Erosion of ITER divertor armour and contamination of SOL after transient events by erosion products
Plasma impact to the divertor expected in the tokamak ITER during ELMs or disruptions can result in a significant surface damage to CFC- and tungsten armours (brittle destruction and melting respectively) as well as in contamination of SOL by evaporated impurities. Numerical investigations for tungs...
Збережено в:
| Опубліковано в: : | Вопросы атомной науки и техники |
|---|---|
| Дата: | 2005 |
| Автори: | , , |
| Формат: | Стаття |
| Мова: | English |
| Опубліковано: |
Національний науковий центр «Харківський фізико-технічний інститут» НАН України
2005
|
| Теми: | |
| Онлайн доступ: | https://nasplib.isofts.kiev.ua/handle/123456789/78650 |
| Теги: |
Додати тег
Немає тегів, Будьте першим, хто поставить тег для цього запису!
|
| Назва журналу: | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| Цитувати: | Erosion of ITER divertor armour and contamination of SOL after transient events by erosion products / B.N. Bazylev, I.S. Landman, S.E. Pestchanyi // Вопросы атомной науки и техники. — 2005. — № 1. — С. 49-53. — Бібліогр.: 19 назв. — англ. |
Репозитарії
Digital Library of Periodicals of National Academy of Sciences of Ukraine| id |
nasplib_isofts_kiev_ua-123456789-78650 |
|---|---|
| record_format |
dspace |
| spelling |
Bazylev, B.N. Landman, I.S. Pestchanyi, S.E. 2015-03-19T17:33:38Z 2015-03-19T17:33:38Z 2005 Erosion of ITER divertor armour and contamination of SOL after transient events by erosion products / B.N. Bazylev, I.S. Landman, S.E. Pestchanyi // Вопросы атомной науки и техники. — 2005. — № 1. — С. 49-53. — Бібліогр.: 19 назв. — англ. 1562-6016 PACS: 52.40Hf https://nasplib.isofts.kiev.ua/handle/123456789/78650 Plasma impact to the divertor expected in the tokamak ITER during ELMs or disruptions can result in a significant surface damage to CFC- and tungsten armours (brittle destruction and melting respectively) as well as in contamination of SOL by evaporated impurities. Numerical investigations for tungsten and CFC targets provide important details of the material erosion process. The simulations carried out in FZK on the material damage, carbon plasma expansion and the radiation fluxes from the carbon impurity are surveyed. Вплив плазми на дивертор, передбачуваний в ИТЕРі при розвитку ЕЛМ нестійкостей у скін-шарі або при зривах, може викликати як значне поверхневе ушкодження (тендітне руйнування і плавлення відповідно) диверторних пластин, виготовлених із графітових композитів і вольфраму, так і забруднення скін-шару домішками, що випарувалися. Чисельні розрахунки з вольфрамовими і композитними мішенями проясняють важливі деталі процесу ерозії матеріалів. Дано огляд розрахунків, проведених у НЦК, по ушкодженню матеріалів, поширенню вуглецевої плазми і випромінюванню вуглецевої домішки. Воздействие плазмы на дивертор, предполагаемое в ИТЕР при развитии ЭЛМ неустойчивостей в скин-слое или при срывах, может вызвать как значительное поверхностное повреждение (хрупкое разрушение и плавление соответственно) диверторных пластин, изготовленных из графитовых композитов и вольфрама, так и загрязнение скин-слоя испарившимися примесями. Численные расчеты с вольфрамовыми и композитными мишенями проясняют важные детали процесса эрозии материалов. Дан обзор расчетов, проводимых в НЦК, по повреждению материалов, распространению углеродной плазмы и излучению углеродной примеси. en Національний науковий центр «Харківський фізико-технічний інститут» НАН України Вопросы атомной науки и техники ITER and fusion reactor aspects Erosion of ITER divertor armour and contamination of SOL after transient events by erosion products Эрозия диверторных пластин в токамаке ИТЕР и загрязнение плазменного скин-слоя продуктами эрозии Ерозія диверторных пластин у токамаке ИТЕР і забруднення плазменного скин-шару продуктами ерозії Article published earlier |
| institution |
Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| collection |
DSpace DC |
| title |
Erosion of ITER divertor armour and contamination of SOL after transient events by erosion products |
| spellingShingle |
Erosion of ITER divertor armour and contamination of SOL after transient events by erosion products Bazylev, B.N. Landman, I.S. Pestchanyi, S.E. ITER and fusion reactor aspects |
| title_short |
Erosion of ITER divertor armour and contamination of SOL after transient events by erosion products |
| title_full |
Erosion of ITER divertor armour and contamination of SOL after transient events by erosion products |
| title_fullStr |
Erosion of ITER divertor armour and contamination of SOL after transient events by erosion products |
| title_full_unstemmed |
Erosion of ITER divertor armour and contamination of SOL after transient events by erosion products |
| title_sort |
erosion of iter divertor armour and contamination of sol after transient events by erosion products |
| author |
Bazylev, B.N. Landman, I.S. Pestchanyi, S.E. |
| author_facet |
Bazylev, B.N. Landman, I.S. Pestchanyi, S.E. |
| topic |
ITER and fusion reactor aspects |
| topic_facet |
ITER and fusion reactor aspects |
| publishDate |
2005 |
| language |
English |
| container_title |
Вопросы атомной науки и техники |
| publisher |
Національний науковий центр «Харківський фізико-технічний інститут» НАН України |
| format |
Article |
| title_alt |
Эрозия диверторных пластин в токамаке ИТЕР и загрязнение плазменного скин-слоя продуктами эрозии Ерозія диверторных пластин у токамаке ИТЕР і забруднення плазменного скин-шару продуктами ерозії |
| description |
Plasma impact to the divertor expected in the tokamak ITER during ELMs or disruptions can result in a significant surface damage to CFC- and tungsten armours (brittle destruction and melting respectively) as well as in contamination of SOL by evaporated impurities. Numerical investigations for tungsten and CFC targets provide important details of the material erosion process. The simulations carried out in FZK on the material damage, carbon plasma expansion and the radiation fluxes from the carbon impurity are surveyed.
Вплив плазми на дивертор, передбачуваний в ИТЕРі при розвитку ЕЛМ нестійкостей у скін-шарі або при зривах, може викликати як значне поверхневе ушкодження (тендітне руйнування і плавлення відповідно) диверторних пластин, виготовлених із графітових композитів і вольфраму, так і забруднення скін-шару домішками, що випарувалися. Чисельні розрахунки з вольфрамовими і композитними мішенями проясняють важливі деталі процесу ерозії матеріалів. Дано огляд розрахунків, проведених у НЦК, по ушкодженню матеріалів, поширенню вуглецевої плазми і випромінюванню вуглецевої домішки.
Воздействие плазмы на дивертор, предполагаемое в ИТЕР при развитии ЭЛМ неустойчивостей в скин-слое или при срывах, может вызвать как значительное поверхностное повреждение (хрупкое разрушение и плавление соответственно) диверторных пластин, изготовленных из графитовых композитов и вольфрама, так и загрязнение скин-слоя испарившимися примесями. Численные расчеты с вольфрамовыми и композитными мишенями проясняют важные детали процесса эрозии материалов. Дан обзор расчетов, проводимых в НЦК, по повреждению материалов, распространению углеродной плазмы и излучению углеродной примеси.
|
| issn |
1562-6016 |
| url |
https://nasplib.isofts.kiev.ua/handle/123456789/78650 |
| citation_txt |
Erosion of ITER divertor armour and contamination of SOL after transient events by erosion products / B.N. Bazylev, I.S. Landman, S.E. Pestchanyi // Вопросы атомной науки и техники. — 2005. — № 1. — С. 49-53. — Бібліогр.: 19 назв. — англ. |
| work_keys_str_mv |
AT bazylevbn erosionofiterdivertorarmourandcontaminationofsolaftertransienteventsbyerosionproducts AT landmanis erosionofiterdivertorarmourandcontaminationofsolaftertransienteventsbyerosionproducts AT pestchanyise erosionofiterdivertorarmourandcontaminationofsolaftertransienteventsbyerosionproducts AT bazylevbn éroziâdivertornyhplastinvtokamakeiterizagrâznenieplazmennogoskinsloâproduktamiérozii AT landmanis éroziâdivertornyhplastinvtokamakeiterizagrâznenieplazmennogoskinsloâproduktamiérozii AT pestchanyise éroziâdivertornyhplastinvtokamakeiterizagrâznenieplazmennogoskinsloâproduktamiérozii AT bazylevbn erozíâdivertornyhplastinutokamakeiterízabrudnennâplazmennogoskinšaruproduktamierozíí AT landmanis erozíâdivertornyhplastinutokamakeiterízabrudnennâplazmennogoskinšaruproduktamierozíí AT pestchanyise erozíâdivertornyhplastinutokamakeiterízabrudnennâplazmennogoskinšaruproduktamierozíí |
| first_indexed |
2025-11-25T22:19:40Z |
| last_indexed |
2025-11-25T22:19:40Z |
| _version_ |
1850562638449213440 |
| fulltext |
EROSION OF ITER DIVERTOR ARMOUR AND CONTAMINATION OF SOL
AFTER TRANSIENT EVENTS BY EROSION PRODUCTS
B.N. Bazylev, I.S. Landman, S.E. Pestchanyi
Forschungszentrum Karlsruhe (FZK), Institute for Pulsed Power and Microwave Technology,
P. B. 3640, 76021 Karlsruhe, Germany
Plasma impact to the divertor expected in the tokamak ITER during ELMs or disruptions can result in a significant surface
damage to CFC- and tungsten armours (brittle destruction and melting respectively) as well as in contamination of SOL by
evaporated impurities. Numerical investigations for tungsten and CFC targets provide important details of the material
erosion process. The simulations carried out in FZK on the material damage, carbon plasma expansion and the radiation
fluxes from the carbon impurity are surveyed.
PACS: 52.40Hf
1. INTRODUCTION
In the future tokamak ITER the high power transient
processes such as bursts of edge localized modes (ELMs)
that accompany with the frequency of 1-102 Hz the normal
tokamak operation and the disruptions that sometimes
interrupt the quasi-stationary discharge, will be probably
much more serious problem than in the available tokamaks.
This is due, for instance, to large plasma energy up to 0.5
GJ assumed in ITER, which is of two orders larger than in
now available tokamaks. The heat deposition Q in the range
about 0.5 − 3 MJ/m2 on the time scale τ about 0.1 − 1 ms for
ELMs and Q ~ 10 − 30 MJ/m2 at τ ~ 1 − 10 ms for the
disruptions are expected at the divertor surface [1]. The
plasma facing components (PFC) of the ITER divertor are
going to be made of carbon fibre composites (CFC) in the
most loaded part at the separatrix strike position (SSP) and
of tungsten (W-brushe) in the other parts. The large loads
can cause surface erosion of the armour and evaporation of
a thin layer of the material that after ionization in the
impacting plasma stream acts as a plasma shield which
simultaneously propagates in the scrape-off layer (SOL).
The erosion of the armour materials and the
contamination of the confined plasma have to be estimated
before ITER itself faces the plasma wall interaction
problems. As the expected fluxes are not achievable at
present tokamaks, it follows that the behaviour of the
involved materials and the impurity plasma dynamics must
be simulated numerically. However, the only way to obtain
confidence of the used assumptions and theoretical models
is a validation of the modelling by experiments, because in
reality unexpected physical processes might play an
important role. Therefore non-tokamak validation
experiments are necessary.
This work surveys latter theoretical investigations
carried out in FZK for the divertor armour damages and the
impurity propagation in SOL applying and upgrading earlier
developed codes. The computational tool applied for direct
calculation of wall loads and plasma processes
is two-dimensional (2D) RMHD (radiation-magneto-
hydrodynamics) code FOREV−2D [2]. ITER relevant
simulations with it for SOL are described in section 2. For
simulation of the damage to tungsten surfaces the
incompressible fluid dynamics code MEMOS-1.5D [3] is
applied. The validation of the code against experiments on
the plasma guns [4-6] is explained. The simulations with
MEMOS-1.5D are presented in sect. 3. For PFC made of
carbon-based materials the main erosion effect under the
high heat fluxes is brittle destruction. In this case the
thermomechanics code PEGASUS-3D [7] was applied. The
brittle destruction simulations are described in sect. 4.
2. PLASMA PROCESSES IN SOL
The code FOREV-2D models the hot deuterium-tritium-
helium plasma lost across the separatrix into SOL during a
transient event, its propagation towards the wall and impact
on the armour surface. It calculates also a self-consistent
evaporation, ionization of carbon-, tungsten- and beryllium
atoms, the radiation transport in the contaminated plasma
and backward propagation of the erosion products into
SOL. The code was created aiming simulations for a dense
plasma shield in front of the surface at the disruption
armour loads. Therefore the algorithm described slab
geometry and one-fluid plasma, which was sufficient at the
density above 1021 m−3 [2]. Since the reference regime of
ITER operation will be probably the ELMy H-mode,
recently FOREV-2D was upgraded for adequate tokamak
relevant simulations at lower heat fluxes when the vapour
shield is rather weak or absent. Now FOREV-2D models
the whole SOL, the x−point and inner- and outer divertor
plates. The ion fluids of D, T, He, He+ and all charge states
of C are simultaneously simulated. The code calculates also
the quite important radiation transport in the carbon
emission lines, accounting for reabsorption of line radiation.
In the ELM scenario the hot plasma losses from the
pedestal region through the separatrix with a velocity
Problems of Atomic Science and Technology. 2005. № 1. Series: Plasma Physics (10). P. 49-53 49
Fig. 1. Carbon plasma density profiles along the tokamak
separatrix at different time moments
determined by a specified Q and τ. It is assumed that the
lost plasma appears then in SOL having an exponential
density profile with maximum density n at the separatrix. In
different simulations the values of n were varied in the
range 2−7×1019 m-3, initial plasma temperature from 1 to 3
keV, and τ in the range 0.1−0.5 ms. The hot plasma
propagates towards the CFC divertor armour legs, heats
them and causes vaporization at their surfaces. The carbon
vapour propagates backwards finally filling the whole SOL.
As an example, Fig. 1 demonstrates the profiles of
carbon plasma density and temperature at several time
moments t after the start of ELM at t = 0. Plasma streams
have expanded from the inner- and outer divertor legs and
then collided at ~3 ms. The density peak after this collision
exists until 7 ms. This simulation corresponds to Q = 0.8
MJ/m2 and τ = 0.3 ms. The vaporisation starts at around 20-
25 µs approximately at the same time on the inner- and
outer divertor plates.
Typical radiation flux distribution calculated at the
first wall at t = 15 ms is shown in Fig. 2. At this moment the
carbon plasma is distributed almost uniformly along the
separatrix in SOL and in the divertor legs. The maximal
radiation heat load is on the dome and adjoining parts of
divertor most filled with the radiative impurity. A smaller
peak of the radiation flux is seen at the top of the main
chamber, where the plasma thickness increased due to a
divergence of the poloidal magnetic field.
3. BEHAVIOUR OF MOLTEN TUNSTEN
3.1. ITER relevant simulations
The erosion of a metallic surface under the loads exceeding
the melting threshold is caused substantially by the melt
motion. The melt motion develops mainly due to the
variations of plasma pressure across the divertor plate of a
width of 0.2 m. For ITER, the maximum pressure at
Fig. 2. Radiation load onto ITER vessel walls calculated
with FOREV
the surface during a disruption can be of the order of 1 bar
and during an ELM of 0.1 – 0.2 bar.
The simulations for tungsten armour under the
disruptions loads are described in [3] where a large crater
depth of a few tenth of µm was obtained. Now the regimes
typical of the type I ELM are discussed following [8]. ITER
operation in the H-mode is the regime with multiple ELMs.
The ELM energy can exceed the melting threshold. After a
single ELM, melt motion produces surface roughness of a
rather small magnitude, which is nevertheless significantly
larger than the thickness of vaporized surface layer (the
evaporation thickness). However, during each ITER
discharge more than 103 ELMs are expected. From one
hand, the total roughness may accumulate and become
rather large. From the other hand, during ELMs the
separatrix strike position (SSP) moves rather unpredictably
across the armour plate [9] which can average the
roughness.
Tungsten target damage for the ITER discharges
interspaced by ELMs has been simulated with the code
MEMOS−1.5D which describes surface melting in the
‘shallow water’ approximation [10]. The melt motion is
modelled taking into account the surface tension, viscosity
of molten metal and the radiative losses from the hot
tungsten surface. The plasma pressure gradient across the
divertor plate, the gradient of surface tension and the
Lorentz force of the currents crossing the melt layer
immersed in the strong magnetic field produce the melt
acceleration. During each ELM the SSP was assumed
motionless but stochastically changing across the plate from
ELM to ELM, obeying the Gaussian distribution. In the
coordinate frame of the random SSP, the time dependent
spatial profiles of heat fluxes and plasma pressure at the
target surface were calculated with FOREV−2D taking into
account the plasma shield effect.
The simulations demonstrated that one single ELM of
τ = 0.3 ms with Q varying from 1 to 2 MJ/m2 produces
50
melting without evaporation. At Q > 2.5 MJ/m2 the vapour
shield forms significantly so that the pressure
Fig. 3. The dependence of crater depth on the number of
ELMs for δ = 0.1 m obtained with MEMOS−1.5D
gradient becomes essential. In the regimes with the melt
motion, the magnitude of surface roughness varies in range
of 0.3 − 0.5 µm, the melt velocity V is of 0.5 m/s, and the
evaporation thickness per single ELM of 0.1 µm.
For multiple ELMs, the Gaussian distribution of SSP
with the dispersion δ = 0.1 m results in a significant
decrease of the crater depth compared to the case with fixed
SSP. It is concluded that in the case of the stochastic SSP
the evaporation mechanism becomes the main reason of
erosion at the ELM number larger than 103, and thus the
erosion rate acquires a linear dependence on the number of
ELMs. For Q < 1 − 2 MJ/m2 the number of ELMs to erode 1
cm of tungsten armour is obtained to be about 105 − 106
(see Fig. 3)
3.2 Validations of MEMOS-1.5D
The tokamak simulators used for validation of the
MEMOS-1.5D so far are the electron beam test facilities
JUDITH [11] and JEBIS [12] and also the plasma guns MK
−200UG [4], QSPA-Kh50 [5] and QSPA−T [6]. The code
was earlier validated comparing numerical results with
JUDITH for beryllium and JEBIS for tungsten [3]. As a
tokamak simulator, each facility has advantages and
disadvantages. Plasma itself, as a natural substance for the
required loads, presents the main advantage of the plasma
guns. Their main drawback is short pulse durations (not
more than 0.5 ms). The electron beam high heat fluxes can
last much longer time covering the required time scales but
the beam pressure is negligibly low, which hampers the
application of e-beams to metallic targets.
Earlier attempts to validate MEMOS-1.5D by the
plasma gun MK-200UG failed. In the calculations with the
load pulses of MK-200UG of 0.05 ms a negligible damage
has been obtained in the simulations, in contrast to the
experiments. Nevertheless, the calculations discovered that
if the load would last longer, even as an energetic tail of
a small magnitude, the melt motion would get substantial
due to a non-zero melt velocity. Even without the energetic
tail the calculated resolidification time was obtained to be
much larger (of 0.35 ms) than the impact duration. The
importance of
Fig. 4. Comparison of experimental- and numerical
tungsten melting threshold
measurements on a longer time scale has been recognized.
Finally in an experiment with copper target a high surface
temperature keeping for at least 0.7 ms was obtained, which
converged theoretical and experimental results.
Other validations of MEMOS-1.5D concerned the
tungsten melting threshold. At such low load regime the
energetic tail of MK-200UG is not important. The tungsten
surface temperature Tw was measured for varied Q. During
the exposition Tw first reaches Tmelt at Q of 0.28 MJ/m2, and
Tw remains less than Tmelt at lower Q. At Q of 0.4 MJ/m2, Tw
fast increases up to Tmelt but then it remains at this value:
evidently the absorbed energy flux is consumed for melting.
After these measurements it was concluded that tungsten
melting starts at Q ≈ 0.30 MJ/m2, which is considered as the
melting threshold.
Calorimetric measurements of Q at the facility QSPA-
Kh50.for multiple (up to 300) pulse irradiations of τ = 0.25
ms provided the determination of the melting threshold.
After initial exposures the melting threshold was
determined as 0.56 MJ/m2. However, after 150 exposures
the melting threshold decreased to 0.45 MJ/m2, which
seems due to material modification and the development of
the bulk cracks parallel to the target surface that cause
decrease of thermal conductivity in a pre-surface layer.
For checking the experimental results obtained at MK-
200UG and QSPA-Kh50 and aiming MEMOS validation,
the tungsten melting threshold was also calculated. In the
calculations the triangle and rectangular shapes of the
impinging plasma load in time are assumed, with the pulse
durations covering the range 0.05-0.25 ms. Fig. 4 illustrates
the experimental results mentioned and the numerical
results obtained with MEMOS: they are in a reasonable
agreement.
4. BRITTLE DESTRUCTION OF CFC
The complicated structure of CFC is modelled in terms of
involved properties: the thermal conductivity, the
51
coefficient of thermal expansion and the Young’s modulus.
A complicated three-dimensional structure of
Fig. 5. Destruction of CFC obtained in the PEGASUS-3D
simulation
different sorts of graphite constitutes CFC. It includes a
framework consisting of bunches of fibres, so called tows
[13], and a matrix that fills the space between the tows.
CFC has good thermomechanical properties at surface
temperatures at least up to 1.5×103 K, for instance a large
thermal conductivity χ, in the fibres up to 103 W/mK.
However, in average χ is strongly anisotropic and varies
from 102 to 3×102 W/mK [14].
At the transient events the behaviour of CFC becomes
problematic. For instance, CFC samples after irradiation by
the plasma pulses of Q = 10 MJ/m2 at τ ~ 0.05 ms at MK−
200UG and QSPA−T showed a drastic surface destruction
with deep caverns of several hundred of µm and a strong
damage along the longitudinal tows [15].
For analyses of the properties of CFC at the high heat
fluxes the code PEGASUS−3D was applied [16]. The
matrix and the tows are described by means of several
millions of numerical cells of one-micrometer size. Some
grains built of different groups of the cells simulate the
graphite structure. Neighbour grains contact by means of
mechanical- and heat conduction bonds.
In the simulations, the following typical parameters
along and across the fibres are used: −1.5×10−6 and 20×10−6
K−1 for the thermal expansion coefficient, 500 and 20 GPa
for the Young’s modulus, 103 and 102 W/mK for the thermal
conductivity, respectively.
A new mechanism of macroscopic erosion was
discovered and named the local overheating erosion. Large
local thermostress reaching a failure threshold removes the
bond between the grains, thus producing a crack. The crack
interrupts the local heat conduction. To maintain the given
heat flux the temperature gradient becomes locally larger,
increasing stress in the vicinity of the crack and producing
new cracks. At the disruptive scale of the load, it is
found that the cracks are mainly concentrated in the matrix
around the tows. At the surface, the cracked out matrix
particles are assumed to emit. Horizontal tows under the
removed matrix and the caverns at the eroded surface
appear, as it is shown in
Fig. 6. Thermal conductivity of CFC matrix with- and
without cracking obtained in the simulation
Fig. 5, which is rather similar to the experiments [15].
The thermostress due to anisotropy of thermal
expansion coefficient and the Young’s modulus of CFC was
obtained in PEGASUS-3D simulations to be much larger
than the thermostress due to the temperature gradient. This
result leaded to a new approach for description of thermal
transport in CFC [17]. The local crack density N is assumed
to depend on T only and the dependence on ∇T is neglected:
N = N(T). From this approximation follows that an average
function of effective macroscopic thermal conductivity κ(T)
exists that accounts for N(T), which enabled to simplify heat
transport modelling in cracked graphite materials. Fig. 6
demonstrates the influence of cracking on the typical
thermal conductivity of CFC matrix obtained in the
calculation.
5. SUMMARY
The described investigations are published and presented
elsewhere (Refs. [2,3,7,8,16-19]). The main achievements
and problems are as follows.
In the radiation transport calculations for carbon
impurity, the line radiation dominates. It is to note that the
data for line shapes need improvements, which demands
significant effort for producing new opacities.
The stochastic distribution of SSP in MEMOS-1.5D
reduces tungsten target erosion because of smoothing the
surface roughness caused by the melt motion at each single
ELM. Validation of MEMOS against plasma gun
experiments at the tungsten melting threshold was
successful. The tungsten melting threshold obtained in the
experiments is in a reasonable agreement with the
calculations of MEMOS, which justifies the application of
the code for investigation of melt motion damage at the
tokamak conditions.
The physical picture for CFC needs further
development. PEGASUS-3D has to be validated against
averaged thermophysical properties of CFC. From
52
comparison of the simulations and the experimental
observations it is concluded that the experiments confirm
the simulation concerning the development of the caverns
and cracking horizontal tows.
REFERENCES
[1] Federici et al. // Journ. Nucl. Mat. (11).2003, p.313-316.
[2] H. Wuerz et al. // Fus. Sci. Techn. (40). 2001, p. 191.
[3] B. Bazylev, H.Wuerz. // Journ. Nucl. Mater. (69). 2002,
p. 307-311.
[4] N.I. Arkhipov et al. // J. Nucl. Mater. (767).1996, p.233-
237.
[5] V.I. Tereshin et al. // J. Nucl. Mater. (686). 2003, p.313-
316.
[6] V. Belan et al. // Proc. 20th SOFT Marseille, France, 7-
11 Sept. 1998. v.1, p. 101
[7] S.E. Pestchanyi, H. Wuerz. // Fus. Eng. Design. 66-68C,
271, 2003.
[8] B. Bazylev et al. Erosion of divertor tungsten armour
after many ELMs. // Proc. 30th EPS Conf. on Contr.
Fus. And Plasma Phys., St. Petersburg, Russia, 7-11
July 2003 // ECA v. 27A, P-2.44
[9] J. Lingertat et al. // J.Nucl.Mater. (402).1997,p.241-243.
[10] L.D. Landau, E.M. Lifshits. Course of theoretical
physics. // Fluid mechanics, Oxford u.a.: Butterworth-
Heinemann, 2000, V.6
[11] J.Linke et al. // J.Nucl.Mater. 290-293, 1102 (2001)
[12] K. Nakamura et al. , Fus. Eng. Des., 39-40 (1998) 285
[13] H. O. Pierson. “Handbook of carbon, graphite,
diamond and fullerenes”. (Noyes publications, New
Jersey, 1993) p. 189
[14] J.P. Bonal, D. Moulinier. Thermal properties of
advanced carbon fiber composites for fusion
application. Rapport DMT/95-495. CEA. Direction des
reacteurs nucleaires. Departement de mecanique et de
technologie (1995)
[15] N.I. Arkhipov et al. Journ. Nucl. Mater. 307-311
(2002) 1364
[16] S.E. Pestchanyi et al., Estimation of carbon fibre
composites as ITER divertor material. The paper will
be published in Journ. Nucl. Mater. in 2004
[17] S.E. Pestchanyi, I.S. Landman, Physica Scripta, T111
(2004) 218
[18] I.S. Landman et al., Physica Scripta, T111 (2004) 206
[19] I.S. Landman et al., Simulation of tokamak armour
erosion and plasma contamination at intense transient
heat fluxes in ITER. A report at the Int. Conf. Plasma
Surface Interactions, Portland Main, U.S.A., May 24-
28, 2004
ЭРОЗИЯ ДИВЕРТОРНЫХ ПЛАСТИН В ТОКАМАКЕ ИТЕР И
ЗАГРЯЗНЕНИЕ ПЛАЗМЕННОГО СКИН-СЛОЯ ПРОДУКТАМИ ЭРОЗИИ
Б.Н. Базилев, И.С. Ландман, С.Е. Песчаный
Воздействие плазмы на дивертор, предполагаемое в ИТЕР при развитии ЭЛМ неустойчивостей в скин-слое или при
срывах, может вызвать как значительное поверхностное повреждение (хрупкое разрушение и плавление
соответственно) диверторных пластин, изготовленных из графитовых композитов и вольфрама, так и загрязнение
скин-слоя испарившимися примесями. Численные расчеты с вольфрамовыми и композитными мишенями
проясняют важные детали процесса эрозии материалов. Дан обзор расчетов, проводимых в НЦК, по повреждению
материалов, распространению углеродной плазмы и излучению углеродной примеси.
ЕРОЗІЯ ДИВЕРТОРНЫХ ПЛАСТИН У ТОКАМАКЕ ИТЕР І
ЗАБРУДНЕННЯ ПЛАЗМЕННОГО СКИН-ШАРУ ПРОДУКТАМИ ЕРОЗІЇ
Б.Н. Базилєв, І.С. Ландман, С.Є. Піщаний
Вплив плазми на дивертор, передбачуваний в ИТЕРі при розвитку ЕЛМ нестійкостей у скін-шарі або при зривах,
може викликати як значне поверхневе ушкодження (тендітне руйнування і плавлення відповідно) диверторних
пластин, виготовлених із графітових композитів і вольфраму, так і забруднення скін-шару домішками, що
випарувалися. Чисельні розрахунки з вольфрамовими і композитними мішенями проясняють важливі деталі
процесу ерозії матеріалів. Дано огляд розрахунків, проведених у НЦК, по ушкодженню матеріалів, поширенню
вуглецевої плазми і випромінюванню вуглецевої домішки.
53
|