Analysis of diffusion and dissolution of two-component hydrogen (H + D) in lead lithium

A lead–lithium eutectic alloy (Pb–Li) is one of the most promising candidate materials for the liquid blanket of an advanced fusion reactor. We have experimentally determined mass-transfer properties by an unsteady permeation method, which data are necessary to design a system to recover tritium (T) from a Pb–Li blanket. An experiment of simultaneous H and D permeation through Li₁₇Pb₈₃ is performed to clarify interactions between atoms in the two-component permeation process. The experimental results are analyzed by a model of one-dimensional or two-dimensional permeation through Li₁₇Pb₈₃. The major permeation proceeds in the longitudinal direction of the present system, and the ratio of hydrogen leak in the radial direction is evaluated using the simulation. As a result, it was found that H and D atoms permeate independently regardless of the H/D component ratio within the present experimental conditions. The permeability and diffusivity of H are 1.4 times higher than that of D. The solubility of H is close to that of D. The isotope effect in diffusivity is in proportion to the square root of the mass ratio of D to H. When these data can be extended to the case of T, T permeability and diffusivity is predicted as 1/1.7 times lower than that of H in the temperature range from 773 K to 973 K.     

Experimental study of efficiency of natural oxide layers for reduction of tritium permeation through Eurofer 97

Tritium permeation from the breeder through the helium coolant is a fundamental safety issue in the design of the HCLL (Helium coolant lithium lead) blanket system. The permeation of hydrogen isotopes through Eurofer in different conditions was deeply studied in the past, demonstrating that it is necessary to reduce this amount using tritium permeation barriers (TPB). A strong effort has been made to select the best technological solution for the realisation of tritium permeation barriers on complex structures not directly accessible after the completion of the manufacturing process, but after many years of activity for the qualification of different materials acting as TPBs it was demonstrated that these technologies are not yet mature for nuclear applications. An easier solution was identified in the nucleation and growth of natural oxides on the helium-exposed surface of cooling system components. The major objective of this work is the evaluation of the Permeation Reduction Factor (PRF) of natural oxides on Eurofer steel adding a known content of water and hydrogen to argon, used in substitution of helium. The PRF was measured on disk shaped specimens, in gas phase, using the PERI II apparatus, at a temperature of 550 °C. The oxide layer was produced in situ, in a well-defined range of hydrogen on water ratios. The obtained results are presented and discussed.     

Tritium management and anti-permeation strategies for three different breeding blanket options foreseen for the European Power Plant Physics and Technology Demonstration reactor study

In DT fusion reactors like DEMO, the commonly accepted tritium (T) losses through the steam generator (SG) shall not exceed about 2 mg/d that are more than 5 orders of magnitude lower than the T production rate of about 360 g/d in the breeding blanket (BB). A very effective mitigation strategy is required balancing the size and efficiency of the processes in the breeding and cooling loops, and the availability and efficiency of anti-permeation barriers. A numerical study is presented using the T permeation code FUS-TPC that computes all T flows and inventories considering the design and operation of the BB, the SG, and the T systems. Many scenarios are numerically analyzed for three breeding blankets concepts – helium cooled pebbles bed (HCPB), helium cooled lithium lead (HCLL), and water cooled lithium lead (WCLL) – varying the T processes throughput and efficiency, and the permeation regimes through the BB and SG to be either surface-limited or diffusion-limited with possible permeation reduction factor. For each BB concept, we discuss workable operation scenarios and suggest specific anti-permeation strategies.     

Tritium permeation in EUROFER in EXOTIC and LIBRETTO irradiation experiments

The reduced activation martensitic steel (RAFM) EUROFER is foreseen as a structural material in test breeder module (TBM) in ITER and breeder blanket in DEMO design. In a number of irradiation experiments conducted in high flux reactor (HFR) in Petten EUROFER was used as a containment wall of the breeder material, through which tritium permeation was monitored on line. Thus in EXOTIC-9/1 (EXtraction Of Tritium In Ceramics) experiment where Li₂TiO₃ pebbles were the breeder material, EUROFER was irradiated up to 1.3 dpa at 340–580 °C. In LIBRETTO experiments (LIBRETTO-4/1, -4/2 and -5) the breeder material was lead lithium eutectic which was in direct contact with the EUROFER containment wall. The neutron damage in steel achieved in the LIBRETTO experiments varied from 2 to 3.5 dpa. The irradiation temperature was 350 °C (LIBRETTO-4/1), 550 °C (LIBRETTO-4/2), and 300–500 °C (LIBRETTO-5).Tritium permeability was studied by varying the irradiation temperature and hydrogen concentration in the purge gas. From the analysis of the temperature transients performed in all four experiments yielded the tritium diffusion coefficients were derived, which appear to be factor ten lower than the literature data obtained in the gas driven permeation experiments.     

Solid solubility of hydrogen in Pb83Li17 alloy: A preliminary investigation

Lead–lithium alloy Pb₈₃Li₁₇ (0.6 wt.% lithium) is a potential candidate to be used as a neutron multiplier, tritium (fuel) breeder and heat transfer agent (coolant) in the International Thermonuclear Experimental Test Reactor (ITER). The tritium produced in the alloy could be soluble in the alloy or appear as a new phase. During reactor shut down condition, Pb₈₃Li₁₇ will be solidified and stored in a storage tank. Investigation on the solubility of tritium in solid Pb₈₃Li₁₇ is essential to quantify the trapped tritium in solid Pb₈₃Li₁₇ alloy to estimate the extent of radioactive contamination (with respect to tritium) and valuable tritium loss. Tritium being the isotope of hydrogen behaves more or less similar to hydrogen. In the present study solid-solubility of hydrogen in Pb₈₃Li₁₇ alloy has been investigated as a function of temperature and pressure. It was found that the hydrogen solubility increases with temperature (373–473 K) and follows the Sieverts law. Hydrogen solution enthalpy has been calculated using Seiverts constant and found to be −4.81 kJ/mole of hydrogen.     

Construction and initial operation of MHD PbLi facility at UCLA

A magnetohydrodynamic flow facility MaPLE (Magnetohydrodynamic PbLi Experiment) that utilizes molten eutectic alloy lead–lithium (PbLi) as working fluid has been constructed and tested at University of California, Los Angeles. The loop operation parameters are: maximum magnetic field 1.8 T, PbLi temperature up to 350 °C, maximum PbLi flow rate with/without a magnetic field 15/50 l/min, maximum pressure head 0.15 MPa. The paper describes the loop itself and its major components, basic operation procedures, experience of handling PbLi, initial loop testing, flow diagnostics and current and near-future experiments. The obtained test results of the loop and its components have demonstrated that the new facility is fully functioning and ready for experimental studies of magnetohydrodynamic, heat and mass transfer phenomena in PbLi flows and also can be used in mock up testing in conditions relevant to fusion applications.     

R&D activities of the liquid breeder blanket in Korea

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/m². 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.     

Evaluating the hydrogen isotope absorption/diffusion coefficient of CVD-SiC at high temperature

Deuterium diffusion coefficient measurements of CVD-SiC were carried out using a solubility and diffusivity measurement apparatus to investigate the permeation mechanism of the hydrogen isotope through CVD-SiC. Experiments were conducted with thin-sheet-type samples with thicknesses of 0.1 mm, 1 mm, and 2 mm at 1073–1183 K. Total amount of occluded gas into or released gas from different thickness but same weight sample were expected to be the same, but unexpectedly differed by more than 50%. As the release rates after sufficient time had passed were almost the same, and the 1-mm-thick sample had twice the surface area of the 2-mm-thick sample, the measurements were probably affected by adsorbed gas on the surface. The value of D/L² (the diffusion coefficient divided by the square of the thickness), obtained by fitting to the theoretical formula but ignoring the early phase of discharge, was in good agreement for samples of different thickness at the same temperature, and was more than 5 orders of magnitude smaller than that obtained from the permeability measurement experiments. Therefore, we believe that the deuterium permeation through CVD-SiC is primarily dependent on the permeation rate through the grain boundaries.     

The engineering sizing of the packed desorption column of hydrogen isotopes from Pb–17Li eutectic alloy. A rate based model using experimental mass transfer coefficients from a Melodie loop

The model of the desorption of hydrogen isotopes from lead lithium alloy in a packed column is derived from the first principles using the plug flow in the liquid phase either the plug flow or ideal mixing in the gas phases. Sievert's law of non-linear equilibrium is followed. The volumetric mass transfer coefficient k_La and its dependence on the liquid metal flow rate are evaluated on the basis of the Melodie loop experiments.The presented model is used for evaluation of the minimum flow rate of the purge gas for which the concentration of the isotope in the gas leaving the column is at its highest, while the driving force of the interfacial transport of the isotope is still not reduced and the tritium desorption efficiency is therefore retained. The potential effect of the axial dispersion in the gas and liquid phase is evaluated. Highlighted are the issues of the optimum packing geometric surface area, above which the efficiency starts to decrease, and of the role of the surface tension and the contact angle with regard to the wettability of the packing. On the basis of the findings related to these factors, the Mellapak 500 Y and Mellapak packings with flat surfaces are recommended for the tests aiming to intensify the tritium desorption efficiency in the packed columns. The models were used for the engineering sizing of the packed columns in two breeding blanket concepts for the DEMO plant – utilizing DCLL (dual coolant lead lithium) and HCLL (helium cooled lithium lead).     

The EU TBM systems: Design and development programme

This paper presents the status of the design and of the development programme of the two test blanket systems (TBSs) based on the blanket concepts supported by the EU, namely the helium cooled lithium lead (HCLL) and helium cooled pebble bed (HCPB) concepts.Both the test blanket modules (TBMs) box design and the associated systems (Helium Cooling Systems, PbLi loop for the HCLL system, helium processing systems for tritium extraction, etc.) have been revised and, where needed, modified according to the assumption that one ITER equatorial port could be available for testing the two European test blanket modules (TBMs).According to EU TBMs programme, two reliable test blanket systems shall be ready for installation on the first day of ITER operation. In order to comply with this ambitious objective, six EURATOM associates who have sustained the TBM program so far have joined themselves in a consortium aiming to ensure an efficient management of the project tasks and exploit specific competences enhancing potential synergies. The consortium objectives and development programme are summarised in the paper.     

Metallurgical analysis of corroded bellows of bellow-sealed valve in lithium flow system

Liquid lithium has been one of the candidates of the tritium breeder and possibly a coolant for the blanket of fusion reactors. The observation of corrosion behavior was conducted on the bellows of the bellow-sealed valve used in a lithium circulation loop at 350–500 °C for about 1500 h. The results were obtained from observation of the surfaces and cross sections. The bellows was found to be fractured and detached. Selective elements were depleted on the surface.     

A model for tritium transport in fusion reactor components: The FUS-TPC code

Hydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of atomic hydrogen in materials in which hydrogen and its isotopes are present. In this work the problem of tritium transport from lead–lithium breeder through different heat transfer surfaces to the environment has been studied and analyzed by means of a computational code. The code (FUS-TPC) is a new fusion-devoted version of the fast-fission one called Sodium-Cooled Fast Reactor Tritium Permeation Code (SFR-TPC). The main features of the model inside the code are described. A simulation, using the code, was performed by adopting the configuration of the European configuration of the Helium Cooled Lead Lithium (HCLL) blanket for DEMO.     

Numerical study of corrosion of ferritic/martensitic steels in the flowing PbLi with and without a magnetic field

A computational suite called TRANSMAG has been developed to address corrosion of ferritic/martensitic steels and associated transport of corrosion products in the eutectic alloy PbLi as applied to blankets of a fusion power reactor. The computational approach is based on simultaneous solution of flow, energy and mass transfer equations with or without a magnetic field, assuming mass transfer controlled corrosion and uniform dissolution of iron in the flowing PbLi. First, the new tool is applied to solve an inverse mass transfer problem, where the saturation concentration of iron in PbLi at temperatures up to 550 °C is reconstructed from the experimental data on corrosion in turbulent flows without a magnetic field. As a result, a new correlation for the saturation concentration C^S has been obtained in the form C^S = e^{13.604–12975/T}, where T is the temperature of PbLi in K and C^S is in wppm. Second, the new correlation is used in the computations of corrosion in laminar flows in a rectangular duct in the presence of a strong transverse magnetic field. As shown, the mass loss increases with the magnetic field such that the corrosion rate in the presence of a magnetic field can be a few times higher compared to purely hydrodynamic flows. In addition, the corrosion behavior was found to be different between the side wall of the duct (parallel to the magnetic field) and the Hartmann wall (perpendicular to the magnetic field) due to formation of high-velocity jets at the side walls. The side walls experience a stronger corrosion attack demonstrating a mass loss up to 2–3 times higher compared to the Hartmann walls. Also, computations of the mass loss are performed to characterize the effect of the temperature (400–550 °C) and the flow velocity (0.1–1 m/s) on corrosion in the presence of a strong 5 T magnetic field prototypic to the outboard blanket conditions.     

Penetration of tritiated water vapor through hydrophobic paints for concrete materials

Behavior of tritium transfer through hydrophobic paints of epoxy and acrylic-silicon resin was investigated experimentally. The amounts of tritium permeating through their paint membranes were measured under the HTO concentration condition of 2–96 Bq/cm³. Most of tritium permeated through the paints as a molecular form of HTO at room temperature. The rate of tritium permeating through the acrylic-silicon paint was correlated in terms of a linear sorption/release model, and that through the epoxy paint was controlled by a diffusion model. Although effective diffusivity estimated by a diffusion model was obtained 1.1 × 10⁻¹³–1.8 × 10⁻¹³ m²/s for epoxy membranes at the temperature of 21–26 °C, its value was found to be hundreds times larger than that for cement-paste coated with epoxy paint. Hence, resistance of tritium diffusion through interface between cement-paste and the epoxy paint was considered to be the most effective in the overall tritium transfer process. Clarification of tritium transfer behavior at the interface is important to understand the mechanism of tritium transfer in concrete walls coated with various paints.     

Development of a hydrogen permeation sensor for future tritium applications

Tritium monitoring in lithium–lead eutectic is of great importance for the performance of liquid blankets in fusion reactors. In addition, tritium measurements will be required in order to proof tritium self-sufficiency in liquid metal breeding systems. On-line hydrogen (isotopes) sensors must be design and tested in order to accomplish these goals.In this work, an experimental set up was designed in order to test the permeation hydrogen sensors at 500 °C. This experimental set-up allowed working with controlled environments (different hydrogen partial pressures) and the temperature was measured using a thermocouple connected to a temperature controller that regulated an electrical heater.In a first set of experiments, a hydrogen sensor was constructed using an α-iron capsule as an active hydrogen area. The sensor was mounted and tested in the experimental set up. In a second set of experiments the α-iron capsule was replaced by a welded thin palladium disk in order to minimize the death volume. The experiments performed using both membranes (α-iron and palladium) showed that the operation of the sensors in the equilibrium mode required at least several hours to reach the hydrogen equilibrium pressure.     

Conceptual design of a water cooled breeder blanket for CFETR

China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by ⁶Li(n,α)T reaction. Li₂TiO₃ pebbles and Be₁₂Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li₂TiO₃ and Be₁₂Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be₁₂Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into account to replace Be plate in viewpoint of safety. In this contribution, study on neutronics and thermal design for a water cooled breeder blanket with superheated steam is reported.     

First-principles study of hydrogen adsorption and permeation in reconstructed cubic erbium oxide surfaces

Understanding surface properties of Er₂O₃, especially in relation to adsorption and permeation of atomic hydrogen, is of considerable importance to the study of tritium permeation barriers. In this work, hydrogen diffusion pathways through the low-index (1 0 0), (1 1 0), and (1 1 1) surfaces of cubic Er₂O₃ have been calculated using density functional theory within the GGA (PBE) + U approach. The dependence of the effective U parameter on lattice constants, bulk moduli, and formation energies of Er₂O₃ has been investigated in detail. The energetics of hydrogen penetration from the surfaces to the solution site in bulk Er₂O₃ were defined using the optimum effective U value of 5.5 eV. For a low surface coverage of hydrogen (0.89 × 10¹⁴ H/cm²), a penetration energy of at least 1.7 eV was found for all the low-index erbium oxide surfaces considered. The results of the present study will provide useful guidance for future studies on modeling defects, such as grain boundaries and vacancies, in tritium permeation barriers.     

Fabrication and characterization of Li3TaO4 ceramic pebbles by wet process

Lithium-containing ceramics have long been recognized as the tritium breeding materials in the fusion–fission or fusion reactor blanket. Li₃TaO₄ (lithium orthotantalate) pebbles, with high melting point (～1406 °C), good thermal stability, and high thermal conductivity, were fabricated by wet process (freeze–drying) as a new potential candidate of tritium breeder. The diameter of ceramic pebbles is 0.7–1.0 mm, density is over 90% (TD), pore diameter is 1.86 μm (a.v), grain size is 15 μm (a.v), crush load is up to 46.7 N (a.v).     

Investigating permeation and transport of H isotopes in tungsten by first-principles

We have investigated permeation and transport of hydrogen (H) isotopes in tungsten (W) single crystal employing first-principles calculations in junction with Fick’ law. Permeability was approximately evaluated according to the solubility and diffusion coefficient of H. The solubility for H in bulk W from present calculation is consistent with the experimental results measured by Frauenfelder. The permeation fluxes of H isotopes are examined at the different thickness of W wall. The permeation fluxes of deuterium with the W thickness of 21 μm at the temperature of 770 K and with the W thickness of 50 μm at the temperature of 893 K were 0.68 × 10¹³ atom/m²s and 0.34 × 10¹⁴ atom/m²s, respectively. The dissociation coefficients of H isotopes are also evaluated. We believe that the present first-principles combined with Fick’ law method can be also generalized to investigate permeation and transport of H isotopes in most metals since such H isotopes behaviors in most metals are similar to those of H isotopes in W.     

Transport of hydrogen and deuterium in the reduced activation martensitic steel ARAA

Advanced reduced activation alloy (ARAA) is a reduced activation ferritic/martensitic (RAFM) steel under development at the Korea Atomic Energy Research Institute. The transport of hydrogen and deuterium in ARAA was investigated in an elevated temperature range of 250–600 °C. A continuous-flow method, a time-dependent gas-phase technique, was used for the measurements. Complete sets of transport parameters (permeability, diffusivity, solubility, trap site density, and trapping energy) of hydrogen and deuterium in ARAA were successfully obtained. We show that appreciable trapping effects are observed only at low temperatures (250–350 °C) and that the isotope effect ratio for the diffusivity differs from the classical prediction. However, the measured values of permeability, effective diffusivity, and effective solubility of ARRA were within the range of results reported for other RAFM steels.