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Net plasma facing components
Fusion Engineering and Design 11 (1989) 125-137 North-Holland, Amsterdam
NET PLASMA FACING COMPONENTS
125
*
G. V I E I D E R , M. H A R R I S O N a n d F. M O O N S
The NET Team, c/o Max-Planck-lnstitut ffw Plasmaphysik, Boltzmannstr. 2, D-8046 Garchin~ Fed. Rep. Germany Submitted 6 October 1988; accepted 14 December 1988 Handling Editor: P. Komarek
The progress in the design and development of the first wall (FW) and divertor plates (DP) for the Next European Torus (NET) are summarized, highlighting the assumed main operating conditions, material choices, design options and their analysis as well as associated manufacturing studies and the ongoing testing programme. As plasma facing armor on both FW and DP, carbon based materials will be used at least during the initial physics phase due to their good performance in current tokamaks in respect to impurity control and disruption resistance. For the FW structure in water cooled austenitic steel, with radiation cooled armor adequate thermo-mechanicai performance is predicted allowing peak heat fluxes of up to 0.8 MW/m 2 at 2)< 104 long duration burn pulses. For divertor concepts with the armor attached by brazing to a water cooled heatsink, the peak heat flux is about 10 MW/m 2. However, the main critical issue for the DP is the lifetime which is critically limited by erosion. The demonstration of the basic feasibifity of FW and DP design is in progress via manufacture and thermo-mechanical testing of prototypical mock-ups.
1. lntrodu~ion The objective of N E T (Next European Toms) is to demonstrate fusion energy production in an apparatus which meets the basic requirements of a reactor including [1,2]: extended D - T plasma burn pulses to be established in the initial 'Physics' Phase, qualification and testing of reactor-like components mainly during the second 'Technology' Phase. Hence N E T has to be designed with great flexibility with respect to the accessible plasma parameters, which can currently only be predicted with significant uncertainty. This is in particular important for the plasma facing components (PFC) whose operating conditions will first be more dearly established in the physics phase, so that an optimisation of these components will only be possible during and after the physics phase. Such improvements of the PFC's is essential for the operation in the technology phase, which could be crucially limited by the performance of these PFC's. -
-
* A version of this paper was presented as an invited paper at the 15th Symposium on Fusion Teclmolog3,,held in Utrecht, The Netherlands, September 19-23, 1988.
Fig. 1. Integration of the blanket (incl. first wall and divertor) in NET with a 15 MA plasma.
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G. Vieider et a L / NETplasma facing components
Fig. 1 shows a cross section of N E T indicating the main plasma facing components: - the first wall (FW) around about 85% of the plasma surface, and - the divertor plates (DP) at top and bottom of the plasma chamber required for the exhaust of particles and power from the plasma edge. By about 1993 the scientific basis for a decision on the construction of N E T will be available from the present tokamaks. For crucial components such as the initial PFC's, design solutions and technologies must therefore be developed with a high confidence in their feasibility by this date. Consequently, in the current R & D programme as well as in this paper emphasis is given to these initial PFC's for the physics phase. The P F C development is co-ordinated by the N E T Team and carried out in collaboration with European industries and research laboratories.
2. Assumed
operating
conditions
for
PFC
- the physics phase with emphasis on ignition studies on plasmas with typically 15 M A current over a total burn time of about 300 h, - the technology phase with long pulses driven inductively or non-inductively at plasma currents down to 11 M A over a total burn time of at least 4000 h. In addition to the required flexibility for different plasma parameters, the current uncertainty concerning the plasma edge physics and p l a s m a - w a l l interaction contribute t o a wide range of possible P F C operating parameters, such as: - For nominal operation, the prediction of the peak surface heat flux at the F W is so far rather uncertain mainly due to neutral beam shine through and because of lack of information on localized fusion alpha particle losses induced by the toroidal field ripple. - Plasma edge modelling gives nominal divertor peak heat fluxes of 3 to 7 M W / m 2 for ideally symmetric double null plasma configurations. For the design, these heat fluxes should be increased by factors of 1.5-2 to allow for uncertainties in the modelling, for single null operation and for geometrical misalignment.
design
Table 1 summarizes the current assumed operating conditions for the P F C design for [1,2,3]:
Table 1 Main operating requirements for plasma facing components of ITER Operation phase components 1. Nominal operation Aver. neutron wall load (MW/m 2) Heat flux - average (MW/m 2) - peak (MW/m 2) Total number of load pulses (10 4) Average pulse duration (s) Typical 'off burn' time (s) Average neutron fluence (MW y r / m 2) Peak particle - flux c (102°/m 2 s) - energy ¢ (eV) 2. Disruptions Total number at full load Thermal quench - time (ms) - peak energy deposition (MJ/m 2) Current quench - time (ms) - radiative energy depos. ( i J / m 2)
Technology
Physics First wall
Divertor plates
First wall
Divertor plates
0.7 0.1 1
0.4 1.5
0.7 0.2
0.4 1.5
10 a
1
1 102 103
>103 >102 0.01 5.103 70
0.02 5 10-200
5.102 500
0.3-0.5 5 10-200
100
0.1-3 5
1
0.5
0.5
5-50 0.5
0.15-0.25 5xlO ~ 5xlO 2 70 700 10
0.1-3 1
10 a 1
5 5-50 0.5
Notes a For symmetric double-null plasma (under L-mode assumptions) on outboard divertor plates with 20 o inclination to the separatrix. b In addition, highly localized energy deposition of up to 500 M J / m 2 due to run-away electrons could occur. c For the divertor the range of operating conditions is indicated.
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G. Vieider et al. / N E T plasma facing components
- The energy of the particles impinging on the divertor can only be predicted with major uncertainty, but in most cases this energy is too high for high Z materials with a low threshold energy for sputtering. - Present day tokamaks experience a high number of disruptions, but for a machine like NET with very high current, a method of controlling the discharge must be developed so that disruptions can be considered as a fault condition. About 100 hard disruptions are assumed for the full life of the machine.
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3.1. Plasma facing F W armor [3]
Based on the experience with present large tokamaks, it is now well established that at least some local armor in refractory low-Z material is required on metal F W surfaces for the following reasons [4,5]: - to cope with the energy deposition by the plasma and run-away electron jets during start-up and disruptions to prevent excessive heat loads and melting, - to improve impurity control, mainly during start-up, to protect against neutral beam shine-through. The assumed disruption energy deposition for NET would cause melting of an unprotected steel F W in a depth of about 0.1 mm [6,7]. This must be avoided because: - repeated melting by disruptions produces cracks and changes in the structure which have been shown to cause a reduction of up to 90~ of the fatigue life [7,8]; a crack propagation analysis indicates similar results [ 9 ] ; damage by run-away electrons has shown that part of the melt is removed by splashing so that hundred disruptions could lead to a loss of material of several millimeters. Even if the F W surface temperatures remain below the melting point, there will be significant additional thermal fatigue of unprotected FW-structures due to disruptions. This is illustrated in fig. 2 which shows the influence of the disruption conditions on the 'disruption fatigue life', i.e. the number of such disruptions allowed according to the design codes without normal nominal fatigue cycles and with different temperature assumptions for the steel fatigue data: - if e.g. 100 disruptions should not reduce the nominal fatigue life according to design codes by more than 10-50~, then the disruption fatigue life should be 1000-200 (the 'target' on fig. 2 is placed typically in. the middle of this range), and the temperature spike due to disruptions should be less than 200-400 o C,
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it follows that the steel F W needs to be protected where more than a few percent of the assumed peak disruption energy could be deposited. In conclusion, because of the highly uncertain disruption conditions and the already severe thermal fatigue fife limitation of the NET F W it is now assumed: - for the physics phase, to protect a major part of the steel F W by an armor, - for the technology phase, to reduce the F W armor coverage on the basis of the experience gained. For the protection armor a low-Z material will be required in order to [5]: - minimize the high-Z impurity in-flux into the plasma, spread the energy deposition from run-away electrons over a larger volume this minimizing local overheating and damage; considering the penetration depth of high energy electrons it is estimated that more than 1 cm thickness of such protection tiles is required. Fine grain graphite and carbon fiber composites (CFC) are the most attractive choices for N E T armor tiles due to: -
128
G. Vieideret al. / NET plasma facing components
- their unique high temperature capability, which for typical FW disruptions yields a material loss of 10-20 # m by sublimation above 3000 o C instead of melting more than 100 # m as, e.g., for beryllium at 1277°C, - their superior thermal shock resistance, which is estimated to permit nearly an order of magnitude higher disruption heat loads than for, e.g., beryllium or silicon carbide [10,11]. These considerations are supported by the good performance of graphite as FW material in present large tokamaks [4,5]. There are, however, also several critical issues associated with the use of carbon based materials as ~ armor: Irradiation damage is expected to limit the armor life due to swelling to a neutron fluence of 1-3 MW y r / m 2, this would, however, only be a concern for future power reactors [11]. Radiation enhanced sublimation sets an upper limit for normal operation of graphite at around 2000 ° C [10] with an estimated total erosion of about 20 mm/year. By erosion substantial quantities of dust are produced which put constraints on vacuum pumping and maintenance. There are indications that the retention of hydrogen and other impurities in graphite tiles could be very high at temperature below 1000 ° C - especially with irradiation damage - while rising the temperature seems to reduce the impurity content in the graphite. This instead may lead to outgassing problems, therefore baking of the armor at about 350 ° C is mandatory. Water a n d / o r air ingress into the plasma chamber could represent potential safety hazards which now are studied in more detail. For the initial extensive protection coverage, mechanically-attached tiles appear at present to be the most credible ~ngineering solution since: their design and performance is reasonably predictable on the basis of similar solutions in present tokamaks [4,5], they can be replaced in-situ by remote maintenance equipment, for bonded tiles, the bonding technology and in-situ replacement procedures for maintenance have not yet been developed, and the frequent removal of complete FW-scgments would require prohibitive maintenance times and waste volumes.
COOLANT
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Fig. 3. Out board 1/48-FW &. blanket segment for NET with poloidal cooling.
-
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3.2. Basic F W design [3] The development of basic FW concepts for the initial NET operation has the following main design objectives:
robust design considering the uncertainty in operating conditions - high reliability and passive safety - credibility in the near term with a minimum of critical issues requiring R & D effort beyond the next few years potential for continued operation into the technology phase. " In addition to the plasma side protection, the following general design features have been adopted for the basic FW: - The general architecture of the FW & blanket segments for vertical maintenance is shown in figs. 3 and 4. The FW panel is integrated into a box enclosing the breeding or shielding units because of the need for: • passive plasma stabilisation via a saddle loop conductor consisting of the FW panel (ca. 10 mm steel) and the box sidewalls (ca. 15 mm copper), • support for the FW capable of withstanding forces due to disruptions: this re.quires a boxwall thickness of more than 10 mm [12], • minimum outgassing and leakage of blanket materials into the plasma. - As structure material, solution annealed austenitic stainless steel type AISI 316L was selected considering the extensive database, the acceptable irradiation resis-
-
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tance, the absence of ferro-magnetic effe,cts, the wide temperature range capability and the ease of manufacture by welding or brazing. - A s coolant for the basic machine water is preferred over helium because of: • the better heat removal capability at low pressure and low pumping power, • the better shielding properties, • the potential for passive shut-down cooling by natural convection. Critical issues for water cooling include hazards associated with leaks into the plasma chamber and limitations of the baking temperature. The nominal F W coolant conditions are 6 0 - I 0 0 o C at less than 1 MPa in order to avoid significant primary stresses and thermal creep in the F W structure. - Double containment o f the coolant m a y be neces-
sary to achieve the permitted leakage rates which are mainly set by considerations of the initial pump down and oxygen contamination to about 10 -8 g / s total leakage into the plasma chamber. These leakage rates are several orders of magnitude lower than those achieved in nuclear heat exchangers. The double containment can be produced by brazing coolant tubes into
el~ffo~ ~ weld~ ( ~ W ) sidepro~fio~.
the outer F W structure. First experiments at KfK Karlsruhe indicate that the brazing joint can act as a crack stopper [13]. - Cooling via poloidal U-tubes, as indicated in figs. 3 and 4, has been selected for the F W panel. Its advantages over the alternative toroidal coolant tube arrangement are:
• 30 times fewer tube joints are required and the welds to collectors are located outside the plasma chamber, which is estimated to result in a coolant leakage frequency of less than 1~ of that for toroidal tubes • the absence of double curvature bending considerably eases the fabrication of the F W and blanket box, whilst toroidal tubes would require a double curved F W panel in order to place the coolant collectors at the backside of the blanket box. Fig. 4 shows the principal design options for the basic F W on which the present development effort by the NET Team is concentrated taking into account the results of industrial study contracts: The steel structure of the F W panel is manufactured via electron beam (EB) welding in areas with low stresses [14]. This process permits a minimum of heataffected zones and welding distortion. Cooling tubes are -
130
G. Vieider et al. / NET plasma facing components
brazed into this structure to achieve the required extremely low leakage. A similar design is also considered where the whole panel is manufactured via brazing [15]. These panel designs can be used for different protection concepts. - Conductioely cooled tiles can be attached via a conical graphite or CFC nut on bolts in poloidal grooves of the FW structure. These attachment grooves are designed so that the associated peak stresses are not raised above the values without tile attachment. A flexible graphite layer (e.g. Papyex) is proposed [22] to compensate for some differential deformation between tile and steel substrate. Such designs have recently been tested up to 250 cycles [16]. The major critical issue to be studied in detail is the variation of the contact pressure under all operating conditions and the associated likely redistribution of the heat flux to the steel FW structure. -Radiation cooled tiles can be attached via poloidal rails in similar grooves of the steel FW structure as for conductive tiles. The attachment rails and pins could be made of carbon fiber composites. Unlike for the conductive tiles, the heat flux to the steel FW surface will be uniform with radiation and this steel surface can be shaped to eliminate hot spots, so that peak temperature gradients and peak thermal stresses will be less than half of that of the conductive tile option. The major critical issue is the high tile temperature required by radiation heat transfer. Blackening coatings such as AI203 and TiO2 are therefore required on the steel surface. These coating also reduce tritium permeation through the FW structure. 3.3. Thermo-mechanical analysis and operating limits
For NET FW steel structures with disruption protection, modest irradiation damage and temperatures below the crrep range, it is expected that thermal fatigue by the nominal cyclic heat loads will be the major life limiting phenomenon. In addition, with radiation cooled graphite tiles, erosion mainly by sublimation is estimated to limit the permitted peak tile temperature to about 2000 and 1800 ° C in the physics and technology phase, respectively. The peak tile temperatures are evaluated assuming reradiation from hot radiative tiles to the surrounding cooler tiles with average heat loads. Fig. 5 shows an estimate of the allowed nominal peak FW heat flux limited by cyclic thermal fatigue of the austenitic steel structure on the following basis: - The French nuclear design code RCC-MR [17], which permits higher thermal stresses than ASME at low primary stresses below 105 cycles. It should be noted
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that the design codes include a contingency margin between allowed fatigue limits and failure of a factor 2 on strain a n d / o r a factor 20 on number of cycles. Numerous 2-D and first 3-D thermo-elastic analyses of the basic FW steel structure with different protection schemes, assuming quasi steady state conditions and uniform heat flux onto the FW structure [18]. More detailed studies are required mainly for the whole FW blanket box and for the conductive tiles to verify the present assumptions. For a FW life of typically 2 × 104 cycles at about 1 M W / m 2 neutron wall load, the allowed peak FW heat flux according to fig. 5 is estimated to about 0.8 and 0.2 M W / m 2 for the radiative and conductive armor tiles, respectively. The significantly higher values for the radiative tiles are mainly explained by: - reradiation to cooler FW surfaces, - b e t t e r possibility for optimization of the FW cross section. Comparing these results with the expected operating conditions in table 1 it is concluded, that the FW nominal heat flux requirements can most likely be satisfied at neutron wall loads up to 1 M W / m 2 by the proposed FW design with: -
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G. Vieider et al. / N E T plasma facing components
analysis a n d lifetime prediction u n d e r realistic elasto-plastic conditions. I n the near future m a n u f a c t u r i n g feasibility studies will also b e started for the different F W protection concepts a n d for the integration of the F W p a n e l in the b l a n k e t segments.
4. T h e divertor plates (DP) 4.1. P l a s m a f a c i n g D P a r m o r
T a b l e 2 summarizes the m a i n properties for candidate D P - a r m o r materials. As D P a r m o r in the physics p h a s e for N E T only c a r b o n based materials are b e i n g considered for the following m a i n reasons: - H i g h - Z impurities in the p l a s m a are to b e avoided. C-based materials have the best possible d i s r u p t i o n resistance as discussed in section 3.1. This is of
particular i m p o r t a n c e in the physics p h a s e with typically 100 m a j o r disruptions, which even for graphire are estimated to remove a b o u t 3 m m b y sublimation. - T h e sputtering erosion of C-based materials is less sensitive to the r a t h e r u n c e r t a i n particle energies t h a n the sputtering of tungsten. T h e total n o m i n a l erosion during u p to 300 h integrated b u r u - t i m e could b e of the same order as the erosion b y disruptions, if one operates in o p t i m u m conditions for low erosion. - These considerations are c o n f i r m e d b y the generally good experience with C-based materials o n divertors a n d limiters in all operating large tokamaks. A m o n g s t the c a r b o n - b a s e d materials c a r b o n fiber composites ( C F C ) are preferred c o m p a r e d to fine grain graphites m a i n l y because of (see table 2): - at least twice higher t h e r m a l conductivity, which permits accordingly thicker D P a r m o r a n d longer life time - better t h e r m a l shock resistance due to the good ther-
Table 2 Typical thermomechanical data for unirradiated first wall or divertor materials T ( ° C) Melt/sublim. temp. Tra ( ° C) Spec. heat C (J/gK) Density g (g/cm 3) Thermal conduct. h (W/mK) Thermal expans. coeff, a
Fine grain graphite
Carbon fibre composite (LCL A05)
Silicon carbide
Beryllium
TZM
Cu-alloy 0.25% Ai 203)
AISI 316
3500
3500
2700
1277
2620
1083
1370
20 1000
0.7 2.0
0.44 1.8
0.7 1.2
1.7 2.9
20
1.7
1.8
3.0
1.8
20 1000 20
75 40
270 100
80 a 40 a
4 to 5
11
8a
11
8 a
160 b 190 b
(10-6/K)
1000
Ultimate tens. stress o (106 N / m 2)
20 1000
30 60
50 60
Youngs modulus E (GPa)
20 1000
12 15
20 24
Figures of merit no melting
-
-
thermal shock
67 25 2.9 4.9
150
11 18.4
0.24 0.30 10.2 125 100 5.3 6.0
0.4
0.5 0.65
8.9
8.0
340 280 16.6
16.2 19.5
360 450
350
1150 700
450
525
430 370
310
300 220
131
192
20 1000
0.65 0.65
1 1
0.59 0.38
0.34
0.83 0.68
0.40
20 1000
0.16 0.25
1 1
0.10 0.06
0.07
0.38 0.23
0.22
Note: a ~ and a in two directions, b o in compression.
14.6 28.5
0.44
0.04
133
G. Vieideret al. / NET plasma facing components
mal conductivity, low thermal expansion and high strength. Most of the critical issues for C-based materials have already been discussed in context with the F W armor: mainly hydrogen and gaseous impurity retention and outgassing as well as potential safety hazards. As for the F W armor baking at 350°C will also be required for the divertor. In addition, chemical erosion by hydrogen and oxygen is a major concern [20]. For the technology phase of next step tokamaks until now, high-Z materials such as tungsten have been considered as the main candidate for a DP armor with adequate lifetime. The present trend to higher fusion powers, lower plasma densities and consequently to higher particle eneergies, however, raises severe doubts on the feasibility of a high-Z DP armor for most of the anticipated plasma scenarios. Hence, low-Z materials such as C-based materials or beryllium have to be considered also for the technology phase: - The main draw back of these low-Z materials are gross erosion rates without redeposition of up to several m / b u r n year - A s s u m i n g some divertor replacements during the technology phase, the net erosion has still to be considerably smaller than the estimated gross erosion. This may be achievable via: • redeposition, • sweeping of the null-point, • reduction of chemical erosion and radiation enhanced sublimation of carbon based materials e.g. by addition of Si or B, and DP surface temperature limitation to less than 1500 ° C. - Beryllium has no chemical erosion and would permit rapid in-situ repair of the divertor plates by plasma spray. However, the melting temperature and thermal shick resistance of Be is rather low. 4.2. DP design concepts
Fig. 8 illustrates a divertor plate integration on blanket segments together with a typical heat flux distribution. The simple fiat plate geometry facilitates manufacture and sweeping of the null point. The replacement of the DP is foreseen by remote maintenance equipment without removal of the blanket segments. The DP design is mainly dictated by the required cyclic peak heat fluxes in the order of 10 M W / m 2 otgether with the constraints by the carbon based armor material. Two basic design concepts are being studied via industrial contracts [21,22,23]. As for the first wall, water at 6 0 - 1 0 0 0 C and 1 MPa is the coolant. The
Fig. 8. Divertor plate with heat flux distribution and integration on inboard blanket segment.
concepts differ mainly in the attachment of the C-based armor on the heat sink: (0 Brazed armor [21] This concept is shown in fig. 9 with the graphite armor brazed to a heat sink in a molybdenum alloy TZM. - TZM has been chosen mainly because of a matching thermal expansion with graphite, which facilitates the brazing operation. Such brazing of graphite on TZM is performed industrially for the manufacture of rotating X-ray anodes of about 100 mm diameter operating at comparable temperatures for very high cycles. ~
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~STM~ES5STEELS~P~TPLATE Fig. ]0. DJvertor plate c o n @ t with c~bo~ composite armor ~]es m ~ h ~ c ~ y attach~ to a Cu-~]oy heat s ~ a ~ r ~ n g Novafime ~ U r ~ - P ~ n e y .
- Draw backs of TZM compared to copper alloys as the the main alternative are a lower thermal conductivity and low fracture toughness especially under irradiation, however, for the physics phase, irradiation damage should be low. - The heat sink is manufactured by joining TZM half shells with coolent tubes in M o - R e alloys via a HIP (Hot Isostatic Pressing) process in the same manufacturing step. Thus a kind of double containment of the coolant is obtained. - Mo-41 ~ Re is compared to TZM more ductile, has a much lower ductile-brittle transition temperature, is better weldable but also more expensive. - the use of TZM backplates on the C-based armor was shown to permit an in-situ tile replacement by de- and rebrazing. Such intermediate refractory metal layers would also facilitate the use of copper alloys as heatsink. 00 Mechanical armor attachment [22] This concept is shown in fig. 10 with CFC-tiles pressed against the heat sink in a copper-alloy via blade springs and a locking bar: this solution would permit an in-situ tile replacement, the feasibility of the concepts is critically depending on an efficient heat transfer between armor and heatsink of at least 1 W / c m 2 K, which under ideal conditions can be achieved by the use of 0.5 mm thin foils of exfoliated graphite (Papyex).
- 1600 ° C for 10 mm CFC assumed to be brazed to TZM, - 2000 ° C for 6 mm CFC mechanically attached via papyex on copper alloy assuming a high contact heat transfer. It should be noted, that radiation enhanced sublimation of C-based materials sets a temperature limit at about 1500°C when sublimation exceeds the peak chemical erosion. It seems therefore, that 10 M W / m 2 static peak DP heat flux represents an upper limit for carbon based armor with 5-10 thickness. The thermo-mechanical analysis also shows that the thermal stresses in the TZM heat sink (without brezing stresses) appear to be acceptable for 104 cycles at 10 M W / m 2. However, the lack of material data and design criteria should be noted. For the assumed low pressure water coolant in simple 12-14 mm diameter tubes, burn-out is predicted at critical heat fluxes of 10-12 M W / m 2. The critical heat fluxes can, however, be increased by more advanced coolant schemes such as vortex promoters.
-
-
Thermo-mechanical analyses were performed for both concepts. Fig. 11 shows that at-a DP heat flux of 10 M W / m 2 the peak carbon armor temperature without irradiation damage would be about [24]:
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G. Vieider et al. / NETplasma facing components
135
The present results indicate that a static divertor heat load of 10 M W / m 2 is already an ambitious development target. For yet higher divertor heat fluxes of 15-20 M W / m 2 probably required in the technology phase, sweeping of the separatrix is now being studied as a promising method to obtain armor temperatures, heat sink fatigue and coolant conditions at acceptable values, i.e. not worse than for static heat loads of 10 M W / m 2. 4.3. Manufacturing and thermo-mechanical testing In the frame of NET study contracts the manufacturing feasibility of various cross-sections and subassenblies is to be demonstrated by producing relevant samples. For the brazed lay-out, some samples are shown in fig. 12. Additionally, gundrilling of TZM bars over a length of more than 2 m was accomplished for holes of 18 mm diameter. The weldability of similar and dissimilar joints of various tube materials was tested. A similar concept based on graphite tiles brazed to thin-walled
Fig. 13. Test of mechanically attached carbon based tiles in the neutral beam test bed at JET.
Fig. 12. Divertor test specimen manufactured by Plansee via brazing and hot isostatic pressing.
molybdenum tubes has been manufactured for Asdex Upgrade [25] and extensively tested at the SANDIA electron beam facility [26]. Extensive bonding trials between different graphites and heat sink materials, copper or molybdenum alloys, revealed that, on thermal cycling, sensible larger scale graphite/copper samples were subject to excessive thermal mismatch strains [23]. A broader and more systematic investigation has been started at the Kernforschungszentrum Karlsruhe in the frame of the NET supporting technology programme. The thermomechanical behaviour of different materials and brazed material combinations shall be studied under thermal cycling conditions, produced by a plasma flame, which can deposit peak heat fluxes up to 15 M W / m 2. During 1987 a not yet optimised test piece of 500 × 100 mm2 was manufactured with 10 graphite and carbon composite tiles mechanically attached on a water cooled copper heatsink - see fig. 13 [16,22]. With this test piece a series of tests has been performed recently at the JET
136
G. Vieider et al. / N E T plasma facing components
neutral beam injectors test bed at short pulsed heat fluxes up to 10 M W / m 2 and up to 250 cycles. The main objective was to study the influence of the intermediate soft 'Papyex' layer (between heatsink and tile) on the thermal bchaviour of the tiles. From these tests it is concluded that: - Both tiles with brazed back plates in Mo-alloy failed by deformation and debrazing, while all the other tiles showed no damage. The use of Papyex improves the heat transfer between tile and heat sink by an order of magnitude, however, this improvement was considerably less than expected. In late 1988, other medium sized mock-ups will become available as well for the mechanically attached concept (320 × 110 mm 2) as for the brazed concept (650 × 200 mm2). Thermo-mechanical testing of these mock-ups is planned at: ion- and dectron beam facilities at SANDIA National Laboratories, Albuquerque, USA, a 1 MW solar furnace at the CNRS center Odeillo, France, capable of delivering an almost in time unlimited peak heat flux of 10 M W / m ~.
5.
Conclusions
The studies on plasma facing components by the NET Team in collaboration with European industries and laboratories have led to a consolidation of the initially rather uncertain situation with the following main conclusions: (i) Carbon based material will be used as the plasma facing armor on both divertor plates and major areas of the first wall at least during the initial physics phase. The main reasons for this choice are the generally good experience with these materials in current large tokamaks concerning plasma impurity control and disruption resistance. (ii) The first wall structure in austenitic stainless steel will be cooled by double contained low pressure water for achieving a reliable 'zero leakage' solution. Several design concepts for the structure and the mechanically attached tiles are being developed, for which analyses predict adequate thermo-mechanical performance, i.e. allowed nominal heat fluxes at 2 × 104 cycles of up to: 0.8 M W / m 2 with radiation cooled tiles at about 1800°C, 0.2 M W / m 2 with conduction cooled tiles at less than 1000 ° C. The demonstration of the feasibility and performance of these designs is in progress via:
the manufacture of prototypical test sections, which is nearly completed, the thermo-mechanical testing of these FW-sections until 1989/90. (iii) Several concepts of dioertor plates are being developed with mechanical attachment or brazing of the armor to a water cooled heat sink in copper- or molybdenum alloys. First thermo-mechanical analyses indicate that nominal cyclic heat fluxes of up to 10 M W / m 2 may be allowed. This appears adequate for the physics phase and would probably require 'sweeping' of the separatrix in the technology phase. The manufacture of medium-size mock-ups and their thermo-mechanical testing are in progress. The main critical issue for the divertor plates is the limited life time, mainly due to erosion and subfimation in the technology phase (assuming only a few disruptions and some DP replacements). The nominal erosion life time would only be adequate: • for carbon armor, if the net erosion would be considerably less than the predicted sputtering erosion, which might be obtained by redeposition, separatix sweeping and improved materials, • for tungsten armor, if the plasma edge temperatures would not exceed 20 eV, which seems difficult to achieve in most of the conceived plasma scenarios especially with non-inductive current drive. Summarizing, there appears to be reasonable prospects to develop plasma facing components with adequate performance for the physics phase of NET, provided that the disruption frequency would be significantly lower than in present tokamaks. In the technology phase, the major crucial issue is the risk for too short divertor life times due to nominal erosion. -
Acknowledgements
The authors would like to thank all colleagues who are participating in the studies and development of plasma facing components for NET, in particular J.L. Boutard, M. Budd, M. Chazalon, A. Cardella, F. Engelmann, B. Libin, J. Raeder, R. Toschi and C.H. Wu from the NET Team as well as F. Farfaletti-Casali and R. Matera from JRC ISPRA.
References
[1] R. Toschi et al., Next European Torus (NET) Fusion Technology (July 1988).
G. Vieider et al. / N E T plasma facing components [2] R. Toschi, NET and the European fusion technology programme, 15th SOFT Utrecht, September 1988, Fusion Engrg. Des. 11 (1989) 47-62, in this issue. [3] G. Vieider et al., Progress in the development of a first wall for NET, Int. Symposium on Fusion Nuclear Technology, Tokyo, April 1988, Fusion Engrg. Des. 8-10 (1989). [4] D.K. Owens and M.A. Ulrichson, Plasma-wall interaction: Recent TFTR results and implications on design and construction of lindters, 14th SOFT, Avignon, September 1986. [5] M.A. Pick, G. Celentano, E. Deksnis, K.J. Dietz, R. Shaw, K. Sonnanberg and M. Walravens, Experience with graphite in JET, 12th SOFE, Monterey, October 1987. [6] H.Th. Klippel, J.G.v.d. Laan, Thermo-mechanical behaviour of bare and coated first walls under plasma disruption conditions, ISFNT Tokyo, April 1988, Fusion Engrg. Des. 8-10 (1989). [7] F. Brossa, M. Cambini, D. Quataert, G. Rigon and P. Schiller, Physical and metallurgical phenomena during simulations of plasma disruptions, 3rd Int. Conference on Fusion Reactor Materials, Karlsruhe, October 1987. [8] B. van der Schaaf, Radiation damage of stainless steel, Fusion Technology Progress Report ECN-185, ECN Petten (May 1986). [9] E. Diesele, T. Fett, D. Munz and H. Stamm, Stress and lifetime calculations for first wall and blanket structural components, KfK-Report 4210, Karlsruhe (March 1987). [10] H. Brinkschulte, E. Deksnis and A.S. Bransden, Heatshock resistance of graphite determined with a CO 2 laser, JET Report 1'(87) 43. [11] M. Birch and J.E. Brocklehurst, A review of the behaviour of graphite under the conditions for protection of the first wall of a fusion reactor, UKAEA Report ND-R-1434 (S), Springfield (Dec. 1987). [12] Y.R. Crutzen and F. Farfaletti-Casali, Structural design problems related to plasma disruption events of the removable NET first wall segments, Int. Symposium on Fusion Nuclear Technology, Tokyo, April 1988, Fusion Engrg. Des. 8-10 (1989). [13] D. Munz et al., Evaluation of crack growth delay in multilayer sheets, NET Study Contract 253/86-11/FUD, KfK Karlsruhe (July 1988).
137
[14] (3. Vieider et al., First wall concepts with welded steel structures for NET, 14th SOFT, Avignon, September 1986. [15] V. Zampaglione et al., First wail for NET: Engineering studies and mock-up design, technological development and manufacturing, Proc. of the 15th SOFT, Utrecht, September 1988 (Elsevier Sc. Pub., Amsterdam, 1989). [16] A. Cardella et al., Testing of divertor samples for NET at the JET N.B.I. test bed, Proc. of the 15th SOFT, Utrecht, September 1988 (Elsevier Sc. Pub., Amsterdam, 1989). [17] Association Franqaise pour les R~les de Construction des Chaodi~res Electro-nucl~ires, Design and Construction Rules for FBR Nuclear Islands, Paris (June 1985). [18] E. Zolti, Comparative thermo-mechanical study of first wall protection concepts with conductively cooled tiles, NET Internal Note NET/IN/88-01. [19] R. Matera, V. Renda, G.L. Morlotti, G. Graglia, First wall design criteria, thermal fatigue and creep experiments and theoretical modelling, 14th SOFT, Avignon, September 1986. [20] C.H. Wu, J.W. Davis, A.A. Hassz, The formation of methane by the interactions of low energy hydrogen ions with graphite, Proc. of 15th EPS Conference on Controlled Fusion and Plasma Heating, Dubrovnik, May 1988. [21] G. Kneringer et al., The design and feasibility of NET divertor target plates; Final Report on Contract NET/87724A, Phase I, Metallwerk Piansee (January 1988). [22] M. Besson et al., Divertor plates with replaceable conductively cooled carbon tiles, Proc. of the 15th SOFT, Utrecht, September 1988 (Elsevier Sc. Pub., Amsterdam, 1989). [23] R. Howard et al., The design and feasibility of NET divertor target plates; Final Report on Contract NET/87724B, Phase I, GEC Energy Systems Ltd. (1988) R(88) 18. [24] E. Zolti, Thermomechanical design evaluation and material properties requirements for NET divertor elements, ICFRM-3 Conference, Karlsruhe, Oct. 1987, to be published in J. Nucl. Mater. [25] H.E. Kotzlowski, Actively cooled graphite limiter, 13th SOFT, Varese, September 1984. [26] J. Bohdansky et al., Behaviour of graphite under heat load and in contact with a hydrogen plasma, Nucl. Instr. and Meths. in Phys. Res. B23 (1987) 527-537.
NET PLASMA FACING COMPONENTS
125
*
G. V I E I D E R , M. H A R R I S O N a n d F. M O O N S
The NET Team, c/o Max-Planck-lnstitut ffw Plasmaphysik, Boltzmannstr. 2, D-8046 Garchin~ Fed. Rep. Germany Submitted 6 October 1988; accepted 14 December 1988 Handling Editor: P. Komarek
The progress in the design and development of the first wall (FW) and divertor plates (DP) for the Next European Torus (NET) are summarized, highlighting the assumed main operating conditions, material choices, design options and their analysis as well as associated manufacturing studies and the ongoing testing programme. As plasma facing armor on both FW and DP, carbon based materials will be used at least during the initial physics phase due to their good performance in current tokamaks in respect to impurity control and disruption resistance. For the FW structure in water cooled austenitic steel, with radiation cooled armor adequate thermo-mechanicai performance is predicted allowing peak heat fluxes of up to 0.8 MW/m 2 at 2)< 104 long duration burn pulses. For divertor concepts with the armor attached by brazing to a water cooled heatsink, the peak heat flux is about 10 MW/m 2. However, the main critical issue for the DP is the lifetime which is critically limited by erosion. The demonstration of the basic feasibifity of FW and DP design is in progress via manufacture and thermo-mechanical testing of prototypical mock-ups.
1. lntrodu~ion The objective of N E T (Next European Toms) is to demonstrate fusion energy production in an apparatus which meets the basic requirements of a reactor including [1,2]: extended D - T plasma burn pulses to be established in the initial 'Physics' Phase, qualification and testing of reactor-like components mainly during the second 'Technology' Phase. Hence N E T has to be designed with great flexibility with respect to the accessible plasma parameters, which can currently only be predicted with significant uncertainty. This is in particular important for the plasma facing components (PFC) whose operating conditions will first be more dearly established in the physics phase, so that an optimisation of these components will only be possible during and after the physics phase. Such improvements of the PFC's is essential for the operation in the technology phase, which could be crucially limited by the performance of these PFC's. -
-
* A version of this paper was presented as an invited paper at the 15th Symposium on Fusion Teclmolog3,,held in Utrecht, The Netherlands, September 19-23, 1988.
Fig. 1. Integration of the blanket (incl. first wall and divertor) in NET with a 15 MA plasma.
0 9 2 0 - 3 7 9 6 / 8 9 / $ 0 3 . 5 0 © Elsevier Science Publishers B.V.
126
G. Vieider et a L / NETplasma facing components
Fig. 1 shows a cross section of N E T indicating the main plasma facing components: - the first wall (FW) around about 85% of the plasma surface, and - the divertor plates (DP) at top and bottom of the plasma chamber required for the exhaust of particles and power from the plasma edge. By about 1993 the scientific basis for a decision on the construction of N E T will be available from the present tokamaks. For crucial components such as the initial PFC's, design solutions and technologies must therefore be developed with a high confidence in their feasibility by this date. Consequently, in the current R & D programme as well as in this paper emphasis is given to these initial PFC's for the physics phase. The P F C development is co-ordinated by the N E T Team and carried out in collaboration with European industries and research laboratories.
2. Assumed
operating
conditions
for
PFC
- the physics phase with emphasis on ignition studies on plasmas with typically 15 M A current over a total burn time of about 300 h, - the technology phase with long pulses driven inductively or non-inductively at plasma currents down to 11 M A over a total burn time of at least 4000 h. In addition to the required flexibility for different plasma parameters, the current uncertainty concerning the plasma edge physics and p l a s m a - w a l l interaction contribute t o a wide range of possible P F C operating parameters, such as: - For nominal operation, the prediction of the peak surface heat flux at the F W is so far rather uncertain mainly due to neutral beam shine through and because of lack of information on localized fusion alpha particle losses induced by the toroidal field ripple. - Plasma edge modelling gives nominal divertor peak heat fluxes of 3 to 7 M W / m 2 for ideally symmetric double null plasma configurations. For the design, these heat fluxes should be increased by factors of 1.5-2 to allow for uncertainties in the modelling, for single null operation and for geometrical misalignment.
design
Table 1 summarizes the current assumed operating conditions for the P F C design for [1,2,3]:
Table 1 Main operating requirements for plasma facing components of ITER Operation phase components 1. Nominal operation Aver. neutron wall load (MW/m 2) Heat flux - average (MW/m 2) - peak (MW/m 2) Total number of load pulses (10 4) Average pulse duration (s) Typical 'off burn' time (s) Average neutron fluence (MW y r / m 2) Peak particle - flux c (102°/m 2 s) - energy ¢ (eV) 2. Disruptions Total number at full load Thermal quench - time (ms) - peak energy deposition (MJ/m 2) Current quench - time (ms) - radiative energy depos. ( i J / m 2)
Technology
Physics First wall
Divertor plates
First wall
Divertor plates
0.7 0.1 1
0.4 1.5
0.7 0.2
0.4 1.5
10 a
1
1 102 103
>103 >102 0.01 5.103 70
0.02 5 10-200
5.102 500
0.3-0.5 5 10-200
100
0.1-3 5
1
0.5
0.5
5-50 0.5
0.15-0.25 5xlO ~ 5xlO 2 70 700 10
0.1-3 1
10 a 1
5 5-50 0.5
Notes a For symmetric double-null plasma (under L-mode assumptions) on outboard divertor plates with 20 o inclination to the separatrix. b In addition, highly localized energy deposition of up to 500 M J / m 2 due to run-away electrons could occur. c For the divertor the range of operating conditions is indicated.
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G. Vieider et al. / N E T plasma facing components
- The energy of the particles impinging on the divertor can only be predicted with major uncertainty, but in most cases this energy is too high for high Z materials with a low threshold energy for sputtering. - Present day tokamaks experience a high number of disruptions, but for a machine like NET with very high current, a method of controlling the discharge must be developed so that disruptions can be considered as a fault condition. About 100 hard disruptions are assumed for the full life of the machine.
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3.1. Plasma facing F W armor [3]
Based on the experience with present large tokamaks, it is now well established that at least some local armor in refractory low-Z material is required on metal F W surfaces for the following reasons [4,5]: - to cope with the energy deposition by the plasma and run-away electron jets during start-up and disruptions to prevent excessive heat loads and melting, - to improve impurity control, mainly during start-up, to protect against neutral beam shine-through. The assumed disruption energy deposition for NET would cause melting of an unprotected steel F W in a depth of about 0.1 mm [6,7]. This must be avoided because: - repeated melting by disruptions produces cracks and changes in the structure which have been shown to cause a reduction of up to 90~ of the fatigue life [7,8]; a crack propagation analysis indicates similar results [ 9 ] ; damage by run-away electrons has shown that part of the melt is removed by splashing so that hundred disruptions could lead to a loss of material of several millimeters. Even if the F W surface temperatures remain below the melting point, there will be significant additional thermal fatigue of unprotected FW-structures due to disruptions. This is illustrated in fig. 2 which shows the influence of the disruption conditions on the 'disruption fatigue life', i.e. the number of such disruptions allowed according to the design codes without normal nominal fatigue cycles and with different temperature assumptions for the steel fatigue data: - if e.g. 100 disruptions should not reduce the nominal fatigue life according to design codes by more than 10-50~, then the disruption fatigue life should be 1000-200 (the 'target' on fig. 2 is placed typically in. the middle of this range), and the temperature spike due to disruptions should be less than 200-400 o C,
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it follows that the steel F W needs to be protected where more than a few percent of the assumed peak disruption energy could be deposited. In conclusion, because of the highly uncertain disruption conditions and the already severe thermal fatigue fife limitation of the NET F W it is now assumed: - for the physics phase, to protect a major part of the steel F W by an armor, - for the technology phase, to reduce the F W armor coverage on the basis of the experience gained. For the protection armor a low-Z material will be required in order to [5]: - minimize the high-Z impurity in-flux into the plasma, spread the energy deposition from run-away electrons over a larger volume this minimizing local overheating and damage; considering the penetration depth of high energy electrons it is estimated that more than 1 cm thickness of such protection tiles is required. Fine grain graphite and carbon fiber composites (CFC) are the most attractive choices for N E T armor tiles due to: -
128
G. Vieideret al. / NET plasma facing components
- their unique high temperature capability, which for typical FW disruptions yields a material loss of 10-20 # m by sublimation above 3000 o C instead of melting more than 100 # m as, e.g., for beryllium at 1277°C, - their superior thermal shock resistance, which is estimated to permit nearly an order of magnitude higher disruption heat loads than for, e.g., beryllium or silicon carbide [10,11]. These considerations are supported by the good performance of graphite as FW material in present large tokamaks [4,5]. There are, however, also several critical issues associated with the use of carbon based materials as ~ armor: Irradiation damage is expected to limit the armor life due to swelling to a neutron fluence of 1-3 MW y r / m 2, this would, however, only be a concern for future power reactors [11]. Radiation enhanced sublimation sets an upper limit for normal operation of graphite at around 2000 ° C [10] with an estimated total erosion of about 20 mm/year. By erosion substantial quantities of dust are produced which put constraints on vacuum pumping and maintenance. There are indications that the retention of hydrogen and other impurities in graphite tiles could be very high at temperature below 1000 ° C - especially with irradiation damage - while rising the temperature seems to reduce the impurity content in the graphite. This instead may lead to outgassing problems, therefore baking of the armor at about 350 ° C is mandatory. Water a n d / o r air ingress into the plasma chamber could represent potential safety hazards which now are studied in more detail. For the initial extensive protection coverage, mechanically-attached tiles appear at present to be the most credible ~ngineering solution since: their design and performance is reasonably predictable on the basis of similar solutions in present tokamaks [4,5], they can be replaced in-situ by remote maintenance equipment, for bonded tiles, the bonding technology and in-situ replacement procedures for maintenance have not yet been developed, and the frequent removal of complete FW-scgments would require prohibitive maintenance times and waste volumes.
COOLANT
S!DE W~ALL
-
Cu-PLATE F01 ~TABILISATI~
-
Fig. 3. Out board 1/48-FW &. blanket segment for NET with poloidal cooling.
-
-
-
-
-
3.2. Basic F W design [3] The development of basic FW concepts for the initial NET operation has the following main design objectives:
robust design considering the uncertainty in operating conditions - high reliability and passive safety - credibility in the near term with a minimum of critical issues requiring R & D effort beyond the next few years potential for continued operation into the technology phase. " In addition to the plasma side protection, the following general design features have been adopted for the basic FW: - The general architecture of the FW & blanket segments for vertical maintenance is shown in figs. 3 and 4. The FW panel is integrated into a box enclosing the breeding or shielding units because of the need for: • passive plasma stabilisation via a saddle loop conductor consisting of the FW panel (ca. 10 mm steel) and the box sidewalls (ca. 15 mm copper), • support for the FW capable of withstanding forces due to disruptions: this re.quires a boxwall thickness of more than 10 mm [12], • minimum outgassing and leakage of blanket materials into the plasma. - As structure material, solution annealed austenitic stainless steel type AISI 316L was selected considering the extensive database, the acceptable irradiation resis-
-
129
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tance, the absence of ferro-magnetic effe,cts, the wide temperature range capability and the ease of manufacture by welding or brazing. - A s coolant for the basic machine water is preferred over helium because of: • the better heat removal capability at low pressure and low pumping power, • the better shielding properties, • the potential for passive shut-down cooling by natural convection. Critical issues for water cooling include hazards associated with leaks into the plasma chamber and limitations of the baking temperature. The nominal F W coolant conditions are 6 0 - I 0 0 o C at less than 1 MPa in order to avoid significant primary stresses and thermal creep in the F W structure. - Double containment o f the coolant m a y be neces-
sary to achieve the permitted leakage rates which are mainly set by considerations of the initial pump down and oxygen contamination to about 10 -8 g / s total leakage into the plasma chamber. These leakage rates are several orders of magnitude lower than those achieved in nuclear heat exchangers. The double containment can be produced by brazing coolant tubes into
el~ffo~ ~ weld~ ( ~ W ) sidepro~fio~.
the outer F W structure. First experiments at KfK Karlsruhe indicate that the brazing joint can act as a crack stopper [13]. - Cooling via poloidal U-tubes, as indicated in figs. 3 and 4, has been selected for the F W panel. Its advantages over the alternative toroidal coolant tube arrangement are:
• 30 times fewer tube joints are required and the welds to collectors are located outside the plasma chamber, which is estimated to result in a coolant leakage frequency of less than 1~ of that for toroidal tubes • the absence of double curvature bending considerably eases the fabrication of the F W and blanket box, whilst toroidal tubes would require a double curved F W panel in order to place the coolant collectors at the backside of the blanket box. Fig. 4 shows the principal design options for the basic F W on which the present development effort by the NET Team is concentrated taking into account the results of industrial study contracts: The steel structure of the F W panel is manufactured via electron beam (EB) welding in areas with low stresses [14]. This process permits a minimum of heataffected zones and welding distortion. Cooling tubes are -
130
G. Vieider et al. / NET plasma facing components
brazed into this structure to achieve the required extremely low leakage. A similar design is also considered where the whole panel is manufactured via brazing [15]. These panel designs can be used for different protection concepts. - Conductioely cooled tiles can be attached via a conical graphite or CFC nut on bolts in poloidal grooves of the FW structure. These attachment grooves are designed so that the associated peak stresses are not raised above the values without tile attachment. A flexible graphite layer (e.g. Papyex) is proposed [22] to compensate for some differential deformation between tile and steel substrate. Such designs have recently been tested up to 250 cycles [16]. The major critical issue to be studied in detail is the variation of the contact pressure under all operating conditions and the associated likely redistribution of the heat flux to the steel FW structure. -Radiation cooled tiles can be attached via poloidal rails in similar grooves of the steel FW structure as for conductive tiles. The attachment rails and pins could be made of carbon fiber composites. Unlike for the conductive tiles, the heat flux to the steel FW surface will be uniform with radiation and this steel surface can be shaped to eliminate hot spots, so that peak temperature gradients and peak thermal stresses will be less than half of that of the conductive tile option. The major critical issue is the high tile temperature required by radiation heat transfer. Blackening coatings such as AI203 and TiO2 are therefore required on the steel surface. These coating also reduce tritium permeation through the FW structure. 3.3. Thermo-mechanical analysis and operating limits
For NET FW steel structures with disruption protection, modest irradiation damage and temperatures below the crrep range, it is expected that thermal fatigue by the nominal cyclic heat loads will be the major life limiting phenomenon. In addition, with radiation cooled graphite tiles, erosion mainly by sublimation is estimated to limit the permitted peak tile temperature to about 2000 and 1800 ° C in the physics and technology phase, respectively. The peak tile temperatures are evaluated assuming reradiation from hot radiative tiles to the surrounding cooler tiles with average heat loads. Fig. 5 shows an estimate of the allowed nominal peak FW heat flux limited by cyclic thermal fatigue of the austenitic steel structure on the following basis: - The French nuclear design code RCC-MR [17], which permits higher thermal stresses than ASME at low primary stresses below 105 cycles. It should be noted
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that the design codes include a contingency margin between allowed fatigue limits and failure of a factor 2 on strain a n d / o r a factor 20 on number of cycles. Numerous 2-D and first 3-D thermo-elastic analyses of the basic FW steel structure with different protection schemes, assuming quasi steady state conditions and uniform heat flux onto the FW structure [18]. More detailed studies are required mainly for the whole FW blanket box and for the conductive tiles to verify the present assumptions. For a FW life of typically 2 × 104 cycles at about 1 M W / m 2 neutron wall load, the allowed peak FW heat flux according to fig. 5 is estimated to about 0.8 and 0.2 M W / m 2 for the radiative and conductive armor tiles, respectively. The significantly higher values for the radiative tiles are mainly explained by: - reradiation to cooler FW surfaces, - b e t t e r possibility for optimization of the FW cross section. Comparing these results with the expected operating conditions in table 1 it is concluded, that the FW nominal heat flux requirements can most likely be satisfied at neutron wall loads up to 1 M W / m 2 by the proposed FW design with: -
132
G. Vieider et al. / N E T plasma facing components
analysis a n d lifetime prediction u n d e r realistic elasto-plastic conditions. I n the near future m a n u f a c t u r i n g feasibility studies will also b e started for the different F W protection concepts a n d for the integration of the F W p a n e l in the b l a n k e t segments.
4. T h e divertor plates (DP) 4.1. P l a s m a f a c i n g D P a r m o r
T a b l e 2 summarizes the m a i n properties for candidate D P - a r m o r materials. As D P a r m o r in the physics p h a s e for N E T only c a r b o n based materials are b e i n g considered for the following m a i n reasons: - H i g h - Z impurities in the p l a s m a are to b e avoided. C-based materials have the best possible d i s r u p t i o n resistance as discussed in section 3.1. This is of
particular i m p o r t a n c e in the physics p h a s e with typically 100 m a j o r disruptions, which even for graphire are estimated to remove a b o u t 3 m m b y sublimation. - T h e sputtering erosion of C-based materials is less sensitive to the r a t h e r u n c e r t a i n particle energies t h a n the sputtering of tungsten. T h e total n o m i n a l erosion during u p to 300 h integrated b u r u - t i m e could b e of the same order as the erosion b y disruptions, if one operates in o p t i m u m conditions for low erosion. - These considerations are c o n f i r m e d b y the generally good experience with C-based materials o n divertors a n d limiters in all operating large tokamaks. A m o n g s t the c a r b o n - b a s e d materials c a r b o n fiber composites ( C F C ) are preferred c o m p a r e d to fine grain graphites m a i n l y because of (see table 2): - at least twice higher t h e r m a l conductivity, which permits accordingly thicker D P a r m o r a n d longer life time - better t h e r m a l shock resistance due to the good ther-
Table 2 Typical thermomechanical data for unirradiated first wall or divertor materials T ( ° C) Melt/sublim. temp. Tra ( ° C) Spec. heat C (J/gK) Density g (g/cm 3) Thermal conduct. h (W/mK) Thermal expans. coeff, a
Fine grain graphite
Carbon fibre composite (LCL A05)
Silicon carbide
Beryllium
TZM
Cu-alloy 0.25% Ai 203)
AISI 316
3500
3500
2700
1277
2620
1083
1370
20 1000
0.7 2.0
0.44 1.8
0.7 1.2
1.7 2.9
20
1.7
1.8
3.0
1.8
20 1000 20
75 40
270 100
80 a 40 a
4 to 5
11
8a
11
8 a
160 b 190 b
(10-6/K)
1000
Ultimate tens. stress o (106 N / m 2)
20 1000
30 60
50 60
Youngs modulus E (GPa)
20 1000
12 15
20 24
Figures of merit no melting
-
-
thermal shock
67 25 2.9 4.9
150
11 18.4
0.24 0.30 10.2 125 100 5.3 6.0
0.4
0.5 0.65
8.9
8.0
340 280 16.6
16.2 19.5
360 450
350
1150 700
450
525
430 370
310
300 220
131
192
20 1000
0.65 0.65
1 1
0.59 0.38
0.34
0.83 0.68
0.40
20 1000
0.16 0.25
1 1
0.10 0.06
0.07
0.38 0.23
0.22
Note: a ~ and a in two directions, b o in compression.
14.6 28.5
0.44
0.04
133
G. Vieideret al. / NET plasma facing components
mal conductivity, low thermal expansion and high strength. Most of the critical issues for C-based materials have already been discussed in context with the F W armor: mainly hydrogen and gaseous impurity retention and outgassing as well as potential safety hazards. As for the F W armor baking at 350°C will also be required for the divertor. In addition, chemical erosion by hydrogen and oxygen is a major concern [20]. For the technology phase of next step tokamaks until now, high-Z materials such as tungsten have been considered as the main candidate for a DP armor with adequate lifetime. The present trend to higher fusion powers, lower plasma densities and consequently to higher particle eneergies, however, raises severe doubts on the feasibility of a high-Z DP armor for most of the anticipated plasma scenarios. Hence, low-Z materials such as C-based materials or beryllium have to be considered also for the technology phase: - The main draw back of these low-Z materials are gross erosion rates without redeposition of up to several m / b u r n year - A s s u m i n g some divertor replacements during the technology phase, the net erosion has still to be considerably smaller than the estimated gross erosion. This may be achievable via: • redeposition, • sweeping of the null-point, • reduction of chemical erosion and radiation enhanced sublimation of carbon based materials e.g. by addition of Si or B, and DP surface temperature limitation to less than 1500 ° C. - Beryllium has no chemical erosion and would permit rapid in-situ repair of the divertor plates by plasma spray. However, the melting temperature and thermal shick resistance of Be is rather low. 4.2. DP design concepts
Fig. 8 illustrates a divertor plate integration on blanket segments together with a typical heat flux distribution. The simple fiat plate geometry facilitates manufacture and sweeping of the null point. The replacement of the DP is foreseen by remote maintenance equipment without removal of the blanket segments. The DP design is mainly dictated by the required cyclic peak heat fluxes in the order of 10 M W / m 2 otgether with the constraints by the carbon based armor material. Two basic design concepts are being studied via industrial contracts [21,22,23]. As for the first wall, water at 6 0 - 1 0 0 0 C and 1 MPa is the coolant. The
Fig. 8. Divertor plate with heat flux distribution and integration on inboard blanket segment.
concepts differ mainly in the attachment of the C-based armor on the heat sink: (0 Brazed armor [21] This concept is shown in fig. 9 with the graphite armor brazed to a heat sink in a molybdenum alloy TZM. - TZM has been chosen mainly because of a matching thermal expansion with graphite, which facilitates the brazing operation. Such brazing of graphite on TZM is performed industrially for the manufacture of rotating X-ray anodes of about 100 mm diameter operating at comparable temperatures for very high cycles. ~
~~
~-~CFC "OLE
H !_ ~
I
P ~
~
JOl~lT
~
Fig. 9. Div~or plate ~ n ~ t ~ c~n ~m~ite ~or br~ on~ a M ~ l o y h ~ t ~ j o i n ~ by the i ~ f i c p r y i n g (HIP) a ~ g PI~.
134
G. Vieider et aL
/
N E T plasma facing components
CFCTILF FXF~.IATED GRAPHITLFAYER COPP~R-~L.OC YOOLANTTUBF 4,7
~
~
~~
~
~
~ L~KI~BAR
~STM~ES5STEELS~P~TPLATE Fig. ]0. DJvertor plate c o n @ t with c~bo~ composite armor ~]es m ~ h ~ c ~ y attach~ to a Cu-~]oy heat s ~ a ~ r ~ n g Novafime ~ U r ~ - P ~ n e y .
- Draw backs of TZM compared to copper alloys as the the main alternative are a lower thermal conductivity and low fracture toughness especially under irradiation, however, for the physics phase, irradiation damage should be low. - The heat sink is manufactured by joining TZM half shells with coolent tubes in M o - R e alloys via a HIP (Hot Isostatic Pressing) process in the same manufacturing step. Thus a kind of double containment of the coolant is obtained. - Mo-41 ~ Re is compared to TZM more ductile, has a much lower ductile-brittle transition temperature, is better weldable but also more expensive. - the use of TZM backplates on the C-based armor was shown to permit an in-situ tile replacement by de- and rebrazing. Such intermediate refractory metal layers would also facilitate the use of copper alloys as heatsink. 00 Mechanical armor attachment [22] This concept is shown in fig. 10 with CFC-tiles pressed against the heat sink in a copper-alloy via blade springs and a locking bar: this solution would permit an in-situ tile replacement, the feasibility of the concepts is critically depending on an efficient heat transfer between armor and heatsink of at least 1 W / c m 2 K, which under ideal conditions can be achieved by the use of 0.5 mm thin foils of exfoliated graphite (Papyex).
- 1600 ° C for 10 mm CFC assumed to be brazed to TZM, - 2000 ° C for 6 mm CFC mechanically attached via papyex on copper alloy assuming a high contact heat transfer. It should be noted, that radiation enhanced sublimation of C-based materials sets a temperature limit at about 1500°C when sublimation exceeds the peak chemical erosion. It seems therefore, that 10 M W / m 2 static peak DP heat flux represents an upper limit for carbon based armor with 5-10 thickness. The thermo-mechanical analysis also shows that the thermal stresses in the TZM heat sink (without brezing stresses) appear to be acceptable for 104 cycles at 10 M W / m 2. However, the lack of material data and design criteria should be noted. For the assumed low pressure water coolant in simple 12-14 mm diameter tubes, burn-out is predicted at critical heat fluxes of 10-12 M W / m 2. The critical heat fluxes can, however, be increased by more advanced coolant schemes such as vortex promoters.
-
-
Thermo-mechanical analyses were performed for both concepts. Fig. 11 shows that at-a DP heat flux of 10 M W / m 2 the peak carbon armor temperature without irradiation damage would be about [24]:
5OO-~
/ 7 ~
/ I~N-OUTFGR L 51B°LET~F.
"--"--'----'-" ~ E X FL~ ~
~0 DI~RT~~ATFL~[~Im~
Fig. 11. Temperatures ~ divertor plates ~ t h carbon-bas~ a ~ o r .
G. Vieider et al. / NETplasma facing components
135
The present results indicate that a static divertor heat load of 10 M W / m 2 is already an ambitious development target. For yet higher divertor heat fluxes of 15-20 M W / m 2 probably required in the technology phase, sweeping of the separatrix is now being studied as a promising method to obtain armor temperatures, heat sink fatigue and coolant conditions at acceptable values, i.e. not worse than for static heat loads of 10 M W / m 2. 4.3. Manufacturing and thermo-mechanical testing In the frame of NET study contracts the manufacturing feasibility of various cross-sections and subassenblies is to be demonstrated by producing relevant samples. For the brazed lay-out, some samples are shown in fig. 12. Additionally, gundrilling of TZM bars over a length of more than 2 m was accomplished for holes of 18 mm diameter. The weldability of similar and dissimilar joints of various tube materials was tested. A similar concept based on graphite tiles brazed to thin-walled
Fig. 13. Test of mechanically attached carbon based tiles in the neutral beam test bed at JET.
Fig. 12. Divertor test specimen manufactured by Plansee via brazing and hot isostatic pressing.
molybdenum tubes has been manufactured for Asdex Upgrade [25] and extensively tested at the SANDIA electron beam facility [26]. Extensive bonding trials between different graphites and heat sink materials, copper or molybdenum alloys, revealed that, on thermal cycling, sensible larger scale graphite/copper samples were subject to excessive thermal mismatch strains [23]. A broader and more systematic investigation has been started at the Kernforschungszentrum Karlsruhe in the frame of the NET supporting technology programme. The thermomechanical behaviour of different materials and brazed material combinations shall be studied under thermal cycling conditions, produced by a plasma flame, which can deposit peak heat fluxes up to 15 M W / m 2. During 1987 a not yet optimised test piece of 500 × 100 mm2 was manufactured with 10 graphite and carbon composite tiles mechanically attached on a water cooled copper heatsink - see fig. 13 [16,22]. With this test piece a series of tests has been performed recently at the JET
136
G. Vieider et al. / N E T plasma facing components
neutral beam injectors test bed at short pulsed heat fluxes up to 10 M W / m 2 and up to 250 cycles. The main objective was to study the influence of the intermediate soft 'Papyex' layer (between heatsink and tile) on the thermal bchaviour of the tiles. From these tests it is concluded that: - Both tiles with brazed back plates in Mo-alloy failed by deformation and debrazing, while all the other tiles showed no damage. The use of Papyex improves the heat transfer between tile and heat sink by an order of magnitude, however, this improvement was considerably less than expected. In late 1988, other medium sized mock-ups will become available as well for the mechanically attached concept (320 × 110 mm 2) as for the brazed concept (650 × 200 mm2). Thermo-mechanical testing of these mock-ups is planned at: ion- and dectron beam facilities at SANDIA National Laboratories, Albuquerque, USA, a 1 MW solar furnace at the CNRS center Odeillo, France, capable of delivering an almost in time unlimited peak heat flux of 10 M W / m ~.
5.
Conclusions
The studies on plasma facing components by the NET Team in collaboration with European industries and laboratories have led to a consolidation of the initially rather uncertain situation with the following main conclusions: (i) Carbon based material will be used as the plasma facing armor on both divertor plates and major areas of the first wall at least during the initial physics phase. The main reasons for this choice are the generally good experience with these materials in current large tokamaks concerning plasma impurity control and disruption resistance. (ii) The first wall structure in austenitic stainless steel will be cooled by double contained low pressure water for achieving a reliable 'zero leakage' solution. Several design concepts for the structure and the mechanically attached tiles are being developed, for which analyses predict adequate thermo-mechanical performance, i.e. allowed nominal heat fluxes at 2 × 104 cycles of up to: 0.8 M W / m 2 with radiation cooled tiles at about 1800°C, 0.2 M W / m 2 with conduction cooled tiles at less than 1000 ° C. The demonstration of the feasibility and performance of these designs is in progress via:
the manufacture of prototypical test sections, which is nearly completed, the thermo-mechanical testing of these FW-sections until 1989/90. (iii) Several concepts of dioertor plates are being developed with mechanical attachment or brazing of the armor to a water cooled heat sink in copper- or molybdenum alloys. First thermo-mechanical analyses indicate that nominal cyclic heat fluxes of up to 10 M W / m 2 may be allowed. This appears adequate for the physics phase and would probably require 'sweeping' of the separatrix in the technology phase. The manufacture of medium-size mock-ups and their thermo-mechanical testing are in progress. The main critical issue for the divertor plates is the limited life time, mainly due to erosion and subfimation in the technology phase (assuming only a few disruptions and some DP replacements). The nominal erosion life time would only be adequate: • for carbon armor, if the net erosion would be considerably less than the predicted sputtering erosion, which might be obtained by redeposition, separatix sweeping and improved materials, • for tungsten armor, if the plasma edge temperatures would not exceed 20 eV, which seems difficult to achieve in most of the conceived plasma scenarios especially with non-inductive current drive. Summarizing, there appears to be reasonable prospects to develop plasma facing components with adequate performance for the physics phase of NET, provided that the disruption frequency would be significantly lower than in present tokamaks. In the technology phase, the major crucial issue is the risk for too short divertor life times due to nominal erosion. -
Acknowledgements
The authors would like to thank all colleagues who are participating in the studies and development of plasma facing components for NET, in particular J.L. Boutard, M. Budd, M. Chazalon, A. Cardella, F. Engelmann, B. Libin, J. Raeder, R. Toschi and C.H. Wu from the NET Team as well as F. Farfaletti-Casali and R. Matera from JRC ISPRA.
References
[1] R. Toschi et al., Next European Torus (NET) Fusion Technology (July 1988).
G. Vieider et al. / N E T plasma facing components [2] R. Toschi, NET and the European fusion technology programme, 15th SOFT Utrecht, September 1988, Fusion Engrg. Des. 11 (1989) 47-62, in this issue. [3] G. Vieider et al., Progress in the development of a first wall for NET, Int. Symposium on Fusion Nuclear Technology, Tokyo, April 1988, Fusion Engrg. Des. 8-10 (1989). [4] D.K. Owens and M.A. Ulrichson, Plasma-wall interaction: Recent TFTR results and implications on design and construction of lindters, 14th SOFT, Avignon, September 1986. [5] M.A. Pick, G. Celentano, E. Deksnis, K.J. Dietz, R. Shaw, K. Sonnanberg and M. Walravens, Experience with graphite in JET, 12th SOFE, Monterey, October 1987. [6] H.Th. Klippel, J.G.v.d. Laan, Thermo-mechanical behaviour of bare and coated first walls under plasma disruption conditions, ISFNT Tokyo, April 1988, Fusion Engrg. Des. 8-10 (1989). [7] F. Brossa, M. Cambini, D. Quataert, G. Rigon and P. Schiller, Physical and metallurgical phenomena during simulations of plasma disruptions, 3rd Int. Conference on Fusion Reactor Materials, Karlsruhe, October 1987. [8] B. van der Schaaf, Radiation damage of stainless steel, Fusion Technology Progress Report ECN-185, ECN Petten (May 1986). [9] E. Diesele, T. Fett, D. Munz and H. Stamm, Stress and lifetime calculations for first wall and blanket structural components, KfK-Report 4210, Karlsruhe (March 1987). [10] H. Brinkschulte, E. Deksnis and A.S. Bransden, Heatshock resistance of graphite determined with a CO 2 laser, JET Report 1'(87) 43. [11] M. Birch and J.E. Brocklehurst, A review of the behaviour of graphite under the conditions for protection of the first wall of a fusion reactor, UKAEA Report ND-R-1434 (S), Springfield (Dec. 1987). [12] Y.R. Crutzen and F. Farfaletti-Casali, Structural design problems related to plasma disruption events of the removable NET first wall segments, Int. Symposium on Fusion Nuclear Technology, Tokyo, April 1988, Fusion Engrg. Des. 8-10 (1989). [13] D. Munz et al., Evaluation of crack growth delay in multilayer sheets, NET Study Contract 253/86-11/FUD, KfK Karlsruhe (July 1988).
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[14] (3. Vieider et al., First wall concepts with welded steel structures for NET, 14th SOFT, Avignon, September 1986. [15] V. Zampaglione et al., First wail for NET: Engineering studies and mock-up design, technological development and manufacturing, Proc. of the 15th SOFT, Utrecht, September 1988 (Elsevier Sc. Pub., Amsterdam, 1989). [16] A. Cardella et al., Testing of divertor samples for NET at the JET N.B.I. test bed, Proc. of the 15th SOFT, Utrecht, September 1988 (Elsevier Sc. Pub., Amsterdam, 1989). [17] Association Franqaise pour les R~les de Construction des Chaodi~res Electro-nucl~ires, Design and Construction Rules for FBR Nuclear Islands, Paris (June 1985). [18] E. Zolti, Comparative thermo-mechanical study of first wall protection concepts with conductively cooled tiles, NET Internal Note NET/IN/88-01. [19] R. Matera, V. Renda, G.L. Morlotti, G. Graglia, First wall design criteria, thermal fatigue and creep experiments and theoretical modelling, 14th SOFT, Avignon, September 1986. [20] C.H. Wu, J.W. Davis, A.A. Hassz, The formation of methane by the interactions of low energy hydrogen ions with graphite, Proc. of 15th EPS Conference on Controlled Fusion and Plasma Heating, Dubrovnik, May 1988. [21] G. Kneringer et al., The design and feasibility of NET divertor target plates; Final Report on Contract NET/87724A, Phase I, Metallwerk Piansee (January 1988). [22] M. Besson et al., Divertor plates with replaceable conductively cooled carbon tiles, Proc. of the 15th SOFT, Utrecht, September 1988 (Elsevier Sc. Pub., Amsterdam, 1989). [23] R. Howard et al., The design and feasibility of NET divertor target plates; Final Report on Contract NET/87724B, Phase I, GEC Energy Systems Ltd. (1988) R(88) 18. [24] E. Zolti, Thermomechanical design evaluation and material properties requirements for NET divertor elements, ICFRM-3 Conference, Karlsruhe, Oct. 1987, to be published in J. Nucl. Mater. [25] H.E. Kotzlowski, Actively cooled graphite limiter, 13th SOFT, Varese, September 1984. [26] J. Bohdansky et al., Behaviour of graphite under heat load and in contact with a hydrogen plasma, Nucl. Instr. and Meths. in Phys. Res. B23 (1987) 527-537.