The development of divertor and first wall armour parts at JAERI, Sandia N.L. and KFA Jülich

Fusion Engineering and Design 16 (1991) 111-125 North-Holland

111

The development of divertor and first wall armour parts at JAERI, Sandia N.L. and KFA Jiilich M. A k i b a a, H. Bolt b, R. W a t s o n c, G. K n e r i n g e r a a n d J. Linke e "Japan Atomic Energy Research Institute, 801-1 Naka-machi, Naka-gun, Ibaraki-ken 311-01, Japan b Nuclear Engineering Research Laboratory, The University of Tokyo, Tokai-mura, Ibaraki-ken 319-11, Japan c Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185, USA d Metallwerk Plansee GmbH, A-6600 Reutte, Austria e Institut fiir Reaktorwerkstoffe, Forsehungszentrum Jiilieh GmbH, Postfaeh 1913, D-5170 Jiilieh, Germany

The development of new armour materials, and fabrication and testings of the divertor and first wall mock-ups have worldwidely been carried out during the Conceptual Design Activities (CDA) of ITER. This paper is a review of the activities on the divertor and first wall armour components which has been performed by JAERI, Sandia National Laboratory, and KFA Jiilich. The design requirements have instantly been reflected in material development. For instance, carbon fiber composites (CFC's) have already been developed to have a thermal conductivity as high as copper at room temperature. Further modification of CFC's has been made. Based on the extensive progress in armour materials, the fabrication and testings of mockups have been started. Divertor mock-ups which are able to endure a stationary heat flux of 8 to 15 MW/m 2 have already been developed. Some new high heat flux test facilities have been constructed and are able to simulate a heat load of plasma disruption. Extensive understanding on disruption erosion of the armour materials has been obtained by experiments with these facilities. Some mock-up tests and disruption simulating tests have been performed under international collaboration.

1. Introduction Extensive R & D activities have been carried out for fusion experimental reactors, such as ITER which is designed under international collaboration. In the ITER design, fusion power and plasma power reach 1000 MW and about 200 MW, respectively. The burning time exceeds several hundred seconds. The plasma facing components (PFC) have to endure severe particle and heat loads from the plasma. The development of PFC's have actively been performed all over the world for several years, starting with basic tests which include material screening, development of bonding techniques, and heating tests with small specimens. Although these elemental tests are still continued, the PFC development has recently advanced to a new stage, in which partial models (mockups) have been fabricated and tested in high heat flux test facilities. In addition the study of disruption damage of armour materials have extensively been performed by using new high heat flux test facilities, such as the JAERI Electron Beam Irradiation Stand (JEBIS), which can simulate disruption heat loads. Some experiments have

already been performed under international collaboration, such as the J A E R I - K F A - N E T collaboration. The international collaboration becomes increasingly important to develop and study PFC's for the next fusion devices. This report summarizes the present status of R & D activities on PFC's in JAERI, KFA, and SNL with emphasis on testing results of the armour parts. Major testing results in NET Team and USSR are also included. In the next section the basic requirements to plasma facing materials in the ITER design are described. Sections 3 and 4 describe the fabrication and testing of PFC's. The response of plasma facing components to off-normal operation conditions is described in section 5.

2. Plasma facing armour materials 2.1. Requirements to plasma facing materials

The plasma facing materials provide the interface between the plasma and the tokamak structure. Most

0 9 2 0 - 3 7 9 6 / 9 1 / $ 0 3 . 5 0 © 1991 - Elsevier Science Publishers B.V. All rights reserved

112

M. Akiba et al. / Development o f divertor and first waft armour parts

of the requirements to these materials are based on the need for compatibility between the plasma facing material and the performance of the tokamak plasma itself. Of highest importance with regard to the plasma-material interaction is the control of the influx of impurities from the material into the plasma. This leads to a strong limitation of the materials in question and also of the allowable surface temperature to avoid excessive material erosion by the incident plasma. The maximum allowable temperature depends on the erosion behavior of the material under particle and heat load impact and on the application of the material, either as divertor or as first wall armour. An additional requirement for next generation tokamaks is to maximize the lifetime of the plasma facing components under normal and off-normal operation conditions which leads to different demands for the divertor and for the first wall armour. In the case of the divertor the high incident particlc flux leads to a strong limitation of the maximum allowable surface temperature for the plasma facing materials. In addition to the erosion during normal operation severe thermal erosion due to disruptions is expected to occur on the divertor. Thus the aim is to maximize the thickness of the plasma facing armour on the divertor so that the erosion life of the component is lengthened. The high incident heat flux on the divcrtor which has to be removed requires that the thermal conductivity of the plasma facing material is as high as possible so that a plasma facing armour with maximum thickness can be realized. The operation conditions of the plasma facing components in a next generation tokamak will be close to the scenario developed for I T E R [1] shown in table 1. The main requirements for the divertor armour of next generation tokamaks can be summarized as follows: optimized composition of the material with regard to the interaction with the plasma and thus allowing high operating temperatures at thc material surface, optimized behavior with regard to thermal erosion and fracture during disruptions, - maximum thermal conductivity to allow h~r maximum armour thickness, stability against neutron induced degradation of material properties, especially of the thermal conductivity, - optimization of the interface for the bonding of the armour material to the actively cooled metal structure. For the first wall the requirements to the materials are less stringent as compared to the divertor. Since -

-

-

Fable 1 Main operation parameters for the plasma facing materials during the physics phase in a next generation tokamak (1TER)

Normal operation incident ions peakflux(m ~s i) energy (eV) Surface heat flux peak (MW/m 2) average (MW/m 2) Av. neutron wall load (MW/m 2) No. of pulses (tokamak) Total burn lime

Divertor

First wall

4×1023 , 100

ix102~ < I(10

15--3{) 0.6


(I.5

1 104

(tokamak, hrs)

Disruptions Number Thermal quench time (ms) Peak energy deposition (MJ/m 2) Current quench time (ms) Energy dep. (MJ/m 2) Runaway electrons peak energy (MeV)

400

500 0.1-3 10-20 5 I00 2

2 5-100 2 300

the direct particle flux from the plasma is much less than on the divertor the operation temperature can be higher. Due to the lower heat flux on the first wall the thermal conductivity of the armour may be lower. Because of the lower disruption heat loads the thermal erosion is less serious than on the divertor. The present development of plasma facing materials is pursued in three main directions: Carbon materials are developed as main option for near term application. Carbon fiber reinforced carbons (CFCs) and pyrolytic graphite with high thermal conductivity are available, but other properties, especially the stability of the thermal conductivity against neutron irradiation and the thermal shock behavior during disruptions are deficient. Other low Z materials like beryllium, carbon based doped materials and boron-carbon compounds are developed to improve the plasma compatibility as compared to that of carbon. The application of these materials in present tokamaks showed remarkable success [2,3]. On the other hand the major drawback of beryllium is the low melting point which limits the

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maximum operation temperature of the material. At present a drawback of most boron-carbon materials is the lower thermal conductivity as compared to high thermal conductivity CFCs or pyrolytic graphite. The third direction is the development, of high Z plasma facing materials like tungsten or tungsten alloys. In a properly adjusted plasma environment the erosion lifetime of these materials would much longer than that of low Z materials. High Z materials in tokamaks do however not allow flexible plasma scenarios during operation. Thus the application of these materials is presently regarded as long term option.

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COHPRESSIONARI~EALEO

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~O~.YrlC GRAP.IT~

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q

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q

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113

Decelopment o f divertor a n d first wall a r m o u r parts

,

technology

of advanced

Among carbon materials CFCs with high thermal conductivity have been developed [4]. In addition highly oriented pyrolytic graphite (HOPG) is available. Figure 1 [1] shows the thermal conductivity of these materials as function of'the temperature. The fiber architecture

I 1400

TEt~PERATURE E°C3 Fig. 1. T h e r m a l conductivity o f candidate materials f o r plasma facing components as function o f the temperature.

,4

/ mixing r q,x"Coat-Mixprocedure"t

moulding pressure: ] "~,, 2 to 16 bar I"

i

1O0 lum

materials

moulded green body x/

carbonization

i porouscarbon bounded ceramic composite of high homogeneity

Fig. 2. Manufacturing process for B4Cand SiC based Coat Mix materials.

114

M. A kiba et al. / Det,elopment of dil,ertor and first wall armour parts

of the C F C materials has not been optimized, yet. Such optimization should create a material with high thermal conductivity, the necessary strength and good interfacial properties for bonding. Presently several types of boron-carbon materials are under development. One group are-conventional graphites or C F C materials which are doped with a boron content between 1 and 30%. The evaluation of these materials is under way [5,6]. A n o t h e r group of materials are compound materials of B4C grains in a carbonized matrix (Coat-Mix materials) [5]. These materials have boron contents of about 60% (figure 2). Due to the limited thermal conductivity of these compound materials they can be applied as first wall tiles rather than as divertor armour. The same holds for related silicon-carbon materials which are also under investigation. The high Z materials which are presently available for fusion application are either based on powder metallurgy or on the vacuum arc melting techniquc. From the materials point of view the brittlencss of these materials is still a major problem. In addition to the bulk materials described abovc plasma spraying techniques to apply coatings of thicknesses of the order of up to an mm are under development. Plasma spraying would eventually providc the possibility for in-situ repair of the coatings inside of the tokamak by remote handling. Plasma spray coating of beryllium [7], B4C [5] and tungsten [7] has been performed with success. Inherent problems with these coatings are internal stress, coating adhesion and limited thermal conductivity.

A common problem which needs to be addressed with all of these materials is the stability under neutron irradiation. With carbon based materials a major concern is the rapid degradation of the thermal conductivity [8] whereas refractory metals may embrittle already under low neutron doses.

3. F a b r i c a t i o n a n d t e s t i n g of d i v e r t o r e l e m e n t s

The design parameters of the I T E R PFC's are summarized in tables I and 2. Low-Z materials, carbon and Be, are proposed as armour materials for the physics phase, and high-Z materials, W and Mo, for the technology phase. As structural materials Mo, Nb and Cu alloys are proposed for the divertor plate. In this section, the fabrication and testing of divertor elements are reviewed. 3.1. Divertor plate with low-Z armour materials

The low-Z armour materials, such as graphite and carbon fiber composites (CFC), are proposed as reference materials in the physics phasc of the ITER. Be armour is considered as alternative. A schematic of thc divertor plate for the physics phase is shown in figure 3 [1]. The divertor plate consists of many blocks brazed on cooling channels. Two types of the divertor blocks are proposed for I T E R . O n e is a heat sink type, and the other is a monoblock type, as shown in figure 4. An armour tile is brazed on the metal heat sink in the heat sink type, and is directly brazed on the cooling tube in

Table 2 Major design parameters for the plasma facing components in 1TER Technology

Operation phase

Physics

Component

First Wall

Materials Armor • reference • alternative • peak temperature °C • baking/conditioning ° C temperature - Structure • reference

-

• alternative Coolant • Inlet temperature • Inlet pressure

°C MPa

Divertor

('-Fiber Composites Be 100t)

1800

First Wall

Divertor

W C < 50(l

W C. Be 1500 200/150

35(I/15{)

Mo-alloy Nb-alloy DS-Cu

SS 316

Water 60

Water 60 1.5

Nb-alloy Mo-alloy DS-Cn

SS 316

3.5

1.5

3.5

M. Akiba et al. / Decelopment of diuertor and first wall armour parts

115

20OO°C 1500°C 1200°C1000°C JLll

25.00-

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Vzx 5s toading time Ix z ls loading time

-

1000 Z~ =E aa

500 -

v

""~,ghtly ~

".~ec~ed~

250 -

brazei n t a c t

x

1O0

v

-

050-

I ! brazeintact slightlyaffected

v

~7

v

v

beginning damage emerging of braze

0.25-

Zr

NiTi

5

6

v CuTi

~

AgCuT

8

9

brazing femperoture/10-~ K -I Fig. 5. Permissible heal flux of graphite/TZM bonds brazed

at various temperatures.

Fig. 3. Schematic of ITER divertor plate. the monoblock type. In the monoblock divertor case the stress distribution around the tube is uniform, however the manufacturing integrity is more complicated than that of the heat sink type. In the case of the heat sink type, however, stress concentrations occur at

Sink Tvoe

Monoblock Tvoe

Fig. 4. Schematic of divertor block.

the edge of the interface between the armour tile and the metal substrate, which will lead to cracks in the tile. The ITER divertor design adopts the monoblock divertor as a reference from the view point of lower stress concentration. The fabrication of the divertor block is quite difficult because the thermal expansion coefficient and Young's modulus of the carbon relatively differ from that of the substrate materials, Mo and Cu alloys. Screening tests on brazing technique have been performed in SNL, NET Team, and JAERI. Comprehensive work on braze materials and brazing temperature with graphite/TZM bonds, as shown in figure 5, was performed by the NET Team. The material properties are summarized in table 3. A major fatigue mechanism comes from residual stress at the bonding interface. The NET Team has been developed several types of the divertor mock-ups; CFC flat tiles brazed on Cu-alloy and Mo-alloy heat sinks, and CFC monoblock divertor. These mock-ups has been tested and under testing at several laboratories, such as CEA Cadarache, ENEA Frascati, KFA Jiilich, KFK Karlsruhe, and SNL under the international collaboration. One of the divertor mock-ups designed by the NET Team with CFC (AO5) flat tiles brazed on TZM heat sink with Mo/Re41 tubes, similar to the one shown in fig. 6, were manufactured by Metallwerk Plansee and has

116

M. Akiba et al. / Decelopment o/'dicertor and/Trst wall armour part.s

Fig. 6. Actively cooled divertor Targets (CC Armour/TZM/Mo-Re) Designed by NET Team, Manufactured by Plansce, to be tested in the Ion Beam Test Stand of KFA Jiilich.

been tested in the electron beam facility in SNL lit;] under the DOE-Sandia-NET agreement. The mock-up have resisted single shots up to about 13 M W / m 2 and to more than 1000 cycles at 8 M W / m e. However thc maximum surface temperaturc exceeds 1000 ° C which is the ITER requirement; an optimization of the design will be done in the future. At SNLA the bonding of pyrolytic graphite ( P G ) / M o and Cu alloy tubes was studied. The results are summarized in table 4. Only the pyrolytic graphite/OFHC Cu bonds do not show damage after the brazing process. A divertor mock-up with P G / O F H C bonds as shown in figure 7 has been manufactured and was tested in the electron beam facility. This mock-up can endure a heat flux of 15 M W / m 2, and maximum surface temperature remains around 1000 ° C. Degradation of the braze joint was observed after 1000 cycles at 15 M W / m 2. At JAERI various divertor mock-ups with C / C u and C F C / C u bonds were produced as shown in figure 8. The mock-ups are tested in an ion and electron beam facility. Thermal cycle tests in the ion beam facility with a short pulse heating, and with stationary heating in the electron beam facility were performed. The results are summarized in table 5. Aftcr 1000 thermal cycles, C F C / C u bonds survive against a heat

flux of II) MW/m-'. while C / C u bonds show cracks at the bonding interface. For a heat flux of 12.5 M W / m ~, C F C / C u bonds show separation and cracks at the bonding interface.

Fig. 7. Overviewof divertor mock-up (SNL).

M. Akiba et al. / Del,elopment of diuertor and.first wall armour parts

I 17

Table 3 Brazes used Nominal composition (wt%) Composition after brazing wt% (EDX), metals only Brazing temperature ( o C) Braze thickness (/xm) prior to brazing after brazing Penetration depth into graphite (~m)

Bulk materials used Materials Trade name Specimen geometry (ram) Composition (wt%) Density (g/cm 3) Open porosity Expansion coefficient at RT ( l / K ) Thermal conductivily at RT (W/re. K) Melting point ( ° C)

70Ag 27Cu 3Ti

90Cu 10Ti

90Ni 10Ti

Zr

30Ag 50Cu 15Ti 5Mo 800-850

83Cu 16Ti 1Mo 950-1020

27Ni 3Ti 70Mo 1330-1400

66Zr 34Mo

100 ~ 40

100 ~ 35

100 400

100 130

max. 600

max. 800

max. 50

max. 100

152(I- 1865

Graphite

Substrate

Molybdenum

CL 1116 PT (fine grain graphite) 80 x 80 x 10 and 25 x 25 x 10 resp. C

-

1.82 (2.27 theor.) 10% 5.5.10 6

TZM (refractory molybdenum alloy) 80 x 80 x 3 and 25 x 25 x 3 resp. 99.3%Mo, 0.5%Ti, 0.08%Zr 10.15 5.3.10 - 6

5 . 3 ' 10 - 6

90

126

142

~ 3700 (subl.)

(2620)

2620

At Efremov Institute in U S S R , small mock-ups of divertor plates have been tested in an electron beam facility. The mock-ups had isotropic graphite armours brazed on O F H C Cu and tested at 26 M W / m 2 for 0.3-0.5 s. The graphite armours lost thermal contact

Mo 10.2(I -

after few cycles. Small monoblock mock-uPs have also been made. Results of the heating tests on the monoblock divertor will be reported by Barabash [21]. It should be noted that the dimensions of the armour tiles in these manufacturing trials were different.

Table 4 Monoblock brazing trials I @

]

Tile PG PG PG PG PG

Tube OFHC Glidcop Moly Glidcop Moly

Interlayer none none none OFHC OFHC

Result no cracks cracks cracks cracks cracks

118

M. Akiba et al. / Det:elopment oldivertor and first wall armour parts

Table 5 Summau of thermal cycling tests of various divertor mock-ups Armor material

(2"FC CX2002U

Interlayer

PCC-2S AO5 JCC

none Mo TZM none none none

Graphite IG430U PD330S

none none

Heat flux ( M W / m 2) 10

12.5

No damage No damage

Separation

No damage

Cracks

.4"

(re-radiation) 4,.

~

~,~"Armour

.4"

Tile i ?~!~{{J

~ g ~ { * } ' ] : : 1 ~, :

,i, ,i, /

I

==

\

=

:

~, ,i, .

=

==1

I~ i~>~{.:!

ee,%~

~

'

~

I md,==,~==,~=,~=l

~- I / ~Substrate /

i,¢ [

Armour ll!.e.~,~:g I I

I

\ \Coolinq Channels

Cracks Cracks Separation Separation

Radiativ~ly Cooled First W~ll

Cracks Cracks

and s e p a r a t i o n of W tiles lk~r a heat flux of 10-30 M W / m 2. A t p r e s e n t W - F G M (functionally g r a d e d material) tiles b r a z e d on Cu tubes are u n d e r d e v e l o p m e n t at J A E R I . U S S R group in Efremov Institute have b e e n tested small mock-ups with t u n g s t e n a r m o u r s c o a t e d on O F H C Cu h e a t sinks. T h e mock-ups could e n d u r e over 2000 t h e r m a l cycles at a heat load of 10 M W / m e.

T h e r e f o r e , the results c a n n o t be c o m p a r e d completely. F u r t h e r heating tests on C F C a n d C / m e t a l b o n d s are n e e d e d to obtain a whole view of the fatigue mechanisms.

Conductively Cooled First Wall

Fig. 9. Schematic of two types of the first wall.

3.2. Dicertor plate with high-Z armour materials"

4. F a b r i c a t i o n

In c o n t r a s t to the wide fabrication experience on c a r b o n a r m o u r divertors elements, only very few metal a r m o u r divertor e l e m e n t s have b e e n m a n u f a c t u r e d and tested. J A E R I m a d e such W / C u divertor mock-ups and tested t h e m in the ion b e a m test facility with a heating d u r a t i o n of 1 s. In these tests W / C u b o n d s e n d u r e a h e a t flux of 8 M W / m 2, a n d show cracks in W

Most of the first wall is exposed to a heat flux of 0.2 M W / m 2. In front of the toroidal coils a heat flux of 0.6 M W / m 2 is locally deposited on the first wall because of ripple losses, T h e a r m o u r tiles of the first wall are mechanically a t t a c h e d on the cooled substrate, which is m a d e of stainless steel. As a r m o u r materials, C F C ' s are c o n s i d e r e d for the physics phase, a n d a c o a t e d

and testing of first wall elements

Fig. 8. Overview of divertor mock-up (JAERI).

M. Akiba et a L / Development of divertor and first wall armour parts

tungsten layer or bare stainless steel are proposed for the technology phase. Several first wall structures have been developed for ITER. In the physics phase, a conductively cooled first wall is adopted for lower heat fluxes, and a radiatively cooled first wall, which can re-radiate a part of incident heat flux to the conductively cooled armour is proposed for higher heat fluxes. For cooling channels in the substrate, rectangular channels are proposed for the outboard first wall, and circular channels are proposed for the inboard first wall. On the basis of numerical predictions without considering the degradation of thermal conductivity induced by neutron damage, the surface temperature of the conductively cooled first wall remains at 300700 ° C, while that of the radiatively cooled first wall is about 1700 ° C. In the technology phase, tungsten is coated on the substrate by plasma spraying, which is a commercially available technology. As alternative a bare stainless steel is considered, which means that no armour tiles are attached to the first wall. First wall elements with conductively and radiatively cooled CFC armour have been manufactured by JAERI and were tested in the JEBIS electron beam test facility. A schematic of both first wall elements is shown in figure 9. A major issue with the conductively cooled first wall concept is to obtain good thermal contact between the armour tile and substrate which also remains during heating. On the other hand, a critical issue with the radiatively cooled first wall concept is

119

whether the tile can be sufficiently thermally insulated from the substrate, and whether the tile attachment mechanism has enough durability. Figure 10 shows the conductively cooled first wall mock-up installed in the test facility. The CFC armour is fixed on the substrate by the bolt and nut. The substrate, bolt and nut are made of stainless steel. A carbon compliant layer of 0.2-0.6 mm in thickness is inserted between the armour and substrate to obtain high thermal conductivity. The conductively cooled first wall is tested at a heat flux of 0.2 M W / m 2 for 20-60 minutes. The surface temperature remains at 350°C for a compliant layer thickness of 0.6 mm. The maximum surface temperature of 700 °C occurs at a cover cap of the attachment nut. No temperature degradation is found for 60 thermal cycles. Radiatively cooled first wall elements have also been fabricated and tested. The carbon armour is fixed on the substrate by a couple of ceramic bolt and nut. The ceramic nut is brazed in a Ti cover which is screwed into a TZM pedestal brazed on the substrate. The surface of the substrate is coated with Cr20 3 to enhance radiation heat transfer between the armour and the substrate. An emissivity of 0.8 is obtained by the Cr203 coating. The substrate has rectangular cooling channels as shown in figure 9 and is fabricated by hot isostatic pressing (HIP). This manufacturing method is one of the realistic technique to fabricate a large first wall. Figure 11 shows the test specimen

Fig. 10. Overview of conductively cooled first wall.

12(}

M. Akiba et al. / Del~elopment of dil:ertor and .[irst wall armour part,s

Fig. 11. Overview of radiatively cooled first wall.

Fig. 12. First wall mock-ups with round cooling tubes (NET).

M. Akiba et al. / Development of divertor and first wall armour parts

121

Fig. 13. First wall mock-ups with carbon tiles (NET).

'

~ 2 mm

, 200~m

t

r

, 50/am

~ 501~m

Fig. ]4. SEM images of pyrolytic carbon specimens exposed to a heat load of ]200 M W / m 2 for 5 ms duration by electron beam and laser beam (top: JEBIS facility; bottom: N d - Y A G laser).

122

M. Akiba et al. / Detrelopment of dit,ertor and first wall armour parts

installed in the test facility. In steady state, the surface temperature reaches over 1000 °C. After 50 thermal cycles, no damage was found on the armour tile and substrate. The Ti cover of the ceramic nut, however, is melted. It means that the thermal insulation by the ceramic bolt and nut results in high temperature at the Ti cover. The NET Team has developed and fabricated the first walls which have the round cooling channels without armour tiles, as shown in fig. 12 [20]. Stainless steel cooling tubes are brazed into a steel panel manufactured from two halves by transparent electron beam welding or brazing by industry. Thermal fatigue tests have been performed at a heat flux of 0.5 M W / m 2 in JRC Ispra. They found no fatigue damage of the heated surface for the thermal cycles of up to 36000 cycles. They also have fabricated the radiative cooled first wall with CFC tiles as shown in fig. 13. Heating tests will be performed at KfK Karlsruhe. Manufacturing development of the NET first wall has been also carried out at CEA Cadarache, FRAMATOME, and ANSALDO.

5. M a t e r i a l a n d c o m p o n e n t

behavior under off-normal

operation conditions

5.1. Bisruptions

During disruptions the plasma facing materials of a tokamak are subjected to very high surface energy depositions from the plasma. The values for a next generation device shown in table a are peak energy depositions during the thermal quench phase of 10-20 M J / m 2 on the divertor and about 2 M J / m 2 on the first wall. During the subsequent thermal quench phase about 2 M J / m 2 radiative energy would be deposited on both first wall and divertor. The response of the plasma facing materials to short pulses of high energy deposition is investigated in experiments using electron beam or laser facilities. Extensive experiments have been carried out in the frame of an collaboration between the Japan Atomic Energy Research Institute (JAERI), the Research Center Jiilich (KFA), and the Next European Torus (NET) Team using the JAERI Electron Beam Irradiation Stand (JEBIS). Graphites, CFCs, pyrolytic carbon, boron-carbon materials, B4C and SiC Coat Mix materials, stainless steel and plasma sprayed B4C and tungsten coatings have been tested. The specimens had been subjected to heat loads ranging from 1.8 to 7

M J / m 2 for pulse durations from 2 to 10 ms. The results in terms of thermal erosion per unit of deposited energy, crack formation and melting are described in [10]. A comparison of the results from the JEBIS clectron beam experiments with results from laser beam experiments at KFA Jiilich has been performed by exposing specimens of the same material to the same heat loads in both facilities. Figure 14 shows SEM images of pyrolytic carbon surfaces after exposure to 1200 M W / m 2 for 5 ms duration. The appearance of the crack formation by delamination is similar in the electron beam and the laser experiment. Also the crater depths of the thermal erosion craters from both experiments are of similar magnitude, fig. 15. From both types of experiments the following preliminary conclusions on the material behavior under high heat depositions can be drawn: - The thermal erosion on carbon materials exceeds the values of numerical predictions by factor of 3-5. The observed heat of ablation for carbon materials is of the order of 10-15 k J / g which is also consistent with laser experiments carried out at ECN Petten [11]. - D e p e n d i n g on the manufacturing process boron doped graphite materials show thermal erosion of similar order or less erosion than pure graphites. - Disabling crack formation has not been observed with CFCs and fine grain graphites. - Pyrolytic carbon shows strong delamination along the basal planes of the material. - B4C Coat Mix material shows surface melting and SiC Coat Mix materials show erosion but no disabling crack formation. - BaC plasma spray coatings on graphite flaked off under 2 ms energy deposition pulses but adhered and melted at longer energy depositions. The effect of neutron irradiation on the disruption behavior of plasma facing materials will be investigated by exposing neutron irradiated material specimens to high heat flux pulses in hot cell electron beam facilities. Such facilities have been constructed at KFA Jfilich and are under construction at SNL Albuquerque. In addition to the investigation of the response of materials to high energy depositions experimental work has begun at the University of Tokyo with the aim to determine the interaction processes of the eroded material with the incident plasma [12]. These processes may cause a considerable reduction of the deposited energy due to shielding of the material surface from the incident energy by the thermally eroded material.

123

M. Akiba et aL / Development of divertor and first wall armour parts

Other experimental facilities in which the response of materials to high heat flux deposition from dense plasmas can be studied and where initial results have been obtained are the Plasma Guns at the University of New Mexico, U S A [13], at the Efremov Institute in Leningrad, USSR, and the Plasma Focus at the University of Stuttgart, F R G [14].

In addition a major part of the eroded material may be redeposited on the heated surface. In the experiments a magneto-plasma-dynamic arc jet has been used to expose carbon specimens to 3.8 M J / m 2 energy depositions from a dense hydrogen plasma. Initial results from spectroscopy of thermally eroded carbon in the hydrogen plasma shows very strong line radiation from excited species which indicates the presence of a vapour shielding effect. The carbon ion concentrations in the plasma show that most of the eroded neutral carbon particles are ionized near the heated surface. Since the ionized species have to follow the magnetic field lines in a tokamak redeposition should occur at the locations were the respective field lines intersect with the component surface.

5.2. Runaway electrons

It is expected that runaway electrons will be accelerated during the current quench phase of disruptions in next generation tokamaks. The energy content of the runaway electrons can be of the order of 100 MJ at an anticipated peak energy of about 300 MeV [15]. In

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contrast to the surface energy deposition during the thermal quench phase of disruptions a large fraction of the runaway electron energy will be deposited in the bulk of the low-Z plasma facing materials. Thus high thermal excursions can occur not only in the plasma facing material, but also in the metal substrate in the vicinity of the coolant channels. Experiments with an electron linear accelerator have been performed to determine the energy deposition from high energy electrons in metal structures which were shielded by graphite of varying thickness [16]. The experimental results were modeled with good accuracy [17]. On this basis further modeling effort using Monte Carlo computer codes was performcd to determine the energy deposition in plasma facing components under runaway electron impact and to derive threshold values for the energy deposition above which component damage would occur [18]. Figure 16 shows results from Monte Carlo calculations of the energy deposition in plasma facing component materials as function of the runaway electron energy.

6. C o n c l u s i o n

The development of now armour materials, and fabrication and testings of the divertor and first wall

mock-ups havc worldwidely been performed during the ITER CDA phase. In this paper, the activities in JAERI, Sandia National Laboratories, and KFA Jfilich are reviewed with including major results of the NET Team and USSR. For carbon based materials, it has been required to have following pmpertics: - high thermal conductivity at higher temperature, - high compatibility with plasma, -- high resistance against neutron damages. One approach to obtain high compatibility with plasma is development of boron coated or doped carbon materials. The application of metal matcrials for the armour depends on experimental results in present plasma devices and il is presently rcgarded as long term option. The fabrication of the divcrtor mock-ups with carbon armour tiles, havc been performed in EC. Japan, USSR, and US. Dimension of the mock-ups are presently limited in several tens cm e and major interesting is the brazing between the armour and the actively cooled substratc. Thc carbon tiles have been successfully brazed on TZM and OFHC Cu substratcs, but further brazing trials should bc required for carbon-DS Cu bonds. Some mock-ups have already tested in the high heat flux test-facilities and they can cndurc a heat flux of 8 to 15 M W / m : for 1000 cycles. The fabrication of the first wall mock-ups have bcen carried out in EC and Japan. The demonstration of the radiatively cooled first wall and the conductivety cooled first wall have been conducted in both parlics. The thermal fatigue tesls of the first wall have also been procceded in EC under the NET Team as the central figure. Matcrial properties under the off-normal plasma conditions have been studied from the both sides of the analyses and experiments. The disruption simulating experiments have been made in thc plasma jet facilities, laser beam facilities, and clectron and ion beam facilities. The material erosion obtained in the cxpcriments is 3 to 5 times larger than that of numerical predictions. It is strongly required lot developing thc plasma facing materials and components to make international collaboration, since thc facilities which can simulate a heat load in the next tokamak devices are limited in the world. The collaborativc cxpcriments of U S / N E T and N E T / K F A / J A E R I have been successfully being carried out. The needs for thc international collaboralion will bc increasing in the Engineering Design Activity of the ITER.

M. Akiba et al. / Development of divertor and first wall armour parts References [1] T. Kuroda, G. Vieider, Eds., ITER Plasma Facing Components, ITER Documentation Series No. 30 (IAEA, Vienna, 1990). [2] M. Keilhacker and the JET Team, Overview of JET results using a beryllium first wall, Joint European Torus, JET-P(89)83 (1989). [3] J. Winter, A comparison of tokamak operation with metallic getters (Ti, Cr, Be) and boronization, J. Nucl. Mater. 176-177 (1990) 14-31. [4] M. Yamamoto, T. Ando, H. Takatsu, M. Shimizu, T. Arai, K. Kodama, K. Horiike, K. Teranuma, A. Kiuchi, Y. Goto, Evaluation tests on first wall and divertor plate materials for JT-60 upgrade, Japan Atomic Energy Research Institute, JAERI-M 90-119 (1990). [5] J. Linke, H. Bolt, R. Doerner, H. Griibmeier, Y. Hirooka, H. Hoven, C. Mingam, H. Schulze, M. Seki, E. Wallura, T. Weber, J. Winter, Performance of boron/ carbon first wall materials under fusion relevant conditions, J.-Nucl. Mater. 176&177 (1990) 856-863. [6] Y. Hirooka, G. Chevalier, R.W. Conn, M. Khandagle, T. Matsuda, H. Ogura, T. Sogabe, H. Sugai, H. Toyoda, Interactions of bulkboronized graphites with deuterium plasmas in the PISCES-B facility, University of California, UCLA-PPG-1313 (1990). [7] M.F. Smith, C.D. Croessmann, F.M. Hosking, R.D. Watson, J.A. Koski, Plasma sprayed materials for magnetic fusion energy devices, in: Proc. of the 2nd Plasma Technic Symposium, Lucerne, 5-7 June 1991, to appear. [8] T. Maruyama, H. Masaaki, Change of thermal properties of graphite by neutron irradiations, in: Proc. of the Japan-US Workshop on "Critical Topics of PFM/PFC Data for the Next Step Fusion Devices", National Institute for Fusion Science, Nagoya, 3-6 Dec. 1990, pp. 221-235. [9] M. Akiba, M. Araki, M. Seki et al., High heat flux experiments at JAERI, in: Proceedings of 13th Symp. on Fusion Engineering, Knoxville, 2-6 Oct. 1989, pp. 529532. [10] J. Linke, M. Akiba, M. Araki, A. Benz, H. Bolt, H. Hoven, K. Koizlik, H. Nickel, M. Seki, E. Wallura, Dis-

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ruption simulation experiments in electron and laser beam facilities, in: Proc. 16th Symp. on Fusion Technology, London, 3-7 Sept. 1990 (Elsevier, Amsterdam, 1991). J.G.v.d. Laan, J. Bakker, R.C.L.v.d. Stad, H.T. Klippel, Experimental simulation and analysis of off-normal heat loads accompanying plasma disruptions, in: Proc. of the 16th Symp. on Fusion Technology, London, 3-7 Sept. 1990 (Elsevier, Amsterdam, 1991). H. Bolt, Y. Ooishi, M. Iida, T. Sukegawa, Study of the plasma-material interaction during simulated plasma disruptions, ISFNT-2, 1991, Karlsruhe, Fusion Engrg. Des. 18 (1991) 117-123, in these Proceedings, Part C. M. Langhoff, G. Hess, J. Gahl, R. Ingrain, Plasma facing component disruption heat flux simulator, Proc. of the 14th Int. Syrup. on Discharges and Electrical Insulation in Vacuum, Santa Fe, 17-20 Sept. 1990. J. Linke, Simulation of disruptions in different HHF test facilities, in: Proc. of the Japan-US Workshop on "Critical Topics of PFM/PFC Data for the Next Step Fusion Devices", National Institute for Fusion Science, Nagoya, 3-6 Dec. 1990, pp. 379-400. D. Post, N.A. Uckan, Eds., ITER Physics, ITER Documentation Series No. 21 (IAEA, Vienna, 1990). H. Bolt, A. Miyahara, M. Miyake, T. Yamamoto, C.D. Croessmann, Runaway-electron simulation by use of an electron linear accelerator, J. Nucl. Mater. 155-157 (1988) 256-260. K.A. Niemer, C.D. Croessmann, J.G. Gilligan, H. Bolt, Sandia National Laboratories, SAND89-2304 (1990). H. Bolt, H. Calen, Evaluation of runaway-electron effects on plasma facing components for NET, J. Nucl. Mater. 179-181 (1991) in print. I. Smid, A. Cardella et al., Response to high heat flux and metallurgical examination of brazed carbon composite/refractory metal divertor mock-up. B. Shaw, G. Vieider, et a., Design and performances of mechanically attached low Z armour tiles for the NET integrated First Wall, Proc. 16th Symp. Fusion Technology, London, 1990 (Elsevier, Amsterdam, 1991). V.R. Barabash, R.N. Giniyatulin, V.L. Komarov, I.V. Mazul, D.L. Miloslavsky, et al., Thermocyclic tests of the divertor plate mock-ups for the ITER rector.