Recent progress of tritium related activities in · PDF fileRecent progress of tritium related...
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Recent progress of tritium related
activities in Japan
Kenji Okuno
Shizuoka University
Radio-science Research Laboratory
3rd-China-Japan WS
Kunming, China, June 20-23,2010
Research Subjects and Institutes
for Tritium Issues
Research Subjects
- Fusion
(Processing, Blanket, First Wall, Safety, Licensing)
- Fission Reactor (Heavy Water Reactor)
- Waste Management
- Environmental Behavior
- Biological Effects
- Fundamental Science
Institutes
- Universities
- Japan Atomic Energy Agency
- National Institute for Fusion Science (NIFS)
- National Institute of Radiological Sciences (NIRS)
Tritium/Material Interaction
- Plasma Facing Materials
- Structural Materials
Blanket Engineering
- Breeding Materials
- Tritium Breeding Ratio
- Tritium Recovery
- Tritium Permeation
Tritium Processing and Safety
Tritium Behavior in Environment
Key tritium issue
for Fusion Engineering Researches
Institutes and Universities where Tritium
Related Studies are conducted in Japan
JAEAJAEA
Akita Univ.
Research Resources
JAEA 50
NIRS 10
Industries many
University Researchers Students
Hokkaido U. 3 4
Akita U. 2 4
Tohoku U. 1 2
Ibaraki U. 6 6
U. of Tokyo 7 15
Shizuoka U. 4 12
Toyama U. 9 7
NIFS 7 0
Nagoya U. 7 10
Kyoto U. 7 7
Osaka U. 2 1
Hiroshima U. 2 2
Kyushu U. 9 8
Total 66 78
Tritium Network in Japan
JAEAUniversities
NIFS
National
InstitutesIndustries
Academic Societies・Atomic Energy Society of Japan
・The Japan Society of Plasma
Science and Nuclear Fusion
Research
・The Japan Radiation Research
Society
Conferences (International
and Domestic)・Tritium Science and Technology
・ISFNT
・Joint Symposium, ・・・・・
Many International Collaborations
-CUP (Core Universities Program)
-MEXT/Korea-STA
-MEXT/US-DOE: TITAN Project
-JAEA and foreign countries
-ITER
BA program
-Between Universities
-International Conferences and Symposiums
Tritium conference at Nara in Oct. 2010
Tritium mini-WS, TOFE, ICFRM, ISFNT, et. al.
International collaboration researches
To obtain fundamental understanding for establishing tritium and
thermofluid control throughout the first wall, blanket, and heat
exchange/T-recovery system of MFE and IFE systems by experiments
under specific conditions to fusion, such as irradiation, pulse high heat
flux, circulation and high magnetic field.
The results will be applied through the integrated modeling to advance-
ment of design for tritium and heat control of MFE and IFE systems.
Objectives of TITAN project
Shield
Magnet
storage
pump
Heat exchanger
14MeV neutrons
turbine
Blanket
HeatTritium
Heat
TritiumBlanket
First Wall
Recovery systemPlasma
Recovery system
JUPITER-II
Next Project
pump
purification
First Wall
HeatTritium
TITAN Project will
focus on consistency
of the blankets with
first wall and recovery
systems with respect
to tritium and heat
control
TITAN Structure (revised on Feb.12, 2009)RepresentativesCoordinators
JP : K. Okuno (Sizuoka U.) US : G. Nardella/B. Sullivan (USDOE)JP : T. Muroga (NIFS) US : D. Sze (UCSD)
Task Subtask Facility TC (JP) STC/Deputy (JP) TC (US) STC/Deputy (US)
Task 1Transport phenomena
1-1 Tritium and mass transfer in first wall
TPEPISCES
T. Terai(U.Tokyo)
Y. Ueda (Osaka U.)/N. Ohno (Nagoya U.)K. Tokunaga (Kyushu U.)
D. Sze(UCSD)
R. Doerner(UCSD)
1-2 Tritium behaviorin blanket systems
STAR T. Terai (U. Tokyo)/S. Fukada (Kyushu U.) S. Konishi (Kyoto U.)
P. Sharpe(INL)/P. Calderoni(INL)
1-3 Flow control and thermofluid modeling
MTOR T. Kunugi (Kyoto U.)/T. Yokomine(kyushu U.)
N. Morley (UCLA)/K. Messadek(UCLA)
Task 2Irradiation synergism
2-1 Irradiation-tritium synergism
HFIRSTAR
A.Kimura(Kyoto U.)
Y. Hatano (Toyama U.)/Y. Oya (Shizuoka U.)
R. Kurtz(PNNL)
M. Sokolov (ORNL)/ Y. Katoh (ORNL)P. Calderoni (INL)
2-2 Joining and coating integrity
HFIRORNL-HL(incl. T-test)
A. Kimura (Kyoto U.)/N. Hashimoto(Hokkaido U.)
T. Yamamoto (UCSB)/M. Sokolov (ORNL)
2-3 Dynamic deformation
A.Hasegawa (Tohoku U.)/T. Hinoki (kyoto U.)
Y. Katoh (ORNL)
Common TaskSystem integration modeling
MFE/IFE system integration modeling
A.Sagara(NIFS)
A. Sagara (NIFS)/H. Hashizume(Tohoku U)T. Norimatsu (Osaka U.)
R. Nygren(SNL)
R. Nygren (SNL)
Laboratory Liaisons
ORNL : R. Stoller (ORNL)INL : P. Sharpe (INL)IMR-Oarai (Tohoku) : T. Shikama (Tohoku U.)
IFE Liaisons
K. Tanaka
(Osaka U.)
Kodama(Osaka U.)
Yoneda(UTC)
M. Tillack
(UCSD)
Research activities at
Laboratory of Plasma Physics
and Engineering, Hokkaido
University
T.Hino
Hokkaido University
Plasma – material interactions and heat &
particle controls in fusion devices have been
investigated in this laboratory since 1980.
The research program includes the
collaboration studies with National Institute
for Fusion Science (Toki) and Japan Atomic
Energy Agency (Naka).
The present major research staffs are
T. Hino (Professor), Y. Yamauchi (Associate
Professor) and Y. Nobuta (Assistant Professor).
Overview
Present Researches
(1)Fuel hydrogen retention and desorption
behavior in LHD
Hydrogen retention and desorption of the first wall
are investigated using numerous material probes
and a technique of thermal desorption.
(2)Reduction of tritium inventory by glow
discharge conditionings
The fuel hydrogen retention in plasma facing
materials (SS, W, C) is reduced by the glow
discharge using inert gas such as He, Ne and Ar.
Reduction of fuel hydrogen retention in W will be
presented in this workshop.
(3)Tritium inventory of carbon dust
Numerous carbon dusts are prepared by a
deuterium arc device with carbon electrodes.
Discussed is the relation of the hydrogen
concentration with preparation conditions such as
discharge gas pressure and substrate temperature.
(4)Plasma – wall interactions of low
activation materials
Fuel hydrogen retention and desorption behavior is
investigated for ferritic steel, vanadium alloy and
SiC/SiC composite.
(5)Diagnostics of radiation power from
boundary plasma
Infrared bolometer has been developed as the
boundary plasma diagnostics.
Activity in Akita University (1)
• There is a facility for handling of radioactive material
in the university. The facility was constructed in 1989.
• A new activity for permission and authorization for
handling of tritium had started in October 2009,
which was finally authorized in March 2010.
• At present, they can handle tritium of 3.7 GBq daily,
27 GBq in three months and 55.5 GBq annually. Our
storage capacity of tritium is 185 GBq.
• They are cooperating with National Institute for
Fusion Science (NIFS), Kyoto University Research
Reactor (KUR), Japan Atomic Energy Agency (JAEA),
Forschungszentrum Karlsruhe (FzK) in Germany and
Shizuoka University.
Activity in Akita University (2)
• (1) R&D related to air cleanup system and
process gas cleanup system (NIFS)
Experimental work with tritium, Honeycomb
catalyst and adsorbent, Process simulation
• (2) R&D related to new neutron multipliers
for fusion blankets (JAEA)
Experimental work, Process simulation,
Quantum chemistry calculation
• (3) Experimental investigation on tritium
behavior in chambers using Caisson
Assembly for Tritium Safety study (CATS)
and its numerical simulation (TPL, JAEA)
Experimental work with tritium, Process
simulation
Activity in Akita University (3)
• (4) Modeling of membrane reactor for
treatment of plasma exhaust gases (FzK)
Process simulation
• (5) Study on behavior of tritium release from
ceramic breeders (Shizuoka University and
KUR)
Experimental work with tritium, Irradiation
in research reactor, Process simulation,
Quantum chemistry calculation
• (6) Study on adsorption of hydrogen
isotopes at cryogenic temperature
Experimental work, New adsorbent, New
process, Process simulation
[Tanaka Lab., Univ. Tokyo] “Tritium in liqiud Li-Pb”
<Calculation condition>
CASTEP code
DFT (GGA-PBE)
Plane-wave base
Pseudopotential (ultrasoft)
Periodic boundary condition
(Li6Pb30 >> Li17Pb83)
NVT ensemble (1 fs, 900 K)
: Hydrogen atom
: Lithium atom
: Lead atom
It is difficult to observe hydrogen in liquid materials by experiment.
Hence, existence states and diffusion mechanism are not well understood.
“Direct observation” by ab-initio molecular dynamics [QM-MD]
(*ab-initio = quantum mechanics)
Flibe (234g, 873K)
He + 0-100%H 2 + 0-2%HF
Molecular sieve bed
Aluminium bed (673K)
Ionization chamber A
Tritium trapping system Ionization chamber B
Gas supply system
Sample container
Reactor core
Polyethylene block
NiF2 bed (673K)
n = 108-11 n/cm2 sec, Tirr = room temp. to 800 C
Materials investigated: Li2O, Li2TiO3, Li, Li-Pb, Flibe (LiF-BeF2)
Tritium production rate (in comparison with MCNP calculation),
tritium chemical species released, kinetic parameters (D, Ks, K,
mass-transfer coefficient from liquid to purge gas), tritium
permeation through structural materials were investigated.
In-situ tritium release behavior from tritium breeders
under neutron irradiation at high temperature
core neutron
- PbLi natural convection loop is operated under neutron irradiation
- Tritium permeated through wall of the loop is detected
Tritium separating
/detecting system
50cm
Fast neutron source
“YAYOI”
Tritium production experiments from PbLi
under flowing conditions
Filtered arc deposition device
(IPP-Garching, Germany)
Development of Tritium Permeation Barrier
Er
Fe
1mm
Material: Erbium oxide (Er2O3)· Thermodynamically stable
· Efficient suppression of hydrogen
· Lower deposition temperature
than other oxides (ex. Al2O3)
SEM
1.0 1.2 1.4 1.6 1.8
10-15
10-14
10-13
10-12
10-11
10-10
Perm
eabili
ty (
mo
l/m
/s/P
a0
.5)
1000/T (1/K)
973 873 773 673 573
T [K]
■ F82H
▲ 1.3 mm one side
▼ 2.6 mm one side
● 1.3/1.3 mm both
◆ 2.6/2.6 mm both
D2
Gauge 1 Gauge 3
QMS
Calibration
Volume
TMP 2TMP 1
Upstream Downstream
Furnace
SampleGauge 2
RP 2RP 1
103~105 Pa <10-6 Pa
Jbare / Jcoated ~105
Er2O3 coatings have a potential
as a tritium permeation barrier
Deuterium permeation setup Permeation results
Hydrogen Isotope Research Center (HRC),
University of Toyama
HRC is one of the largest tritium research facilities in Japanese universities and
licensed to handle 8 TBq (217 Ci) tritium per day and 555 TBq (15 kCi) per year.
The center was established at 1980, and this year is in celebration of 30 years
anniversary. Safety equipments including ventilation, waste water processing and
tritium monitoring systems in the radiation facility are currently under
full-reconstruction for stable operation in the future.
Uniqueness of Facility
(1) Handling of tritium in any chemical/physical
form
(2) Various instruments for tritium measurements
(gas, liquid and solid)
(3) Various tools for material characterization in
tritium laboratory.
7 Full, associate and assistant professors,
1 Posdoc, 2 Technicians, Secretaries
(2) Tritium ad/absorption, diffusion, permeation and release in/from fusion
materials
Tungsten, Austenitic and Ferritic Steels, Be alloys, Concretes etc.
Influence of ion/neutron irradiations (Hatano, this workshop)
(3) Development of T permeation barrier coatings and low
adsorption/solubility/diffusivity materials (Zhang, this workshop)
(4) Application of tritium to materials science (e. g. visualization of hydrogen
isotopes in solids by autoradiography)
(5) Development of functional materials for hydrogen energy systems
(hydrogen separation membranes, photocatalysts for hydrogen production etc. ).
Current Research Topics
(1) Tritium measurements
Development of a high sensitivity calorimeter for
absolute measurement (Matsuyama, this workshop)
b-ray-induced X-ray spectrometry (BIXS) and imaging
plate (IP) technique for tritium measurements in liquids
and solids, and near-IR spectroscopy for HTO detection
HRC started new collaboration systems
International Collaboration
(1) Foreign Guest Professor Position
(2) Visitors by Personal Exchange Program including Japan-China
Core University Program and Japan-US Project TITAN.
Domestic Collaboration
(1) Tritium for Fusion
Bidrectional collaborative research program of NIFS (NIFS, the
national user facility for fusion research financially supports
collaboration in HRC)
(2) Tritium for Physics, Chemistry and Material Science, Hydrogen
Energy
Joint Usage/Research Program of HRC
Research activities of Okuno &Oya Lab.
at Shizuoka University
Prof. Kenji Okuno Assoc. Prof. Yasuhisa Oya
Radiochemistry Research Laboratory,
Faculty of Science, Shizuoka University
Typical Research topics at Shizuoka University
Wall conditioning
→Boronization
- Impurity effect on fuel retention
D,T, He
Impurities
PWI issues
- Simultaneous ion implantation
(D, T, He, C) effects on fuel retention
neutron
Blanket issues
- Tritium recovery from lithium oxides
- Behavior of irradiation defects
Tritium permeation behavior
through vacuum vessel
Tritium behavior in fusion related materials
based on radiochemical aspects
Tritium retention behavior in tungsten under simultaneous ion
implantation conditions
300 500 700 900 11000
2
4
6
8Simultaneous
implantation
C, D, He
D, He
C, D
D
Des
orp
tio
n r
ate
/ 1
01
7 D
2 m
-2 s
-1
Tempreture / K
0 5 10 15 20
0.0
0.5
1.0
1.5
2.0
2.5
3.0
3.5
4.0
C in C-D imp. W
C in C-He-D imp. W
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
D in D imp. W
D in He-D imp. W
D in C-D imp. W
D in C-He-D imp. W
Depth / nm
Inte
nsi
ty /
arb
. u
nit
TEM micrographs for the simultaneous C+-D2
+ and C+-D2+-He+ implanted samples.
D2 TDS spectra for the simultaneous C+-D2
+, D2+-He+ and C+-D2
+-He+ implanted
tungsten samples
To simulate actual environment of plasma facing materials under fusion
relevant conditions, simultaneous ion implantation of C+, He+, and D2+
were performed with various ion implantation conditions.
Simultaneous ion implantation system
Depth profiles of C and D in tungsten with various ion
implantation conditions
・The D retention was increased by simultaneous ion implantation.
・Most of D would be trapped by interstitial site and/or irradiation
defects.
・For the D trapping by C, C sputtering by C+ and/or He + and He +
accumulation should be taken into consideration This is a collaboration research with
Kyushu University, University of
Toyama and NIFS.
Plasma CVD system for boron deposition
Tritium retention behavior in boron for wall conditioning
400 600 800 1000 12000.0
1.0
2.0
3.0
4.0
5.0
Temperature / K
Des
orp
tion
rate
/ 1
018 D
2 m
-2 s
-1 TDS spectrum B-D-B bond B-D bond
400 600 800 1000 12000.0
0.5
1.0
1.5
2.0
2.5
3.0
Des
orp
tio
n r
ate
/ 1
01
7 D
2 m
-2 s
-1
Temperature / K
TDS spectrum
B-D-B bond
B-D bond
B-O-D bond
400 600 800 1000 12000.0
0.5
1.0
1.5
2.0 TDS spectrum
B-D-B bond
B-D bond
B-C-D bond
Temperature / K
Deso
rpti
on
rate
/ 1
01
8 D
2 m
-2 s
-1
- Boronization has been applied as one of the most effective techniques for first wall
conditioning for various fusion devices like LHD (NIFS), EAST (ASIPP) and so on.
- Boron is easily bonded to impurities, such as oxygen and carbon, in vacuum vessel and
the elucidation of impurity effect on tritium retention in boron is one of critical issues.
(a) Pure boron film (b) 25% oxygen contained boron (c) 35% carbon contained boron
400 600 800 1000 12000.0
1.0
2.0
3.0
4.0 Pure boron film
LHD boron film
Deso
rpti
on
rate
/ 1
01
8 D
2 m
-2 s-1
Temperature / K
No.11 Campaign
First wall position(2800 mm) Inlet gas: B2H6
Substrate temperature: 373 K
Substrate: Si(99.999%)
LHD sample : 30%, O: 27%, C: 39% N: 3%
Comparison of D2 TDS spectra for various boron films
Comparison of D2 TDS spectra for pure boron film and LHD boron film
The D was bound to impurities as B-O-D and B-C-D and higher temperature is required for desorption.
The D retention for LHD boron film was clearly decreased, which indicated that some of D was quickly re-emitted as H2O and
CDx during the operation.
g-ray irradiation
excitation state
Formation processes of defects
nth
6Li 6Li(n,a)T
He T
6Li
He T
e-
OM
M
hO
MO
- Elucidation of tritium recovery mechanism is one of important
issues for fuel management in fusion.
- Irradiation defects would play an important role in the
trapping and detrapping of tritium
300 400 500 600 7000.0
0.2
0.4
0.6
0.8
1.0
The a
mount of irra
dia
tion d
efe
ct / -
Annealing temperature / K
O-related defects(Thermal neutron irr.)
E'-center(Thermal neutron irr.)
O-related defects(g ray irr.)
E'-center(g ray irr.)
The normalized intensities for E’-center and O--center for thermal neutron and g-ray irradiated samples as a function of annealing
temperature.
O--center
E’-center2.06 2.04 2.02 2.00 1.98 1.96
-0.2
-0.1
0.0
0.1
0.2
g-ray irr.
Un-irr.
Inte
nsity / a
rb.u
nit
g value / -
ESR spectra for the un-irradiated and the g-ray irradiated samples
Fundamental study for the annihilation of irradiation
defects in Li2TiO3 was performed.
The defects induced by g-ray annealed in higher
temperature region than those by thermal neutron,
indicating that formation processes were important in
the trapping and detrapping of tritium.
Correlation between tritium release and annihilation of
irradiation defects in lithium oxide (Li2TiO3)
Kyushu University tritium activity : (1) Clarifying behavior of tritium release from
blanket materials and developing mew tritium recovery method
tritium extraction
1st wall
tritium generation
rate
Flibe
Be
reactor core
leak rate < 10Ci/day
FFHR-2
6.5m3/s (12.8t/s)
500o
C
600o
C
3GWt
pump
heat exchanger
4.5MCi/day
gas turbine
Generator~
cooler
compressor
He (3.3 t/s)
tritium permeation
220oC
395oC
Flibe loop for FFHR-2 and He Brayton
cycle
99.9998% T removal
permeation barrier
Flibe
Radioactivity of tritium released from neutron-irradiated Flibe
(2LiF+BeF2 mixed molten salt) under under Ar + H2 purge at
300oC. Experiment and calculation are in good agreement.
•FFHR is a conceptual design of
stellarator-type fusion reactor designed
by NIFS.
•This is operated under steady-state
high beta plasma confinement.
Flibe has low tritium solubility
and counter-current tower for
Flibe-He bubble extraction is
under design.
Kyushu university activity, 3rd china-Japan tritium workshop in Kunming China, June 21,
2010
leak leak
envi
ron
men
t
enviro
nm
ent
environmen
t
Porous concrete wall
leak
Containment cell
T
TD
T
D
T
HHO
Kyushu University activity : (2) Tritium containment in concrete block with hydrophobic
paints
HTHTO
CH3T
Concrete pellets with or without paints
are exposed to HTO vapor atmosphere
Globe box
HTO vapor exposure experiment
Hydrophobic paint coating
HTO dissolution from concrete pellets to water
•HTO transfer in porous concrete materials can be predicted by the diffusion model.•Epoxy paint coating is preferable to reduce HTO transfer in concrete.•The reduction factor is around 1/10.
results
•It is necessary to contain tritium safely in the tertiary enclosure composed of concrete walls in case of accidental tritium release.•The present experiment clarifies how the released tritium is transferred through concrete walls and how well hydrophobic paints coated on concrete can protect tritium transfer in walls.
Kyushu university activity, 3rd china-Japan tritium workshop in Kunming China, June 21, 2010
Behavior of Environmental Tritium
and Assessment of Influence on
Environment
Research groups
Sugihara, S., Momoshima, N. (RIC, Kyushu Univ.)
Amano, H., Ando, M. (JAEA)
Miyamoto, K. (NIRS)
Takahashi, T.,Fukutani, S. (Kyoto Univ.)
Shimada, J. (Kumamoto Univ.)
Tamari, T. (KEEA)
Uda, T., Sakuma, Y., Yamanishi, H., Tanaka, M (NIFS)
Purposes
- to develop the technique to evaluate the
environmental tritium behavior of the facility
origin.
seasonal variation
year change
climate change in the environmental tritium
- to verify safety than the level of the other
domestic area in the change level with tritium
concentration around the facilities.
Determination of tritium concentration of 34 river
waters and 6 lake waters in Japan
45
~ 12 PBq (2009 March)
4.91 EBq/year
Tritium Process Laboratory (TPL) of JAEA
Building of TPL
Glove Box
46
0
20
40
60
80
88 90 92 94 96 98 00 02 04 06Am
ount
of
trit
ium
rel
ease
d
(GB
q)
Fiscal year
HTO
HT
Average concentration of tritium at stack =
60 Bq/m3 (1/100 of regulation value).
From experimental apparatus
CompressorFrom glove box
Catalysts
473 K and
773 KMolecular sieve bed
To stack
Cooler Pre-heater
Vessel
Recycle operationConceptual flow diagram of
tritium removal system
Amount of tritium released from stack
47
1) A series of safe operation of tritium has been demonstrated at TPL.
2) The design studies of ITER Detritiation system, and ITER TBM
system have been continued at TPL.
3)Some basic studies on tritium safety technologies have also been
carried out at TPL.
Detailed information will be presented by Dr. Yamanishi later.
Schematic of Blanket Tritium Recovery
System using Hydrogen Pump for ITER-
TBM
Basic research for tritium
safety related to BA program
–tritium permeation behavior
into cooling water
oxide layer growth
(magnetite, 1.7mm)
on iron after 20 h in
water jacket at 423K