Plasma-wall interaction in NET

Journal

of Nuclear

Materials

14S147

North-Holland.

154

PLASMA-WALL F. ENGELMANN and G. VIEIDER’

INTERACTION

(19X7) 154-164 Amaterdarn

IN NET

‘, M. CHAZALON

‘, M.F.A.

HARRISON

2. E.S. HOTSTON

l. F. MOONS

1

’ The NET Team c/o Max-Plan&Institut

ftir Plusmaphysrk. Bolt;munnstr. 2. D-8046 Gurching hei Miinchen, Fed. Rep. Germanq ’ Culhant Luhoratory. Ahingdon, Oxon OX 14 3 DB. United Kingdom (UKA EA/Euratom Fwon Association)

Key words:

plasma-wall

interaction,

plasma

edge modelling.

tokamak

reactor

design

NET is conceived as an experimental reactor with the aim of demonstrating reactor-relevant plasma performance and reliable operation of the device as well as developing and testing components for a demonstration reactor. For power and particle exhaust both a single-null and a double-null poloidal divertor configuration are under consideration. An intense modelling effort is undertaken to predict the heat load and erosion characteristics for these configurations. Under bum conditions, the divertor will operate in the high-recycling regime. The resulting heat loads on the divertor plates are predicted to be somewhat more demanding in the case of a single-null divertor. If one excludes working under conditions where a large part of the power is exhausted by radiation from the plasma edge, refractory metals (W. MO) have to be used for the plasma-facing surface of the divertor plates, the radial heat and particle transport in the scrape-off layer must be large (XT. = 4 m’/s; D, - 1 m’/s) and the plasma density at the edge of the discharge must be high ( nS - 5 x 10” mm3). Erosion of a bare stainless steel first wall, under normal working conditions, appears to be within acceptable limits. but the use of graphite armouring is considered in order to avoid wall damage due to localized loads of highly energetic particles and to protect against disruptions. Such a solution would also be consistent with the anticipated requirements during start-up. For both the first wall and the divertor plates various concepts are under consideration. Using replaceable tiles as plasma-facing components throughout appears attractive.

1. Introduction

1.1. Objectives of NET The Next European Torus (NET) is conceived as an experimental fusion reactor to follow JET and aiming at the demonstration of the feasibility of fusion energy production in a plant which integrates the essential components and technologies of a fusion reactor [l]. This implies that NET has both physics and technological objectives. The physics objectives are to produce a plasma with parameters and performance relevant to a fusion reactor. This includes - generating an ignited plasma in a controlled way, and _ operation with extended (up to 1000 s) and reproducible burn pulses. The technological objectives, in general terms, consist in an integrated test of reactor components and systems as well as engineering tests of in-vessel components (plasma-facing components like the first wall and power exhaust devices as well as blanket modules) and, specifically, can be summarized as - selection and qualification of design concepts which meet the basic performance requirements of a demonstration reactor, - testing of materials and components in an integrated way in a fusion environment, _ testing of tritium and power extraction under reactorrelevant conditions, - detionstration of the maintainability of a fusion reactor,

0022-3115/87/$03.50 0 Elsevier Science Publishers (North-Holland Physics Publishing Division)

B.V.

- demonstration of plant reliability at levels relevant for a demonstration reactor, and - demonstration of the safe and environmentally acceptable operation of a fusion reactor-like plant. A total neutron fluence of about 0.8 MW a/m2 is required to achieve these objectives. The number of pulses considered as a target is 105. 1.2. Reference concept and parameters

Extensive optimization studies have lead to the selection of a reference concept for NET which is characterized by a system integration scheme that allows assembly and disassembly of in-vessel components from the top of the device (see fig. 1). Both a single-null (NET SN) and a double-null (NET DN) poloidal divertor configuration are under consideration. Their main parameters are given in table 1. 1.3. Approach to plasma-wall

interaction in NET

Because there is as yet no satisfactory data base for plasma-wall interactions under reactor conditions it is, to a large extent, necessary that the analyses performed to evolve concepts for the plasma-facing components of NET be based on working hypotheses. It is anticipated that a considerably more solid data base will be available when a detailed design of NET will start around 1990. The present approach to providing physics specifications for plasma-wall interaction is based on the following elements:

155

F. Engelmann et al. / Plasma - wall interaction in NET NET D.N.

BLANkET

TOROIDALFIELD COILS INNERPOLOIDALCOILS ,PLASMA

Fig. 1. Reference

concept

(i) For bum conditions, modelling studies are performed (see section 2), adopting as a reference case a situation of low impurity content in the plasma (i.e., ) of the cc-particle power is radiated to the first wall and s channelled to the divertor plates). Operation at a high plasma edge density (n s = 5 X 1019 rnm3) is envisaged so that the divertor works in the high-recycling regime and divertor plate surfaces of tungsten or molybdenum are appropriate. Attention is also paid to the possibility that ripple-induced losses of fusion a-particles might lead to appreciable localized power loads (e.g., see [2]). (ii) For start-up, the conditions characteristic of present-day large tokamaks are taken to be typical; this implies that start-up has to take place in a low-Z environment. Table 1 Main parameters

of NET DN and NET SN.

Parameter

Plasma minor radius Plasma major radius Magnetic field on plasma axis maximum on coils Plasma current Total beta DT density (total) Murakami parameter Fusion power a-particle power Average neutron wall loading

NET DN a(m) R(m) B(T) B,(T) I,(M) B,,, (W) EDT (10” mm3) M (lOI Wb-‘) Pfu,

WY

p, 04.W)

Q, (m/m*)

NET SN

1.35 5.18

1.53 5.19

5.0 10.4 10.8 5.6 1.20 16 600 120

5.8 11.6 9.9 3.9 1.13 15 625 125

1

1

NET DN (cutaway

model).

(iii) For disruptions, the guidelines adopted by the INTOR Workshop [2,3] are used, supplemented with a specification of a localized wall load due to beams of runaway electrons such as are observed in JET [4]. The resulting disruption specification is summarized in table 2. If loads due to fusion a-particles are disregarded, Table 2 Specification

of disruptions

for NET

Frequency per pulse (experimental phase/routine operation) Fast phase (energy quench): duration Slow phase (current decay): duration First wall: fast energy deposition slow energy deposition peaking.factor Note: (i) This energy is mainly carried by radiation (ii) Protruding parts of the first wall (sublimiters) are subject to a high local power deposition: here also the energy of beams of runaway electrons is deposited which may be tentatively taken to be 0.1 MJ (deposition time: 2 ms) Divertor plates (all targets) fast energy deposition peaking factor Note: The energy is mainly carried by plasma particles.

10-2/10~” ms2 ms 20 MJ 100 MJ 150 3

MJ 100 3

156

F.

Engelmann

et al. / Plasma

then under normal working conditions, a bare stainless steel wall appears to be a possible solution. However, partial (“sublimiters”) or close to complete coverage by graphite in the form of tiles is now being considered in order to cope with localized loads and with disruptions. This is also consistent with the anticipated requirements during start-up. However, erosion and the mixing of surface materials arising from migration is an unresolved concern. 2. Modelling of plasma-wail

wall interactron

in NET

the outer target, and in some cases the inner, must be inclined in order to reduce the peak power load delivered by plasma particles. The location accepted for the targets is such that a suitable inclination (e.g. 10” to 15’ to the magnetic surfaces) can be comfortably accomodated within the outer divertor. The spacing between the null-point and the nearest part of the target is not less than about 0.4 m and in addition the distance (measured along the separatrix in the poloidal plane) from the null-point to the point of interception of the outer target is about 1 m. More precise location of the targets is the subject of ongoing studies.

interaction

2.1. Modelling of the boundary plasma -7.2. Application of the two-dimensional plasma model Modelling of the edge and divertor plasma has been undertaken in order to evolve physics specifications for plasma-facing surfaces and to establish helium exhaust requirements of both single-null (SN) and double-null (DN) divertor configurations of NET under steady-state bum conditions. Details of the two-dimensional fluid model used to assess plasma transport are given in [5]; the manner in which the model is applied is extensively discussed in [6]. The equilibrium diagram of NET SN is shown in fig. 2(a) and that of NET DN in fig. 3(a): further details can be found in table 3. In the case of NET SN, the region modelled is shown in fig. 2(b). It includes both the inner and outer divertor channels and their inner-connecting scrape-off layer together with a thin layer of plasma inboard of the separatrix. Modelling of NET DN is, as can be seen in fig. 3(b), at present restricted to the outer scrape-off layer together with its top and bottom divertor channels. For the purposes of modelling, the divertor targets are assumed to have a tungsten surface and to lie at right angles to the magnetic surfaces although actually

The plasma model requires, as an input parameter, a prescribed value of the non-radiated plasma power, Q which flows across the separatrix into the scrape-off la;er. This is assumed to be 5 of the a-particle heating power (see table 3) and hence it corresponds to rather clean conditions in the ignited main plasma. In the case of the double-null configuration, the power flowing to the outer scrape-off layer is taken to be 0.8 Q, in accord with experimental data from ASDEX [7]. The model also requires a prescribed value of plasma denDivertor performance is sity, n,, at the separatrix. strongly sensitive to both Q, and n, and a survey of the consequences of varying n, around the value n, = 0.3 n, (i.e. in the regime where n, = 5 x 1019/m3) has been performed. Additional input parameters needed for the model are the coefficients for transverse transport of energy (x:) and of particles (Dl). These are chosen to be in accord with present experiments; typical values are x; = 4 m2/s and D, = 1 m2/s. These values are used to predict the reference conditions in the edge

(b) .off

Inner jivertor

Outer ‘Central

divertor regions

Reflecting barrier

Fig. 2.’ Single-null configuration of NET. (a) Equilibrium diagram indicated. (b) The region of plasma modelled in a single-null

for NET SN. A typical target location as used in the modelling divertor configuration lies between the two dashed surfaces.

is

F. Engelmann et al. / Plasma-

Fig. 3. Double-null indicated.

wall interaction in NET

configuration of NET. (a) Equilibrium diagram for NET DN. A typical (b) The outer region of plasma modelled in the double-null configuration

157

target location as used in the modelling lies between the dashed surfaces.

is

Table 3 Physics specification for tungsten divertor targets in NET during bum conditions: (i) Operational conditions are selected on the basis of a balance between n, and the release of tungsten by physical sputtering. In all cases the transverse transport coefficients are at the divertor is very small X’l = 4 m*/s and D, = 1 m*/s. (ii) Erosion assessed on the basis that oxygen impurity ion concentration ( <10m3), i.e. chemical sputtering is neglected. (iii) Sputtering by He *+ ions is tentative: it depends upon plasma density at the separatrix, plasma transport and the performance of the pumping and fuelling scheme

Main plasma Fusion power [MW] Power to scrape-off layer [MW] (for clean plasma conditions) Power radiated [MW] Temperature at separatrix [eV] Density at separatrix [1019/m3]

Plasma at targets Power deposited on target total [h4W] Via kinetic energy of electrons and ions Via atomic processes ‘) Peak power loading [MW/m2] (perpendicular to magnetic surfaces) Full width-half height of power profile [lo- 2 m] Uniform radiation from main plasma [MW] Peak plasma temperature [eV] Ion flux (D+/T+) [1O24/s] Physical sputtering rate of tungsten target [lot9 W at/s] (includes contributions from He2+ ) Corresponding peak erosion [lo-* m/a] (neglecting redeposition; 100% availability; perpendicular inclination) Profile of this erosion (full width half height) [10K2 m] a) Includes

D+/T+

ion recombination

NET DN

NET SN

600 64 (outer scrape-off)

625 83

40 T,=lOO; 5.5

42 T,=126; 6.5

T,=125

T,=144

Top as well as bottom outer targets

Outer target

Inner target

28 21 7 16

46 31 15 18

27.5 15.0 12.5 11

2.8 2.5 T, = 36; T, = 20 2.3 8.5 at unpumped target (5% He*+ ) 1.7 at pumped target (1% He2+) 7 at unpumped target 1.5 at pumped target

4.2 2.5 T,=20; T,=13 5 0.3 (1% He*+ )

4.2 2.5 r,=lO; T,=9 4.3 nil (5% He’+)

0.3

nil

2

2.5

at the target plus half of radiation

from recycling

D/T

neutrals.

15X

F. Engelmann

et al. / Plusmu - wdl inteructwn

plasma (presented in table 3) but the consequences of reducing transverse transport have also been explored. It should be noted that both the peak power load and the peak plasma temperature close to the target increase substantially if the transverse transport coefficients fall significantly below the reference values. In such conditions it is unlikely that the main plasma will remain clean. In the present modelling, pumping of D/T gas from the divertor is simulated by assigning an albedo (A > 0.99) for plasma ions incident upon the divertor target. A fraction (1 - A) of the incident ions is thus lost from the system. In principle, different values of A are assigned to He and D/T ions, these values being derived from stand-alone Monte Carlo analyses [6,8] of neutral gas transport to the divertor vacuum pumps. However, at present the plasma transport code can only be applied to D+/T’ ions, the simulated throughput of D/T being based on the exhaust requirements of helium (2 x 10” He atoms/s) assuming that the concentration of helium gas in the exhaust flow is 5%. In the case of NET SN, exhaust of D/T gas is simulated only at the outer target (in accord with neutral particle transport predicted by Monte Carlo calculations). For NET DN pumping of D/T gas has been simulated (a) in the bottom divertor only and (b) in both bottom and top divertors. The difference in pumping conditions causes but slight changes in the D/T plasma. The fraction of He’+ required in the incident plasma flux onto the divertor target is 1% in the case of a pumped target: it is taken to be 5% in the case of the unpumped target, but this number is expected to be appreciable reduced if gas puffing is performed adjacent to this target. 2.3. Comparison tungsten targets

of single and double null-divertors

with

The data presented in table 3 have been selected to show a favourable balance between the release rate of tungsten (by physical sputtering) and the contention that n, should not grossly exceed 5 X lO”/rr?. Results from modelling of many single-null configurations show that there is an asymmetry of power flow in favour of the outer divertor channel (by a factor of at least 1.7). This arises predominantly as a consequence of the shape of the poloidal cross section (i.e. a combination of aspect ratio and elongation/triangularity) but appears to be rather insensitive to the plasma poloidal beta. The peak power loads at the outer divertor target are high but they decrease with increasing plasma triangularity to, for example, 18 MW/m’ for NET SN (with a perpendicularly inclined target). Thus inclination of the NET SN target at about 15O is required. Plasma conditions for the double-null configuration tend to be superior with regard to power loading because the power flow is symmetric above and below the equatorial plane. Hence the power is shared

rn .VET

equally between the upper and lower targets. The high peak loads, which are characteristic of the asymmetrical power flow in single-null divertor configurations, are thereby reduced. Target inclination at about 18” appears to be adequate for NET DN. The radial profile of the target power load is rather narrow for both configurations. The profile for NET DN. which is shown in fig. 4, can be regarded as typical; it has a scale length (full width-half height) of only a few lo-’ m. An oscillatory motion of the divertor plasma across the target surface, not considered in the model, would widen the effective deposition profile and hence reduce the peak lower loading. 2.4. Impurity control It is difficult to quantify the maximum release rate of tungsten atoms from the divertor target which is acceptable from the point of view of plasma contamination, but about 1019 atoms/s appears to’ be a conservative estimate. Tungsten is chosen as the target material because it has a high threshold energy for physical sputtering. Plasma modelling predicts that powerful, localised recycling occurs within the divertor. As a result, for NET SN and NET DN, the incident energy of D/T ions (i.e. E, = 5 kT,) lies only just above the threshold for physical sputtering if n, is about 5.5 x 10’“/m3. Here T, is the plasma electron temperature adjacent to the target which is strongly sensitive to the upstream density, n,, of the scrape-off plasma. To keep r, low, it is desirable to operate with the highest practicable value of n,. Reduction of T, by means of puffing gas through the surface divertor target into the divertor plasma has been simulated in a single-null configuration. A significant reduction in T1 (and hence in physical sputtering) results with a gas puffing rate as low as half of the net flow of plasma particles across the separatrix into the scrape-off layer, but the effect is to be attributed predominantly to an overall enhancement of the density in the scrape-off layer, both of the

I

Fig. 4. Radial profile of power loading on the outer divertor target of NET DN. Data are for a tungsten target which is inclined perpendicularly to the magnetic surfaces.

F. Engelmann et al. / Plasma - wall inieraction in NET upstream area.

plasma

density,

n,, and within

159

the divertor

2.5. Target erosion The radial profiles predicted for erosion due to physical sputtering of the tungsten targets are very narrow. A typical example (for NET DN) is shown in fig. 5). It is evident that even small movements of the divertor plasma across the target surface will substantially reduce the average erosion. The effects of redeposition are neglected in the model and it is not unreasonable to expect that they would reduce the peak erosion by a factor 10 to 100. Erosion and the commensurate release of target atoms could be enhanced by chemical sputtering due to plasma impurities such as oxygen. The physical issues pertinent to the situation are not yet adequately understood and the experimental data base is fragmentary (indeed chemical sputtering of powerfully bombarded surfaces does not appear to be a substantial effect in present-day tokamak experiments). A second issue of importance is physical sputtering of the divertor target by energetic impurity ions. Impurities can reach high velocities as a consequence of friction forces and, if the ions are also raised to high charge states, by the acceleration of such ions in the electric field of the plasma sheath. Despite the inability to quantify these issues, present evidence indicates the desirability of minimising the impurity content of the divertor plasma and this need is particularly strong in the case of oxygen.

-0

Radiation (uniform) [MW] Note: Radiation based on “clean” plasma From charge exchange atoms [MW] Flux density of these atoms [102’/m2/s]

Distribution of sputtering: some peaking around Physical sputtering by D+/T+ ions: nil

Analysis of first wall sputtering has not yet been undertaken specifically for NET DN or NET SN. Nevertheless the scrape-off plasma conditions for these concepts are similar to those described in [6] for which the stand-alone calculation of charge exchange atom fluxes reported in [9] applies with reasonable accuracy. Specifications based on these fluxes are presented in table 4.

increase

bum (charge

exchange

data are taken from [9])

to 100 MW

Sputtering rate 0.5 2 8.5

the equatorial

1

[m]

2.7. First wall

35 - could otherwise 1 5

Sputtering by DT atoms [10’s at./m’/s] Erosion [10m3 m/a] (redeposition neglected; 100% availability) Stainless steel wall Graphite wall at 300 K Graphite wall at 1800 K

0.30

temperature increases. Such effects have not yet been adequately analysed but they appear to be due to a combination of (a) the variation in ( Br,,,i/Btor) and (b) the decrease in connection length. Cross field diffusion into the central region of the divertor appears to be a minor effect in the parameter range so far investigated.

for NET DN and NET SN during

conditions

O-20

0.10

Fig. 5. Radial profile of erosion due to physical sputtering of the outer divertor target of NET DN. Data relate to a tungsten target inclined perpendicularly to the magnetic surfaces and redeposition is neglected.

If the divertor target is moved towards the null-point, the peak power loading decreases but the plasma sheath

of first wall conditions

0.00 t Separotrix

2.6. Location of the divertor target

Table 4 Physical specification

05

plane and possibly

Erosion rate 0.2 0.6 2.4

near to the entrance

of the divertor.

Note: (i)

Chemical sputtering by oxygen not included in above assessment: for graphite at 1800 K and 1% oxygen, additional erosion of 10 mm/a, neglecting redeposition. (ii) There may be localised loads of fusion alpha-particles due to ripple-induced losses (not yet estimated).

there

could

be

160

F. Engelmann

et al. / Plasma-wall

interaction

in NET

3. First wail concepts

secbon

A-A:

3.1. First wall plasma side protection Although erosion of a bare stainless steel wall due to fluxes of D/T atoms which arise due to charge exchange is anticipated to be modest over the total life of NET (about 0.2 mm; see table 4), it is nevertheless considered that armouring by graphite could be necessary to avoid first wall damage due to localized loads of highly energetic particles and to disruptions [l]. The reliability and easy maintainability of such a protection is an important issue and requires further study. Two alternative concepts are under consideration; _ a partial protection by poloidal sublimiters which protrude up to 50 mm inboard of the metal surface of the first wall. and - a close to complete coverage of the first wall by graphite. While preferable from the point of view of performance under normal working conditions and maintenance, the first option may not provide a sufficient protection against the effect of disruptions. In fact, for the disruption specification of table 2 and not considering the normal cyclic load, thermal fatigue already appears to limit the lifetime of the first wall to no more than about 100 disruptions, due to the formation of a melt layer and, concomitantly, cracks in the adjacent solid wall material. Also loss of molten material could be critical. Using mechanically attached graphite tiles, similar in principle to those used in JET, is attractive for maintenance reasons. A simple “snap-in rail” concept which relies upon radiation cooling appears to be promising (see fig. 6). The tiles have a thickness of about 15 mm and must be bonded to a substrate of refractory alloys (e.g. TZM) to avoid carburization of the steel. This solution also facilitates attachment to the first wall and provides an optimum protection of the first wall against runaway electrons. It is estimated that the metal substrate of a tile will reach temperatures in the range 1500 to 1900 K, depending on the incident heat flux and the emissivity of the surface. A TZM substrate would be satisfactory for pulsed operation of up to lo5 cycles provided that the temperature does not exceed 1500 K [lo]. Refractory alloys appropriate for higher temperatures are available, but component development is required. The emissivity can be improved, e.g., by coatings with Al 2O, or TiO?. Graphite is an attractive material for wall protection because, due to its thermo-mechanical properties, it can accept high localized heat loads. On the other hand. erosion is rather fast. For NET conditions, taking the availability of the device to be 100%. the erosion rate of graphite at 1800 K, due to physical sputtering, is estimated (see table 4) to be 2.4 mm/a; chemical sputtering which is difficult to quantify on the basis of present

Fig. 6. Graphite tile with refractory metal support fixed on the first wall [lo].

knowledge. could increase this rate considerably (e.g. to 10 mm/a), in particular if the oxygen content of the edge plasma is not kept very low ( +C 1%). The main limitation of graphite, however, is its swelling under neutron irradiation which for the working temperatures considered is expected to lead to cracking at fluences below 1 MW a/m2. Also tritium retention is a concern. 3.2 First wall structure The first wall concepts of NET are being developed for either water or helium cooling used in conjunction with austenitic or ferritic/martensic steel as structure materials. There are many uncertainties and foreseen difficulties so that several alternative design concepts for the first wall are being studied. Most of the concepts have the following common features (see also figs. 7 and 8): _ a thin front plate facing the plasma to minimise thermal stresses, _ channels with double contained coolant for minimum leakage, _ a thicker back plate to provide structural rigidity and some passive plasma stabilisation due to its electrical conductivity, _ integration of the first wall in a box enclosing the breeding units of the blanket in order to support the first wall structure in a simple way which offers the maximum resistance to disruption forces, to improve the passive plasma stabilisation features, and to facilitate outgassing and minimize gas leakage from the

F. Engelmann et al. / Plasma-wall interaction in NET

I

EL.BEAHYELDSI/

PROTECTION

B~=LlPUlO

LiPb OP

SOL10 Pb

\EL.EfAH

UELDS

b

161

breeding units into the vacuum chamber. The first wall concepts differ mainly in the method of manufacture (the feasibility of electron beam welding or brazing is being studied), and in the orientation of the coolant tubes (poloidal U-tubes appear to be more suitable for water cooling whereas toroidal tubes would be required for helium cooling). In the case of NET, which has a modest neutron fluence, the major limitations to lifetime of a graphite protected first wall structure arise from phenomena caused by thermal fatigue. On the basis of 2D and 3D thermo-elastic analyses the total thermal strain in the first wall structure has been estimated for various concepts based on austenitic stainless steel (see fig. 9). A comparison with the permitted strain range of AISI 316 L steel shows that the target objective of lo5 burn cycles could be achieved with the expected base wall load and a maximum front plate thickness of 2 to 3 mm if there are 15 mm thick protective tiles: without protection, either a higher wall load (up to 0.4 MW/m*) or a

Fig. 7. First wall concepts for NET with double contained coolant in an electron beam welded or brazed steel structure, distinguished by 1; 1’: 2; 3; 4.

OElAlL - B -

EBU=ELECTR(IY SEAI!YLDlN6

Fig. 8 Cross section of first wall integrated in a box enclosing an inboard blanket for liquid (LiPb) breeder material: the water cooling pipes are in poloidal direction. A and B refer to two different variants of concept 4.

162

F. Engelmnnn FW. - CROSS

FW - BOX

et al. / Plasma - wull interaction

SECTION

TYPES 0

in NET MAIN

(3

l

ASSUMPTIONS

FW

HEAT LOADS

=&E

55 “,;$ C W/cm1

l

PLASMA

SIDE

STRESS

OUTB

INB

;;

,; 9

12

CONCENTRATION

I,

LIMITS AT 0 OPA

I

!

1

I

O3

3

MIN

Fig. 9. Estimated

I

thermal

’

jiW

- THICEESS

t

mm

”

NUMBER

‘f:

lo‘ OF CYCLES

fatigue limits for NET first walls (FW) of austenitic stainless steel (average average surface heat load: 0.1 MW/m’: concepts 1: 2: 3: 4)

higher number of cycles (up to 106) or a thicker wall (up to 6 mm) would be possible. A front plate thickness of 2 to 3 mm is considered to be the minimum acceptable with regard to manufacturing constraints and to containment of coolant in the event of tube rupture. It is thus concluded, that a peak surface heat flux of 0.2 h4W/m2 as assumed for NET reference conditions is close to the maximum permitted for stainless steel first wall structures that have to achieve a fatigue life of lo5 bum cycles if wall armouring is necessary. Using martensitic steel the permitted peak heat flux could approach 0.5 MW/m’. However, in this case there are major uncertainties, for example, concerning the welding properties, ferromagnetic effects and loss of toughness under irradiation. 4. Divertor plate concepts 4.1.

lo

neutron

lob

wall load: 1 MW/m’;

brazed. welded or bolded to a cooled (copper) heat sink: (ii) a molybdenum plate design [12,13], see fig. 11, in which the plate is manufactured as an integral block which contains the coolant channels (molybdenum or

W-5%Re

cu

‘\

Fig. 10. W-Cu

Design concepts I

Three variants of NET divertor plates are under consideration: (i) a tungsten-copper plate design [ll], see fig. 10, which was initially proposed for INTOR and which involves the use of an alloy with a low sputtering coefficient (W/5% Re) for the armour plates that are

TARGET

/

divertor

plate.

Nb 173

-

no uztc

24.0 _i

Fig. 11. Molybdenum

(TZM) divertor

plate.

F. Engelmonn et al. / Plasma- wall interaction

alternatively TZM alloy is considered for the block to facilitate manufacture whilst the coolant tubes could be made of Nb/l’% Zr); (iii) a plate design with tiles as plasma facing elements [13], see fig. 12, in which the plate consists of a box of MO alloy filled with a heat sink material (Cu) through which the coolant channels are drilled and with a surface armoured with tiles of refractory material (e.g. of W) which are mechanically attached to the box (the contact pressure being provided by the different thermal expansion of the materials) and potentially can be replaced, if necessary. All these concepts require further investigation before their viability can be definitely judged. 4.2. Lifetime estimations During normal operation, the divertor plates are subjected to cyclic heat loads and fluxes of particles (see table 3) which cause material fatigue and erosion. In addition, further material loss as well as crack initiation and growth is to be expected as a consequence of disruptions. Present NET divertor plate designs provide a minimum sacrificial layer of 5 mm for erosion. As a working hypothesis adopted for NET DN, considering an availability of loo%, peak erosion rates of respectively 100 mm/a at the top plate (70 mm/a for physical sputtering and 35 mm/a for chemical sputtering due to oxygen) and 50 mm/a at the bottom plate (15 and 35 mm/a, respectively) are taken for targets perpendicular to the magnetic surfaces, not considering redeposition [l]. By inclining the plate at an angle of, e.g., 10” these values can be reduced by a factor of 6. Redeposition might reduce the peak erosion by a further factor 10 to 100. Hence, erosion might not limit the lifetime of the plates to less than that of the machine, but one has also to be prepared for several exchanges of the plates to be necessary. Preliminary calculations, performed for the molybdenum plate design, revealed that thermal fatigue limits 02Omm

Fig.

12. Divertor

plate

concept with tiles.

mechanically

attached

in NET

163

the lifetime of the plates to (1 to 3) X lo4 cycles. Thermal fatigue, hence, could be a more severe limiting factor for the lifetime of the divertor plates than erosion. The specification of disruptions given in table 2 implies a peak energy flux onto the divertor plates of 2.5 MJ/m* lasting for 2 ms. Assuming that there is no redeposition, a sacrificial layer having a thickness of 5 mm is consumed by evaporation after 100 to 200 disruptions under these conditions. While these expectations for the lifetime of the divertor plates are rather promising, one must keep in mind that one has to expect cross effects of the various phenomena to appear. These require further analysis. Whereas erosion has a beneficial effect on the fatigue behaviour, crack initiation during resolidification in the zones that are molten during a disruption would be detrimental. 5. Conclusions While for bum conditions a bare stainless steel first wall would appear to be appropriate for NET provided that localized loads due to fusion a-particle losses can be kept small, a partial or close to complete coverage of the first wall by graphite could be necessary to cope with disruptions. A low-Z environment is also expected to be needed for start-up. For the divertor plate surfaces, anticipating that the NET divertor will work in the high-recycling regime, using tungsten or molybdenum is considered. Mixing of surface materials due to migration is an unresolved concern. Both for the first wall and for the divertor plates various concepts are under consideration. For the first wall armouring, mechanically attached graphite tiles which allow for remote replacement are being studied. For the first wall structure, both austenitic and ferritic/martensitic steel are considered. For NET conditions, the dominant phenomenon in limiting the liftime of the first wall is expected to be thermal fatigue: for a target of lo5 full power bum cycles, the allowable peak heat fluxes onto the first wall are estimated to be at most 0.2 MW/m’ for austenitic and 0.5 MW/m*for ferritic/martensitic steel if the wall is armoured by graphite tiles of 15 mm thickness. For the divertor plates, the tungsten armour has to be brazed, welded or bonded onto a copper heat sink; also mechanical attachment of tiles appears to be a possibility. Altematively, the plates could be made as integral, actively cooled blocks of molybdenum. Erosion of the plate surface is considerable, but again thermal fatigue appears to be the factor imposing the most stringent limit on the lifetime: for the molybdenum option the lifetime has been estimated to be a few lo4 ‘full power bum cycles. Thus one has to be prepared for several exchanges of the divertor plates during NET operation.

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Both for the first wall and for the divertor plates the most demanding requirements derive from the anticipated occurrence of diruptions. They are the driving element in considering a complete armouring of the first wall and could appreciably decrease the lifetime of the divertor plates by causing cracks to be formed and grow. Hence, finding a way to avoid disruptions in tokamaks would be of greatest interest. The content of this paper is based on work of the NET Team and in the Associations. The authors gratefully acknowledge the many contributions by their colleagues. References

PI The PI [31 [41

is1 [61

NET Team, NET Report Nr. EUR-FU/XII80/86/51 (February 1986). INTOR Workshop, Report on Phase Two A, Part I. IAEA, Vienna, STI/PUB/638 (1983). INTOR Workshop, Report on Phase Two A. Part II. IAEA, Vienna, STI/PUB/714 (1986). M. Huguet. J.-A. Booth, G. Celentano, E. Deksnis, K.J. Dietz. P.H. Rebut, R. Shaw and K. Somienberg, Limiters and First Wall on JET, 11 th Symp. on Fusion Engineering, Austin (September 1985). B.J. Braams, CuIham Laboratory Preprint CLM-P725 (1984). M.F.A. Harrison. E.S. Hotston and A. De Matteis. NET

rn XET

Report Nr. EUR-FU/XII-361/86/50 (19X5); also Culham Laboratory Preprint, CLM-P 761 (1985). 171 The ASDEX Team. Proc. IAEA Tech. Comm. Meeting on Divertors and Impurity Control. Garching, paper I. R? (1981) 23. PI E. Cupini, A. De Matteis, R. Simon%. E.S. Hotston and M.F.A. Harrison, NET Report Nr. EUR XII 324/26 (1984). in: European Contributions to the IN[91 W.J. Goedheer, TOR Workshop, Phase Two A, Part II, Report Nr. EURFU-BRU/XII-268/85/EDVI, Brussels, Vol. II III-107 (1985). Design of Mechanical AtUOI H. Gassler et al., Conceptual tachments of First Wall Protective Tiles for NET, NET Study Contract, MetaIlwerk Plansee, Reutte (October 1985). A. Inzaghi Pll M. Biggie, A. Cardella, F. Farfalletti-Casali, and M. Turri, Fusion Technology. Proc. 13th SOFT. Varese (1984) 1231; G. Casini and F. Farfalletti-Casah. 8th Int. Conf. on Structural Mechanics in Reactor Technology, 1985, Brussels, Paper N4/1 (1985); F. Brossa. G. Federici, V. Renda and L. Papa, 11th Symp. on Fusion Engineering, Austin, Texas (1985). to the INTOR [=I The NET Team in European Contributions Workshop, Phase Two A, Part II, Report Nr. EUR-FUBRU/XII-268/85/EDVl, Brussels (1985) Vol. I X11-58. F. Moons, NET Internal Note NET/85/IN-071 (1985). Concept Innovations, INTOR-Phase Two A, P31 Tokamak Part III, Report Nr. EUR-FU-BRU/XII-52/86/EDVl (January 1986).