掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

Development of an efficient tritium recovery method is indispensable in order to compose a liquid blanket system of a D-T fusion reactor in the near future. Here, tritium recovery using a bubbling tower is focused on, and the behavior of H transfer between fluidized lithium-lead (Li-Pb) and gas bubbles of Ar-H<inf>2</inf> or pure Ar is examined analytically and experimentally under isothermal conditions. Gas of Ar-H<inf>2</inf> or pure Ar is injected into fluidized Li-Pb through an I-shape nozzle made from SS-316. Time variations of the H<inf>2</inf> concentration in gas bubbles that come out from fluidized Li-Pb are measured by gas chromatography. Mass-transfer coefficients to correlate rates of H atom transfer between Li-Pb and gas bubbles are obtained by fitting analytical equations to experimental results. The solution is derived under conditions where H transfer between bubbles and liquid Li-Pb is limited by diffusion in the Li-Pb boundary layer. The parameters such as bubble diameter and terminal rising velocity which are used in order to derive analytic formula are estimated from balance among several forces such as gravity, surface tension, inertia force and so on. The behavior of hydrogen transfer at gas-liquid interfaces in liquid blanket is investigated in terms of the mass-transfer coefficient obtained under various conditions.

DOI： 10.13182/FST14-903

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Informa UK Limited  

Hydrogen transfer under a fluidized condition of Li-Pb is investigated experimentally to design a Li-Pb blanket system. Li-Pb eutectic alloy flows through inside of a Ni tube in the experimental sy...

DOI： 10.13182/fst14-t104

Scopus

CiNii Research

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

The present study aims at obtaining fundamental data on tritium migration in porous materials, which include soaking effect, interaction between tritium and cement paste coated with paints and transient tritium sorption in porous cement. The amounts of tritium penetrated into or released from cement paste with epoxy and urethane paint coatings were measured. The tritium penetration amounts were increased with the HTO exposure time. Time to achieve a saturated value of tritium sorption was more than 60 days for cement paste coated with epoxy paint and with urethane paint, while that for cement paste without any paint coating took 2 days to achieve it. The effect of tritium permeation reduction by the epoxy paint was higher than that of the urethane. Although their paint coatings were effective for reduction of tritium penetration through the cement paste which was exposed to HTO for a short period, it was found that the amount of tritium trapped in the paints became large for a long period. Tritium penetration rates were estimated by an analysis of one-dimensional diffusion in the axial direction of a thickness of a sample. Obtained data were helpful for evaluation of tritium contamination and decontamination.

DOI： 10.13182/FST14-T20

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

What affects tritium migration through porous concrete walls coated with a hydrophobic paint is reviewed from the viewpoint of tritium safety. Being taken into consideration of multi-structural concrete composed of aggregates, sand, water and cement which contents are CaO, SiO<inf>2</inf>, Al<inf>2</inf>O<inf>3</inf>, Fe<inf>2</inf>O<inf>3</inf>, MgO, CaSO<inf>4</inf> and so on, tritium path is discussed in terms of the HTO diffusivity and adsorption coeffcient on porous walls. Measures to predict rates of tritium leak from laboratory walls to the environment and residual tritium amounts in concrete are estimated based on previous data. Three cases of accidental or chronic tritium release to laboratory air are discussed using the diffusion-adsorption model.

DOI： 10.13182/FST14-T25

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Elsevier BV  

Recovery of D dissolved in a liquid Li flow at low D concentration is experimentally investigated using a Y metal absorber under the two fluidized conditions: (a) in a vertical cylindrical tube and (b) in an agitated vessel. The target concentration is 1 appm in Li around at 300 degrees C. The two concentrations of D remaining in Li and recovered by Y are detected by a dissolution method using H2O with depleted-D and HNO3. The main released species is HD. A small amount of HDO released is reduced to HD by a Mg particle bed. It is found that HF-treated Y can absorb H isotopes at the target temperature and concentration. The chemical dissolution technique is found to be useful to specify the two absolute concentrations of D recovered by Y and D remaining in Li. (C) 2014 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2014.02.030

Scopus

CiNii Research

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Engineering and Design  

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. © 2014 Elsevier B.V.

DOI： 10.1016/j.fusengdes.2014.02.059

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

EVEDA Lithium Test Loop (ELTL) has been designed and constructed, has operated a liquid lithium flow test facility with the world's highest flow rate and has succeeded in generating a 100-mm-wide and 25-mmthick free-surface lithium flow along a concave back plate steadily at a high speed of 20 m/s at 300°C for the first time in the world. This result will greatly advance the development of an accelerator-based neutron source to high energy and high density, one of the key objectives of the fusion reactor materials development under the BA (Broader Approach) Activities. Recent related engineering validation and engineering design of the lithium facility has been evaluated.

DOI： 10.13182/FST13-770

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

Hydrogen permeation is investigated in a conjugated system of Li <inf>15.8</inf>Pb<inf>84.2</inf> eutectic alloy or Ar under natural convection inside a cylindrical vertical sus316 tube. Two different temperature conditions are made between the upper and lower parts of the vertical tube. When thermal convection in Pb-Li or Ar with inside temperature distribution is fast, the rate-determining step for overall permeation rate is considered H<inf>2</inf> permeation through the sus316 tube wall maintained at the lower temperature. On the contrary, when the inside convection is slow, the convection affects the transient behavior of overall H<inf>2</inf> permeation rates. The analytical result was consistent with experimental ones of the Ar convection system. However, the steady-state overall H<inf>2</inf> permeation rate for the Pb-Li system was much higher than the value estimated from the intrinsic permeability of sus316, and the values changed gradually with elapsed time. This may be because hydrogen is present as an atomic state in the Li<inf>17</inf>Pb <inf>83</inf> convection layer with different temperatures, and bubbles on interfaces affect permeation. The overall H<inf>2</inf> permeation rates under various temperature conditions are compared, and the effect of thermal convection is discussed.

DOI： 10.13182/FST13-A19163

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

IFMIF-EVEDA progresses in Japan as one of the EU-Japan Broader Approach Activities. The research is performed to decide whether or not IFMIF is constructed after some uncertainties included in the design are clarified. One of the uncertainties included in the Li purification process is to prove experimentally the removal of 1 weight ppm (wppm) T and 10 wppm D from flowing Li for safety. Our research group is experimentally investigating the recovery of hydrogen isotopes including T not only in static Li but also in fluidized Li. In the past study, hydrofluoric acid (HF) treatment of Y is successful in removing oxide inevitably formed on its surfaces. The recovery of hydrogen isotopes including T less than 1 wppm is successfully proved with use of the HF-treated Y at 300°C, which is the IFMIF hot-trap temperature. Mass-transfer rates of hydrogen isotopes in the liquid Li and Y under stirred conditions were determined. In addition, we developed a way to determine an amount of D or T dissolved in Li and Y by using a dissolution method. The quantitative D analysis is performed by using techniques of HNO3 solution for Y and H2O one for Li. The distribution coefficient between Li and Y is determined as a function of temperature and contact time.

DOI： 10.13182/FST13-A19148

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Materials Transactions  

In order to clarify the tritium behavior in PbLi blanket system, the solubility of hydrogen isotopes in the PbLi eutectic alloy was experimentally investigated. The effects of (i Li preferential evaporation from PbLi, (ii) thermal convection in a quartz-glass reaction tube and (iii) difference in crucible materials between A<inf>22</inf>O<inf>3</inf> and W on the solubilities were discussed. Another important issue of the PbLi blanket system is to develop a new ceramic coating to decrease tritium permeation rate. Tritium permeation test was performed for Er<inf>2</inf>O<inf>3</inf> coating on F82H. The permeation reduction factor achieved here was higher than 103. Precise experiments were carried out to determine the solubility, diffusivity and permeability of hydrogen isotopes in PbLi eutectic alloy by means of a transient permeation method. The isotope effects in solubility and diffusivity between H and D were clarified. No isotope effect in solubility was observed between H and D, and the H diffusivity was 1.4 times larger than the D one. The ratio was near to a value predicted from the classical theory. Experiment using tritium is designed under the collaboration work with INL. © 2013 The Japan Institute of Metals.

DOI： 10.2320/matertrans.MG201203

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Engineering and Design  

The effects of hydrophobic paint coating on a concrete material of cement paste on the tritium transport are investigated. The cement paste is coated with two kinds of paints, acrylic-silicon resin paint and epoxy resin paint. We investigated the amount of tritium trapped in the samples exposed to tritiated water vapor by means of sorption and release. It was found that both the hydrophobic paints could reduce effectively tritium permeation during 50 days exposure of tritiated water vapor. The effect of tritium reduction of the epoxy paint was higher than that of silicon while the amount of tritium trapped in the epoxy paint was larger than that of silicon due to difference of the structure. Based on an analysis of a diffusion model, the rate-determining step of tritium migration through cement paste coated with the paints is diffusion through the paints respectively. It was found that tritium was easy to penetrate through silicon because there were many pores or voids in the silicon comparatively. In the case of tritium released from the epoxy paint, it is considered that tritium diffusion in epoxy is slow due to retardation by isotope exchange reaction to water included in epoxy paint. © 2012 Elsevier B.V.

DOI： 10.1016/j.fusengdes.2012.02.058

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Elsevier BV  

Abstract 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.

DOI： 10.1016/j.fusengdes.2012.03.004

Scopus

CiNii Research

掲載種別：研究論文（学術雑誌）   出版者・発行元：Elsevier BV  

Abstract The use of D–Li stripping reaction to generate multiple neutron is expected in the International Fusion Materials Irradiated Facility (IFMIF). Tritium generated by its side reaction needs to be recovered from liquid Li for safety. Y hot trap is expected as an effective purification system to recover tritium from the liquid Li loop. In IFMIF, since liquid Li circulates in a loop, it is important to trace tritium behavior in Y and liquid Li under dynamic conditions. In this study, the authors investigate the effectiveness of tritium recovery from liquid Li under dynamic conditions of Y hot trap. Hydrogen absorption in Li instead of tritium is investigated under dynamically fluidized conditions by stirring liquid Li with an arbitrary rotational rate. The experimental results prove that the hydrogen absorption rate from liquid Li to Y is independent of the rotational rate. This means that the rate-determining step is hydrogen diffusion in Y.

DOI： 10.1016/j.fusengdes.2012.03.032

Scopus

CiNii Research

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

Tritium release behavior for thermal neutronirradiated Li <inf>0</inf>17Pb<inf>0.83</inf> eutectic alloy was studied. Main tritium release peak was observed in the temperature just a little higher than melting point in a thermal desorption spectrometry (TDS) experiment. Most of tritium release from Li<inf>0.17</inf>Pb<inf>0.83</inf> eutectic alloy was found to be governed by diffusion process from the results of isothermal annealing experiments. Tritium diffusivity in a liquid state of Li<inf>0.17</inf>Pb<inf>0.83</inf> eutectic alloy was evaluated to be D =4.7×IO^-8exp(-0.13 eV/kT)m²s¹-1¹. Tritium diffusivity was increased by the phase transition of Li<inf>0</inf>17Pb<inf>0.83</inf>eutectic alloy from a solid state to a liquid state, resulting in the sharp tritium release peak that appeared in TDS spectrum. In addition, about 4% of tritium was trapped in Li<inf>017</inf>Pb<inf>083</inf> eutectic alloy as Li-T bond.

DOI： 10.13182/FST12-A14112

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

Systems to Recover Tritium (T) from Fusion Liquid Blankets of LiPb (Li <inf>17</inf>Pb<inf>18</inf>) and Flibe (Li<inf>2</inf>BeF<inf>4</inf> under safety conditions are discussed based on available data of T transport properties. We recently performed experiments on hydrogen isotopes permeating through LiPb in Kyushu University and introduced them in the present paper. Solubilities, diffusivities and permeabilities of H and D through Li <inf>17</inf>Pb<inf>83</inf> are determined by means of a transient permeation method. Their isotope effects between H and D in the Li-Pb eutectic alloy are found to be independent of D/H atomic ratio and a function of only temperature. Tritium recovery apparatuses of a liquid-gas counter-current flowing bed or a T permeation window are designed in the liquid blanket system composed of the primary coolant of Flibe or LiPb and the secondary coolant of He or water. Simultaneous transfer of heat and T needs new configuration in order to satisfy the necessary conditions of low T leak and high heat transfer, which make it possible to operate the blanket effectively and safely. Tritium permeation rates through the primary LiPb or Flibe breeder and secondary water or He coolant loops are estimated, which is the most important factor to operate the blanket safely.

DOI： 10.13182/FST12-A13397

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Engineering and Design  

An experimental study on tritium (T) transfer in porous concrete for the tertiary T safety containment is performed to investigate (i) how fast HTO penetrates through concrete walls, (ii) how well concrete walls contaminated with water-soluble T are decontaminated by a solution-in-water technique, and (iii) how well hydrophobic paint coating works as a protecting film against HTO migrating through concrete walls. The experiment is comparatively carried out using disks of cement paste which W(water)/C(cement) weight ratio is 0.6:1 with or without hydrophobic paints, and mortar disks which W/C/sand ratio is 0.6:1:2 with or without the paints. The hydrophobic paints tested in the present study are an epoxy polymer resin paint and an acrylic-silicon polymer resin one. After T exposure during specified time under a constant HTO vapor pressure in an acrylic box, the amount of water-soluble HTO on/in the disks is determined using a technique of H <inf>2</inf>O dissolution during specified time. The results obtained here are summarized as follows: (1) HTO penetration in porous concrete can be correlated in terms of the effective diffusivity. (2) Its value in porous cement without coating is 1.2 × 10 ^-11 m ²/s at 25 °C. (3) HTO penetrates only through pores in cement, and there is no path for HTO transfer in non-porous sand. (4) Rates of sorption and dissolution of HTO in disks of cement and mortar coated with the epoxy resin paint are correlated in terms of the effective diffusivity through the paint film which value is D <inf>T</inf> = 1.0 × 10 ^-16 m ²/s. The rate-determining step is diffusion through the paint. (5) The epoxy resin paint works more effectively as an anti-HTO diffusion coating. (6) Another acrylic-silicon resin paint does not work well as anti-HTO diffusion coating. This may be because the hydrophobic property of the silicon resin paint is deteriorated with elongating the contact time with H <inf>2</inf>O vapor or liquid. (7) The HTO uptake inside the epoxy paint is greater than that of the silicon one. (8) The permeation reduction factor (PRF) of HTO for the epoxy paint at steady-state is expected large, if HTO vapor only contributes to diffusion. However, when concrete surfaces coated with the epoxy paint are under wet conditions, the PRF value becomes smaller. All those results can be used to estimate the effect of HTO soaking in concrete walls in case of accidental T release in a fusion reactor room and to decontaminate wastes of tritiated concrete. © 2011 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2011.09.001

Scopus

記述言語：英語   出版者・発行元：一般社団法人 プラズマ・核融合学会  

Materials to compose a fusion reactor chamber are subjected to neutron irradiation. International Fusion Materials Irradiated Facility (IFMIF) will be used to analyze how neutron damage affects material durability. The material is irradiated by high-intensity neutron beam, which is generated by the D-Li stripping reaction. Tritium (T) generated by by-pass reaction needs to be recovered from liquid Li for safety. Recovery of tritium by an yttrium (Y) hot trap is one of the ways to recover T from liquid Li in IFMIF. Therefore, it is necessary to analyze the behavior of T in the liquid Li. In this study, the authors measured H2 absorption rates under stirred conditions of liquid Li and analyzed the effect of hydrogen absorption rate of Y by elevating temperature from 250°C to 400°C to establish the way to recover T from liquid Li with the Y hot trap in IFMIF. Judging from the comparison, we considered the rate-determining step is H diffusion in Y. Mass transfer coefficients at each temperature were determined by fitting the results between calculation and experiment. © 2012 The Japan Society of Plasma Science and Nuclear Fusion Research.

DOI： 10.1585/pfr.7.2405080

Scopus

CiNii Research

記述言語：英語   出版者・発行元：一般社団法人 プラズマ・核融合学会  

Clarification of tritium transfer in blanket is an important issue for realizing fusion reactor. We perform an experiment of simultaneous H and D permeation through Li<inf>17</inf>Pb<inf>83</inf> by means of an unsteady permeation method in order to clarify the interactions and isotope effects between H and D atoms. The experiment is conducted under the condition where the Sieverts' law holds. As a result, it is found that (i) H and D atoms permeate independently regardless of the H/D component ratio in the upstream gas and (ii) diffusion process is the rate-determining step in the overall permeation process. The diffusivity of H is around 1.4 times larger than that of D. The solubility of H is close to that of D. We estimate two vibration modes at an absorption site and a saddle point of H in Li<inf>17</inf>Pb<inf>83</inf> based on the ratio of isotope effect. It is considered that the zero-point vibration energy of H at the absorption site in Li<inf>17</inf>Pb<inf>83</inf> is around 0.173 eV and that at the saddle point is around 0.235 eV. The ratio of isotope effect is almost in proportion to the square root of mass ratio of D to H. The diffusivity of T can be estimated as 1/1.7 times of H. © 2012 The Japan Society of Plasma Science and Nuclear Fusion Research.

DOI： 10.1585/pfr.7.2405074

Scopus

CiNii Research

掲載種別：研究論文（学術雑誌）   出版者・発行元：Journal of Nuclear Materials  

In order to estimate mass-transfer properties of tritium (T) in Li + Pb binary alloys as a blanket material for fusion reactors, variations of the H solubility with different Li and Pb compositions are presented. Deviation of the Li + Pb binary liquid mixture from an ideal solution model is discussed, and a correlation is proposed for predicting properties of various Li + Pb alloys. Isotope effects on solubility and diffusivity of Li<inf>17</inf>Pb<inf>83</inf> between H and D are correlated based on an isotropic harmonic oscillation model. The zero-point energy of H atoms dissolved in Li<inf>17</inf>Pb<inf>83</inf> is determined from the comparison with the model. Wettability at the interface between liquid Li<inf>17</inf>Pb<inf>83</inf> and metallic walls and effects of impurities on H solubility are discussed. © 2010 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2010.12.127

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Journal of Nuclear Materials  

The permeability, diffusivity and solubility of hydrogen isotopes (H, D) in a Li-Pb eutectic alloy are measured in the temperature range from 573 to 973 K using a permeation method. Their isotope effects between H and D are obtained. It is shown that the solubility of hydrogen isotopes in Li-Pb obeys the Sieverts' law. The solubility of D is around 1.4 times larger than that of H. The isotope effect of the diffusivity between H and D is very small. © 2010 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.jnucmat.2010.12.126

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

An experimental study on tritium transfer in porous concrete materials for the tertiary tritium safety containment is performed to investigate; (i) how fast tritium is transferred through porous concrete walls coated with or without a hydrophobic paint, and (ii) how well the hydrophobic paint coating works as a film protecting against tritium migrating through concrete. The experiment is comparatively carried out using two types of cement-paste and mortar disks with or without two kinds of paints. The results obtained here are summarized as follows: (1) Tritium transfer can be correlated in terms of the effective tritium diffusivity of DT=1.2xl0^-11 m²/s in porous cement. (2) Adsorbed or condensed liquid HTO itself is transferred only through pores in cement, and no tritium transfer path is present in non-porous sand. (3) Rates of tritium sorption and dissolution in cement and mortar coated with an epoxyresin paint is correlated in terms of the diffusivity through the paint film of D<inf>T</inf>=1.0x101^-16 m²/s. (4) The epoxy paint works more effectively as an anti-tritium diffusion coating than the acrylic-silicon resin paint. (5) The hydrophobic property of the silicon resin paint is deteriorated with elongating the contact time with H<inf>2</inf>O.

DOI： 10.13182/FST11-A12599

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

We performed an experiment of H and D permeation through Li <inf>17</inf>Pb<inf>83</inf> in the two-component (H-VD) system by an unsteady permeation method to clarify interactions between H and D atoms. It was found that H and D permeate independently regardless of the H/D composition ratio in the upstream gas. Dissolution of H and D atoms into Li<inf>17</inf>Pb <inf>83</inf> follows the Sieverts' law in the two-component system in a similar way to the single one. Diffusion of H and D in Li<inf>17</inf>Pb<inf>83</inf> was the rate-determining step in the overall permeation process. An isotope effect of permeability between H and D was around 1.4 in the temperature range from 400°C to 700°C. The ratio of the isotope effect was in proportion to the square root of mass ratio of D to H. Tritium permeation can be estimated in a similar way.

DOI： 10.13182/FST60-1163

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Engineering and Design  

In IFMIF/EVEDA, tasks for lithium target system consist of 5 validation tasks (LF1-5) and a design task (LF6), and are shared by Japan and Europe. Japan is covering LF1, LF2, LF4, LF5 and LF6, and JAEA as an Implementing Agency is organizing these tasks. Japanese universities contribute these tasks. The purpose of LF1 task is to construct and operate EVEDA Li test loop (ELTL) with diagnostics and purification systems for the validation of Li target facility. ELTL will be constructed at the end of February in 2011 and the operation will be started at the summer in 2011. LF2 and LF4 are basic research for the diagnostics and purification system, respectively. And LF2 and LF4 have also contributed the construction of the ELTL (LF1). LF5 is focusing on the remote handling for the replacement of the target assembly. Major deliverable of LF6 is finally engineering design report, which is assembled with all validation tasks, and this activity will be gradually enhanced to the end of IFMIF/EVEDA. © 2011 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2011.04.033

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Engineering and Design  

The International Fusion Materials Irradiation Facility (IFMIF) is designed to provide an intense neutron source for the development of fusion reactor materials. Neutron is supplied by the D-⁷Li reaction, and at the same time tritium (T) is generated as its by-product. Therefore, the removal of T generated in Li becomes inevitable. It is proved in the present study that HF pretreatment of Y getter materials leads to enhance hydrogenating rates of Y placed in Li. The ratio of the H removal rate of HF-treated Y to that of non-treated Y is determined under the conditions of three different temperatures of 573K, 553K and 523K. The effect of enhancement of hydrogenating rates is quantitatively determined. © 2010 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2010.04.022

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Engineering and Design  

Transport of tritium (T) and heat is calculated to estimate the performance of liquid Li<inf>17</inf>Pb<inf>83</inf> (Li-Pb) or Li<inf>2</inf>BeF <inf>4</inf> (Flibe) as a T-breeder in a fusion reactor blanket. This bred in such a way of low leak to facilities outside and continuous recovery by a removal system, and heat is transferred through structural walls to He coolant efficiently. In this paper, T permeation in a blanket composed of structural materials and liquid Li-Pb or Flibe is calculated based on data of previous experiment. The effects of T recovery ratio by the outside removal apparatus and fluid-film T diffusion resistance in the liquid blanket on overall T permeation rates are analytically clarified. Design of a liquid blanket with low T leak and high T recovery is discussed here. In addition, possibility in that microbubbles may be generated at interfaces between a liquid blanket and a structural wall is investigated. © 2010 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2010.03.032

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Engineering and Design  

The permeation process of hydrogen isotopes through Li <inf>0.17</inf>Pb<inf>0.83</inf> has been investigated experimentally. We obtained the overall D permeation rates that take into consideration of the effects of the H-D isotopic exchange reactionon the surface and diffusion inthe Li-Pb layer. It was proved that the rate-determining step was not the surface reaction but the diffusion in the Li-Pb bulk. We proposed a system of gas-liquid counter-current extraction using a packed column for continuous tritium recovery. The concentration profiles through the column and tritium extraction rates are calculated. It is estimated that the column height of5m is sufficient to achieve the tritium recovery ratio of 99.999%. This counter-current extraction method is easy to operate and efficient to recover tritium from Li-Pb. © 2010 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2010.03.010

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Engineering and Design  

The use of Li is expected for a flowing target in the International Fusion Materials Irradiation Facility (IFMIF) for high-intensity neutron generation. Tritium (T) of by-product is removed by a Y hot trap placed in a by-path position of the flowing Li target. Since the equilibrium pressure of hydrogen isotopes is extremely low, the removal rate of H isotopes is affected by coexisting other impurities such as N, C and O dissolved in Li. In the present study, the removal rate of H isotopes in Li is experimentally investigated under the three different methods or conditions: (i) measuring gravimetrically in a static Li + Y system with the supply of low-concentration H<inf>2</inf> in Ar purge, (ii) counting the change of T radioactivity in a static Li + Y system after neutron irradiation and (iii) measuring the change of the H<inf>2</inf> concentration in a stirred Li + Y system with the supply of low-concentration H<inf>2</inf> in Ar purge. There was no corrosive action in the Li-Y interface observed after contact at 500 °C for 100 h. The present results are extended to design a Y hot trap for the IFMIF-EVEDA Li loop in order to remove H isotopes less than a specified concentration. © 2009 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2009.06.002

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掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

Rates and amounts of absorption and desorption of hydrogen and deuterium in a Li<inf>0.17</inf>Pb<inf>0.83</inf> eutectic alloy are determined under the conditions of temperature of 400-700°C and the upstream H<inf>2</inf> or D<inf>2</inf> partial pressure of 10³Pa-10⁵Pa by using a one-dimensional permeation pot. Because of small interaction between the alloy and dissolved atoms of hydrogen isotopes, the temperature dependence of the Sieverts' solubility constant for the Li<inf>0.17</inf>Pb<inf>0.83</inf>-H or -D system, i.e., the enthalpy change of absorption or desorption, is small, and the absolute value of D solubility is higher than that of H. The isotope effect of diffusivity between H and D is very small. The generation rate and inventory of tritium (T) in a fusion blanket is estimated under an assumption of one-dimensional Li<inf>0.17</inf>Pb<inf>0.83</inf> blanket system with a constant and uniform neutron flux.

DOI： 10.13182/FST56-831

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

The rate of tritium released from temperaturecontrolled Flibe (a mixed molten salt of 2LiF + BeF<inf>2</inf>) after neutron irradiation was determined comparatively under two different conditions of Ar-H<inf>2</inf> (10%) or Ar gas purge at a constant or linearly elevated temperature. Experimental rates of tritium release were analyzed based on its diffusion in Flibe and isotopic exchange between T atoms on surfaces and H atoms included in gaseous components. Gas released from Flibe had compositions of various ratios of HT to TF depending on the different conditions of Ar-H<inf>2</inf> or Ar purge gas. The major molecular species of tritium released from Flibe after neutron irradiation was HT under the condition of the Ar-H<inf>2</inf> purge and 300°C The rate of tritium release under the Ar-H<inf>2</inf> purge was simulated well by the present analytical model. Although its chemical form immediately after the release was TF under the condition of Ar purge, it was changed to HT partly by interaction with metallic surfaces.

DOI： 10.13182/FST09-A4067

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：International Journal of Hydrogen Energy  

A chemical heat-pump system using two hydrogen-absorbing alloys is proposed to utilize heat exhausted from a high-temperature source such as high-temperature-gas-cooled reactor, HTGR, which is designed to produce H<inf>2</inf> more efficiently. The overall system proposed here consists of HTGR, He gas turbines, chemical heat pumps and reaction vessels corresponding to the three-step decomposition reactions comprising the I-S process. A fundamental research is performed experimentally on heat generation in a single bed packed with a hydrogen-absorbing alloy that works at the H<inf>2</inf> production temperature. The hydrogen-absorbing alloy of Zr(V<inf>1-x</inf>Fe<inf>x</inf>)<inf>2</inf> is selected as a material that has a proper plateau pressure for the heat-pump system operated between the input and output temperatures of HTGR. Temperature jump due to heat generated when the alloy absorbs H<inf>2</inf> proves that the alloy-H<inf>2</inf> system can heat up the exhaust gas even at 600 °C without any external mechanical force. © 2008 International Association for Hydrogen Energy.

DOI： 10.1016/j.ijhydene.2008.10.064

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Engineering and Design  

The best material for a wet wall and blanket of an inertial fusion reactor is selected among Li, eutectic alloys of Li-Pb, Li-Sn and a 2LiF + BeF<inf>2</inf> molten salt mixture called Flibe, judging from their chemical, nuclear and heat-transfer properties. Li<inf>0.17</inf>Pb<inf>0.83</inf> is found to be the most promising one because of low Li vapor pressure, moderate melting temperature, good heat-transfer properties under the condition of a KOYO-fast circulation loop operated between 300 and 500 °C. A counter-current extraction tower packed with metallic rashig rings is proposed to extract tritium generated and dissolved in the Li-Pb eutectic alloy. Mass-transfer parameters when He and Li-Pb flow counter-currently through the tower packed with the rings are determined to satisfy the two necessary conditions of a self-sufficient tritium generation rate of 1.8 MCi/day and a target tritium leak rate of 10 Ci/day. It is found that the height of a tower to achieve the 99.999% recovery is comparatively low because of the promising property of a large equilibrium pressure of tritium. In order to mitigate the disadvantage of its high density, which needs a large pumping power, a porous packing material that keeps good contact between He and Li-Pb should be developed in the future. It is found experimentally that D<inf>2</inf> addition in He purge gas is effective to achieve a faster rate of tritium recovery from the Li-Pb flow. The rate-determining step of tritium permeation through a steam generator is determined as a function of a Li-Pb flow rate in a stainless-steel heat-transfer tube. © 2008 Elsevier B.V. All rights reserved.

DOI： 10.1016/j.fusengdes.2008.05.030

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

The diffusion, solution and permeation coefficients of hydrogen isotopes in liquid Li-Pb which is a candidate liquid blanket material for fusion reactors were determined in the temperature range 573-973K using an unsteady permeation method. Each coefficiens was correlated to temperature as follows: D <inf>Li-Pb</inf> = 1.8 × 10^-8 exp(-11590/RT) [m²/s] (1) K<inf>Li-Pb</inf> = 2.1×10^-6 exp(-18700/RT) [1/Pa <inf>0.5</inf>] (2) P<inf>Li-Pb</inf> = 1.8×10<inf>-9</inf> exp(-30290/RT) [mol/msPa^0.5] (3) The hydrogen permeation flux depends on the square root of pressure at 773-973K. Although the power of pressure declined below 0.4 when temperature was below 673K, the effects of surface resistance were neglected above 673K. The hydrogen solubility in liquid Li-Pbwas found to correlate with a Sievert's constant. We calculated a height-equivalent to theoretical-plate of a gas-liquid countercurrent extraction tower for tritium recovery rates in liquid Li-Pb to be H<inf>L</inf>= 7.0×10^-2 [m] (4).

DOI： 10.13182/FST54-131

Scopus

掲載種別：研究論文（学術雑誌）   出版者・発行元：Fusion Science and Technology  

The recovery of tritium, T, from neutron-irradiated Li by a Y plate was experimentally investigated, and a Y hot trap was designed to remove T from a flowing Li target loop of IFMIF. T generated in Li ranging from 0.01 to O.OSppm in molar fraction was successfully removed by a Y plate at 300-500°C. The ratio of recovered T to generated one increased with temperature and immersion time. HT was the main chemical species of T that was released from Li, while HTO was the main one without Y absorption. SEM-EDX analysis revealed the transfer of not only T but also O originally included in Li as an impurity. Treatment of Y by HF was effective to remove oxides formed on the as-received Y plate and enhanced its hydrogenating rate. Conditions of a Y hot trap designed for the flowing Li target loop of IFMIF are given in the present paper.

DOI： 10.13182/FST08-A1777

Scopus