Preface
The International Tritium Workshop on the Present Status of Tritium-Material Interaction Studies was held on July 18-19th, 1996 at Toyama University in the frame work of "Japan-EU Forum '96 in Toyama" under the cosponsorship of Hydrogen Isotope Research Center, Toyama University, Atomic Energy Society of Japan and The Japan Society of Plasma and Nuclear Fusion Research. The workshop focused the attention to tritium behavior and its control in thermonuclear fusion reactors, on account that this subject attracts growing interest from both technological and public viewpoints with the progress of research and development of the experimental burning reactor.
Fifteen papers were presented by research groups of Japanese universities, Japan Atomic Energy Research Insitute and companies, and five papers by foreign invited speakers from EU, US, Canada and China. The meeting was in an appropriate size of about 50 attendants, giving a good opportunity to keep close contact for exchanging views and to deepen an acquaintance each other.
The topics presented covered a wide range from fundamentals to technological aspects. The relevant phenomena were ad/desorption, catalysis/surface reactions, solution/hydride formation, diffusion/permeation, trapping/detrapping, and isotope effect/radiochemical effects. The technological aspects were contamination/decontamination, purification/separation, storage/recovery, detritiation/waste handling, measurement/monitoring and so on. Tritium handling experience in large systems also attracted attention of the participants.
Each paper commanded strong interest of the attendants and aroused intense discussions. The discussions were liveliest in a very frank atmosphere, resulting in fruitful new ideas to investigate further the phenomena and the applocations of new findings to tritium handling systems. The workshop highlighted worldwide strong advances in finding and understanding of tritiun-material interactions and in tritium management techniques. Continuing the progress and worldwide cooperation, there is a good reason to believe that the tritium management and safety will not pose a limiting issue of not only the public acceptation but also the commercialization of nuclear fusion power stations.
All of the papers were carefully peer reviewed to ensure high quality. A number of participants of the workshop served as reviewers. Their diligence was essential for publishing the proceedings and many thanks must be given for carrying out this vital job.
In closing, thanks should be extended to all the participants who made the workshop enjoyable and fruitful. Special thanks are due to Drs. Causey, Penzhorn, Perevezentsev, Shmayda and Prof. Zhao from overseas as invited speakers.
Toyama University, Hydrogen Isotope Research Center
Kuniaki WATANABE
[workshop 1996_01]
Tritium Retention in the Codeposited Beryllium-Tritium Layer
Rion A. Causey*, David Walsh**, and Wally Harbin***
*Sandia National Laboratories, Livermore, California 94550 USA
**Sandia National Laboratories, Albuquerque, New Mexico 87185 USA
***Los Alamos National Laboratory, Los Alamos, New Mexico 87545 USA
Abstract
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.
[workshop 1996_02]
Studies on Tritium Retention in Plasma Facing Materials
Kenji Okuno, Shigeru O'hira, Yumi Yaita* and AAndre' Steiner
Tritium Engineering Laboratory, Japan Atomic Energy Research Institute, Toki, Ibaraki, 319-11, Japan
+81-29-282-6390
Abstract
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.
[workshop 1996_03]
Management of Radioactive Substance at Shika Nuclear Power Plant
Toshinao Furuta
Nuclear Power Department, Hokuriku Electric Power Company, 15-1 Ushijima-cho, Toyama 930, Japan
Abstract
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.
[workshop 1996_04]
Plasma Exhaust Gas Processing for Fusion Reactors: Developments and Problems
R. -D. Penzhorn, M. Glugla
Forschungszentrum Karlsruhe, Hauptabteilung Verfahrenstechnik, Tritium Labor
Karlsruhe, P. O. Box 3640, D-76021 Karlsruhe, Germany
Abstract
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.
[workshop 1996_05]
Tritium Purification, Storage and Supply in the Tritium Plant of a Fusion Reactor
A. N. Perevezentsev and J. L. Hemmerich
JET Joint Undertaking, Abingdon, OXON, OX14 3EA, United Kingdom
Abstract
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.
[workshop 1996_06]
Tritiated Stainless Steel Surfaces
W. T. Shmayda and A. B. Antoniazzi
800 Kipling Ave., Toronto, Ontario, M8Z 5s4, Canada
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.
[workshop 1996_07]
Interactions of Hydrogen Isotopes with Metals and Alloys
P. J. Zhao, T. Hirabayashi*
Southwest Institute of Nuclear Physics and Chemistry, P. O. Box 525, Chengdu, Sichuan, 610003, P. R. China
*Japan Atomic Energy Institute, Tokai, Ibaraki 319-11, Japan
Abstract
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.
[workshop 1996_08]
Status of Fusion Blanket Irradiation Study in JAERI
H. Kawamura, H. Segawa, E. Ishitsuka, K. Tsuchiya and M. Nakamichi
Oarai Research Establishment, Japan Atomic Energy Research Institute
Natita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki-ken, 311-13, Japan
Abstract
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.
[workshop 1996_09]
Tritium Behavior on the Surface of Solid Breeding Materials
Satoru Tanaka and Masaki Taniguchi
Department of Quantum Engineering and Systems Science, The University of Tokyo, 7-3-1 Hongo,
Bunkyo-ku, Tokyo 133, Japan
Abstract
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.
[workshop 1996_10]
Study on Tritium Recovery from Breeder Materials
H. Moriyama1 and K. Moritani2
1Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-04, Japan
2Department of Nuclear Engineering, Kyoto University, Yoshida, Sakyo-ku, Kyoto 606-01, Japan
Abstract
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.
[workshop 1996_11]
Study on the System Effect of Tritium
Kenzo Munakata, Nobuyuki Nakashio, Satoshi Fukada and Masabumi Nishikawa
Department of Nuclear Engineering, Faculty of Engineering, Kyushu University
6-10-1, Hakozaki, Higashi-ku, Fukuoka 812-81, Japan
+81-92-642-3784
Abstract
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.
[workshop 1996_12]
Hydrogen Plasma Driven Permeation through Selected Metals
Y. Fujii, M. Takizawa, Y. Sokawa, T. Endo and M. Okamoto
Research Laboratory for Nuclear Reactors, Tokyo Instirute of Technology, O-okayama, Meguroku,
Tokyo 152, Japan
Abstract
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.
[workshop 1996_13]
Recent Studies of Production, Purification and Utilization of Tritium in JAERI
M. Tanase1, K. Kurosawa1, M. Kato1, M. Hashimoto1,
T. Masuzaki2, K. Ishida2 and K. Nagamine2, 3
1Department of Radioisotopes, Japan Atomic Energy Research Institute (JAERI), Tokai-mura, Ibaraki 319-11, Japan
2Institute of Physical and Chemical Research (RIKEN), Wako-shi, Saitama, 351-01, Japan
3Meson Science Laboratory, University of Tokyo (UT-MSL), Bunkyo-ku, Tokyo, 113, Japan
Abstract
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.
[workshop 1996_14]
Hydrogen Isotope Separation by Advanced Gas Chromatography and Fractionation
Kuniaki Watanabe and Masao Matsuyama
Hydrogen Isotope Research Center, Toyama University, Gofuku 3190, Toyama 930, Japan
Abstract
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.
[workshop 1996_15]
Present Status of Hydrogen Isotope Separation at Nagoya University
Ichiro Yamamoto, Noboru Kobayashi and Takahiko Sugiyama
Department of Nuclear Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-01, Japan
Abstract
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.
[workshop 1996_16]
Preliminary Design of Tritium Cleanup System for the Large Hellical Device
Y. Sakuma, H. Yamanishi, T. Uda and H. Hirabayashi
Natioanl Institute for Fusion Science, Furo-cho, Chikisa-ku, Nagoya 464-01, Japan
Abstract
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.
[workshop 1996_17]
Current Status and Issues in Tritium Technology to Produce Laser Fusion Target
T. Norimatsu, M. Takagi and K. Mima
Institute of Laser Engineering, Osaka University, 2-6 Yamada-oka, Suita, Osaka 565, Japan
Abstract
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.
[workshop 1996_18]
R&D of Tritium Technology at SHI
N. Yokogawa
Sumitomo Heavy Industry, Yato-cho 2-1-1, Tanashi-shi, Tokyo 188, Japan
Abstract
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.
[workshop 1996_19]
MHI's Activities on Tritium Technology
Masaaki Nagakura, Taisei Naitou, Hideki Imaizumi, Koichi Kurita
Niclear Application Technoligy Department, Mitsubishi Heavy Industries, LTD.
1-297, Kitabukuro-cho, Omiya-City, Saitama, Japan
Abstract
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.
[workshop 1996_20]
R&D Activities of Tritium and Breeder Technology at KHI
Tatsushi Suzuki
Nuclear Systems Division, Kawasaki Heavy Industries, Ltd., Minamisuna 2-6-5, Koto-ku, Tokyo 136,
Japan
Abstract
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.