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Tritium accountancy in the fusion DEMO plant has been studied at JAERI. Tritium accountancy is indispensable from the viewpoints of safety and plant operation because of the large amount of tritium used as fuel. DEMO is the quasi-steady continuous operation machine to demonstrate the feasibility of a fusion reactor as a commercial power production machine and it has functions of power production and tritium breeding. In addition to the basic accounting techniques for large amount of tritium in the tritium plant of the fusion system, which will be established in the ITER project, it is necessary to establish accounting techniques for tritium consumption and production. Dynamic accounting techniques have to be much improved or developed for optimum and safe control of the continuously operating fuel system of DEMO.

Two nuclear microprobe techniques; deuteron-induced nuclear reaction analysis (d-NRA) and neutron elastic recoil detection analysis (NERDA), are presented to obtain hydrogen isotope depth profiles of fusion reactor materials in a shallow and deep region, respectively. Tritium and deuterium depth profiles in a bumper limiter of TFTR used in the D–T experiments were measured as an application of d-NRA. The T(d, α)n and D(d, t)H reactions were used to determine the depth profiles. The tritium concentration had a peak, whereas deuterium showed a broad distribution in the profile. Also, the application of NERDA is demonstrated. A proof-of-principle experiment was performed using a standard sample of deuterated polyethylene. The depth profile and resolution for NERDA are discussed.

Characterization of Tokamak exhaust gas, gas species and their concentrations, is important for the design of the tritium fuel processing system. The exhaust gas from JT-60U during the experimental campaign has been investigated with micro gas chromatography (MGC). Gas species, such as hydrogen isotopes, CH 4 , C 2 H 2 + C 2 H 4 , C 2 H 6 and CO 2 , in the exhaust gas from a plasma discharge could be identified successfully. In this campaign, the total ratio of hydrogen isotopes and impurity species in exhaust gas was estimated to be 39 to 1. From the comparison between both amount of gas species and some plasma parameters, there was a tendency observed indicating that the amount of hydrogen isotopes and carbon compounds increased with maximum electron density. This relation suggests that the amount of hydrogen isotopes and carbon compounds in the exhaust gas is increases with the extension of plasma performance.

The influence of blistering on deuterium retention in polycrystalline tungsten (W) has been studied at incident energies of tens of eV and a fixed incident flux of 1 × 1022 D/m2/s. The surface morphology was observed using scanning electron microscopy (SEM) and the deuterium retention in the irradiated samples measured using thermal desorption spectrometry (TDS). The results indicate that at substrate temperature around 400–500 K dense and high dome-shaped blisters appeared at the surface and the deuterium retention also reached its maxima. The blisters then became lower and sparse and finally disappeared at 900 K. Additionally, the D retention decreased to a quite low level above 700 K. Thermal desorption spectra of the deuterium in the irradiated samples showed four peak temperatures as a function of the substrate temperature during irradiation, indicating different types of trapping sites.

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.

The radiation durability of a Nafion ® ion exchange membrane in the solid-polymer-electrolyte (SPE) water electrolyzer component of the ITER water detritiation system was investigated from various viewpoints. Nafion ® is composed of a PTFE (polytetrafluoroethylene) backbone and side chains terminating with sulfonic acid. Nevertheless under the condition that polymer specimens are soaked in water, degradation of Nafion ® by γ-ray irradiation differs significantly from that of PTFE. It is made clear that the durability of Nafion ® is distinctively higher than that judged from its backbone structure. No serious damage was observed in the membrane properties up to ITER-WDS requirement (530 kGy).

For developing an alternative mean for on-line and real-time tritium monitoring, the method of detecting bremsstrahlung X-rays was examined by measuring the counting rate of bremsstrahlung X-rays as a function of pressure in two mixed tritium gases diluted with hydrogen or helium. At a constant tritium partial-pressure ratio of 0.010, the relationship between the counting rate of bremsstrahlung X-rays, X (cpm) and the total pressure, P (Pa) can be expressed as X = 4.1 × 105 (1 − e−4.0 P/100,000). The counting rate depends only on tritium partial-pressure and the total pressure if the mixed gases contain no other species but hydrogen isotopes and helium isotopes. In addition, the relation between X and P can be simplified to a linear function in the low-pressure region of P ≪ 2.5 × 104 Pa.

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.

Polycrystalline tungsten samples have been irradiated at near room temperature by high flux (1 × 1022 D/m2/s) deuterium plasma beams with incident ion energies ranging 7–98 eV/D. Surface blistering occurred at all energies as observed by means of scanning electron microscopy. At all energies, the blisters increased in their size and number with fluence within the corresponding low fluence ranges. The size increase tended to saturate at certain fluences within the experimental fluence ranges, which might be attributed to rupturing of blisters. The critical fluence for blistering Φcr was found to increase with decreasing the incident energy. At energies

In order to eliminate atmospheric tritium gas (HT) released from tritium handling apparatus, we proposed to use the HT oxidizing ability (hydrogenase enzyme) of bacterial strains isolated from surface soils instead of a high temperature precious metal catalyst. Among the isolated strains with high HT oxidation activity, several strains were selected to develop a tritium elimination (detritiation) system. Bioreactors were made of bacterial cells grown on agar medium on a cartridge filter and stored in a refrigerator until use. The detritiation ability of these bioreactors at room temperature was investigated during the intentional HT release experiments carried out in the Cassion Assembly for Tritium Safety Study (CATS) in TPL/JAERI. When HT contaminated air from the CATS was introduced into the biological detritiation system, in which three bioreactors were connected in series, 86% of HT in air was removed as tritiated water in these bioreactors at a flow rate of 100 cm 3 /min for 2 hours.

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.

To obtain performance data of atmosphere detritiation system at the off normal events such as fire for the safety of ITER, the detritiation experiment was planned and performed at Tritium Process Laboratory (TPL) in Japan Atomic Energy Research Institute (JAERI) using a new scaled detritiation system for the oxidation performance test which can process gas flow rate of {approx}2.64 m{sup 3}/hr in circulation through 2m{sup 3} tank. The detritiation system consists of two oxidation catalyst beds (473K and 773K) for converting hydrogen isotopes and tritiated methane in compounds to water vapor and a molecular sieve drying absorber for removing water vapor as the usual detritiation system. In this time, the performance of oxidation catalyst bed of the detritiation system for hydrogen and methane under existence of carbon monoxide or carbon dioxide which are produced in the fire was investigated.Basic performance of the detritiation system for hydrogen (1.9%) and methane (1.3%) in air was evaluated under maximum ventilation flow rate (2.64m{sup 3}/h). Obtained oxidation efficiency was more than 99.99% for hydrogen in the catalyst bed at 473K and more than 99.9% for methane in the 773K one, respectively. It was confirmed that these performances were maintained even under carbon dioxide ofmore » up to 20% , carbon monoxide of up to 10% if sufficient oxygen remained in the process gas, and that the existence of carbon monoxide and carbon dioxide at the fire would not influence the performance of the oxidation catalyst bed in the detritiation system.« less

At the Tritium Process Laboratory (TPL) in Japan Atomic Energy Research Institute (JAERI), the three-dimensional 'TBEHAVIOR' code has been developed and improved to understand initial tritium behavior and tritium confinement performance in a ventilated room of a fusion reactor in case of tritium leak event. The purpose of this study was mainly focused to; 1) investigate the effect of atmospheric exchange time per hour on the tritium confinement performance in an actual scaled tritium handling room after off-normal tritium release; 2) investigate the effect of atmospheric exchange time per hour on the time necessary for detecting tritium release; 3) investigate the suitable location of exhaust ducts and alarm monitors. The simulated room used in the present analysis is approximately 3000 m{sup 3} of tritium handling room (12.00 m{sup W}, 29.00 m{sup D} and 8.50 m{sup H}) with six supply ducts and six exhaust ducts. Atmospheric exchange time per hour is changed as a parameter from 0.67 to 3.33 times per hour.

A new conceptual design of a tritium permeation test module assembly was developed, for simulation of tritium permeation in the real plasma facing components and validation of the models and codes for evaluation of the tritium permeation. The assembly was designed for tests using powerful ion sources, which has a capability to simulate condition relevant to that of the ITER divertor. The heat load test of the prototype module has been performed using an electron beam to verify thermal and mechanical behavior of the bonded structure and to assess its structural integrity under the cyclic heat loads. Then, the first tests using tritium ion beam generated by the TPE apparatus at TSTA/LANL with the prototype module was performed and procedure to measure tritium permeated was established. Considerations for tests using the target module with defects generated by neutron irradiation or accelerated ion beam irradiation will be also taken in the new module design.

Addition of gas separation membrane process into the usual tritium removal process from an indoor atmosphere is attractive for a fusion plant, where a large amount of atmosphere should be processed. As a manner to improve the partial pressure difference between feed and permeated side, intended reflux of vapor and the hydrogen concentrated at permeated side is conceived to enlarge the partial pressure difference. Membrane separation with reflux flow has been proposed as an attractive process to enhance the recovery ratio of tritium component. Effect of reflux on the recovery ratio of tritium component was evaluated by numerical analysis. The effect of reflux on separation performance becomes striking as the target species have higher permeability coefficients. Hence, the gas separation by membrane with reflux flow is favorable for tritium recovery.

To have better understanding of radiochemical reactions among oxygen baking products in a fusion reactor, reactions in equimolar tritium molecule (T{sub 2}) and carbon dioxide (CO{sub 2}) were examined by laser Raman spectroscopy and mass spectrometry. After mixing them at room temperature, T{sub 2} and CO{sub 2} decreased rapidly in the first 30 minutes and then the reactions between them became much slower. As the predominant products of the reactions, carbon monoxide (CO) and tritiated water (T{sub 2}O) were found in gaseous phase and condensed phase, respectively. However, there likely existed also some solid products that were thermally decomposed into CO, CO{sub 2}, T{sub 2}, T{sub 2}O, etc. during baking up to 523 K.

Tritium behavior released in the ITER hot cell has been investigated numerically using a combined analytical methods of a tritium transport analysis in the multi-layer wall (concrete and epoxy paint) with the one dimensional diffusion model and a tritium concentration analysis in the hot cell with the complete mixing model by the ventilation. As the results, it is revealed that tritium concentration decay and permeation issues are not serious problem in a viewpoint of safety, since it is expected that tritium concentration in the hot cell decrease rapidly within several days just after removing the tritium release source, and tritium permeation through the epoxy painted concrete wall will be negligible as long as the averaged realistic diffusion coefficient is ensured in the concrete wall. It is also revealed that the epoxy paint on the concrete wall prevents the tritium inventory increase in the concrete wall greatly (two orders of magnitudes), but still, the inventory in the wall is estimated to reach about 0.1 PBq for 20 years operation.

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.

Pressure swing adsorption (PSA) has been studied as a new water processing technique, detritiation of tritiated water, for a future fusion power plant. It will be a new tritium removal method having an additional function of isotope separation and quick dehydration, and it is expected to become the first stage of the system processing a large amount of tritiated water generated in a fusion plant. A series of the adsorption and dehydration experiments was carried out for a typical adsorbent of NaX zeolite as fundamental investigation to realize HTO/H2O separation system by PSA. It was clearly observed that break through time differs in H2O and HTO concerning the isotope separation function of NaX zeolite. It is certain that NaX zeolite can separate into the tritium concentrated water and the tritium reduced water by this difference of the break through time. The quick dehydration is attained by decompression and purge gas flowing. It was observed that a part of amount of the water released by the decompres...

Japan Atomic Energy Research Institute (JAERI) has been performing blanket development based on the long-term research program of fusion blankets in Japan, which was approved by the Fusion Council of Japan in 1999. The blanket development consists of out-pile R and D, In-pile R and D, TBM Neutronics and TPR Tests and Tritium Recovery System R and D. Based on the achievements of element technology development, the R and D program is now stepping to the engineering testing phase, in which scalable mockup tests will be performed for obtaining engineering data unique to the specific structure of the components, with the objective to define the fabrication specification of test blanket modules for ITER. This paper presents the major achievements of the element technology development of solid breeder blanket in JAERI.

Tritium and deuterium depth profiles of the TFTR tile exposed to deuterium-tritium plasmas have been measured to reveal the hydrogen isotope behavior at the surface region by means of deuteron induced nuclear reaction analysis. The analyzed sample was a part of a tile made of a four-dimensional carbon fiber composite, which was placed at K bay, column C, row 16 of the TFTR vacuum vessel. Four kinds of elements, deuterium, tritium, lithium -6 and lithium -7, were identified. The tritium concentration had a peak at 0.5 μm with an atomic density of 7.4×1025 T/m3 in depth profile, whereas the deuterium showed a broad distribution up to the depth of 1.5 μm with atomic densities of 3.4×1027 D/m3. It is found that a fraction of the retained tritium from the surface to 1.5 μm, 8.1×1019 T/m2, corresponded to 2% of that from the surface to 1 mm, 1.0×1021 T/m2, which was measured for the KC-18 tile using the combustion method.

Tritium permeation through heat exchanger from primary cooling water to secondary cooling water has been investigated experimentally with SS316L heat exchanger under simulated ITER (international thermonuclear experimental reactor) operation condition in order to establish the tritium permeation evaluation method through the heat exchanger. As the result, the permeation rate of aqueous tritium was found to be about three orders smaller than that of the gaseous tritium. Tritium permeation through the heat exchanger in ITER was then evaluated, and it was revealed that total tritium permeation amount based on obtained aqueous permeability was about one order less than that with the former method with the gaseous permeability and putting the permeation reduction factor as 1000. Evaluated tritium permeation amount into secondary water during 20 years was quite small, which could be considered as negligible from the safety viewpoint.

The contribution of groundwater output and input to lake chemistry was examined by estimating the hydrological and chemical budgets of a volcanic caldera lake, Lake Kussharo, Hokkaido, Japan. The lake level, meteorology, river water discharge and water properties were measured in the ice-covered period of February–March and in the open-water period of June–October in 2000. The inorganic chemistry was then analyzed for sporadically sampled surface water and hot spring water. The chemistry of lake water at pH of 6.91–7.57 and EC25 (electric conductivity at 25 °C) of 29.2–32.7 mS/m appears to be controlled by the input of two types of hot spring water: the inflowing Yunokawa River (pH of 2.27–2.54 and EC25 of 197.8–258.0 mS/m) and groundwater discharging directly on the shore (pH of 7.13–8.32, water temperature of 35.0–46.5 °C and EC25 of 53.1–152.0 mS/m). Excluding the days with rainfall or a great change in lake level, the water budget in June–October gave a net groundwater input of −7.41 to 2.97 m3/s. A combination of the water budget with the chemical budget of two solutes, Na+ and Cl−, led to the best estimate of groundwater output, Gout, at 3.82±3.02 m3/s, the total fresh groundwater input, ∑Gfresh, at 2.14±1.00 m3/s, and the total groundwater input of hot springs, ∑Gspa, at 0.46±0.05 m3/s. This is comparable to G out =3.87 m 3 / s , ∑G fresh =1.49 m 3 / s and ∑G spa =0.41 m 3 / s during the ice-covered period. The chemical flux by the freshwater input plays an important role in the alkalinity of lake water, as does the chemical flux by the shoreline hot springs. The large groundwater output could occur by the leakage through the highly permeable, underground pumice, distributed from the east-to-south lake basin to southeast of the outlet.

To better understand radiochemical reactions in the process of tritium recovery from the in-vessel components of future deuterium–tritium fusion machines like international thermonuclear experimental reactor (ITER), reactions in a N2-balanced system of tritium oxides (T2O and possible T2O2) with carbon monoxide (CO) were examined by laser Raman, Fourier transform infrared (FT-IR) and mass spectrometry. The main products after the reactions were completed at 290 K are carbon dioxide (CO2) and tritiated water (T2O). Tritiated acids as found in T2–CO system were not detected in the reactions between T2O/T2O2 and CO at the same detection limit.

As part of study to establish in-vessel tritium reduction methods in carbon plasma facing fusion devices such as ITER, measurement of tritium removal characteristics of glow discharge methods, usually used for wall conditioning, have been applied to and examined in the vacuum vessel of JT-60U. Release rates of hydrogen isotopes (tritium and deuterium) as well as hydrocarbons from the JT-60U vacuum vessel induced by glow discharge cleaning (GDC) with He and H2 were measured. Release characteristics of hydrogen isotopes could be classified into three release processes, and each process was well described by a simple exponential decrease with time. It was found that H2 GDC showed the superior hydrogen isotope release characteristics to the He GDC, probably because of chemical processes, such as isotope exchanges assisted by the chemical sputtering process between discharged hydrogen and hydrogen isotopes in the plasma facing carbon tiles were enhanced by the H2 glow discharge. Based on the release kinetics observed in the present work, it is estimated that it takes several days to reduce tritium (deuterium) inventory within the removal depth by the H2 GDC in JT-60U to a half by continuous H2 GDC at 573 K.

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...

Permeation behavior of tritium (T) implanted into tungsten and nickel has studied by using pure tritium ion beam. Tritium permeation characteristics of steady and transient states were obtained and were compared with the result of deuterium (D) under the same experimental conditions. It was concluded from the steady state results of T and D that the permeation was controlled by diffusion in implantation side – diffusion in permeation side process for tungsten and recombination at the implantation surface – diffusion in the permeation side process for nickel. Effective diffusivities were evaluated from the permeation behavior of pure T and D. Isotope effect on the diffusivity between D and T did not agree with the classical diffusion theory, but could be explained by a quantum statistical diffusion model for nickel. The results indicated that diffusivity of T could be larger than that of D in W and Ni, respectively.

The codeposits on JT-60 tiles experienced hydrogen plasma burning were irradiated by focused beams of an excimer laser. The removal rate of the JT-60 codeposits was low when the laser energy density was smaller than the ablation threshold (1.0 J/cm 2 ), but reached to 1.1 μm/pulse at the laser energy density of 7.6 J/cm 2 . The effective absorption coefficient k in the JT-60 codeposits at ArF excimer laser wavelength was determined to be 1.9 μm -1 , which is almost one order smaller than the optical absorption coefficient at the same wavelength in graphite (16.4 μm -1 ). In the process of ablative removal of the codeposits, hydrogen was released predominantly in the form of hydrogen molecule and water formation could be ruled out. The temperature rise on the surface was measured on the basis of Planck's law of radiation, and the temperature during the irradiation at the laser energy density of 0.5 J/cm 2 decreased from 3570 K at the beginning of the irradiation to 2550 K at 1000th pulse of the irradiation.

Fuel processing technology and tritium safe-handling technology have been developed through US/DOE-JAERI collaboration from 1987 till 2001, and the technologies to construct the tritium plant system of ITER have been made currently available. This paper overviews the major achievements of this collaborative researches over fourteen years, which were performed mainly at the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory (LANL). The tritium plant system consists mainly of a fuel processing system, which includes a fuel cycle system and a blanket tritium recovery system, and a tritium confinement/removal system. The fuel cycle system recovers fuel from plasma exhaust gas and recycles it. In the collaboration, major key components and subsystems were developed, and the performance of the integrated system was successfully demonstrated over its one-month operation in which plasma exhaust model gas was processed at a processing rate of up to 1/6 level of the ITER. The technological basis of the fuel cycle system was thus established. Blanket tritium recovery technology was also successfully demonstrated using the TSTA system. Through the successful safeoperation of the TSTA, reliability of tritium confinement/removal system was verified basically. In addition, much data to confirm or enhance safety were accumulated by experiments such as intentional tritium release in a large room. Furthermore,distribution of tritium contamination in the vacuum vessel of the TFTR, a large tokamak of the Princeton Plasma Physics Laboratory (PPPL), was investigated in this work.

The exchange of tritium, implanted into an oxide ceramic with a pure tritium ion beam, for protium by exposure to normal air at room temperature has been measured in a collaboration experiment with the Tritium Laboratory group, JAERI. It is shown that the release of THO produced by the exchange H 2 O+T→THO+H at the surface of the crystallites decreases monotonically with exposure time, while the release of TH gas increases in the beginning, reaches a maximum and decreases monotonically as the time increases. The latter behavior of TH gas release gives an important evidence for bulk recombination of H uptaken from the surface with T implanted. These results are discussed based on the one-way diffusion model proposed previously for the explanation of the D–H exchange experiment.

Tritium and deuterium are expected to be the fuel for the first fusion power reactors. Being radioactive, tritium is a health, safety and environment concern. Room air tritium clean systems can be used to handle tritium that has been lost to the room from primary or secondary containment. Such a system called the Experimental Tritium Cleanup (ETC) systems is installed at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. The ETC consists of (1) two compressors which draw air from the room, (2) a catalyst bed for conversion of tritium to tritiated water, and (3) molecular sieve beds for collection of the water. The exhaust from this system can be returned to the room or vented to the stack. As part of the US–Japan fusion collaboration, on two separate occasions, tritium was released into the 3000 m3 TSTA test cell, and the ETC was used to handle these releases. Each release consisted of about one Curie of tritium. Tritium concentrations in the room were monitored at numerous locations. Also recorded were the HT and HTO concentrations at the inlet and outlet of the catalyst bed. Tritium surface concentrations in the test cell were measured before and at a series of times after the releases. Surfaces included normal test cell equipment as well as idealized test specimens. The results showed that the tritium became well-mixed in the test cell after about 45 min. When the ETC was turned on, the tritium in the TSTA test cell decreased exponentially as was expected. The test cell air tritium concentration was reduced to below one DAC (derived air concentration) in about 260 min. For the catalyst bed, at startup when the bed was at ambient temperature, there was little conversion of tritium to HTO. However, once the bed warmed to about 420 K, all of the tritium that entered the bed was converted to HTO. Immediately after the experiment, surfaces in the room initially showed moderately elevated tritium concentrations. However, with normal ventilation, these concentrations soon returned to routine levels. The data collected and reported here should be useful for planning for the operation of existing and future tritium facilities.

This paper describes the design of the ITER tritium plant subsystems, layout in the tritium building and the construction plan. The tritium plant comprises tokamak fuel cycle processing systems, as well as tritium confinement and detritation systems. The plant processes tritiated gases received from the tokamak and other sources to produce the D, T gas streams for fuelling, and detritiates various waste streams including tritiated water before discharge to the environment. The plant has been designed to meet not only all anticipated plasma operation scenarios in the DD and DT phases with a wide range of burn pulse durations from short pulse (450 s) and long pulse (3000 s), but also safety requirements (minimization of equipment tritium inventory and environmental tritium release from different accidental events in tokamak and tritium processing subsystems, and reduction of workers’ tritium exposure, etc).

A series of tests involving 10 h baking under the current ITER design conditions (240 °C with 933 Pa O 2 ) was performed using a cube of a carbon fiber composite tile that had been used in Tokamak Fusion Test Reactor (TFTR) during its deuterium–tritium burning operation. The removal rate of the codeposits was about 3 μm/h near the surface and 0.9 μm/h in the deeper region. Total amount of tritium released from the cube during 10 h baking was 202 MBq, while remaining tritium in the cube after baking was 403 MBq. Thus 10 h baking at 240 °C with 933 Pa O 2 removed 1/3 of tritium from the cube. After 10 h baking, the tritium concentration on the cube surface also dropped by about 1/3. In addition, some tritium was released from another cube of the tile during baking at 240 °C in pure Ar, and a rapid increase of tritium release was observed when the purging gas was shifted from pure Ar to Ar–1%O 2 . When a whole TFTR tile was baked in air at 350 °C for 1 h and then at 500 °C for 1 h, the ratios of tritium released were 53 and 47%, respectively. Oxygen reacted with carbon to produce carbon monoxide during baking in air.

The hydrogen isotope separation system (ISS) using cryogenic distillation method is one of the key systems in a fuel cycle loop of a fusion reactor. The ISS for the International Thermonuclear Experimental Reactor (ITER), is characterized by its required flexibility for the variability of feed composition, which is not considered in the existing industrial distillation columns, so that the development of evaluation method of dynamic behavior, which is the base for the establishment of ISS control system, is essential as original technology. At the Tritium Process Laboratory in Japan Atomic Energy Research Institute (TPL/JAERI), H-D-T cryogenic distillation experiments have been carried out to acquire the systematic data for the design of the ISS for the ITER. The value of height equivalent to a theoretical plate (HETP) was estimated to be 5 cm by the results of present experiments and it was adopted in the ITER–ISS design. The in situ, multi-points and reliable high-speed gas analytical system with Laser Raman spectroscopy was successfully demonstrated in the experiment to analyze the hydrogen isotopes of H2-HD-HT-D2-DT-T2 within 1 min with the analysis limit of 1000 ppm, which indicates that this system enables fine control of the ISS. The reliability of this system was also proved through its use without any mechanical trouble for more than 2 years. The analysis codes were developed and improved through the experiments, and it was succeeded to simulate the dynamic behavior of the distillation column. Through a series of the ISS experiments at the TPL/JAERI, effective design and effective operation of the ISS became to be possible.

An overview of the present status of development of fusion nuclear technologies at Japan Atomic Energy Research Institute is presented. A tritium handling system for the ITER was designed, and the ...

In the fuel cycle system of the ITER, a large fraction of tritium inventory is expected to be in the cryogenic distillation columns of the hydrogen isotope separation system (ISS). Therefore, the numerical estimation method of hydrogen isotopes inventory in the ISS with high precision is strongly required from safety point of view. Two series of experiments were performed to establish the numerical estimation method of the overall hydrogen isotope inventory in the ISS at steady state using ITER-scale large cryogenic distillation columns at the Tritium Systems Test Assembly in the Los Alamos National Laboratory under the US-Japan collaboration on tritium safety engineering. As a result of experiments, it was confirmed that the hydrogen isotope inventory in a cryogenic distillation column was estimated by the numerical estimation method proposed in this work with enough high precision from the engineering point of view, and it was proved that this method was applied for the ITER-scale cryogenic distillation...

To construct the ITER with high safety and acceptability, it is very important to grasp the removal behavior of tritium happened to leak in the room, the final confinement barrier. In order to obtain data on tritium removal behavior from atmosphere in a room under the various conditions (humidity, ventilation flow rate), intentional tritium release experiments have been carried out with the Caisson Assembly for Tritium Safety Study (CATS) which consists of 12 m 3 gas-tight box (Caisson) for the study of tritium behavior in large space. Effect of adding water vapor has also investigated for effective removal. When the tritiated water existed in the released tritium, residual contamination on the wall of the Caisson was detected under the various ventilation flow rate and it was found that it depended on the initial humidity in the Caisson. On the other hand, when the water vapor was added into the Caisson after found the residual contamination, the residual contamination was removed quickly on the wall of the Caisson. The adding water vapor into the Caisson, it was effective for the tritium removal. Analytical work have also progressed and analyzed tritium removal behavior became to be in good agreement with the experimenta' results by considering the adsorption and desorption reaction rate of tritiated water on the wall.

Typical materials and elements such as carbon steel, grease lubricant, electric cable and AC servomotor for the ITER remote maintenance equipment were exposed to a tritiated moisture environment to choose the appropriate materials from the viewpoint of tritium contamination / decontamination and to contribute to the structural and maintenance design of the remote maintenance equipment. After the test samples were exposed, the concentrations of tritium adsorbed on the samples were measured and decontamination experiments using gas purges with three different moisture concentrations were performed. It was found that metallic oxidized layer (Fe 3 O 4 coated on S45C, slightly rusted SS400, rusted Cu of electric connector of the AC servomotor) adsorbs larger amount of tritiated moisture. Grease lubricant was highly contaminated in proportion to the exposed surface area of the pasted layer. Cable jacket (cross-linked polyethylene) was also highly contaminated in spite of hydrophobicity. This is probably because the jacket contains the filler white carbon (SiO 2 . nH 2 O) which adsorbs large amount of moisture. Internal parts of the AC servomotor were contaminated in the same level as the outer surface, because tritiated moisture goes into the inside through the sealing gap between casing and brackets.

A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium plasmas in the Tokamak Fusion Test Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focussed to an intensity, typically 8 kW/cm 2 , and rapidly scanned over the tile surface by galvanometer driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 °C were recorded by an optical pyrometer. Tritium was released and circulated in a closed loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next step fusion device will be discussed.

Assuming a practical use in ITER facility, rapid recovery and supply of tritium to accommodate pulsed plasma operation cycle with minimal inventory was tested. For this purpose, tritium will be supplied from heated bed with vacuum pump while heat of reaction is supplied externally. For recovery, hydriding reaction occurs at elevated temperature spontaneously. Kinetic behavior of the bed at the temperature around 300 degree-C was studied, and practical operation was successfully demonstrated. Isotopic composition change due to the difference of equilibrium temperature was concerned in supplying mixture, but the effect was found to be negligible. For rapid accountancy, ITER requirement of accuracy (± 1 %) was demonstrated by 25 g tritium storage ZrCo bed with In-bed gas flowing calorimetry. It was revealed that the accuracy is affected by the surrounding temperature, that could readily be controlled for better measurement. Thus technology and experience on storage and transport of large amount of tritium, that are inevitable in fusion tritium facility such as ITER Tritium Plant have been established by Japanese research facility and industry.