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A water-cooled ceramic breeder blanket has been developed. Heat flux and coolant pressure are applied to the first wall of the blanket in a fusion reactor. A square cooling channel is adopted in the first wall in Quantum and Radiological Science and Technology (QST). This study specifically addressed the relation between the thermal mechanical responses with different cooling channel shapes and the allowable load of the first wall. Thermal mechanical responses of the first wall were evaluated using finite element method (FEM). Based on those results, the allowable loads for the structural integrity of the first wall were assessed in accordance with the requirements of ASME Sec. III NH. Among the failure modes the creep-fatigue damage was the most dominant on the allowable loads of the first wall. The allowable loads were determined by characteristics of the thermal mechanical responses with the cooling channel shape. The major determinants were the surface heat flux and the coolant pressure, respectively, for the circular and the square channels.

In this study, creep-fatigue (CF) tests on F82H at 823 K were conducted according to the standard specimen and requirement of ASTM. Tension holding (TH) showed higher lifetime than compression holding (CH). Compared with 9Cr-1Mo-V steels, CF lifetime degradation was not observed at the holding time less than 0.05 h in case of CH and 0.17 h in case of TH. CF damages were summarized from the CF test results by Campbell method. The obtained damages distributed around CF damage envelope of Grade 91 in ASME NH. CF damage of cylindrical water-cooled ceramic breeder (WCCB) test blanket module (TBM) container at the zenith of hemisphere body under the ITER thermo-mechanical load was calculated. The calculated damage located within of the damage envelope of Grade 91. It was indicated that the present design of cylindrical WCCB TBM might resist CF interaction under the ITER thermo-mechanical load conditions.

It is known that the chemical forms of tritium released from a ceramic breeder blanket are hydrogen form and water form. To recover tritiated water vapor, adoption of dryer that is packed column of synthetic zeolite has been proposed. On the other hand, synthetic zeolite is often used as a support of precious metal catalyst. Such catalysts usually have a capability of hydrogen isotope exchange between gas and water vapor. If this catalyst is used to dryer, the dryer may obtain a preferable function for tritium recovery by isotopic exchange reaction. To assess such functions, reaction rate should be estimated. The results of water adsorption experiment on cation-exchanged mordenite-type zeolite suggested the possibility that state of adsorbed water varied by exchanged cation. So, in this work, precious metal catalyst based on cation-exchanged mordenite was prepared, and the reaction rate of chemical exchange between hydrogen and tritiated water was investigated under temperature range between 30 °C and 80 °C by the steady-state approximation. In the case of platinum on Na-mordenite, the reaction between gaseous hydrogen and tritiated water vapor was almost expressed as first-order reaction concerning tritiated water vapor concentration.

We have performed the tritium release experiment on the fusion reactor blanket at JAEA/FNS since 2009, and then clarified the ratio of tritium release and the recovered tritium chemical form. In order to acquire the detailed tritium recovery performances, we have started a new blanket tritium recovery experiment with ionization chamber (IC) at JAEA/FNS. For the appropriate tritium measurement with IC, we improved the experimental container and carried out with an intense DT neutron source at JAEA/FNS. From our new experiment, the tritium recovery radioactivity from the LSC measurement corresponds with the calculation within 6%. However, it was pointed out that further improvement in the quantitative tritium measurement by IC method was needed.

Li2TiO3 with excess Li (Li2+xTiO3+y) has been developed as an advanced tritium breeder. With respect to the tritium release characteristics of the blanket, the optimum grain size after sintering was less than 5 μm. Therefore, an emulsion method was developed to fabricate pebbles with this target grain size. The predominant factor affecting grain growth was assumed to be the presence of binder in the gel particles; this remaining binder was hypothesized to react with the excess Li, thereby generating Li2CO3, which promotes grain growth. To inhibit the generation of Li2CO3, calcined Li2+xTiO3+y pebbles were sintered under vacuum and subsequently under a 1% H2–He atmosphere. The average grain size of the sintered Li2+xTiO3+y pebbles was less than 5 μm. Furthermore, the tritium release properties of Li2+xTiO3+y pebbles were evaluated, and deuterium–tritium (DT) neutron irradiation experiments were performed at the Fusion Neutronics Source facility in the Japan Atomic Energy Agency. To remove the tritium produced by neutron irradiation, 1% H2–He purge gas was passed through the Li2+xTiO3+y pebbles. The Li2+xTiO3+y pebbles exhibited good tritium release properties, similar to those of Li2TiO3 pebbles. In particular, the released amount of tritiated hydrogen gas for easier tritium handling was greater than the released amount of tritiated water.

The accurate measurement of behavior of bred tritium released from a tritium breeder is indispensable to understand the behavior for a design of a tritium extraction system. The tritium collection method combined a CuO bed and water bubbles was not suitable to measure transient behavior of tritium released from Li2TiO3 during neutron irradiation because tritium released behavior was changed to be delayed due to adsorption of oxidized tritium on the CuO. Hence, the tritium collection method with hydrophobic catalyst instead of the CuO was demonstrated and succeeded the accurate release measurement of tritium from Li2TiO3. With the method, we assessed the behavior of tritium release under the various conditions since tritium should be released from Li2TiO3 as the form of HT as much as possible from the view point of the fuel cycle. Our results indicated; promotion of isotopic exchange reaction on the surface of Li2TiO3 by addition of hydrogen in sweep gas is mandatory in order to release tritium smoothly from Li2TiO3 irradiated with neutrons; the favorable sweep gas to release as the form of HT was hydrogen added inert gas; and the temperature of Li2TiO3 was the dominant parameter to control the chemical form of tritium released from the Li2TiO3.

Damage cause and mechanism of the tungsten specimen under ELM-like heat pulse were examined by using a Nd:YAG laser experiment assuming the ITER divertor allowable heat load, 0.5 MJ/m2 and finite-element analysis in terms of thermal mechanical response. The thermal stress distribution with the heat load area was compared by FEM and the stress distribution by a Nd:YAG laser corresponded to that expected in the actual divertor target. The crack growth direction was changed from a perpendicular to parallel and this was caused by the thermal stress distribution in isotropic microstructure under ELM-like heat pulse. In addition the change of crack growth direction caused the deterioration of thermal conduction near the surface and it led to the local surface melting near cracks even at 0.51 MJ/m2 with 3000 heat load cycles.

In test blanket module (TBM) program blanket major functions, which are tritium production, heat extraction for electric power production, and neutron shielding, will be demonstrated under fusion reactor environments in ITER. National Institutes for Quantum and Radiological Science and Technology (QST) has performed developments of water cooled ceramic breeder (WCCB) TBM as a primary concept in Japan. The container design of WCCB TBM was changed from the Box-shaped to a cylindrical structure to increase tritium breeding capability while maintaining pressure resistance under In-Box LOCA caused by water ingress into the container. The nuclear responses related to the blanket major functions would be changed with the container design change. The nuclear responses of WCCB TBM with the container design change were evaluated and the effects of the changed nuclear responses were described in this study. Tritium production rate increased by about 2 times with the container design changes. In contrast neutron shielding performance decreased due to the container design change. Degradation of neutron shielding in the cylindrical TBM led to an increase in nuclear heating of the TBM frame as well as an average of neuron flux behind TBM-set.

The temperature of Li 2 TiO 3 pebble breeder in a fusion DEMO blanket is assumed to be more than 1000 K. For the investigation of tritium release from a Li 2 TiO 3 pebble breeder blanket at such a high temperature, we have carried out a tritium release experiment with the DT neutron source at the JAEA-FNS. The Li 2 TiO 3 pebble (1.0–1.2 mm in diameter) of 70 g was put into a stainless steel container and installed into an assembly stratified with beryllium and Li 2 TiO 3 layers. During the DT neutron irradiation, the temperature was kept at 1073 K with wire heaters in the blanket container. Helium gas including 1% hydrogen gas (H 2 /He) mainly flowed inside the container as the purge gas. Two chemical forms, HT and HTO, of extracted tritium were separately collected during the DT neutron irradiation by using water bubblers and CuO bed. The tritium activity in the water bubbler was measured by a liquid scintillation counter. To investigate the effect of moisture in the purge gas, we also performed the same experiments with H 2 O/He gas (H 2 O content: 1%) or pure helium gas. From our experiment at 1073 K, in the case of the purge gas includes H 2 , it is indicated that the increasing tendency of HT release is similar to that of the dry H 2 /He.

The authors have applied the isotherm equations derived from the reactive vacancy solution theory (RVST) to correlation of experimental and highly non-ideal adsorption isotherms of hydrogen and deuterium on a mordenite adsorbent, and have examined the ability of the isotherm equations to match this correlation. Several isotherm equations such as Langmuir, Freundlich, Toth, Vacancy Solution Theory and so forth were also tested, but they did not work. For the Langmuir-Freundlich equation tests have indicated that its 'ability to correlate' of the adsorption isotherms is not satisfactory. For the multi-site Langmuir-Freundlich (MSLF) equation the correlation of the isotherms appears to be somewhat improved but remains unsatisfactory. The results show that the isotherm equations derived from RVST can better correlate the experimental isotherms.

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 con...

Li2TiO3 advanced tritium breeders with excess Li (Li2+xTiO3+y) are stable in a reducing atmosphere at high temperatures. Although the tritium release properties of tritium breeders are documented in databases for DEMO blanket design, no in situ examination under fusion neutron (DT neutron) irradiation has been performed. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed, and DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. Considering the tritium release characteristics, the optimum grain size after sintering is >5 μm. From the results of the optimization of granulation conditions, prototype Li2+xTiO3+y pebbles with optimum grain size (>5 μm) were successfully fabricated. The Li2+xTiO3+y pebbles exhibited good tritium release properties similar to the Li2TiO3 pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water.

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 tran...

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/cm 3 . 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 m 2 /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.

The Li2TiO3 is one of candidate breeding materials of a water cooled ceramic breeding (WCCB) blanket. In order to clarify the tritium recovery property of the WCCB blanket with the Li2TiO3 breeding material, we have performed the tritium recovery online experiment with the DT neutron source at the Fusion Neutronics Source facility in Japan Atomic Energy Agency (JAEA-FNS). We irradiated an experimental assembly simulating the WCCB blanket and recovered the tritium recovered from the Li2TiO3 pebbles put into the assembly with a heater system, sweep gases and bubblers. The activity of the recovered tritium was measured with a liquid scintillation counter. From our tritium recovery online experiment and calculation, the followings were found out: (1) the recovered tritium corresponded to the calculated tritium production within the experimental error in the range of 573–1073 K and (2) the recovered HTO tended to be easily recovered at lower temperature and high water moisture. The recovered HT increases at higher temperature and dry hydrogen circumstance. However, the maximum level of the tritium gas recovery is around 90% even at higher temperature and 1% H2 circumstance.

Tritium recovery system using adsorption or catalytic isotope exchange has already been proposed for a solid breeding blanket system of a nuclear fusion reactor. Synthetic zeolite is often used as an adsorbent or a substrate of chemical exchange catalyst. And, it is well known that its property is changed easily by exchanging its cations. Synthetic mordenite is one of zeolites having fairly large silica/alumina ratio. There are many reports about hydrogen adsorption properties of cation-exchanged mordenite so far. And, the present authors also have reported that cation-exchanged mordenite with Ca ion (Ca–MOR) indicated fairly large hydrogen adsorption capacity at 77 K in comparison with Molecular Sieves 5A (MS5A). So, in this work, hydrogen adsorption properties of cation-exchanged mordenite with transition metal ion were investigated mainly. The cation-exchanged mordenite with Ag ion (Ag–MOR) has indicated considerably large hydrogen adsorption capacity in lower pressure range at 77 K in comparison with Ca–MOR. The discussion from the viewpoint of adsorption rate is still remaining, but more compact cryosorption column for tritium recovery system is possible to design if Ag–MOR is adopted.

The multi-purpose RI facility has been constructed at Rokkasho site in DEMO R&D building until 2011. The facility is the first and quite unique facility in Japan, where tritium, beta and gamma RI species, and beryllium (Be) can simultaneously be used. The amounts of tritium used and stored are 3.7 TBq per glove box and 7.4 TBq, respectively. Some tritium water samples of 38 GBq has been stored at the equipment on March 2012. The material of the column of the micro gas chromatograph has been studied to develop a real time analysis tool for the hydrogen isotope composition in gas phase. The calorimeter has also been studied as a possible tritium measurement method in solid waste. A set of basic data on the interaction between materials and tritium has been measured by various methods. The behavior of tritium in Fe and W has been studied as a typical subject. As a study for the tritium durability, the endurance of ion exchange membrane has been tested by using high concentration tritium water. The curves of strength vs. dose for the Nafion membranes in tritium water were well consistent with those by gamma rays and electron beams irradiations.

Hydrogen isotope separation is important process in fusion reactor. Cryosorption is one of the methods for hydrogen isotope separation, and is often employed to gas chromatograph (GC) for hydrogen isotope analysis. Synthetic zeolite is used as the packing material of separation column for GC. And, it is well known that the retention time becomes long at 77 K. A zeolite having good separation capability at higher temperature may bring short retention time with good sensitivity. Mordenite type zeolite (MOR) in which cation was exchanged has been reported having comparatively good isotope separation capability. The present authors have investigated adsorption capacity of hydrogen isotope on cation-exchanged MORs at various temperatures. And, it has been shown that MOR with Ca 2+ as exchanged cation (Ca-MOR) has fairly large adsorption capacity. In this work, the breakthrough curves of hydrogen adsorption on Ca-MOR at 194, 175 K have been observed, and mass transfer coefficients or effective diffusivity in Ca-MOR particle have been estimated. The rate-controlling step is the diffusion process in the adsorbent particle. The isotope effect on hydrogen diffusion in Ca-MOR particle is larger than raw material (Na-MOR). Therefore, Ca-MOR is more suitable for the separation column than Na-MOR.

Since tritium (T) handled in fusion environment distributes so widely in its concentration and chemical form, no single T-measuring method can cover the wide concentration ranging from environment level (a few Bq) to carrier free level (GBq or above) and distinguish various tritiated compounds (gas, water, and organics ). In principle, any methods used for hydrogen measurements can be used for T measurements. However, safety requirements owing to the radioactivity of T give limitation in the measurements. Furthermore, electric noises caused by β-electrons emitted at T decay often disturb the measurements. On the other hand, they are allowed to use the radioactivity measurement. Nevertheless, their energy is so low to make their detection difficult. In this chapter, the principle of T detection and measurements is introduced, and its applications are described targeting quantitative analyses of wide ranges of T in gas, liquid, and solid separately. In a fusion reactor, T in plasma is a new target to be quantified, which is described in Chap. 8, separately.

In a fusion reactor, the prediction of tritium release behavior from breeder blanket is important to design the tritium recovery system, but the amount of tritium generated is necessary information to do that. Hence, tritium generation and recovery studies on lithium ceramics packed bed have been started by using fusion neutron source (FNS) in Japan Atomic Energy Agency (JAEA). Lithium titanate (Li 2 TiO 3 ) was selected as tritium breeding material, and its packed bed was enclosed by the beryllium blocks, and was kept at certain temperature during fusion neutron irradiation. During irradiation, the packed bed was purged with the sweep gas continuously, and tritium released was trapped in each gas absorber selectively by chemical form. In this work, the effect of sweep gas species on tritium release behavior was investigated. In the case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in the case of sweep by helium without water vapor, tritium in gaseous form was released first, and release of tritium in water form was delayed from gaseous tritium and was gradually increased.

In JAEA, the tritium processing and handling technologies have been studied at TPL (Tritium Process Laboratory). The main RD the basic tritium behavior in confinement materials; and detritiation and decontamination. The R&D activities on tritium processing and handling technologies for a demonstration reactor (DEMO) are also planned to be carried out in the broader approach (BA) program by JAEA with Japanese universities. The ceramic proton conductor has been studied as a possible tritium processing method for the blanket system. The BIXS method has also been studied as a monitoring of tritium in the blanket system. The hydrogen transfer behavior from water to metal has been studied as a function of temperature. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). A new hydrophobic catalyst has been developed for the conversion of tritium to water. The catalyst could convert tritium to water at room temperature. A new Nafion membrane has also been developed by gamma ray irradiation to get the strong durability for tritium.

Cryogenic adsorption is effective for the recovery of low-concentration hydrogen isotopes in bulk helium gases. In a fusion power plant, application of this process is foreseen for the recovery of ...

Beta ray induced X-ray spectrometry (BIXS) is one of the methods applicable to tritium gas monitor. It can measure tritium by counting the X-ray that is induced by interaction between the beta ray ...

Electrochemical hydrogen pump can extract hydrogen isotopes selectively from a blanket sweep gas, because one of its driving forces of hydrogen transportation is electric potential difference. In this work, transportation properties of multi-component hydrogen isotopes including tritium were investigated experimentally. Ceramic proton conductor membrane used in this work was SrCe 0.95 Yb 0.05 O 3− α made by TYK Co. H 2 –D 2 mixture balanced with He was supplied to the hydrogen pump, that was kept at 873 K and applied the voltage between 0 and 1000 mV, and the permeated gas was analyzed with a gas chromatograph. The rate-determining step of hydrogen transfer via ceramic membrane seems to be close to the diffusion step in the bulk of ceramic with increase of the applied voltage. HT–H 2 mixture balanced with He was also supplied to the hydrogen pump, and recovery of tritium via ceramic membrane was done. Diffusivity of hydrogen in the proton conductor ceramic is larger than that of tritium. Hence, hydrogen is enriched in the gas of the permeation side. In this experimental condition, about 40% of tritium supplied was recovered and the decontamination factor was 1.5.

Cryogenic adsorption is effective for the separative recovery of hydrogen isotopes of small concentrations from the bulk helium gas. Thus, the cryogenic adsorption method is considered to b...

Cryogenic adsorption is effective for the separative recovery of hydrogen isotopes of small concentrations from the bulk helium gas. The authors performed a screening test to find candidate adsorbents for the recovery of hydrogen isotopes from the bulk helium gas at liquid nitrogen temperature. The screening test indicates that a natural mordenite adsorbent has a quite high adsorption capacity for hydrogen under the helium atmosphere. The effect of the ion exchange for the natural mordenite on the adsorption capacity of hydrogen was also investigated using protium and silver as well. With regard to the adsorbent examined in the screening test, the adsorption characteristics of deuterium were also investigated. For the adsorption of deuterium, it was found that the natural mordenite adsorbent have a high adsorption capacity. The isotope effect on the adsorption of hydrogen isotopes on the natural mordenite adsorbent is not large compared with the MS5A adsorbent.

Synthetic zeolite is the candidate material of the separation column of the gas chromatograph for the hydrogen isotope analysis. Mordenite (MOR) is one of the zeolite, and has been reported that th...

A gas chromatography using a cryogenic separation column is one of the methods for hydrogen isotope analysis. However, the use of the refrigerant such as liquid nitrogen is the cause of long analysis time and is not suitable for easy installation. The development of the column material having separation capability at comparatively high temperature region is one of the solutions for these weak points. Mordenite (MOR) is a kind of a zeolite, and it has been reported that the separation column using MOR has possibility to separate hydrogen isotope mixture at comparatively high temperature. In this work, the separation columns using MOR were made and tested. The peaks of H2 and D2 were mostly separated at 144 K, but they were not separated at 195 K. MOR column adjusted in this work was still not for the practical use. However, this result suggests the possibility of the existence of the synthetic zeolite that can separate hydrogen isotope mixture at comparatively high temperature.

Electrochemical hydrogen pump with ceramic proton conductor membrane has been proposed to apply for a blanket tritium recovery system (BTR) of a fusion reactor. The advantage of this system is that it can apply to the system of low hydrogen pressure, because the driving force of hydrogen permeation is the electric potential difference. Perovskite-type ceramic such as SrCe0.95Yb0.05O3−α (SCO) is one of the candidates of membrane. To apply hydrogen pump to BTR, the enhancement of the hydrogen transportation capability is necessary. Modification of electrode is one of the methods to enhance the hydrogen transportation capability. In this work, the electrodes of platinum (Pt) and palladium (Pd) were attached to the SCO sample by the sputtering method (sputtering electrode), and its electric conductivity and proton conductivity were measured. Then, they were compared with that of the usual Pt paste electrode. Hydrogen transportation capability was enhanced when the sputtering electrode was applied. Especially, in case of the Pd sputtering electrode, the current density which was about 4 or 5 times larger than the usual Pt paste electrode was observed at 0.1% of H2 concentration.

In a fusion reactor system, a monitoring of hydrogen isotopes including tritium is necessary for the safety of system control and operation. A gas chromatography using a cryogenic separation column is one of the methods for hydrogen isotope analysis. Synthesis zeolite such as molecular sieve 5A (CaA) is a candidate material of the separation column, and its property varies by the ratio of silica to alumina, the kinds of cation and so on. If the factor affected the hydrogen adsorption property of the synthesis zeolite is clarified, it may lead to the development of the new zeolite optimized to the separation column. So, in this work, adsorption capacity of hydrogen (H2) and deuterium (D2) for mordenite (MOR) and NaY type zeolite (NaY) were investigated at various temperatures, and were compared with CaA. The amount of adsorption per unit weight of MOR was larger than that of CaA, and that of NaY was smaller than that of CaA. The adsorption isotherms were expressed by sum of two Langmuir equations, and the Langmuir coefficients of H2 and D2 were proposed. Furthermore, the Langmuir coefficients of HD, HT, DT and T2 were estimated by the reduced mass. The correlation between the adsorption properties and the physical parameters of the zeolite were not confirmed.

The blanket tritium recovery system using the electrochemical hydrogen pump with proton conductor membrane has been proposed. The feature of the electrochemical hydrogen pump is that the driving force of hydrogen transportation is a potential difference. Therefore, it might be effective to apply the hydrogen pump to the blanket sweep gas (the low hydrogen and water vapor pressure). Perovskite-type ceramic such as SrCe 0.95 Yb 0.05 O 3− α , is one of the candidate proton conductor for hydrogen pump and its ionic hydrogen transportation properties have been investigated. In this work, the basic mass transfer equation for hydrogen, in which the apparent proton conductivity is used as the over-all mass transfer coefficient, is proposed. And then, the apparent proton conductivities were estimated from experimental data using these equations, and mass transfer of hydrogen via proton conductor membrane was discussed by using the apparent conductivity.

Our idea on the DEMO plant is that it must demonstrate (1) an electric power generation of one GW level, (2) self-sufficiency of tritium fuel (TBR is more than 1.05), (3) year-long continuous operation, etc. At the same time, DEMO is expected to use technologies to be proven by 2020 and present an economical prospect of fusion energy in the operational time of the reactor. The design guidelines for the blanket are defined in order to meet the mission of the DEMO plant as mentioned above. Major design conditions are surface heat flux of 0.5 MW/m2 with peaking factor of 2, a neutron wall load of 3.5 MW/m2 with peaking factor of 1.5 and a neutron fluence of about 10 MW/m2. To moderate the technological extrapolation, reduced activation ferritic steel (F82H) structural material, Li2TiO3 and Be neutron multiplier are considered. To improve the economical aspect, supercritical water with inlet/outlet temperatures of 280/510 °C is chosen as coolant material, with coolant pressure of 25 MPa. As a result, a thermal efficiency of 41% is achieved. To obtain higher plasma performance, MHD instabilities suppressing shell structure is adopted with structural compatibility to the blanket structure. To meet higher plant availability requirements (more than 75%), the hot cell maintenance approach is selected for the replaceable power core components.

Several systems for the tritium recovery from cooling water of a blanket of a fusion power plant have been designed. The tritium concentration in the cooling water and the tritium permeation rate to the coolant is assumed to be 370 GBq(10 Ci)/kg, 13 and 130 g/day, respectively. For the case of 13 g/day, the system can be composed of a water distillation (WD: 2.6 m inner diameter and ∼50 m height) and a catalytic exchange column with an electrolysis cell (0.7 m inner diameter and 22 m in height). The WD column can be replaced by a system of catalytic exchange columns in vapor and liquid phases. For the case of 130 g/day, no solution can be found as long as quite a large flow rate to a main fuel cycle for its final treatment is permitted. Otherwise, an electrolysis cell that can be used under a high concentration of tritium water will need to be developed. The tritium inventory of the WD column is appreciably large, so that it is desirable to develop a system having a high separation factor and with no liquid phase.

Tritium bred in the solid breeder blanket of a fusion reactor is extracted by passing H 2 added helium sweep gas through the blanket itself. In the blanket tritium recovery system (BTR), tritium is separated from sweep gas. Palladium (Pd) membrane diffuser is one of the applicable processes for BTR. Recently, the conceptual design study on the demonstration reactor with supercritical water cooling blanket, that is named ‘DEMO2001’, has been carried out in JAERI. In this report, the application of the Pd diffuser to the blanket sweep gas condition is discussed based on DEMO2001 conditions. The counter flow may be the most opportune flow type for the Pd diffuser. However, Pd diffuser is suitable for the secondary process, like the purification process, after the tritium in the sweep gas has been concentrated by another method.

A simulated fuel processing (cryogenic distillation columns and a palladium diffuser) and CMSB (cryogenic molecular sieve bed) systems were linked together, and were operated. The validity of the CMSB was discussed through this experiment as an integrated system for the recovery of blanket tritium. A gas stream of hydrogen isotopes and He was supplied to the CMSB as the He sweep gas in blanket of a fusion reactor. After the breakthrough of tritium was observed, regeneration of the CMSB was carried out by evacuating and heating. The hydrogen isotopes were finally recovered by the diffuser. At first, only He gas was sent by the evacuating. The hydrogen isotopes gas was then rapidly released by the heating. The system worked well against the above drastic change of conditions. The amount of hydrogen isotopes gas finally recovered by the diffuser was in good agreement with that adsorbed by the CMSB. The dynamic behaviors (breakthrough and regeneration) of the system were explained well by a set of basic codes.

In the solid breeder blanket of fusion reactor, bred tritium is planned to be extracted from the blanket as HT by passing of H{sub 2}-added sweep gas in general. In that case, tritium leakage by permeation to coolant can not be ignored. So, the application of H{sub 2}O-added sweep gas is discussed, with which tritium leakage to coolant can be much reduced. As the result of discussion, H{sub 2}O-added sweep gas is probable method of tritium recovery. For the further detailed discussion, it is important to enrich the data correlated to the interaction of H{sub 2}, H{sub 2}O, breeder, multiplier and structures.

A blanket tritium recovery system that uses an electrochemical hydrogen pump with a protonic conductor membrane is proposed. One of the advantages of this system is the potential for processing the...

This paper presents results of conceptual design activities and associated RD neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was successfully fabricated. It withstood the high heat flux test at 2.7 MW m−2. Also, a correlation parameter of the Li2TiO3 pebble bed made by the sol–gel method was verified by measurement of the thermal conductivity of the breeder pebble bed, which is one of the most important design data.

Various analytical techniques exist to determine the compositions of gases handled in the fuel cycle of future fusion machines. Gas chromatography was found to be the most appropriate method. The main disadvantages of conventional gas chromatography were the long retention times for the heavy hydrogen species of >30 min. Recent progress in the development of micro-gas chromatography has reduced these retention times to ≈3 min. The usefulness of micro-gas chromatography for the analysis of hydrogen and impurity gas mixtures in the fuel cycle of future fusion machines is presented and the advantages and drawbacks are discussed.

JAERI has been developing an improved compact tritium gas enrichment/recycling system to handle relatively smaller amount of tritium for efficient researches on tritium processing and development work. An electrochemical hydrogen pump based on CaZr 0.9 In 0.1 O 3- α proton conductor is investigated to separate hydrogen isotopes directly from inert gas carrier after the enrichment process of the system. The prototype hydrogen pump was developed for the first time to operate under vacuum condition at the cathode compartment, and the performance test was carried out.

The blanket tritium recovery system using the hydrogen pump with solid electrolyte membrane has been proposed by the present authors. Perovskite-type ceramic such as SrCe 0.95 Yb 0.05 O 3-a , is one of the candidate protonic conductor for hydrogen pump and its ionic hydrogen transportation properties are being investigated. Deuterium transportation properties were investigated and were compared with H 2 to understand the isotope effect. The basic hydrogen isotope transportation property of SrCe 0.95 Yb 0.05 O 3-a , its technical feasibility, and issues for further development toward practical devices were revealed.

In this design work, a combination of the hydrogen pumps in charge of the function of hydrogen isotopes recovery and the oxygen pump was adopted to the blanket tritium recovery system for the prototype fusion reactor designed by Japan Atomic Energy Research Institute (JAERI). The main functions of this system are described below. 1) Transport of tritium with helium purge gas: tritium released from ceramic breeding material in the blanket is transported to the tritium recovery system by the helium purge gas which contains a small amount of hydrogen gas. 2) Steam electrolysis and removal of oxygen gas: the oxygen pump with electrolyte of oxygen ionic conductors electrolyzes the steam (H 2 O and HTO) contained in the purge gas into hydrogen isotopes and oxygen, and simultaneously removes impurity of oxygen by electrical membrane permeation. 3) Recovery of hydrogen isotopes: the hydrogen pump with electrolyte of protonic conductors electrically recovers the pure hydrogen isotopes (HT and H 2 ) from the purge gas. Based on the experimental data obtained by feasibility study and the present design effort, it was revealed that the simple and continuous tritium recovery system for gaseous stream is possible and attractive for fusion power reactors.

Analysis of hydrogen isotopes is very important in the fuel cycle system of fusion reactors. Gas chromatography with a cryogenic separation column is one of the most extensively used methods for the analysis of hydrogen isotopes. Micro gas chromatograph (micro GC) with a cryogenic column is expected to improve the analysis time, that is a major disadvantage of conventional GC. The present authors have modified the micro GC to use its separation column at cryogenic temperature. In previous work a micro packed column has been tested and indicated fairly good performance. In this work a capillary column was tested at cryogenic temperature for more improvement of analysis time. Obtained retention time of H 2 , HD and D 2 were about 43, 47 and 54 s, respectively, for the column with 0.53 mm of I.D., 4.0 m of length and 0.08 mm of film thickness. Peak resolution between H 2 and HD was about 1.12. These results suggest that the column developed in this work attained the practical level for the separation and short analysis time. The estimation of the retention time was carried out using the dispersion model. The retention times of HT, DT and T 2 were also estimated using the reduced mass.

Electrochemical properties of the ceramic protonic conductor cell were investigated to evaluate its feasibility of hydrogen pumping for the purpose of tritium extraction in fusion fuel system. Expe...

In the fuel cycle system of fusion reactors, analysis of hydrogen isotopes is very important from the view point of system control. The gas chromatograph (GC) with cryogenic separation column (cryogenic GC) is one of the most extensively used methods for the analysis of hydrogen isotopes. The micro GC with cryogenic column is expected to improve analysis time, that is a major disadvantage of conventional GC. The present authors have modified the micro GC to use its separation column at cryogenic temperature for H2, HD and D2 mixture analysis. Obtained retention time of H2, HD and D2 was about 85, 100 and 130 s, respectively. Peak resolution between H2 and HD, these are nearest each other, was about 1.0. These result suggests that the column developed in this work attained the practical level for the separation of hydrogen isotopes without tritium. Present detection limit of hydrogen isotopes was about 100–200 p.p.m., and it can be improved further by adjustment of separation column.

In order to confirm tritium confinement ability in the deuterium–tritium (DT) fusion reactor, intentional tritium release experiments have been started in a specially fabricated test stand called ‘Caisson’, at Tritium Process Laboratory in Japan Atomic Energy Research Institute. The Caisson is a stainless steel leak–tight vessel of 12 m3, simulating a reactor room or a tritium handling room. In the first stage experiments, about 260 MBq of pure tritium was put into the Caisson under simulated constant ventilation of four times air exchanges per h. The tritium mixing and migration in the Caisson was investigated with tritium contamination measurement and detritiation behavior measurement. The experimental tritium migration and removal behavior was almost perfectly reproduced and could almost be simulated by a three-dimensional flow analysis code.

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The cryosorption method is useful for extracting hydrogen isotopes from a helium gas stream with a small amount of hydrogen isotopes. Therefore, in fusion reactors, this method is expected to be applied for the helium glow discharge exhaust gas processing system and the blanket tritium recovery system. To design these systems, adsorption isotherms for each hydrogen isotope are needed, making it possible to estimate the amount of adsorption in a wide pressure range. The amount of tritium adsorption on molecular sieve 5A, molecular sieve 4A, and activated carbon, which are potential adsorbents in the cryosorption bed, at liquid nitrogen temperature were quantified using the volumetric method, It was found that adsorption isotherms of tritium were also expressed with the two-site Langmuir model and that the obtained isotherms were close to the reported isotherms, the Langmuir coefficients for which were estimated using a reduced mass of hydrogen isotopes.

An integrated fusion fuel loop was constructed. A palladium diffuser first treated a simulated plasma exhaust gas, that is a mixture of hydrogen isotopes and impurities such as helium and methane. ...

Although lithium ceramic materials such as Li 2 O, LiAlO 2 , Li 2 ZrO 3 and Li 4 SiO 4 are considered as breeding materials in the blanket of a D-T fusion reactor, the release behavior of the bred tritium in these solid breeder materials is not yet fully understood. Most results of in situ tritium release experiments are analyzed assuming that the overall release process of tritium is mainly controlled with tritium diffusion in the crystal grain of a solid breeder material. However, the diffusivities by various authors for solid breeder materials do not agree with each other. An insufficient estimation of the surface reactions, the effect due to irradiation defects or the system effect may be the reason. Effects of diffusion of tritium in the grain, absorption of water in the bulk of grain, and adsorption of water on the surface of grain, together with two types of isotope exchange reaction on the tritium inventory in the steady-state condition, are discussed in this study. Comparisons of tritium inventory estimated in this study with data obtained in several in-situ experiments are also performed, and good agreement is obtained when the existence of some water vapor is assumed in the purge gas.