layer almost disappeared, while the outermost ZrO2 layer remained, indicating that ZrO2 is more suitable as the outermost layer. However, many cracks and peelings were observed on the outermost ZrO2 layer in the cases of the samples having four coating interfaces. The optimization of layer combination and the control of adhesion between the coatings are required to reduce the degradation of multi-layer coatings.

Experimental study on corrosion-erosion behavior of ITER blanket structure materials in flowing Pb-17Li

Abstract A rotating corrosion test device has been developed to study the compatibility between liquid tritium breeder material and structure materials for fusion facility. The corrosion behaviors of three samples made of reduced activation ferritic/martensitic steel (RAFMS), 316 L stainless steel and oxide dispersion strengthened steel (ODS), are studied in liquid Pb-17Li at 500℃ for an exposure of 1000 h. The weight changes of the specimens are obtained before and after test. The analyses of surface morphology and component are characterized by SEM, EDS and XRD methods. Results show that weight-loss rate of RAFMS is lower than that of 316 L and ODS in the experiment. It indicates that the corrosion mechanism is governed by some or all dissolution of alloy elements (such as Cr, Ni) from the sample surface into liquid Pb-17Li and the exfoliation from the corroded surface by the flowing Pb-17Li. Specifically, the dispersed Al2O3 particles in ODS is easy to be corroded in liquid Pb-17Li and then provide corrosion points to accelerate dissolution of Ni-Fe austenite. The compatibility of RAFMS with liquid Pb-17Li is slightly better than that of 316 L and ODS in the current experimental period.

Microstructure change and deuterium permeation behavior of ceramic-metal multi-layer coatings after immersion in liquid lithium-lead alloy

For the establishment of liquid tritium breeding concepts, static lithium-lead corrosion tests for single-layer erbium oxide coatings and erbium oxide-metal multi-layer coatings were carried out, followed by deuterium permeation measurements. Grain boundary corrosion of erbium oxide coatings was confirmed by the static immersion tests at 550 and 600 °C. An erbium oxide-iron two-layer coating sustained its structure after lithium-lead immersion at 600 °C for up to 3000 h. The results of gas-driven deuterium permeation measurements for multi-layer coatings immersed at 550 and 600 °C indicated that crystallization and grain growth of erbium oxide would sufficiently occur during the immersion at 600 °C, and then the sample showed lower permeabilities in the first measurements at lower temperature. On the other hand, permeation reduction factors of the sample immersed at 550 °C were estimated to be 100‒200 in the temperature range of 400‒550 °C. A corrosion layer formed on the coating surface might work as an additional diffusion barrier after the permeation test at 600 °C.

Hypothetical daily operation model of fuel cycle in tritium plant

This paper describes a design model for the fuel cycle of a tritium plant; the model considers both the fueling line and the Neutral Beam Injector line. The fuel cycle typically consists of a 16-h of burn-and-dwell operations followed by an 8-h silent shift, so the behavioral changes of fuel cycle of the tritium plant are influenced by both an hourly cycle of burn-and-dwell operations and 24-h cycle. To provide a tool to optimize the tritium inventory level, and to gain insight into the behavior changes of the fuel cycle, a mathematical model is developed. The model mainly focuses on the Isotope Separation System, which holds the largest amount of gaseous and liquid tritium. The optimization problem is formulated as a mixed integer linear program model with linear constraints. An inductive operation scenario is presented to illustrate the applicability of the proposed model. The obtained minimum tritium working inventory in ISS is 103.498 g in the inductive operation scenario.

Lithographic Printing Via Two-Photon Polymerization of Engineered Foams

Understanding deuterium-tritium mix in capsules is critical to achieving fusion within inertial confined fusion experiments. One method of understanding how the mix of hydrogen fuels can be controlled is by creating various structured deuterated foams and filling the capsule with liquid tritium. Historically, these materials have been a stochastically structured gas-blown foam. Later, to improve the uniformity of this material, pore formers have been used which are then chemically removed, leaving behind a foam of monodisperse voids. However, this technique is still imperfect in that fragments of the pore templating particles may not be completely removed and the void distribution may not be uniform over the size scale of the capsule. Recently, advances in three-dimensional printing suggest that it can be used to create microlattices and capsule walls in one single print. Demonstrated here are proof-of-concept microlattices produced using two-photon polymerization with submicrometer resolution of various structures as well as a microlattice-containing capsule. With this technology, complete control of the mixing structure is possible, amenable to modeling and easily modified for tailored target design.

Transfer of Tritium Released into the Marine Environment by French Nuclear Facilities Bordering the English Channel

Controlled amounts of liquid tritium are discharged as tritiated water (HTO) by the nuclear industry into the English Channel. Because the isotopic discrimination between 3H and H is small, organically bound tritium (OBT) and HTO should show the same T/H ratio under steady-state conditions. We report data collected from the environment in the English Channel. Tritium concentrations measured in seawater HTO, as well as in biota HTO and OBT, confirm that tritium transfers from HTO to OBT result in conservation of the T/H ratio (ca. 1 × 10(-16)). The kinetics of the turnover of tritium between seawater HTO, biota HTO, and OBT was investigated. HTO in two algae and a mollusk is shown to exchange rapidly with seawater HTO. However, the overall tritium turnover between HTO and the whole-organism OBT is a slow process with a tritium biological half-life on the order of months. Nonsteady-state conditions exist where there are sharp changes in seawater HTO. As a consequence, for kinetic reasons, the T/H ratio in OBT may deviate transiently from that observed in HTO of samples from the marine ecosystem. Dynamic modeling is thus more realistic for predicting tritium transfers to biota OBT under nonsteady-state conditions.

R&amp;D activities of the liquid breeder blanket in Korea

Abstract A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the common fields of a solid TBM such as design tools, structural material, fabrication methods, and He cooling technology to support this concept for the ITER. Also, other fields such as a liquid breeder technology and tritium extraction have been developed from the designed liquid TBM. For design tools, system codes for safety analysis such as Multi-dimensional Analysis of Reactor Safety (MARS) and GAs Multi-component Mixture Analysis (GAMMA) were developed for He coolant and liquid breeder. For the fabrication methods, Ferritic Martensitic Steel (FMS) to FMS and Be to FMS joinings with a Hot Isostatic Pressing (HIP) were developed and verified with a high heat flux test of up to 0.5–1.0 MW/m 2 . Moreover, three mockups were successfully fabricated and a 10-channel prototype is being fabricated to make a rectangular channel FW. For the integrity of the joining, two high heat flux test facilities were constructed, and one using an electron beam has been constructed. With the 6 MPa nitrogen loop, a basic heat transfer experiment for code validation was performed. From the verification of the components such as preheater and circulator, a 9 MPa He loop was constructed, and it supplies high temperature (500 °C) and pressure (8 MPa) He to the high heat flux test facility. For an electromagnetic (EM) pump development for circulating the liquid breeder, magnetohydrodynamic (MHD) experiment, and flow corrosion test, a PbLi breeder loop was constructed. From the performance test, the EM pump and magnet show their capability, and flow and static corrosion tests including oxide coating for corrosion protection were performed. For tritium extraction from the liquid breeder, a gas–liquid contact method was adopted and a tritium extraction chamber was constructed. For measurement of the tritium amount in the liquid breeder, permeation sensors have been developed.