Hydrogen Isotope Research Center - Toyama Univ.
Editors
K. Watanabe, M. Matsuyama and Y. Hatano
November 18-19, 2001
Toyama Koseinenkin Resort Center, Awasuno, Toyama, Japan
Ten papers were presented by invited speakers from overseas and Japan. The foreign invited speakers were from Germany, Italy, Russia, the United Kingdom and the United States. The Japanese speakers were from the
Japan Atomic Energy Research Institute, Kyoto University, Nagoya University, Kaken Co., and Toyama University. The meeting was in an appropriate size of about 50 attendants (including ten attendees from overseas), giving a good opportunity to keep close contact for exchanging views and to deepen acquaintance each other.
By focusing attention to the material contamination and the decontamination, up-to-date information were given from both basic and technological viewpoints as kinetics of ad/desorption, trapping/detrapping, diffusion and permeation of hydrogen isotopes on/in metals and non-metals, tritium distributions on/in first walls used in large machines as JET and TFTR, and effectiveness of conventional detritiation techniques and development of new methods.
Each paper commanded strong interest of the attendants and aroused intense discussions. The discussions were liveliest in a very frank atmosphere. The workshop confirmed the importance of further studies on fundamental processes and the development of detritiation techniques. There is a good reason to believe that the tritium management and safety control techniques can be developed extensively through worldwide systematic studies and cooperation.
In closing, thanks should be extended to all of the participants who made the workshop enjoyable and fruitful. Special thanks are due to Drs. Aiello, Bell, Glugla, Livshits and Willms visiting Toyama from overseas for presenting invaluable papers.
Toyama University, Hydrogen Isotope Research Center
Kuniaki Watanabe
[workshop 2001_01]
TRITIUM ISSUES AT JET
A. C. Bell
EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, UK
+44 (0)1235 464746
INTRODUCTION
The JET machine was originally planned to be capable of operation with DT plasmas with neutron production up to 10×1023 per year and the torus was therfore constructed within a massive concrete biological shield wall. As the experimental programme developed, numerous enhancements were made to the machine and its ancillary systems. Many of these were associated with tritium operation such as the construction of a plant to supply tritium and recycle it for re-use (The Active Gas Handling System, AGHS) and facilities to enable maintenance and modification to be carried out remotely.
Tritium experiments took place in 1991 and 1997 and required additional measures to be put in place to control radiological hazards and to handle tritiated waste.
Following these experiments, tritium retention in JET machine and other components and materials has been an important consideration and detritiation has been a key issue for study.
The following sections examine these issues.
[workshop 2001_02]
Tritium depth profiles and accelerator mass spectroscopy for JET tiles
(graphite/CFC tiles)
N. Bekris1, M. Glugla1, U. Berndt1, R. -D. Penzhorn1, M. Friedrich2 W. Pilz2
1Forschungszentrum Karlsruhe, Tritium Labor (TLK) P.O.Box 3640, 76021 Karlsruhe, Germany
2Forschungszentrum Rossendorf, P.O.Box 510119, 01314 Dresden, Germany
Introduction
Tritium retention in carbon first wall tiles is an important issue not only from a
radiological point of view but also for inventory purposes. As the divertor vertical
targets of the ITER machine are constituted from carbon tiles, it is of
paramount importance to evaluate the amount of tritium that may be retained by
these tiles. For that purpose a total of 24 tiles were removed from the divertor (10),
the inner guard limiter (6) and the poloidal limiter (8) from the JET fusion machine
after the first Deuterium-Tritium Experiment (DTE1) and sent to the Tritium
Laboratory in Karlsruhe (TLK) for a surface and bulk tritium analysis.
Due to plasma wall interactions a fraction of the fuelled tritium becomes
incorporated into the first wall materials of fusion machines, so it is expected that they
will be tritium-contaminated.
In addition, during plasma discharges hydrogen isotopes are implanted into the
graphite tiles as ions or as energetic neutral atoms having lost their charges after one or
several electron captures (charge-exchange mechanism). Complex processes, different
to some extent from those known in metals, govern the transport of hydrogen in
graphite [1]. Plasma particles interact not only with the outer surfaces of the first
wall but also after diffusion with the inner surfaces of the interconnected pores.
Hydrogen atoms may also combine with the eroded carbon to form a hydrocarbon
film which is co-deposited on the surface of the tiles. In other areas of the tiles the
implanted hydrogen combines with the carbon to form a hydrogenated carbon
layer whose thickness depends on the energy of the more energetic particles.
Retention via co-deposition and implantation of energetic ions into the near
surface layers as well as bulk effects including migration through the network of
interconnected pores and diffusion across the grains are dependent on the plasma
operation conditions (electron temperature, electron density) and on the degree of
neutron and alpha irradiation damage [2]. It is known that hydrogen implanted at
room temperature is retained at the near surface until a saturation is reached [3, 4].
The saturation concentration has been reported to be in the H/C ratio range of
0.3~0.6 [5] and depends on the prevailing temperature. The films found in the
relatively cool (< 200°C), shadowed area (not subjected to plasma bombardment) are
very rich in hydrogen content. The H/C ratio measured by Ion Beam Analysis
(IBA), showed a value between 0.7~0.8 [6] It depends also on the implantation
energy (incident ion energy at JET < 50 eV) but seems to be relatively
insensitive to the microstructure of the carbon [7].
[workshop 2001_03]
Tritium Retention in Plasma Facing Graphite Tiles
T. Tanabe1, K. Miyasaka2, T. Shibahara2, R. Ishida2, N. Miya3, K. Masaki3, V. Philipps4 and M. Rubel5
1CIRSE, Nagoya University
2Graduate school of Eng. Nagoya Univ.
3JAERI, Naka
4Juelich Research Center, Germany
5Van Alfven Laboratory, Sweden
Abstract
The present work, at first, summarizes the basic knowledge of hydrogen behavior in graphite
and then introduces recent investigations how tritium is distributed in plasma facing walls of TEXTOR and
JT-60U in terms of material temperature (or incident flux), local plasma behavior, global plasma behavior
and so on.
In TEXTOR, it was found that tritium distributions was quite different from deuterium
distribution, i.e. deuterium retention was higher at the deposited area, whereas tritium retention was
higher at the erosion dominated area. This is because tritium produced by the D-D reaction, initially
having 1 MeV, did not fully lose its energy in the TEXTOR plasma and implanted into the plasma facing
materials nearly homogeneously, whereas deuterium was codeposited with carbon and boron, the main
impurities in the TEXTOR plasma. This is also confirmed by the finding that high level of tritium was
detected beneath the deposited layer.
In JT-60U, tritium distribution, however, was modified by the temperature increase due to
plasma heat load. The highest tritium level was found at the top of the dome or the private region and the
outer baffle plates, where the plasma did not hit but the distance from the plasma was the shortest. For the
divertor tiles, the tritium retention was very small. Such tritium distribution observed in JT-60U tiles
can be well explained by the homogeneous implantation of rather high energy tritium and thermal release due
to the heat load.
Thus the comparison of tritium profiles with the deuterium profile gives a large amount
of important and new information on PMI, and may be used as a new diagnostic technique for PMI.
[workshop 2001_04]
Tritium Tracking by BIXS in Contamination and Decontamination Processes
M. Matsuyama, Y. Torikai, K. Yoshida, S. Nakagawa and K. Watanabe
Hydrogen Isotope Research Center, Toyama University, Gofuku 3190, Toyama 930-8555, Japan
ABSTRACT
The β-ray-induced X-ray spectrometry(BIXS) was applied to evaluate contamination and decontamination
behavior of tritium for four kinds of metallic materials. Two kinds of charging methods were employed
for tritium contamination: irradiation with tritium ions at room temperature and exposure to molecular
tritium at elevated temperatures.
The order of contamination level due to the irradiation of tritium ions was as follows:
Hastelloy < SS-316 ≒ Tungsten < B/SS-316,
where most of the irradiated tritium was retained on the surface and/or in subsurface layers. For the decontamination
tests of the B/SS-316 sample, it was recognized that moisture in the air plays an important role
for decontamination process at room temperature.
On the other hand, it was estimated from computer
simulation of the X-ray spectra that the SS-316 sample exposed to molecular tritium at 473K was uniformly
contaminated up to the bulk. The dissolved tritium into the bulk could be removed about 40% by heating for 2
hours at 473K, although the surface kept high tritium activity.
[workshop 2001_05]
Codeposit removal by UV laser irradiation
W.M. Shu, T. Tadokoro, Y. Oya, Y. Kawakubo and M.F. Nishi
Tritium Engineering Laboratory, Department of Fusion Engineering Research
Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken 319-1195, Japan
ABSTRACT
In the development work of an effective method for decontamination of tritium retained in codeposits on plasma facing components of D-T fusion reactors such as ITER, ultra violet (UV) laser irradiation test was carried out using simulated codeposits formed by C2H2 or C2D2 glow discharge. The UV laser applied was an ArF excimer laser with a wavelength of 193 nm, a pulse energy of 200 mJ and a pulse duration of 25 ns. A removal rate of 13.5 nm/pulse was observed at the laser irradiation with the energy density of 0.05 J/cm2 for the simulated codeposits whose atomic ratio of hydrogen or deuterium against carbon (H/C or D/C) at the surface was 0.53. Hydrogen isotopes were removed from the codeposits predominantly in the form of gases of hydrogen isotopes by the UV laser irradiation.
[workshop 2001_06]
Newly Developed Gaseous Decontamination Technology
Katsuyoshi Tatenuma
President of KAKEN Inc. and Invited Prof. of Toyama University
Kaken, Inc. Mito-Institute; 1044 Hori-machi, Mito, Ibaraki 310-0903 Japan
Phone +81-29-227-4485, Fax +81-29-227-4082, e-mail: tatenuma@kakenlabo.co.jp
ABSTRACT
Decontamination and volume reduction of the radioactive wastes generated by atomic power
plants, nuclear fuel reprocessing plants and nuclear research institutes has proven to be extremely difficult.
With regard to two cases of heavy metals and hydrogen isotope (tritium), two kinds of
newly developed gas-phase decontamination technology based on gaseous reactions are introduced.
One is based on utilizing volatile properties of carbonyl compounds and fluoric compounds of radioactive
transition elements and actinides (corrosion products: CP, fission products: FP, trans-uranium: TRUs),
and another is based on the treatment by ozone gas to decontaminate the tritiated wastes.
If gas-phase decontamination technology will be practicable, it will not only be
convenient, but economically advantageous as well, since it is currently very difficult to decontaminate
and treat the large volume of nuclear wastes; especially non-incinerable radioactive wastes.
[workshop 2001_07]
Deploying New Tritium D&D Technologies in the US--Recent Experience and Plans
S. Willms1, E. Stallings1, J. McFee2, R. Blauvelt3, D. Krause4
1Los Alamos National Laboratory, Los Alamos NM 87545, USA
2Shaw Environmental and Infrastructure, Inc., 7600 East Orchard Rd, Suite 320 S, Greenwood
Village, CO 80111, USA
3WPI, Fairborn, OH 45324, USA
4BWXT of Ohio, Inc., Miamisburg, OH 45343, USA
Introduction
The US Department of Energy (DOE) Office of Environmental Management (EM) sponsors projects called Large Scale Demonstration and Deployment Projects (LSDDP). These projects seek to identify and deploy technologies that will allow decontamination and decommissioning (D&D) of surplus facilities more quickly, at less cost and with reduced risk to personnel and the environment. Such a project was started centered on the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL) in 2001. TSTA is a tritium contaminated facility that is being shutdown. This LANL LSDDP is also performing deployments at other LANL facilities and at Mound, Princeton Plasma Physics Laboratory, Lawrence Livermore National Laboratory, the Savannah River Site and Lawrence Berkeley National Laboratory. The project encourages international participation and representative from the Joint European Torus are working with the project.
[workshop 2001_08]
HYDROGEN ISOTOPES PERMEATION THROUGH FUSION ORIENTED MATERIALS
A. Aiello, C. Fazio, I. Ricapito, G. Benamati
ENEA Fusion Division, CR Brasimone, 40032 Camugnano Italy
ABSTRACT
The evaluation of hydrogen and its isotopes permeability and solubility in fusion oriented materials is a key issue in fusion technology concerning safety, fuelling and reliability. Different devices have been designed and constructed at ENEA Brasimone Centre, with the aim to characterize hydrogen isotopes transport and inventory phenomena in structural and breeding materials, in the temperature range between 473K~1073K and in liquid metal and gas phase. Several materials were examined, martensitic steels, tungsten, Tritium Permeation Barriers (TPB), and the main results, together with a brief description of experimental devices and procedure, are herein presented and discussed.
[workshop 2001_09]
EXPERIMENTS ON HYDROGEN TRAPPING WITH ION BEAM ANALYSIS
Ikuji Takagi
Department of Nuclear Engineering Kyoto University Sakyo-ku, Kyoto 606-8501, Japan
+81 (75) 753-5838
ABSTRACT
Depth profiles of deuterium in nickel, continuously exposed to deuterium plasma, were observed by use of the nuclear reaction analysis. Trapping energy for deuterium in nickel bombarded with energetic ions was 0.24eV, which was determined under equilibrium between trapping and solution sites. From the shape of the depth profile and the trapping energy in cases of hydrogen and helium bombardment, the traps were considered to be associated with radiation damages. Experiments on kinetics of deuterium on metal surface were also conducted.
[workshop 2001_10]
Anomalous isotope effect in the permeation, retention, and reemission at interaction of energetic hydrogen with niobium
A.I. Livshits1, M.E. Notkin1, and M. Bacal2
1 Bonch-Bruyevich University of Telecommunications, 61 Moika, St Petersburg 191186, Russia
2Laboratoire de Physique et de Technologie des Plasmas, U.M.R. 7648 du C.N.R.S., Ecole
Polytechnique, 91128 Palaiseau, France
Abstract
A niobium membrane sample was placed in H or D plasma and electrically biased. Isotope effects for H vs. D in factors of 20, √40 and 40, respectively, were observed in plasma driven permeation, retention, and in the reemission, within a narrow range of bias voltages (4080 V) at the lowest metal temperature investigated (910 K). The phenomenon occurred at the "superpermeation" of suprathermal hydrogen arising from an oxygen monolayer at the metal surface. The phenomenon is supposed to be caused by dynamics of the oxygen monolayer under the action of ion sputtering and surface segregation of dissolved O. Such and even much stronger isotope effects are also expected on other metals with a similar "real" surface. This isotope effect may be important for D/T-mixture recycling, retention and permeation at its interactions with plasma facing components of fusion reactors as well as for the applications of superpermeable membranes for pumping of hydrogen isotopes and their separation from He.