Development of a plasma driven permeation experiment for TPE

Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1], [2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes.     

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.     

Development of the IFMIF Tritium Release Test Module in the EVEDA phase

This paper presents the engineering design of the IFMIF (International Fusion Materials Irradiation Facility) Tritium Release Test Module (TRTM). The objectives of the TRTM are: (i) in-situ measurements of the tritium released from lithium ceramics and beryllium pebble beds during irradiation, (ii) studying the chemical compatibility between lithium ceramics and structural materials under irradiation, and (iii) performing post irradiation examinations for the irradiated materials. The TRTM has eight rigs which are arranged in two rows (2 × 4) perpendicular to the beam axis and enclosed by a structural container. Each rig includes one capsule that contains lithium ceramic or beryllium pebbles for irradiation. Neutrons reflectors are implemented at different locations to reflect the scattered neutrons back to the active region aiming to improve the tritium production. The TRTM is required to provide irradiation temperature range of 400–900 °C for the capsules filled with lithium ceramics and 300–700 °C for the ones packed with beryllium. The engineering design of the TRTM components such as container, rigs, capsules, pebble beds, neutrons reflectors, and purge gas and coolant tubes are presented. In addition a test matrix for the irradiation campaign is proposed.     

Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3–148 dpa at 378–504 °C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 °C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa √m occurred in room temperature tests when irradiation temperature was below 400 °C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa √m was measured when the irradiation temperature was above 430 °C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3–148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 °C) irradiation cases, which indicates that the ductile–brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.     

Quantifying helium distribution in dual-ion beam irradiated SiCf/SiC composites by electron energy loss spectroscopy

In this study, we report a method to quantify the helium distribution in the SiC_f/SiC composites, which are used as the first-wall materials of fusion reactor. The helium-bubble formation in Hi-Nicalon Type-S (HNS) was observed in the irradiated SiC_f/SiC composites at a level of 100 dpa and at 800 °C and 1000 °C, respectively. We applied transmission electron microscopy and electron energy loss spectroscopy to investigate the helium-gas-bubbles-formation mechanisms. To simulate the practical first-wall environment of Deuterium–Tritium (D–T) fusion reactor, a dual-ion beam (6 MeV Si³⁺ and 1.13 MeV He⁺) was performed to irradiate the SiC_f/SiC composites. The relationship between the energy shift of He K-edge and the radius of the bubble of the SiC composites was estimated by electron energy loss spectroscopy analysis. The results show that all of the helium atoms irradiated at 1000 °C and formed the bubbles. On the other hand, at 800 °C, only 25.5% of the helium atoms form the helium bubbles. A clear thermal-dependent formation mechanism is found.     

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.     

Study of ceramic pebble beds in Post Irradiation Examination of the Pebble Bed Assemblies irradiation experiment

The irradiation experiment Pebble Bed Assemblies (PBA) consists of four mock-up representations (test elements) of the EU Helium Cooled Pebble Bed (HCPB) concept. The four test elements contain a ceramic breeder pebble bed sandwiched between two beryllium pebble beds and are regarded as one of the first DEMO representative HCPB blanket irradiation tests, with respect to temperatures and power densities. The design value of the PBA were to irradiate pebble beds at a power density of 20–26 W/cc in the ceramic breeder, to a maximum temperature of 800 °C.Two test elements contain lithium orthosilicate pebbles (Li₄SiO₄; FZK/KIT) and were irradiated with target temperatures of 600 and 800 °C, respectively. The other test elements have lithium metatitanate (Li₂TiO₃; CEA) with different grain sizes and were both irradiated with a target temperature of 800 °C. The PBA have been irradiated for 294 Full Power Days (12 cycles) in the High Flux Reactor (HFR) in Petten to a total neutron dose of 2–3 dpa in Eurofer, and an estimated (total) lithium burnup of 2–3% in the ceramic pebbles.This work presents results of Post Irradiation Examinations (PIE) on the four HCPB test elements. Using e.g. SEM, the evolution of compressed pebble beds and pebble interactions like swelling, creep, sintering, etc., under irradiation and thermal loads are studied for the candidate pebble materials Li₂TiO₃ and Li₄SiO₄. (Chemical) interactions between ceramic pebbles and Eurofer (e.g. chrome diffusion) are observed. Looking at different sections of the pebble beds, correlations between temperatures and thermal–mechanical behaviour are clearly observed.     

Tritium permeation experiments using reduced activation ferritic/martensitic steel tube and erbium oxide coating

Low concentration tritium permeation experiments have been performed on uncoated F82H and Er₂O₃-coated tubular samples in the framework of the Japan-US TITAN collaborative program. Tritium permeability of the uncoated sample with 1.2 ppm tritium showed one order of magnitude lower than that with 100% deuterium. The permeability of the sample with 40 ppm tritium was more than twice higher than that of 1.2 ppm, indicating a surface contribution at the lower tritium concentration. The Er₂O₃-coated sample showed two orders of magnitude lower permeability than the uncoated sample, and lower permeability than that of the coated plate sample with 100% deuterium. It was also indicated that the memory effect of ion chambers in the primary and secondary circuits was caused by absorption of tritiated water vapor that was generated by isotope exchange reactions between tritium and surface water on the coating.     

Progress in compatibility experiments on lithium–lead with candidate structural materials for fusion in China

Liquid LiPb eutectic is one of the promising candidate tritium breeder materials for fusion reactors. This paper presents the progress in compatibility experiments with liquid LiPb achieved up to now in China for some candidate structural materials. The results showed that CLAM steel had good compatibility with flowing LiPb at 480 °C with the velocity of 0.08 m/s after 5000 h in DRAGON-I loop. On the other hand, after exposed in static LiPb at 700 °C for 500 h in a SiC crucible, the W and Mo specimens suffered much more weight loss compared with Nb specimen, and a thin reaction product layer was visible on the surface of all the three refractory metals. Preliminary analysis on SiC_f/SiC composite specimens indicated that there was no penetration of LiPb and no reaction products on the surface with CVD SiC coating, which showed SiCf/SiC composite were stable and compatible with static LiPb under 700 °C after 500 h exposure.     

Conceptual design of Tritium Extraction System for the European HCPB Test Blanket Module

The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant.In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing.In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.     

Overview of the US–Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten

The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. A recent MIT report (PSFC/RR-10-4, An assessment of the current data affecting tritium retention and its use to project towards T retention in ITER, Lipschultz et al., 2010) summarizes the observations from high-energy ion bombardment studies and illustrates the saturation trend in deuterium concentration due to damage from ion irradiation in tungsten and molybdenum above 1 displacement per atom (dpa). While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in plasma facing components (PFCs), it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high primary knock-on atom (PKA) energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influences the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron-irradiation. Under the framework of the US–Japan TITAN program, tungsten samples (99.99 at.% purity from A.L.M.T. Co.) were irradiated by fission neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory (ORNL), at 50 and 300 °C to 0.025, 0.3, and 2.4 dpa, and the investigation of deuterium retention in neutron-irradiated tungsten was performed in the Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe²⁺, 20 MeV W²⁺, and 700 keV H^?) with that from neutron-irradiated tungsten to identify the similarities and differences among them.     

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).     

An advance process of aluminum rich coating as tritium permeation barrier on 321 steel workpiece

We have proposed an advance three-step process, Al-electroplating in ionic liquid followed by heat treating and selectively oxidation, preparing aluminum rich coating as tritium permeation barrier (TPB). In present work, the advance process was applied to 321 steel workpieces. In the Al-electroplating, pieces were coated by galvanostatic electrodeposition at 20 mA/cm² in aluminum chloride (AlCl₃)–1-ethyl-3-methylimidazolium chloride (EMIC) ionic liquid. The Al coating on those pieces all displayed attractive brightness and well adhered to surface of pieces. Within the aluminizing time from 1 to 30 h, a series of experiments were carried out to aluminize 321 steel pieces with Al 20 μm coating at 700 °C. After heat treated for 8 h, a 30 μm thick aluminized coating on piece appeared homogeneous, free of porosity, and mainly consisted of (Fe, Cr, Ni)Al₂, and then was selectively oxidized in argon gas at 700 °C for 50 h to form Al₂O₃ scale. The finally fabricated aluminum rich coating, without any visible defects, had a double-layered structure consisting of an outer γ-Al₂O₃ layer with thickness of 0.2 μm and inner (Fe, Cr, Ni)Al/(Fe, Cr, Ni)₃Al layer of 50 μm thickness. The deuterium permeation reduction factor, PRF, of piece (Φ 80 × 2, L 150 mm) with such coating increased by 2 orders of magnitude at 600–727 °C. The reproducibility of the process was also showed.     

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.     

Preparation and investigation of aluminized coating and subsequent heat treatment on 9Cr–1Mo Grade 91 steel

Iron aluminide inner coating with alumina top layer is being considered as a potential solution for tritium permeation barrier and mitigating MHD pressure drop for liquid metal blanket concepts in the fusion reactor systems. Hot-dip aluminizing with subsequent heat treatment seems to offer a good possibility to produce aluminized coating with alumina top layer. 9Cr–1Mo Grade 91 steel samples were hot dipped in Al melt containing 2.25 wt% of Si at 750 °C for 3 min. Heat treatment was performed at 650, 750 and 950 °C for 5 h; samples were either air cooled or furnace cooled. Coatings have been evaluated by SEM, EDX, X-ray diffraction, microhardness, scratch adhesion and Raman spectroscopy. The thickness of the layers and phases formed were influenced by the heat treatment adopted. Fe₂Al₅ was the major phase present in the samples heat treated at 650/750 °C, whereas FeAl and α-Fe(Al) primarily made up the outer and inner layers respectively in the samples heat treated at 950 °C. Cooling method deployed affected the hardness. Air cooled samples had comparatively higher hardness than furnace cooled samples. The scratch test showed the adhesion for the samples heat treated at 950 °C was much better as compared to the samples heat treated at 650/750 °C. Raman spectroscopy analysis showed the presence of both α-Al₂O₃ and γ-Al₂O₃ on the surface of the samples heat treated at 950 °C, while Fe₃O₄ was present in the furnace cooled sample only.     

Evaluation of tritium diffusion through the Neutral Beam Injector calorimeter panel

The Neutral Beam Test Facility (NBTF) to be realized in Padoa will test the Neutral Beam Injection (NBI), one of the Heating and Current Drive Systems foreseen for ITER. The NBI is based on the acceleration of hydrogen or deuterium negative ions up to 1 MeV. This work has been aimed at assessing the tritium release from the NBTF in order to provide data for the safety analysis. In particular, the diffusion of the tritium through the neutral beam target material (the CuCrZr alloy calorimeter panels) has been assessed by using literature data of the diffusion coefficient. The tritium generated inside the calorimeter panels moves into both the vacuum and water side: the tritium diffusion flux has been evaluated during the beam-on (200 °C) and the beam-off (20 °C) phases of the NBTF experiments consisting of an interim campaign and a final test. The penetration depth of the tritium through the 2 mm thick CuCrZr alloy material has been also evaluated by using a Monte-Carlo code. As main result, the assessed diffusion flux of tritium during both the beam-on and the beam-off phases are modest. In fact, at the end of the interim campaign (100 days), about the 96% of the all generated tritium (626.5 MBq) exits the calorimeter while the residual tritium inventory (25 MBq) leaves the copper alloy with a diffusion time of about 1 month. At the end of the final test (14 days) about the 99% of the total generated tritium (1.023 × 10⁴ MBq) leaves the copper alloy and the remaining tritium inventory (152.2 MBq) is released by about 32 days. In both the interim campaign and the final test, more than the 99% of the total tritium is transferred into the vacuum side of the calorimeter panel while negligible tritium amounts enter the cooling water system thus showing a very low impact on the environment.     

Ion-irradiation-induced damage of steels characterized by means of nanoindentation

Self-ion irradiation was used to simulate the damage caused by fast neutrons in the austenitic stainless steel SS 304 SA, the ferritic/martensitic steel Eurofer’97 and a Fe–9 at.%Cr model alloy. The irradiation-induced hardness change in the damage layer was evaluated by means of nanoindentation. Three-step irradiations were performed at room temperature and 300 °C up to 1 and 10 dpa. An irradiation-induced hardness change was shown for all materials. No influence of irradiation temperature could be resolved. Irradiation-induced hardening exhibits different fluence dependencies in Eurofer’97 and Fe–9 at.%Cr. While the data indicate a saturation-like behaviour for Fe–9 at.%Cr, an increase of hardness with fluence up to 10 dpa was found for Eurofer’97.     

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.     

Analysis of hardening limits of oxide dispersion strengthened steel

Oxide dispersion strengthened (ODS) ferritic/martensitic (F/M) steels are promising materials for high temperature applications. The hardening limits from room temperature to 1000 °C of one of such steel, ODS EUROFER97, together with the impact of the material production steps, are investigated at a microstructural level by coupling hardness, tensile tests and transmission microscopy, including in situ heating experiments. The oxides, ytttria and complex yttrium titanium oxides, reinforce the material by forming more or less stable obstacles to dislocations, and by promoting grain refinement by pinning grain boundaries. It appears that part of the yttrium titanium oxides particles dissolves from about 600 °C while pure yttria particles are stable at least to 1000 °C in the steel. The concurrent roles of the oxides and the dislocation structure in the hardening are rationalized using the dispersion barrier hardening model. It appears that hardening due to dislocations can overcome the one due to oxides but is more sensitive to temperature than the one due to oxides, and that the main limiting factor is the thermal stability of the oxides.     

Irradiation creep relaxation of void swelling-driven stresses

Swelling-driven-creep test specimens are used to measure the compressive stresses that develop due to constraint of irradiation void swelling. These specimens use a previously non-irradiated 20% CW Type 316 stainless steel holder to axially restrain two Type 304 stainless steel tubular specimens that were previously irradiated in the US Experimental Breeder Reactor (EBR-II) at 490 °C. One specimen was previously irradiated to fluence levels in the void nucleation regime (9 dpa) and the other in the quasi-steady void growth regime (28 dpa). A lift-off compliance measurement technique was used post-irradiation to determine compressive stresses developed during reirradiation of the two specimen assemblies in Row 7 of EBR-II at temperatures of 547 °C and 504 °C, respectively, to additional damage levels each of about 5 dpa. Results obtained on the higher fluence swelling-driven-creep specimen show that compressive stress due to constraint of swelling retards void swelling to a degree that is consistent with active load uniaxial compression specimens that were irradiated as part of a previously reported multiaxial in-reactor creep experiment. Swelling results obtained on the lower fluence swelling-driven creep specimen show a much larger effect of compressive stress in reducing swelling, demonstrating that the larger effect of stress on swelling is on void nucleation as compared to void growth. Test results are analyzed using a recently proposed multiaxial creep-swelling model.