Status of engineering design of liquid lithium target in IFMIF-EVEDA

In International Fusion Materials Irradiation Facility (IFMIF), intense neutron flux (4.5 × 10¹⁷ n/m² s) with a peak energy of 14 MeV are produced by means of two deuteron beams with a total current of 250 mA and maximum energy of 40 MeV that strike a liquid Li target circulating in a Li loop. Major design requirement is to provide a stable Li jet at a speed of 10–20 m/s with a surface wave amplitude on the Li flow less than 1 mm for handling of an averaged heat flux of 1 GW/m² under a continuous 10 MW deuterium beam deposition. The target system consists of a target assembly, a replaceable back-plate, a Li main loop and a Li purification loop. In July 2007, Engineering Validation and Engineering Design Activities (EVEDA) started under Broader Approach. In this paper, status of the engineering design of the IFMIF Li target system performed in 2007/2008 is described. The future EVEDA tasks to develop the target system are also summarized.     

Characteristics of surface oscillation on high speed liquid Li jet

In the International Fusion Materials Irradiation Facility (IFMIF), high speed liquid lithium (Li) wall jet will be used as target irradiated by two deuteron beams of 125 mA at 40 MeV. To obtain knowledge of Li flow behavior, we have been studying on the surface wave characteristics experimentally using the liquid metal Li circulation loop at Osaka University. In this present study, the characteristic of surface oscillation on high speed liquid Li jet were examined. The free surface oscillation of Li flow was measured by an electro-contact probe apparatus, which detects electric contacts between a probe tip and Li surface. It was installed at 175 mm and 15 mm downstream from the nozzle exit to see influence of the initial growth of surface waves. The wave height of free surface waves was obtained from contact signal. While at 15 mm region the flow surface is very smooth covered with small waves in amplitude, the surface waves are developed sufficiently at the 175 mm. In the case of the velocity of 15 m/s, the maximum wave height reaches 4.8 mm. Heat deposition was estimated on the target back-plate with using the present statistical wave data.     

Vacuum sieve tray for tritium extraction from liquid Pb–17Li

Formation of droplet of liquid Li–17Pb released from a nozzle into vacuum was studied for the evaluation of the feasibility as a tritium extraction process. Size of droplets formed from the nozzles was estimated by theoretical and experimental methods. For the theoretical estimation, the non-dimensional comparison of the physical bulk property of liquid Pb–17Li with water (H₂O) at ambient temperature was applied. It was found to be reasonable to apply the Plateau-Rayleigh-Instability theory for the droplet size formula of the fluid Pb–17Li for the nozzle diameter 0.4 mm–1.0 mm, temperature 400 °C–500 °C, at initial velocity of 3 m/s. The experimental results of the droplet size showed good agreement with the theory. This device was used for the parametric study of extraction of deuterium during their free fall in vacuum. The scaling of the device suggests the engineering feasibility of the process.     

Y(BD4)3, an efficient store of deuterium,and impact of isotope effects on its thermal decomposition

Y(BD₄)₃, which stores as much as 16.6 wt.% and 252 kg/m³ D, has been synthesized via high-energy disk milling. The thermal decomposition of Y(BD₄)₃ has been investigated using thermogravimetric and calorimetric analyses combined with the spectroscopic evolved gas analysis. Two major endothermic events corresponding to thermal decomposition could be distinguished in the DSC profile up to 400 °C at ca. 231 and 285 °C, preceded by a phase transition (at ca. 198 °C) from the low-temperature Pa-3 form to a high-temperature polymorph of Y(BD₄)₃ (F-43c). The high-temperature phase forming at the onset of thermal decomposition may be prepared quantitatively by heating of the low-temperature phase to ca. 216 °C followed by rapid quenching.Effects of isotope H→D substitution on various properties of yttrium borohydride have been analyzed. Y(BD₄)₃ constitutes a very efficient low-temperature source of deuterium gas on the laboratory scale.     

Compatibility of erbium oxide coating with liquid lithium–lead alloy and corrosion protection effect of iron layer

Li–Pb compatibility of Er₂O₃ and Er₂O₃-Fe two-layer coatings has been explored for an understanding of corrosion behaviors and effects of the protection layer. The coatings were peeled off after static Li–Pb immersion test at 600 °C due to the degradation of adhesion between the coating–substrate interface. A loss of Er and then subsequent corrosion of Er₂O₃ were shown after immersion at 500 °C for 500 and 1505 h. However, the outer Fe layer played a role to decrease corrosion rate of the coatings by comparing with the results of Er₂O₃ single layer coatings. Deuterium permeation measurements after corrosion tests at 500 °C showed that the Er₂O₃ coatings kept permeation reduction factors of 10²–10³ after 500 h immersion, but seriously degraded after 1505 h immersion. Corrosion mechanisms suggest that corrosion protection properties will be modified by an optimization of the outer Fe layer and a control of oxygen concentration in Li–Pb.     

Assessment of tritium levels in rivers and precipitation in north-western Greece before the ITER operation

The project ITER aims to demonstrate that fusion is the energy source of the future. The prototype Tokamak machine is intended to start operation at about 2019 and tritium is one of the major contaminants that can be accidentally released in the environment. Nowadays environmental tritium levels are of natural origin except in the vicinity of nuclear facilities. The evaluation of background tritium levels is essential in the context of a future possibility of accidental tritium releases. For this purpose and also because of the lack of relevant information, an extended programme of river and rain water sampling was implemented in north-western Greece. Water samples from six major rivers in this area and rain water samples were analysed for tritium content. The rivers under investigation were Aliakmonas River, Pinios River, Arachthos River, Kalamas River, Aoos River and Louros River, which originate from the central Greek mountain range Pindos, and flow to Aegean and Ionian Sea.The tritium concentrations were determined by the Liquid Scintillator Analyser Tri-Carb 3170TR/SL. The statistical analysis of data revealed that there is a seasonal variation of tritium concentration in rain samples and a less pronounced seasonal variation in river samples. The weighted mean tritium concentration for rain samples was determined equal to 0.90 ± 0.08 Bq L^?1 (7.6 ± 0.7 TU) and the respective mean value for river samples was 0.94 ± 0.04 Bq L^?1 (7.9 ± 0.3 TU). Further analysis has demonstrated that river waters tend to show lower tritium concentrations than the concurrently measured tritium concentrations in rain samples, during spring and summer (at 47% and 71% of the sampling stations, respectively), while this observation is reversed during autumn and winter (at 44% and 35% of the sampling stations, respectively). This may be attributed to rain water remaining underground for a long period allowing tritium to decay and when it reappears as river water, the tritium concentration values are lower when compared to the rain water concentrations. Rough estimates of the residence time of underground waters in the study area provided values, which ranged from 0.5 to 11.7 years, with a mean value of 5.2 ± 0.9 years.     

Uptake of deuterium in partially D-depleted residual codeposits left on DIII-D tiles after laboratory thermo-oxidation experiments

Recent evidence has shown that tokamak carbon-based codeposits may become partially or fully depleted of hydrogen through thermo-oxidation, as the hydrogen content of the codeposits is removed more rapidly than the carbon content. In this study we examine the ability of such partially-depleted residual DIII-D divertor codeposits to uptake deuterium upon subsequent exposure to deuterium gas or deuterium plasmas. The partially D-depleted specimens used here were obtained from a previous study where DIII-D codeposits were oxidized for 2 h at 623 K (350 °C) and 267 Pa (2 Torr) O₂ [J.W. Davis et al., Thermo-oxidation of DIII-D codeposits on open surfaces and in simulated tile gaps, J. Nucl. Mater. 415 (2011) S789–S792]. In the present study some of these specimens, having undergone prior oxidation, were exposed to D₂ glow discharge plasmas or D₂ gas at 20 kPa (150 Torr) at 300 or 523 K. In the case of plasma exposure, no uptake of D was observed, while an increase in D content was seen following D₂ gas exposures. When the gas exposure took place at 300 K, heating the specimens in vacuum to 623 K for 15 min led to the release of all of the increased D content. For the gas exposure at 523 K, the increase in D content was found to require longer (8 h) vacuum baking to remove. However, in a reference codeposit specimen (from a closeby location on the tile), which had not been previously oxidized, there was a similar increase in D content following D₂ exposure at 523 K, but it could not be released even following 8 h vacuum baking at 623 K.     

Preparation,characterisation and dissolution of a CeO2 analogue for UO2 nuclear fuel

The behaviour of spent nuclear fuel under geological conditions is a major issue underpinning the safety case for final disposal. This work describes the preparation and characterisation of a non-radioactive UO₂ fuel analogue, CeO₂, to be used to investigate nuclear fuel dissolution under realistic repository conditions as part of a developing EU research programme. The densification behaviour of several cerium dioxide powders, derived from cerium oxalate, were investigated to aid the selection of a suitable powder for fabrication of fuel analogues for powder dissolution tests. CeO₂ powders prepared by calcination of cerium oxalate at 800 °C and sintering at 1700 °C gave samples with similar microstructure to UO₂ fuel and SIMFUEL. The suitability of the optimised synthesis route for dissolution was tested in a dissolution experiment conducted at 90 °C in 0.01 M HNO₃.     

Completion of IFMIF/EVEDA lithium test loop construction

The EVEDA Li test loop (ELTL) successfully completed its construction and installation of a total of 2.5-ton Li in the frame work of the IFMIF/EVEDA as one of the ITER-BA. Design for the ELTL had been done from March 2009 to December 2009 in large part, and then the construction was started on November 2009 in the O-arai site of the Japan Atomic Energy Agency and completed on the middle of November 2010 after passing an authority inspection by a fire department in O-arai town. Subsequently, the 2.5-ton Li was installed to the ELTL by using a glove box in the form of ingots which is 240 mm long and 125 mm in diameter. The nitrogen concentration in the 2.5-ton Li was found to be 127 wppm. During the installation, the oxygen concentration and the humidity in the glove box were almost kept less than 20 wppm, and any large contamination by air was prevented during the handling of Li.     

Application of a boron doped diamond (BDD) electrode as an anode for the electrolytic reduction of UO2 in Li2O–LiCl–KCl molten salt

A boron doped diamond thin film electrode was employed as an inert anode to replace a platinum electrode in a conventional electrolytic reduction process for UO₂ reduction in Li₂O–LiCl molten salt at 650 °C. The molten salt was changed into Li₂O–LiCl–KCl to decrease the operation temperature to 550 °C at which the boron doped diamond was chemically stable. The potential for oxygen evolution on the boron doped diamond electrode was determined to be approximately 2.2 V vs. a Li–Pb reference electrode whereas that for Li deposition was around ?0.58 V. The density of the anodic current was low compared to that of the cathodic current. Thus the potential of the cathode might not reach the potential for Li deposition if the surface area of the cathode is too wide compared to that of the anode. Therefore, the ratio of the surface areas of the cathode and anode should be precisely controlled. Because the reduction of UO₂ is dependent on the reaction with Li, the deposition of Li is a prerequisite in the reduction process. In a consecutive reduction run, it was proved that the boron doped diamond could be employed as an inert anode.     

Hydrogen isotope exchange in tungsten: Discussion as removal method for tritium

Hydrogen isotope exchange in re-crystallized polycrystalline tungsten was investigated at 320 and 450 K. In a first step the tungsten samples were loaded with deuterium to a fluence of 10²⁴ D/m² from a low-temperature plasma at 200 eV/D particle energy. In a second step, H was implanted at the same particle energy and similar target temperature with a mass-separated ion beam at different ion fluences ranging from 2 × 10²⁰ to 7.5 × 10²³ H/m². The analytic methods used were nuclear reaction analysis with D(³He,p)α reaction and elastic recoil detection analysis with ⁴He. In order to determine the D concentration at depths of up to 7.4 μm the ³He energy was varied from 0.5 to 4.5 MeV. It was found that already at an H fluence of 2 × 10²⁰ H/m², i.e. at 1/5000 of the initial D fluence, about 30% of the retained D was released. Depth profiling of D without and with subsequent H implantation shows strong replacement close to the surface at 320 K, but extending to all analyzable depths at 450 K especially at high fluences, leading to higher release efficiency. The reverse sequence of hydrogen isotopes allowed the analysis of the replacing isotope and showed that the release of D is balanced by the uptake of H. It also shows that hydrogen does not diffuse through a region of filled traps into a region were unfilled traps can be encounter but transport is rather a dynamic process of trapping and de-trapping even at 320 K. Initial D retention in H loaded W is an order of magnitude higher than in pristine W, indicating that every H-containing trap is a potential trap for D. In consequence, hydrogen isotope exchange is not a viable method to significantly enhance the operation time before the tritium inventory limit is reached but should be considered an option to reduce the tritium inventory in ITER before major interventions at the end of an operation period.     

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.     

Carbon film growth and hydrogenic retention of tungsten exposed to carbon-seeded high density deuterium plasmas

Tungsten (W) targets have been exposed to high density (n_e ? 4 × 10¹⁹ m^?3), low temperature (T_e ? 3 eV) CH₄-seeded deuterium (D) plasma in Pilot-PSI. The surface temperature of the target was ～1220 K at the center and decreased radially to ～650 K at the edges. Carbon film growth was found to only occur in regions where there was a clear CII emission line, corresponding to regions in the plasma with T_e ? 2 eV. The maximum film thickness was ～2.1 μm after a plasma exposure time of 120 s. ³He nuclear reaction (NRA) analysis and thermal desorption spectroscopy (TDS) determine that the presence of a thin carbon film dominates the hydrogenic retention properties of the W substrate. Thermal desorption spectroscopy analysis shows retention increasing roughly linearly with incident plasma fluence. NRA measures a C/D ratio of ～0.002 in these films deposited at high surface temperatures.     

Lithium concentration dependence of implanted helium retention in lithium silicates

Helium ions of 500 keV were implanted with a fluence of 1.4 × 10¹⁷ ion/cm² into various lithium silicates to investigate whether a threshold level of helium retention exists in Li-containing silicate ceramics similar to that found in SiO_x in previous work. The composition and phases of the as prepared lithium silicates were determined by proton backscattering spectrometry (p-BS) and X-ray diffraction (XRD) methods with an average error of ±10%. Electrostatic charging of the samples was successfully eliminated by wrapping the samples in Al foil. The amounts of the retained helium within the samples were determined by subtracting the non-implanted spectra from the implanted ones. The experimental results show a threshold in helium retention depending on the Li concentration. Under 20 at.% all He is able to escape from the material; at around 30 at.% nearly half of the He, while over 65 at.% all implanted He is retained. With compositions expressed in SiO₂ volume percentages, a trend similar to those reported of SiO_x previously is found.     

The chemical durability of glass and graphite–glass composite doped with cesium oxide

The role of temperature in determining the chemical stability of a waste form, as well as its leach rate, is very complex. This is because the dissolution kinetics is dependent both on temperature and possibility of different rate-controlling mechanisms that appear at different temperature regions. The chemical durability of Alumina-Borosilicate Glass (ABG) and Glass–Graphite Composite (GGC), bearing Tristructural Isotropic (TRISO) fuel particles impregnated with cesium oxide, were compared using a static leach test. The purpose of this study is to examine the chemical durability of glass–graphite composite to encapsulate coated fuel particles, and as a possible alternative for recycling of irradiated graphite. The test was based on the ASTM C1220-98 methodology, where the leaching condition was set at a temperature varying from 298 K to 363 K for 28 days. The release of cesium from ABG was in the permissible limit and followed the Arrhenius’s law of a surface controlled reaction; its activation energy (E_a) was 65.6 ± 0.5 kJ/mol. Similar values of Ea were obtained for Boron (64.3 ± 0.5) and Silicon (69.6 ± 0.5 kJ/mol) as the main glass network formers. In contrast, the dissolution mechanism of cesium from GGC was a rapid release, with increasing temperature, and the activation energy of Cs (91.0 ± 5 kJ/mol) did not follow any model related to carbon kinetic dissolution in water. Microstructure analysis confirmed the formation of Crystobalite SiO₂ as a gel layer and Cs⁺¹ valence state on the ABG surface.     

Boron lattice site location in (BGa)As and (BGa)P thin films studied using RBS and NRA with a channeled He+ ion beam

In this study the boron lattice site location in ternary B_xGa_1?xAs and B_xGa_1?xP thin films grown on (0 0 1) GaAs and (0 0 1) GaP, respectively, using low pressure metal-organic vapour-phase epitaxy (MOVPE) with boron concentrations between x = 0.8% and x = 3.2% was investigated with RBS and the ¹⁰B(α,p)¹³C nuclear reaction using a 2.3 MeV He⁺ ion beam. For this purpose, the ion beam was aligned with the [0 0 1], [0 1 1] and [1 1 1] axis and the RBS and proton yield from the nuclear reaction compared with random ion incidence. For comparison, theoretical proton yields which assume boron to be located on substitutional lattice sites only were calculated for each sample/axis combination and compared with the experimental yields. The RBS/channeling measurements show a very good crystal quality of the films with χ_min being in the range of 3–5% for the [0 1 1] axis. The best crystal qualities, i.e. the lowest χ_min values and dechanneling rates, are achieved for low boron concentrations. From NRA/channeling it can be deduced that in the B_xGa_1?xAs films the fraction of interstitial boron is approximately 5% for low boron concentrations of x = 1% and 6–10% for concentrations up to x = 3.2%, whereas the fraction of interstitial boron is less than 3% in the B_xGa_1?xP film studied despite a concentration of x = 2.0%. This indicates that antisite effects of the boron incorporation are more likely in GaAs compared to GaP.     

On the study of catalytic membrane reactor for water detritiation: Membrane characterization

Tritium waste recycling is a real economic and ecological issue. Generally under the non-valuable Q₂O form (Q = H, D or T), waste can be converted into fuel Q₂ for a fusion machine (e.g. JET, ITER) by isotope exchange reaction Q₂O + H₂ = H₂O + Q₂. Such a reaction is carried out over Ni-based catalyst bed packed in a thin wall hydrogen permselective membrane tube. This catalytic membrane reactor can achieve higher conversion ratios than conventional fixed bed reactors by selective removal of reaction product Q₂ by the membrane according to Le Chatelier's Law.This paper presents some preliminary permeation tests performed on a catalytic membrane reactor. Permeabilities of pure hydrogen and deuterium as well as those of binary mixtures of hydrogen, deuterium and nitrogen have been estimated by measuring permeation fluxes at temperatures ranging from 573 to 673 K, and pressure differences up to 1.5 bar. Pure component global fluxes were linked to permeation coefficient by means of Sieverts’ law. The thin membrane (150 μm), made of Pd–Ag alloy (23 wt.%_Ag), showed good permeability and infinite selectivity toward protium and deuterium. Lower permeability values were obtained with mixtures containing non permeable gases highlighting the existence of gas phase resistance. The sensitivity of this concentration polarization phenomenon to the composition and the flow rate of the inlet was evaluated and fitted by a two-dimensional model.     

CdTe detector use for PIXE characterization of TbCoFe thin films

Peltier cooled CdTe detectors have good efficiency beyond the range of energies normally covered by Si(Li) detectors, the most common detectors in PIXE applications. An important advantage of CdTe detectors is the possibility of studying K X-rays lines instead the L X-rays lines in various cases since CdTe detectors present an energy efficiency plateau reaching 70 keV or more. The ITN CdTe useful energy range starts at K-K_α (3.312 keV) and goes up to 120 keV, just above the energy of the lowest γ-ray of the ¹⁹F(p, p’γ)¹⁹F reaction. In the new ITN HRHE-PIXE line, a CdTe detector is associated to a POLARIS microcalorimeter X-ray detector built by Vericold Technologies GmbH (an Oxford Instruments Group Company). The ITN POLARIS has a resolution of 15 eV at 1.486 keV (Al-K_α) and 24 eV at 10.550 keV (Pb-L_α1). In the present work, a TbCoFe thin film deposited on a Si substrate was analysed at the HRHE-PIXE system. The good efficiency of the CdTe detector at 45 keV (Tb-K_α), and the excellent resolution of POLARIS microcalorimeter at 6.403 keV (Fe-K_α), are presented and the new possibilities open to the IBA analysis of systems with traditionally overlapping X-rays and near mass elements are discussed.     

The effect of displacement damage on deuterium retention in ITER-grade tungsten exposed to low-energy,high-flux pure and helium-seeded deuterium plasmas

Samples prepared from polycrystalline ITER-grade tungsten were damaged by irradiation with 20 MeV W ions at room temperature to a fluence of 1.4 × 10¹⁸ W/m². Due to the irradiation, displacement damage peaked near the end-of-range, 1.35 μm beneath the surface, at 0.89 displacements per atom. The damaged as well as undamaged W samples were then exposed to low-energy, high-flux (10²² D/m² s) pure D and helium-seeded D plasmas to an ion fluence of 3 × 10²⁶ D/m² at various temperatures. Trapping of deuterium was examined by the D(³He,p)⁴He nuclear reaction at ³He energies varied from 0.69 to 4.0 MeV allowing determination of the D concentration at depths up to 6 μm. It has been found that (i) addition of 10% helium ions into the D plasma at exposure temperatures of 440–650 K significantly reduces the D concentration at depths of 0.5–6 μm compared to that for the pure plasma exposure; (ii) generation of the W-ion-induced displacement damage significantly increases the D concentration at depths up to 2 μm (i.e., in the damage zone) under subsequent exposures to both pure D and D–He plasmas.     

Study on flow instability for feasibility of a thin liquid film first wall

This study proposes a probability of the evaporated gas that agitates a growing instability wave in a thin liquid film first wall. The liquid first wall was considered to be in vacuum and the effect of the ambient gas was neglected but the evaporated gas by the high energy fluxes is a probable cause of unstable wave agitation. The criterion is approximately expressed by the density ratio (Q₂) and the Weber number (We) as Q₂ × We^0.5 ≈ 5 × 10⁻⁴. Performed indirect experimental supported this criterion. For a case study of liquid Pb-17Li film with a velocity of 10 m/s, the evaporated gas pressure must be below 6.2 × 10³ Pa to maintain stable conditions. By recent study, this pressure is generated at 1600 K temperature and it is believed to be attainable by the energy fluxes on the first wall. This result is so far not confirmed so the full verification by experimental is to be performed.