International Tritium Workshop
on
Present Status and Prospect of Tritium-Material Interaction Studies

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Preface

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Contents

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[workshop 1996_01]

　The Tritium Plasma Experiment (TPE) has been used to measure the amount of tritium codeposited with beryllium in a film produced by the energetic sputtering of beryllium by a tritium plasma. In these experiments, a plasma consisting of 90% deuterium and 10% tritium was directed onto a biased beryllium disk. A catcher plate located approximately 7 cm away from the beryllium disk was used to collect the sputtered beryllium along with the deposited hydrogen isotopes. A typical experiment used an impinging flux of 3.3×10¹⁷(D+T)/cm²-s and lasted for 1 hour. After the collection of the layer, the small catcher plate was removed and outgassed in a special system designed for thermal spectroscopy of tritium laden samples. Data was collected for the catcher plate heated to 373, 473, and 573 K. At 373 K, the ratio of hydrogen isotopes to beryllium was approximately 0.35. At 573 K, this ratio had dropped to avout 0.03. Implications for tritium inventory in the ITER reactor are discussed.

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[workshop 1996_02]

　Tritium retention behavior on an isotropic graphite was studied by exposure with high flux atomized D/T particles. From analysis of thermal desorption spectra of retained tritium it was shown that tritium implanted into graphite existed in two different states. One was trapping to defects and the other was C-T bond. The amount of tritium retained in graphite was in proportion to a half power of total incident fluence and no saturation was observed up to 10²⁶ atomes·m^-2. The total tritium retention in the sample exposed to atomized D/T particles with total incident fluences of 10²⁵ –10²⁶ atoms·m^-2 were estimated 1×10²²–4×10 ²² atoms·m^-2.
    Morover, the existing sates of deuterium implanted into Be and W have been investigated using the X-ray photoelectron Spectroscopy (XPS). The XPS experiments showed some plasmon energy shifts for Be and W, which implied that deuterium implanted exist in the lattice of those metals.

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[workshop 1996_03]

　Various radioactive substances are generated in the nuclear reactor, and cause the management of radioactive substances as an important job in the nuclear power plant. The radioactive substance of the fission product is remained inside the fuel rods, because the fuel failure seldom occurs. The other radioactive substance generated by the neutron absorption reactions in the coolant water such as Co-60, which is called the radioactive corrosion products, is the main subject of the chemical management in the nuclear power plant, because it causes high radiation exposure during the periodical inspection. In the Shika Nuclear Power Plant, the several techniques for the radiation exposure reduction are adopted. In addition, the radioactive waste management is performed strictly to decrease the waste generation amounts and to minimize the discharge of radioactive quantities. The effect of the radioactive substances discharged from the Shika Nuclear Power Plant for public has been negligible.

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[workshop 1996_04]

　Recent progress at the Tritium Laboratory Karlsruhe is presented on the development of tritium technology for ITER in areas such as fuel clean-up of plasma exhaust gas, qualification of components by long-term testing under a relevant tritium environment, and tritium accountancy based on calorimetry.

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[workshop 1996_05]

　The techniques used at the JET tritium plant for separation of torus exhaust gas and impurity processing are discussed in view of the requirements for the tritium plant of a next generation fusion reactor. A chemical method to convert tritiated hydrocarbons and water with a liberation of molecular hydrogen is presented. The efficiency and safety of uranium hydride containers used at JET for pumping, storage and supply of tritium and deuterium are analyzed and the possible use of non-radioactive intermetallic alloys for metal hydride storage systems is investigated.

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[workshop 1996_06]

　Stainless steel is widely used in the fabrication of tritium gas handling systems. Its popularity as a construction material resides in part with the good mechanical properties stainless steel possesses in the presence of hydrogen and in part with the availability of an established industrial base to build components for high pressure and/or ultrahigh vacuum applications. In fact, it is quite common in tritium gas handling applications to design components to meet design criteria which satisfy both hard and high pressure requirements simultaneously. While the influence of hydrogen on stainless steel has been studied in detail, the impact of tritium with the metal surface is limited. This review paper summarizes studies which have been underway at Ontario Hydro Technologies for several years.

　Surfaces exposed to tritium gas release tritium labeled species continuously for prolonged periods. While the release of these gases depends on several factors including surface condition and exposure conditions, the make up of the outgassed species changes with time. Shortly after an exposure T₂ and HTO dominate the outgassing spectrum. However as time passes, that is to say, as the tritiated surface ages, tritium labeled organics contribute significantly to the spectrum. Compounds such as formaldehyde have been observed in the outgassing spectrum. These compounds tend to be water soluble, rapidly contaminate downstream surfaces in a flowing system and deliver doses to personnel which appear to differ from HTO vapour exposures.

　Decontamination is used to mitigate the spread of activity within a facility and to permit safe handling of items which have seen tritium service. Several techniques are available to reduce tritium contamination of surfaces. These include washing with a solution comprising water and a surfactant with or without the aid of an ultrasonic bath, thermal desorption, purging with a humid stream and exposure to room temperature plasmas. The effectiveness of each of these techniques has been compared using coupons which have been identically treated during the manufacturing, exposure to tritium and storage phases. Washing is convenient however the approach produces low level mixed waste. In addition, the decontamination is temporary; the surface activity returns within a few days as tritium in the subsurface diffuse to the surface. Thermal desorption permits removal approximately 95% of the tritium inventory on a surface provide the metal can be heated to 350°C. Plasma decontamination, where applicable, is the most effective technique achieving nearly complete removal of tritium from a surface within a few minutes with negligible 're-growth' of surface activity in the following weeks. Items in many cases can be decontaminated to background levels.

　This paper will characterize emission species from stainless steel surfaces outgassing at room temperature and at elevated temperatures, in dry and humid streams, demonstrate that the emission spectrum changes with time, and compare the effectiveness of several decontamination techniques.

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[workshop 1996_07]

　Tritium decontamination is one of the important problems for the safety of D-T fusion reactors. The experimental results showed that UV light combined with IR irradiation was effective in decontaminating tritium from metals and alloys. The desorption rate of 96% was achieved at 150°C. Desorption at the temperatures of 75°C to 150°C was enhanced further by replacing argon with dry air. The behavior of tritium sorption, thermal desorption and photo-desorption was investigated for varying metal surfaces.

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[workshop 1996_08]

　An experimental reactor such as ITER is planning various tests using several kinds of blanket designs in addition to demonstrating the physics of burning D-T plasmas. The data of neutron irradiation performance of a blanket is needed for the fusion blanket design. Studies for an in-pile functional test of a blanket mock-up and the development of blanket materials are being continued in the JMTR of Oarai Establishment in JAERI. The present status of these studies is briefly described in this report.

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[workshop 1996_09]

Hydroxyl groups on the surface of Li₂O were studied by using a diffuse reflectance method with Fourier transform infrared absorption spectroscopy at high temperature up to 973K under controlled D₂O or D₂ partial pressure. It was found that hydroxyl groups could exsit on Li₂O surface up to 973K under Ar atmosphere. Under D₂O containing atmosphere, only the sharp peak at 2520cm^-1 was observed at 973K in the O–D stretching vibration region. Below 973K, multiple peaks of the surface -OD were observed and they showed different behavior with temperature and atmosphere. Multiple peaks mean that surface is not homogeneous for D₂O adsorption. Assignment of the observed peaks to the surface bonding structure was also discussed. Tritium release behavior from each observed surface site was discussed. Peculiar tritium release behavior which was observed during temperature increase was also discussed.

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[workshop 1996_10]

　For the development of fusion reactor blanket systems, some of the key issues on the tritium recovery performance of solid and liquid breeder materials were studied. In the case of solid breeder materials, a special attention was focussed on the effects of irradiation on the tritium recovery performance, and tritium release experiments, luminescence measurements of irradiation defects and modeling studies were systematically performed. For liquid breeder materials, tritium recovery experiments from molten salt and liquid lithium were performed, and the technical feasibility of tritium recovery methods was discussed.

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[workshop 1996_11]

　Tritium is sorbed not only on the surface of the specimen or material used in experiments but also on the surfaces of tubes, joints and so forth, which constitute the experimental apparatus. This could lead to incorrect experimental results and erroneous understanding of observed phenomena. The authors call phenomenon "system effect". Therefore, it is indispensable to investigate this effect before the behavior of tritium on a certain material is studied. In this work, the system effect was studied using an experimental piping system. Moreover, numerical simulations of tritium behavior in an aluminum-coated box were carried out using results obtained in the experiments.

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[workshop 1996_12]

　Hydrogen permeation through membranes of selected metals was studied with low temperature plasma generation apparatus. The permeation was found to increase by applying plus voltages to the membrane. Under the plus bias conditions, permeation of hydrogen isotopes was measured with the membranes of SS-304, Ni, Cu, Ti, Pd and Fe. The permeation enhanced by the plus bias is due to the dissociation of neutral hydrogen molecules into atoms on the membrane by incident electrons.

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[workshop 1996_13]

　Using a little modified facility for producing tritium in 40 TBq per batch, 21 TBq and 30 TBq of gaseous tritium for utilization to μ-catalyzed fusion were extracted from neutron-irradiated ⁶Li-Al alloy targets by heating at 500 °C under vacuum and collected in a uranium getter. The recovery yields of tritium were about 100%. Through the production run, no leakage of tritium from the facility was observed. In order to use the tritium gas in the μ-catalyzed fusion experiments, hydrogen (H) in the tritium gas should be removed. Therefore, the tritium gas obtained was purified with an improved gas chromatograph, which enabled handling of about 5 TBq of tritium gas per batch using a separation column of 10 mmφ × 9m under the He flow of 300 ml/min. The old column had been 6 m in length and the flow rate 100 ml/min. About 60 TBq of tritium gas in a high purity was effectively obtained by the batch processing with the improved gas chromatograph. About 50 TBq of the purified tritium gas was transported to RIKEN-RAL Branch in England to be used for muon catalyzed fusion (μCF) experiments. At the Branch, a tritium handling system (THS) was installed for removing ³He of a decay product of tritium. Before the fusion experiments, performances of U-getters, Ti-getters, Pd-filter, and all other parts were tested and the fusion experiments have been carried out.

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[workshop 1996_14]

　A kind of gas chromatography and a tritium counting device were developed for the separation of hydrogen isotopes. The active materials examined for chromatography were Pd/Al₂O₃, and Pd and Pd-Pt(8 at%) alloy powders. The Pd-Pt alloy showed the best separation efficiency among the three materials. It could separate a 50%H₂-50%₂ mixture to H₂ and D₂ of 97.5% purity with 80% recovery at 274K without using any replacement gas. The bremsstrahlung X-ray counting device developed for measuring high concentration tritium showed a good linearity between the counting rate and the tritium pressure, the specific counting rate being evaluated as 70.8 cps/Pa. The combined use of these two devices is expected to be applicable to the recovery of tritium from the flow of fuel gas in thermonuclear fusion reactor.

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[workshop 1996_15]

　In Isotope Separation Laboratory at Nagoya University, we have been studying water distillation and thermal diffusion for hydrogen isotope separation. The present paper describes some recent developments of the separative analyses on these thechnologies.

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[workshop 1996_16]

　At the National Institute for Fusion Science, we are planning to carry out DD experiments using the Large Helical Device, in the near future. The device will generate a small amount of tritium, as a fusion product. In order to remove it from the exhaust, we have designed a tritium cleanup system based on a new concept. This system is mainly composed of a palladium permeater, a decomposer and hydrogen absorbing alloys. It could perfectly recover the tritium without oxidizing it.

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[workshop 1996_17]

　Current status of tritium technology to fabricate laser fusion targets is described. Tritium facilities for laser irradiation, deuterium-tritium (DT) fuel loading system, fabrication of deuterated-tritiated polystyene shells, measurement of partial pressure of tritium and cryogenic technologies to fabricate a thick solid DT layer are briefly mentioned. Issues to fabricate cryogenic DT target for the coming upgraded laser system is also discussed.

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[workshop 1996_18]

　Sumitomo Heavy Industries (SHI) participated in an R&D programme on tritum processing for the first time in 1967 by joining the advanced thermal reactor project. (The thermal reactor is cooled by light water and moderated by heavy water.) From that time SHI has developed various kind of tritium handling technologies. On the basis of cooperation with Sulzer (Sulzer Chemtech Ltd. Switzerland), SHI developed a system for removing waste water for fuel reprocessing plants by water distillation technology. In the field of fusion technology, SHI has developed a hydrogen isotope separation system by cryogenic distillation and thermal diffusion methods, and a tritium storage bed. Fundamental data required for the system design were obtained through the production and operation of the above prototype systems. Recently, SHI has also been taking part in the design and planning of ITER. In the future, along with ITER design, SHI will aim at developing tritium measuring technology.

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[workshop 1996_19]

　MHI has been developing tritium technology for more than 20 years, mainly in the following fields concerning thermonuclear fusion reactors.

1.Isotope separation system by cryogenic distillation.
2.Fuel clean up system by palladium permeation and electrolysis cell method.
3.Tritium recovery system from the blanket by palladium permeation method.
4.Blanket materials, mainly the development and characterization of Li ceramic.
5.Tritium removal system by tritium oxidation catalysis.

Based on the tritium technology recently attained through the above developments, MHI has built a tritium treatment installation where MHI can treat 2.2×10¹²Bq (6 Ci) of tritium per year.

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[workshop 1996_20]

　Kawasaki Heavy Industries, Ltd. (KHI) had developed tritium breeding blanket and tritium handling systems. In current design studies of tritium breeding blanket, small spherical solid breeding material has been desired. A mass-production process of small spherical breeder was developed by using a rotating-granulation/sintering method. The fracture toughness of fabricated solid breeder of 1 mm φ was measured.
Several tritium implantation apparatus has been developed and constructed to investigate the implanted tritium behavior on first wall and divertor plate. The characteristics of these apparatus are also examined.