Study on Maintenance, Recycle, and Radioactive Waste Management ofFusion Reactor
Part III. Accumulation and removal of tritium
核融合炉の保守・リサイクル・バックエンド対策に関する検討
3.トリチウムの蓄積と除去
Yuji Hatano
波多野雄治
Hydrogen Isotope Research Center, University of Toyama
Gofuku 3190, Toyama 930-8555, Japan
富山大学水素同位体科学研究センター 〒930-8555 富山市五福3190
Retention and release of hydrogen isotopes by tungsten depsotion layer and neutron-irradiated tungsten are
dicussed in this presentation to obtain better understanding of tritium accumulation and removal in/fromfirst walls and divertor components of DEMO reactors. The results of laboratory-scale experimens showedthat tritium retention up to relatively high cocnetration is possible under expected reactor conditions.
Evaluation of decay heat induced by neutron irradiation and temerature change due to it are important topredict tritium release behavior from first wall and divertor components during maintenance and disposal
processes.

1. BackgroundTungsten is currently recognized as a primary
candidate of armor materials for first-walls anddivertor components of DEMO reactors due tofavorable physical properties such as highmelting point and high thermal conductivity. It is,however, known that tungsten deposition layer
can accommodate hydrogen isotopes up to
relatively high concentration (e.g. H/W = 0.7 [1])depending on substrate temperature and contentsof impurities such as oxygen and He [1-3]. In
addition, retention of hydrogen isotopes in thebulk of tungsten increases with neutron
irradiation due to trapping effects by defects. The
results obtained in Japan-US Joint Project TITAN
showed that deuterium retention in tungsten at773 K increased from 50 appm to 2000 appm
after neutron irradiation to 0.025 dpa in a fission
reactor [4-6]. Simulation studies by ion
irradiation indicated that concentration ofhydrogen isotopes in damaged tungsten can reach
1-2 at.% [7,8]. Therefore, in spite of low tritium
solubility in well-annealed tungsten, tritium
retention in armor materials could be highest
among fusion reactor materials except getters fortritium storage and supply (U or ZrCo).
Temperature of armor materials is expected tobe relatively high even after termination ofplasma due to decay heat induced by extensive
neutron irradiation [9]. Hence, tritium releasefrom hot armor materials could be a critical safetyissue during the maintenance and disposal

processes. From these viewpoints, retention andrelease of tritium by tungsten deposition layer
and neutron-irradiated tungsten are discussed inthis presentation.
2. Tritium Retention
Fujiki et al. [1] reported that hydrogen retention
in tungsten deposition layer formed by hydrogen
RF plasma reached H/W = 0.7 at ambient
temperature and oxygen content of O/W = 1.69.
Reduction in deuterium retention with increasing
substrate temperature was observed by Temmerman
and Doerner [2] and Alimov et al. [3]. The extent ofreduction, however, was strongly dependent onvacuum conditions and presence or absence of He.
The deuterium retention at 623 K was D/W =2×10-4 when the deposition layer was prepared inpoor vacuum conditions with pure D plasma, but it
was 4×10-2 under high vacuum conditions andaddition of He. Detailed understanding of retention
mechanisms is necessary for the evaluation oftritium retention under reactor conditions. It wasfound in thermal desorption measurements thatdeuterium desorption continued up to 900 K.
In TITAN Project Task 2-1, disk-type specimensof pure tungsten were prepared in Japan andirradiated in High Flux Isotope Reactor (HFIR),
Oak Ridge National Laboratory (ORNL) at 50 oC to
0.025 dpa. Then the specimens were shipped toIdaho National Laboratory (INL) to expose highflux deuterium/tritium plasma with the linear
plasma machine called Tritium Plasma Experiment

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D2 desorption rate (10 D/m s)(TPE). Figure 1 shows thermal desorption spectrumof deuterium thus obtained for neutron-irradiatedtungsten after exposure to TPE plasma at 473 K tofluence of 2×1025 D m-2 together with spectrum
from unirradiated specimen [4-6]. The desorption ofdeuterium from neutron-irradiated tungsten
continued up to 1000 K, while that fromunirradiated specimen was completed at 700 K. In
addition, the total deuterium retention in the former
was significantly larger than the latter.
Measurements of depth profiles of D after exposure
to plasma at 773 K showed that deuterium retention
was 2000 appm in neutron-irradiated tungsten and50 ppm in unirradiated specimen as describedabove [6]. These observations indicated theformation of strong trapping sites by irradiation.
The trapping energy was evaluated to be 2 eVfrom the desorption spectrum, and this value
corresponds to the trapping by voids [7,8].
1098
7
6
5
4
3
2
1
0300

Neutron-irradiatedUnirradiated
500

700

900

1100

1300

Temperature (K)Fig.1. Thermal desorption spectra of deuterium fromunirradiated and neutron-irradiated (0.025 dpa) tungsten
after exposure to high flux plasma at 473 K.

3. Issues Needs To Be Clarified
The above-mentioned observations indicatedthat tritium retention in armor materials could berelatively high under reactor conditions. During
replacement and decomposition of blanket anddivertor components, circulation of coolant should
be stopped. Then temperature of armor materials
should increase by decay heat [9], andthermally-activated tritium release would start. For
the prediction of release behavior of tritium, precise
evaluation of temperature change during
maintenance and disposal processes is necessary inaddition to the detailed understanding of retentionmechanisms in deposition layer and radiation
defects.
On the other hand, controlled temperature riseby using decay heat may allow passive and/oractive tritium removal treatments such as pumping

and glow-discharge cleaning at high wall
temperatures before opening of vacuum chamber.
Such treatments at high temperatures would help
greatly to remove tritium not only from surfaces but
also bulk and significantly reduce uncontrolled
tritium release during maintenance and disposal
processes. Precise evaluation of decay heat andtemperature change is again required to design suchdecontamination scenarios.
Acknowledgments
This work has been partly supported byJapan-US Joint Project TITAN and Grant-in-Aidfor Scientific Research on Priority Areas, 476,Tritium for Fusion.
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