Relation of surface interactions to first-wall and in-vessel-component (IVC) design and materials performance in fusion devices

Joumal of Nuclear Materials 103 & 104 (1981) 7-18 North-Holland Publishing Company

RELATION

OF SURFACE

INTERACTIONS AND MATERIALS

TO FIRST-WALL AND IN-VESSEP-COFlPONENT PERFORMANCE IN FUSION DEVICES

(IVC)

DESIGN

R. W. Conn School of Engineering and Applied Science University of California, Los Angeles Los Angeles, CA 90024 Surface interactions affect the choice of in-vessel-component (IVC) materials and their coatings. Limiters, divertor components, special armor, and the first wall are subject to plasma and neutral atom bombardment at rates up to 5x10z3 particles/s with particle energies ranging from several tens of eV to 2-5 keV, depending on plasma edge conditions. Neutral beam injection leads to bombardment of special beam dumps with lo-200 keV parti les. Special events such as a plasma disruption can deposit hundreds of joules/cm h in a time of 2 to 20 ms on the limiter and first wall. Wall-produced medium or high Z impurities are detrimental to the plasma energy balance and to plasma MHD stability. Quantitative results are described to illustrate anticipated component response. Criteria are summarized which affect the design and choices of materials and coatings for in-vessel-components and the first wall. Illustrative design approaches are drawn from recent tokamak research, particularly from work on the international tokamak reactor, INTOR, and the U.S. fusion engineering device, FED.

1.

INTRODUCTION

It is well known and accepted that plasma-wall interactions (PWI) play a major role in present magnetic fusion experiments [1], particularly the closed-field-line devices such as tokamaks, stellarators, torsatrons, reversed field pinches, Even in open-ended magand elmo bumpy tori. netic mirror devices, in which impurities are expelled by collisions through the ends, plasmawall interactions have been important because Furof the need to control hydrogen recycle. ther, it is clear that plasma-wall interactions are central to the design of both new experiments (the Tokamak Fusion Test Reactor (TFTR), the Joint European Torus (JET), and the Japanese Tokamak, JT-60 [2]) and the next generation of fusion reactors. The international tokamak reactor study, INTOR [3], and the U.S. research on its Fusion Engineering Device, FED [4], are Of course, PWI has been imporgood examples. tant for commercial fusion reactor designs (e.g. STARFIRE [5]) and will be important to all classes of fusion machines, including tandem mirrors. The general reasons for the importance of PWI havebeenwell documented elsewhere [6-lo]. To summarize briefly, impurities in plasma radiate via several processes including bound-bound, free-bound, and free-free (bremsstrahlung) tranThe radiation intensity depends on the sitions. balance between these processes [9,10] which in turn depends on the local electron temperature and whether or not an equilibrium among excited The radiation intensity states can be assumed. is nonuniform spatially. One can therefore expect two effects: 1. The enhanced radiation will negatively influence the plasma energy For a given percentage impurity conbalance. tent, lower atomic number (Z) impurities radi-

0022-3

11S/8 I /OOOO-0000/%02.75

0

198 1 North-Holland

ate less intensely. 2. Intensified radiation in cooler regions of the plasma (due to bound-bound and bound-free processes) modify the electron temperature profile. This in turn effects the spatial profile on both the ion temperature and the pressure. Profile modification is often detrimental to plasma E?HD stability. In the tokamak, as an example, strong edge cooling can induce plasma disruption by changing the plasma current profile with the consequence of a rapid thermal quench. For these reasons, control of both the amount and type of impurity in the plasma (and thus control of plasma-wall interactions), is central to modern fusion research. Control of plasma-wall interactions can take several forms. t!agnetically, field lines can be diverted away from the primary chamber to a secondary region where the plasma interacts with a divertor collector plate [ll]. In tokamaks and other devices, a physical plasma limiter is used to control the plasma size and becomes the It is main area of plasma-wall interactions. essential for plasma operation with limiters that the edge temperature be low enough that the energy of particles is below the sputtering threshold to avoid excessive erosion. Usually, lasma edge this requires a strongly radiating and/or strong gas puffing [12,13,14 li . It may also be critical to protect limiters and the first wall with low Z coatings or low Z armor It is now commonplace to construct limitiles. ters for experiments from graphite. In this paper, we describe the nature of the plasma and heat loadings on the first wall and in-vessel-components (IVC) such as limiters, divertor collector plates, or special first wall We will describe how these loadings inarmor. fluence IVC design and the criteria for the choice of materials for construction, armor pro-

8

R. W. Corm /Surface

interactions

to first-wall

and IVC design

tection, and coatings. As such, this paper is a companion piece to two earlier papers [15,16] which dealt with reactor design and the criteria for selecting first wall and blanket structural materials. Examples of in-vessel-components are drawn from recent tokamak reactor research although most of the problems are comparable in other fusion concepts. Finally, we describe some immediate research problems.

2.

IN-VESSEL-COMPONENTS

Fusion machines are complicated and have many special in-vessel-components, as well as the first wall. The simplest way to describe the most relevant components is to examine some actual machine designs. For this purpose. we will use the tokamaks FED, INTOR, and STARFIRE. The materials choices for the main IVC’s in these designs are summarized in Table 1. In Fig. 1, we shows a cross section view of the right-half of the FED, a toroidal device. The important IVC’s are labeled in the figure. A mechanical limiter capable of pumping a fraction (5-6%) of the particles incident upon it is shown at the bottom of the vacuum vessel. Although made of segments, it forms a toroidal belt around the machine. The pump channel is shown below the limiter. The limiter design itself is illustrated in Fig. 2. The core is constructed of 316 stainless steel and cooled with water. The surface of the limiter facing the plasma is protected by graphite tiles coated with TiC to control chemical reactions of graphite with hydroThe tiles are 15 cm x 15 cm x gen isotopes.

Table 1.

Materials

Component

Choices

FED REFERENCE

Figure 1

VIEW

: Cross section view of the right-hand side of the toroidal fusion engineering device (FED), a tokamak.

1.25 cm (thickness) and are brazed to a copper sleeve attached to the core structure. The first wall at the top, bottom, and in-board sides are protected from sudden high heat loads (or from neutral beam shine-through, should beams rather than RF be used for heating) by

for In-Vessel-Components

STARFIRE (ANL et al.,

CONFIGURATION

ELEVATION

in Recent Reactor

Designs

INTOR (IAEA, 181)

FED (U.S., '81)

PCA (SS)

316 SS

316 SS

Disruption/Beam Armor

None

None Required

Coating

Be

None

TiC on C

Ripple/Runaways Armor

None

316 SS Panels

316 SS Panels

Cu, Nb, V, Ta

SS (N/A)

316 SS

First Wall

Limiter

Structure

'80)

(?)

Graphite Tiles (?)

Armor

None

__

Graphite Tiles

Coating

Be

__

TiC on C

Divertor Armor

Structure

N/A -_

316 SS W Tiles (No Coating)

N/A --

R. W. Conn /Surface

9

interactions to first-wall and IVC design

.

y

OH 8. EF COILS CRYOSTAT -CONTAINMENT

Figure 2

: Perspective views of the mechanical "pump limiter"

in FED.

graphite tiles (15 cm x 15 cm x 5 cm thickness) mechanically attached to the 316 stainless steel first wall. An illustration of this disruption armor is given in Fig. 3. Stainless steel watercooled panels of 0.8 cm thickness are used as armor for the outboard first wall to protect it against the effects of runaway electrons or local magnetic-ripple-induced ion transport.

DIVERTOR COLLECTOR PLATES Figure 4

om SCALE

(ml

: Cross section view of INTOR, a tokamak reactor.

Figure 3

: Perspective view of the in-board, top, and bottom disruption protection armor for the first wall in FED.

INTOR, a tokamak reactor developed by an international design team [3], is shown in cross section in Fig. 4. In contrast with FED, it incorporates a magnetic divertor which guides particles to specially-designed divertor plates at the The pump duct is also bottom of the device. The divertor collector plates have tungshown. sten tiles attached to a water-cooled stainless steel backing plate, as illustrated in Fig. 5. The first wall is a bare, water-cooled, stainThe panel is made thicker less steel panel. (up to 1.56 cm) in regions where higher erosion is expected (in-board first wall and beam shine-

Figure 5

: perspective view of the divertor collector plate and the tungsten protection armor in INTOR.

tile

through areas). For STARFIRE [5], a design developed as a conceptual commercial reactor, the first wall is an advanced stainless steel a lloy, The designated PCA, with a beryllium coating. cross sectional view of STARFIRE is similar to

10

K. W. Conrz I Suyface

-2Ocm

-r

interactions

to first-wall

and IVC desigrr

------7Ocm

SHIELD

SECTION A-A

Figure 6

: Cross section view of the toroidal belt limiter design for the STARFIRE tokamak [S].

that of FED, except for dimensions and the deAs shown in sign and placement of the limiter. Fig. 6, the limiter is a water-cooled toroidal belt placed at the mid-plane on the outside. Particles neutralized on the central stem are pumped from the rear of the limiter. A beryllium coating is used on the limiter but no protective tiles are employed. Summarized in Table 2 is a generic list of invessel-components and typical materials which have been selected in designs for each componPractically all near-term reactor designs ent. have used stainless steel as the construction TiC is the material and water as the coolant. most widely selected coating, particularly for graphite or steel, while graphite or refractory metals are selected for protective tiles. 3.

PLASMA AND SURFACE

HEAT LOADINGS

Particles and electromagnetic radiation produce a surface heat load on IVC’s and interact with the surface in ways that affect low and high Z impurity production, recycle of hydrogen gas, Mechanisms and gas trapping and permeation. which can specifically lead to these effects are manifold and are extensively discussed in reference [l]. We have previously described [17] the plasma and atomic physics that is genFrom comerally included in plasma models. puter codes based on these physics models, one

Table 2.

In-Vessel-Components

Component

Typical

Materials

First Wall

SS, Ni; Al, Ti, V, MO, Nb Alloys

Limiter

SS, Cu, MO, Nb, Ni Alloys

Armor-Tiles

Refractory Metals MO, W, Graphite, Sic

Coatings

Tic, TiB2, Be, others

can calculate the surface loading on components during a typical plasma burn. Actual loads will depend on specifics of the design (limiter vs. divertor, RF vs. beam heating, protective tiles vs. bare walls, etc.). As listed in Table 3, the loads consist of particles in the form of plasma ions or energetic neutral atoms (which result from neutral gas penetration, charge exchange, and transport in the plasma), electrons, and various forms of electromagnetic radiation, depending on the composition and temperature profiles in the In addition, loss of plasma confineplasma. ment may result from an external system failure An example of or some natural plasma process.

R. W. Conn / Surface interactions to first-wall and IVC design

Table 3.

Nature of the Heat and Particle Load Environment for In-Vessel-Components

11

tion transports most of the energy from the plasma to the wall. The high value of 250 W/cm2 may prove to be more typical.

Particles

Ions,

mainly

H (or D and T), also C, 0, high

Z Cold Edge: T1 T lo-100 eV Hot Edge:

T, c 100-1000

Neutrals from charge D and T)

eV

exchange--mainly

H

Cold Edge:

Maxwellian,

Teff c 100 eV

Hot Edge:

Maxwellian,

Teff ?J 2-5 keV

(or

Radiation Line, Recombination, tron Off-Normal

Bremsstrahlung,

Condition--e.g.

Ions to limiter/diverter;

Synchro-

Disruption induced currents

the latter is the disruptive instability in the tokamak. (For an overall review of the tokamak, including a general discussion of the disruptive instability, see reference [18].) In this case, heat typically flows as plasma mainly to the limiter in a very short time (~2-10 ms for machines the size and design of FED or INTOR). The loss of plasma current induces current to flow in near-plasma components and can lead to strong (J x B) mechanical forces. Careful design is required to insure that eddy current paths are broken up into small emements or that structural support is adequate. A summary of the surface loading environment during the normal burn pulse for the three tokamaks being used as illustrations is given in Table 4. The rate of particle flow to the limiter in STARFIRE is low because, in their analysis, a broad plasma temperature profile and high plasma edge temperature are.assumed. This is unlikelv to be the case and the FED and INTOR rates and ledge temperatures should be taken as representative. The erosion rates listed are based on the assumption that sputtering is No net erosion of the caused by D or T ions. beryllium coating in STARFIRE results from assuming that an equilibrium will be reached in which the rate of erosion by sputtering is balanced (in space and time) by redeposition of sputtering atoms. The rate of neutral atom bombardment listed for INTOR is likely to be typical of FED as well (with a lower effective neutral spectral temperature) although the latter case is being calculated. The total heatinq rate of kev IVC’s given in Table 4 includes electromagnetic radiA range is given for FED. The low value ation. (20 W/cm2) corresponds to the case where radia-

The heating rate along the limiter is a function of the angle between the flux surface and the limiter itself. General expressions have been developed previously [19,20]. The variation of heat flux across the FED limiter is shown in Fig. 7 r21.221. For the divertor taraet olates in INTOR; higher heating rates (4001800'W/cm2) are typical because the heat is channeled from the plasma along field lines o&side the separatrix.

In addition to the loading during a normal plasma burn, the plasma may infrequently dump its energy rapidly to in-vessel-components (typicallv to the limiter). The disruotive instability is a good example. Based on experimental data and physics modeling [23], characteristic parameters of a disruption for the FED have been developed and are given in Table 5. The available energy of 80 MJ is divided between the limiter and the wall as indicated. A peaking factor of 2 is suggested for the limiter loading. This characterization remains uncertain because of uncertainties in the model of the disruptive instability. Thus a range is given for all parameters associated with both the thermal quench (which occurs first) and the plasma current decay (which is slower because of inductive effects). 4.

CRITERIA FOR SELECTION OF FIRST WALL AND IVC MATERIAL AND DESIGN SELECTION

Criteria for selecting in-vessel materials and for designing a component are based on the function of the component, its location relative to the plasma, and the nature of the surface loading. Criteria for selection of blanket structural materials has been discussed by the author in two orevious papers r15,16]. The first wall, limiters, and'armor for beam shinethrough, ripple loss, or runaway electrons, all Impurities generated from face the plasma. these surfaces are ionized and enter the plasma edge. Control of PWI is required for reasons of plasma performance and surface erosion in all devices from the tokamak, stellarator, and elmo bumpy torus to the tandem mirror or other openended machines.

In devices such as magnetic mirrors, the dominant plasma-wall interaction takes place beyond the ends of the main confinement region. The types of surface loadings, particularly of particles and heat, is thus shifted to a direct converter or end-plate where surface erosion, control of hydrogen isotope refluxing and, most importantly, control of secondary electron emission are critical [24]. The surface heat load on the first wall is typically low (5 to 20 W/ cm2) in TMR's, erosion is essentially negligible, and sudden heat dumps are improbable (since plasma flow would be out the ends).

I?. W. Corm / Surf&a inteructions to first-wall und I VC design

Table 4.

Normal Surface STARFIRE (ANL et al.,

Loading

'80)

Environment INTOR (IAEA, 181)

FED (U.S., '81)

Ions Rate (s-l)

3.5x10z2 (to limiter)

5.5x1023 (to divertor)

5x1o23 (to limiter)

Ion Temp (eV)

1200

400

16-160

Erosion

0 (W. redeposition)

9.7 mmly For W tiles

3-70 mm/y For G tiles

6.5~10'~

1.6~10~~ (in throat)

N/C

5000

200

N/C

First Wall or Outboard Armor

80 (1.4 peaking)

11

30

Limiter

335 (peak) 90 (ave.)

30

20-250 (cold-hot edge)

N/A

400-800 (inner-outer plates)

N/A

Rate (mm/y)

C-X Neutrals Rate (s-l)

Effective

Temp (eV)

Total Surfase Heating Rates (W/cm )

Divertor

Plates

Table 5.

Thermal

Quench

Parameters

for Plasma Disruptions

Nominal

Value

in FED Range of Value

Time, ms Energy, MJ To limiter To first wall

5 80 75% 25%

2-10 70-90 50-100% O-50%

Peaking Factor Limiter First Wall

2 1 (uniform deposition)

l-3 1

10 20 30 Inboard, 20% 2

5-10 10-30 O-60 __ lo-30% l-3

Current

Quench

Time, ms Thermal energy, MJ Magnetic energy, MJ Region of deposition Extent of region Peaking factor

top, & bottom

R. W. Corm /Surface

interactions to first-wall and IVCdesign

13

,210

-200

-160

- 100

-

so

-80 POLOIIYd_ POSITION

20 CMFOR Figure 7

FROM LIMITER

MIDPlANE,

CM,

TOFJlII39L DISTANCE FROM POINT OF PLASM4 TANGENCY IS

HEAT FLUX AND 30 CM FOR PARTICLE ‘FLUX,

: Variation of the heat flux along the face of the toroidal pump-limiter mum between the two maxima arises one point along it [Zl].

because the plasma is

A list of general criteria is given in Table 6. Surface characteristics are clearly crucial since they will determine the rate of impurity production and hydrogen isotope recycle. Physical characteristics will determine the response of the material to the loadings described in the previous section. Specifically, one desires high values of thermal stress parameter, specifit heat, and melting point. Thermomechanical properties and radiation damage characteristics determine the performance of the component as a function of time and, ultimately, its lifetime. For nearer-term reactors such as FED, thermomechanical properties deserve a higher priority than radiation damage characteristics. For demonstration and commercial reactors, the ordering is reversed [15,16]. Compatibility of materials and coatings with coolants is impor-

tangent

in FED. The minito the limiter only at

tant primarily because of corrosion, transport of radioactive corrosion products, the implications of such transport for secondary component performance, accident hazards and remote maintenance, and failure mechanisms caused by corrosion. Fabricability is crucial to all applications but joining criteria vary with component function. For example, armor tiles of graphite or tungsten may be mechanically attached to the first wall, limiter or divertor plate, or they can be brazed to a substrate. Joining by welding is not required. For a structural material, the joining method may be a critical issue, as it is for molybdenum alloys.

14

R. W. Corm / Surface

Table 6.

1.

interactions

Criteria Affecting Choices of First Wall and In-Vessel-Component Materials and Design

Surface

Characteristics

Sputtering Atomic

Yield

Properties

Recycle-Desorption-Retention Chemical

Activity

Arcingand Coatings

Physical

Electron

Emission

Properties

Characteristics

Thermal (aE)

Stress

Specific

Heat

Melting 3.

Properties

and Properties

Secondary

2.

with H, C, 0

Blistering

Parameter,

M = Eayk(l-v)/

Point

Thermomechanical

Properties

and I VC design

Criteria for coatings on materials include bonding strength, chemical activity, sputtering yield, thermal shock resistance, effective Z of the coating compound, and the availability of a process for in-situ recoating. 5.

Number

H/He Reflection

to first-wall

RESPONSE COATINGS

OF IN-VESSEL-COMPONENTS

TO SURFACE

The criteria just discussed have been used in making the decisions indicated in Table 1 regarding materials for the first wall and IVC’s in the FED, INTOR, and STARFIRE studies. (For INTOR and especially for FED, the designs remain at an early stage and present choices can change as new results from analysis or experiment are developed.) We present in this section calculational results on the response of particular in-vessel-components which will illustrate the range of conditions that materials The remust show for acceptable performance. sults are taken from the recently published U.S. contribution to the international INTOR study [3]. Many of the results are also relevant to the U.S. FED project and have been developed by some of the same researchers.

Uniform and Total Elongation Yield and Ultimate

Tensile

Fracture Toughness,

Strength

Crack Growth

Creep Strength Cyclic 4.

Fatigue Characteristics

Radiation

Damage Characteristics

Swelling Embrittlement Radiation 5.

Creep

Compatibility Corrosion, Hydrogen

with Coolants

and H Isotopes

Stress Corrosion

Cracking

Permeability

6.

Fabricability

7.

Industrial

8.

Induced Radioactivity,

The time-dependent temperature response of the tungsten tiles chosen to protect the divertor plate structure in INTOR is shown in Fig. 8. The surface temperature reaches approximately ZOOO"C, at which point radiation alone transfers 100 W/cm2 away from the surface. With subsequent pulses,-the peak temperature would not change much but the.minimum temperature between pulses will be hither than 1100°C. The temperature swing between-dwell and burn is nevertheless substantial. The thermal stress levels developed in the tiles depends on whether or not the tile edges are constrained against rotation. Results are shown in Fig. 9. Reversal of thermal stresses during dwell or shutdown is the major cause of cyclic fatigue in the tile.

and Joining

Capability,

Data Base, Cost Afterheat

Levels

The existence of an industrial capability to produce and manufacture the material, the data base on material characteristics and performance (especially in a radiation environment), and cost are important criteria for both nearterm and potential commercial designs. Neutron-induced-levels of radioactivity and afterheat can be very critical in choosing a material, particularly for IVC’s where maintenance For and frequent replacement is anticipated. example, the lifetime predictions for the graphite armor tile in FED may require annual replacement. It will be most advantageous if the materials associated with the limiter have low radioactivity levels at the end of 7-30 days after removal from the reactor.

100 0

,

I

I

I

I

40

80

120

160

200

TIME

Figure 8

240

(SECONDS)

: Variation of the tungsten tile surface temperature during the first pulse of the INTOR reactor [3].

R. W. Corm / Surface interactions to first-wall and IVC design

a = 300 w/w42

EDGES CONSTRAINED AGAINST

FREE TO

L_I THERMAL STRESS (MPa)

Figure 9

in a 2.5 cm thick tungsten tile as a function of whether tile-edges are allowed to freely rotate or are constrained in place [3].

TIME ISECI

: Thermal stress distribution

We summarized in Table 5 the nominal parameters of a plasma disruption in the FED reactor. For INTOR, a longer characteristic time for thermal quench was assumed, namely, 20 ms. This was based on a less complete physics analysis than is available now. Nevertheless, parametric studies performed for INTOR are illustrative and important conclusions can be drawn from them. For example, the response of an unprotected stainless steel wall to a 20 ms thermal quench disruption assuming different values for the total energy deposited per unit area is shown in Fig. 10. The particular value of energy deposition (joules/cm2) will depend upon the fraction of the first wall receiving the energy dump. For INTOR, the in-board first wall is assumed to receive essentially all the thermal energy. Recent tokamak experiments show that most (50-75%) of the thermal energy during a disruption is deposited on the limiter. (It is not clear if a divertor plate would also receive the bulk of the thermal energy but this is likely.) Therefore, even though the thermal quench time may be shorter than is assumed in INTOR, the maximum surface temperature of the first wall may still be below the melting point. The INTOR group argues that if melting does not occur, a bare wall (no armor) is the simpler and preferred wall-design approach. The limiter or divertor plate, on the other hand, may be subjected to very high pulse loads in For FED. the limiter is about 40x104 th's case. cm j so that the 80 MJ which is assumed to reach the limit r leads to an energy load of 270 J/cm2 (540 J/cm % for a peaking factor of 2). For the response of a graphite tile to a INTOR [3 200 J/cm 6 dump over 20 ms is shown in Fig. 11. Thermal conduction and radiation play an important role in limiting the maximum temperature. If the thermal quench time is 2-5 ms, less heat can be dissipated by conduction and melting be-

Figure 10

0.000

Figure 11

: The resp nse of graphite. tile to a

ZOO J/cm f! disruption thermal load where the energy deposition is uniform over the tile for 20 ms [3]. The thermal disruption time constant may be as short as 2-10 ms.

0.005

0.015 0.010 TIME ISECI

0.020

0.025

: Response of a bare stainless steel first wall panel to various energy fluxes incident for 20 ms [3]. The thermal disruption quench time may be as short as 2-10 ms but experimental indications are that more than 25% of the thermal energy will flow to the limiter rather than the wall.

low the surface. This will lead to significant material loss by sublimation and an alternate design approach may be required. A complete assessment is not yet available. Erosion of surfaces by sputtering has been calculated for INTQR and FED. The gross rate ranges from 3 to 74 mm/y for the FED limiter, depending upon assumptions'about plasma edge conditions. These estimates include the contribution from periodic disruptions and chemical

16

N. W. Corm

/ Surfhce

interactions

erosion. Graphite is subject to erosion by methane production when the graphite temperature is 400-800°C and acetylene-production when the temperature exceeds about 1500°C ill. Tunasten may be subject to erosion by oxidation at" high temperature and experimental data is required. This is an area of critical need. 6.

SUMMARY AND SOME KEY OPEN PROBLEMS

We have described the surface loadings from plasma-wall interactions and their influence on the design and selection of materials for the first wall and IVC's in magnetic fusion reactors. Criteria usually examined before selecting a material or coating have been summarized in a manner similar to earlier work regarding selection of blanket structural materials [15,16]. The nature of the surface loads and the illustrations given of component response as a function of loading type (normal vs. disruption) implicitly suggest the major problem areas requiring additional work. Recent research and the results reported here have, however, produced a set of key near-term issues requiring resolution. It is appropriate to close this paper with a brief discussion of such problems. Experimental data on the chemical reactivity of hydrogen with bare or coated graphites and of oxygen with refractory metals like tungsten under conditions simulating the true plasma environmentarr! needed. Measurements with graphite have been summarized by McCracken and Stott [1] and by Bohdansky [25] (who provides a comprehensive summary of physical sputtering). The role of particle bombardment at plasma-edge energies simultaneous with temperature increases has not been directly assessed. Coatings may be used on tiles on the first wall or limiter to control chemical activity and to provide a low-Z material facing the plasma. Mechanical and thermal properties of these coatings are required along with techniques (and tests thereof) for redepositing coatings. For tiles themselves, a key engineering question relates to one or more methods for attaching them to a substrate structure. Hydrogen and helium reflection and desorption properties are essential to understanding gas recycling and the enhanced pumping characteristics associated with ballistic scatterinq of ions into pump ducts associated with pumplimiters [22,26]. A further problem relates to diffusion-of-hydrogen (particularly tritium) through in-vessel structures and first walls. Recent results C271 suqqest that tritium permeability in stainless sfeel can be significantly hither when the hvdroaen (tritium) is imolanted via energetic particle bombardments Resolution of this question is crucial to evaluating safety criteria, tritium inventory, and tritium leakage rates. A similar problem relates to tritium trapping in graphite, which, with the wide use of tiles, could lead to a very large in-vessel inventory.

to Ji‘rst-wall ad

11’c‘des~g1r

Finally, two design approaches for the first wall are to protect it with armor tiles or to design a bare wall thick enough to withstand off-normal loads. Knowledge of the response of both tiles and stainless steel panels to very high, short pulse (l-10 ms) heat loads is needed before a design decision is possible. For bare steel panels, the formation of a thin melt layer may occur. Radiation from the plasma in front of the surface during a disruption has not been assessed and will lower the peak energy flux by dissipating energy away from the affected surface. Further, recent experimental results [28] show that most of the thermal energy in a disruption goes to the limiter rather than the first wall. This may allow first wall designs without armor to be used but will make the desiqn of protection for the limiter or divertor-collectorplate even more difficult. A resolution based upon data from both plasma and high-heat-flux simulation experiments and their concomitant analysis is required. ACKNOWLEDGMENT The author would like to thank colleagues in the community associated with FED, INTOR, and STARFIRE research for their helpful input and comments. This research has been supported by DOE Contract DE-AS03-76SF00034. REFERENCES 1.

McCracken, G. M., Stott, P. E., Nucl. Fusion 19 (1979) 889.

2.

Jassby,

3.

"U.S. INTOR, The U.S. Contribution to the Int. Tokamak Reactor Phase I Workshop, Conceptual Design", USA INTOR/81-1. Vols. I and II (June 1981).

4.

Fusion Engineering Design Center, Oak Ridge Nat. Lab., private communication (1981).

5.

"STARFIRE - A Commercial Tokamak Fusion Power Plant Study", Joint Report of Argonne Nat. Lab., McDonnell-Douglas Astro. Co. - East, General Atomic Co., and Ralph M. Parsons Co., Argonne Nat. Lab. Report ANL/FPP-80-1 (1980).

6.

Behrisch, R., Kadomtsev, B. B., in Plasma Physics and Controlled Nuclear Fusion search 1974, Vol. II, IAEA, Vienna (1975) 229.

7.

Meade,

8.

Conn, R. W., Kesner, J., Nucl. Fusion 15 (1975) 775.

9.

TFR Group, (1976).

10.

0. L., Nucl. Fusion 17 (1977) 309.

D., Nucl. Fusion 14 (1974) 289.

Fontenay

Report EUR-CEA-FC-861

Jensen, R., Post, D. E., Nucl. Fusion 17 (1977) 1187.

R. W. Conn /Surface

interactions

11.

Owens, D. K. et al., J. Nucl. Mat. 93/94 (1980) 213.

12.

Gibson, A., J. Nucl. Mat. 76/77 (1978) 92.

13.

Gordiner, M. R., Conn, R. W., J. Nucl. Mat. 93/94 (1980) 420.

14.

Neuhauser, J. et al., "Self-Limitation of Impurity Production by Radiation Cooling at the Edge of a Fusion Plasma", ZEPHYR Report No. 23 (April 1981).

15.

Conn, R. W., J. Nucl. Mat. 76/77

(1978) 103.

16.

Conn, R. W., J. Nucl. Mat. 85/86

(1979) 9.

17.

Conn, R. W., Kesner, (1976) 1.

18.

Furth, H. P., Nucl. Fusion 15 (1975) 487.

19.

Schmidt, J., Comments Fusion 5 (1980) 225.

20.

Conn, R. W., Sviatoslavsky, I. N., Sze, D. K., "Limiter Pumping System for Divertorless Tokamaks", in Proc. 8th Symp Engr. Problems of Fusion Research, IEE:' (1979).

21.

Prinja, A. K., Grotz, S. P., Conn, R. W., UCLA private communication (1981).

22.

Ulrickson, M., "A Preliminary Conceptual Design of a Pump Limiter for FED", to be published as part of U.S. Fusion Engineering Design Project (1981); also Prinja, A. K ., Grotz, S. P., Talmadge, S., Conn, R. W., "Large Area Pump-Limiters for Tokamak Reactors: Applications to FED", ibid.

23.

Carreras, B., Oak Ridge Nat. Lab. and Peng, Fusion Engineering Design Center, priM vaie communication (1981).

24.

Post, R. F., J. Nucl. Mat. 76/77 (1978) 112.

25.

Bohdansky, 44.

26.

Talmadge, S., Taylor, R. J., Bull. Am. Phys. Sot. 25 (1980) 1033; also Overskei, D., Phys. Rev. Letts. 46 (1981) 177.

27.

Wienhold, P., Profant, M., Waelbroeck, F., Winter, J., J. Nucl. Mat. 93/94 (1980) 866.

28.

~%~~~;c~tron

J., J. Nucl. Mat. 63

on Plasma Phys. Cont.

J., J. Nucl. Mat. 93/94 (1980)

General Atomic Company, (July 1981).

private

to first-wall

and IVC design

17