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Impurity control in the varennes tokamak
332
Journal
IMPURITY
CONTROL
IN THE VARENNES
of Nuclear
Materials I I1 & I I2 (19X.3) 332-334 North-Holland Publishing Company
TOKAMAK
C. BOUCHER ‘, P. COUTURE ‘, V. FUCHS ‘, J. GEOFFRION ‘, B.C. GREGORY T.W. JOHNSTON ‘, I.P. SHKAROFSKY ‘, M.M. SHOUCRI ‘, B.L. STANSFIELD B. TERREAULT ’ and J.C.A. VITAL1 ’
‘, ‘.
I INRS-Energie, Universit6 du Quibec. C.P. 1020. Vurennes. QuPbec. Canada. JOL 2PO .’ Institut de recherche d’HvdmQu6bec C.P. 1000, Varennes, QuPbec, Cunadu, JOL -7PO .I MBP Technologies In<,., C.P. 160. Ste-Anne-de-Bellecue. QuPbec, Canodu. H9X 3L5
1. Machine concept
ity line radiation).
The Tokamak de Varennes (TV) will be devoted to study quasisteady state machine technology during long pulses. It is designed for 30 s long periods of high duty cycle operation, repeatable every 15 min (“mode C” operation [I]). The plasma current is to be reversed within a short interval (variable from 8 to 40 ms). every 0.1 to 1.0 s (depending on flux consumption). Impurity control over the 30 s period is an essential objective of the program. and for this purpose the TV will have two symmetrically positioned poloidal divertor triplet coils [I]. The experience of Wesner et al. [2] on ASDEX showed the importance of initiating the plasma far from material walls to obtain clean plasmas. The TV has a large (0.2 m diameter) centrally located region of low poloidal field (G 2 X lop3 T) and a multipole start-up field provided by a carefully designed OH circuit and by the divertor triplets. A powerful positional feedback control system is also essential to avoid wall contact. It is designed to centre the plasma as much as possible during ramp-down and current reversal. Horizontal positional feedback control is provided by internal coils [I] having an instantaneous effect on the plasma. The advantages of feedforward [3] have been demonstrated on ASDEX and will be incorporated on the TV. For divertor operation, the equilibrium field has been carefully designed [4] so that the separatrix clears the divertor coil casing by 2 cm. The theory for the behavior of the plasma in the TV is being developed [5,6]. A I-D tokamak diffusion code, Vartok, which incorporates neo-classical ion and anomalous electron heat transport, and anomalous particle loss, is calibrated to describe normal tokamak operation during the current plateau (Alcator-scaling for the energy confinement time, ohmic heating, impur-
0022-3115/82/0000-0000/$02.75
Neutral particle dynamics are treated using FRANTIC [20]. During current cross-over, j = 0 and q = 00 surfaces develop within the plasma crosssection. In the simplest diffusion code this situation is approximated by using a constant form factor for the current so that j - 0 and q - cc globally over the crosssection as the current goes to zero. Hence, in the model, ohmic heating is underestimated while diffusion and drift losses are overestimated. With MHD equilibrium assumed, and with gas puffing, the diffusion issue is whether during cross-over the electron temperature will drop below the “radiation barrier” level. For instance it is calculated that with 10% oxygen impurity the radiated power loss exceeds the power input for 100 kA plasma current for T, between 20 and 58 eV [6] (see fig. 1). This would necessitate reignition at each cross-over. Our results so far indicate that such a catastrophic heat loss does not occur, the temperature staying above 50 eV. even at low values of plateau current (100 kA). provided
Fig. I. Radiation barrier (arrows) for 100 kA plasma current and 10% oxygen. Poh is the ohmic power and P, is the power loss (radiation +diffusion) dominated by radiation for T G 100 eV.
C. Boucher et al. / Impurity
that (e.g. mize end TV.
the impurity concentration remains low enough B 5% oxygen). It is therefore imperative to minithe amount of impurities in the plasma. To this thorough cleaning and gettering will be used in the
2. Discharge cleaning In order to reduce the amount of low Z impurities (C and 0) in the plasma, the liner of the TV will be discharge-cleaned with hydrogen plasmas, using ECR (2.45 GHz, 0.0875 T) CW discharge for “rinsing” and tokamak-type discharges for “scrubbing”. During “rinsing” the hydrogen ions and neutrals from the plasma react with the weakly bound carbon and oxygen of the surface, forming CH, and H,O. These may be evacuated by the pumping system provided that the electron temperature is low enough (T, < 4 eV). During “scrubbing”, higher energy H ions bombard the wall and reduce oxides of greater chemical stability or more deeply implanted. Because of the higher T,, the H,O molecules are mostly dissociated and the oxygen is readsorbed on the wall with relatively lower binding energy. Alternate rinsing and scrubbing will therefore result in thorough cleaning by increasing the power in the “scrubbing” plasma until it becomes a high power tokamak discharge. lnconel was chosen for the liner; nickel oxide has lower chemical stability than the chromium oxide present in stainless steel. In view of recent results obtained by other research groups [7,8], the liner will be maintained at 300°C during discharge cleaning. An ECR CW plasma has been compared to Taylor discharge cleaning in JFT-2 [9] where it was found that the two methods lead to equivalent cleaning effects. The advantages of an ECR plasma are its low operation cost (using magnetrons in the industrial heating band) and a better control of the plasma parameters. The power requirement to maintain an ECR plasma may be estimated as in ref. 9, using the power balance equation with the added assumption that the ions and electrons are lost to the walls by curvature drift [lo]. We get a requirement of - 25 kW for the TV for a filling pressure of = 10m3 mbar. The neutral and ion fluxes are estimated to be of the order of lOI and lOI cm-* s-’ respectively, showing that the walls would be cleaned uniformly, mainly by the neutrals. In addition, the outside of the liner and the vacuum vessel will be cleaned by glow discharge. This can be done with the liner hot (- 300°C) and the vessel and accessories (ports, etc.) at moderate temperature (- 1OO’C) to avoid con-
control in the Varennes
333
densation on cold surfaces, and to avoid arcs, as recommended by Winter et al. [8].
3. Gettering in the divertor region Strict control of the hydrogen density is also necessary for plasma sustainment during reversal [6]. In addition to fast gas puffing, we plan to achieve this through (1) quick but reduced wall recycling and (2) high thoughput pumping. Indeed, the hot (3OO’C) Inconel liner will speed up thermal recycling but it will contain at any instant less surface hydrogen subject to desorption than the room-temperature saturation dose of - 1 X 10” H/cm2 [21]. Concerning pumping, Princeton [ 1 l] and DOUBLET [ 121 groups have demonstrated a density enhancement in the divertor region, which allows high throughput at modest pumping speed. However, the effect depends nonlinearly on central plasma parameters and it is not certain that the proper conditions would be obtained in the TV. Flexibility then dictates that high speed pumping be available. Furthermore, ASDEX experience [ 131 has shown the importance of divertor gettering for impurity control. Out of the three possibilities: Ti surface gettering, cryopumping, and Zr/Al bulk gettering, the last appears at this stage to fulfill best all the requirements and constraints which follow: hydrogen throughput of 50 mbar l/s; pumpdown time from 1O-4 to 10e6 mbar of 1 s; pumping speed of lo5 l/s; hydrogen sorption capacity (reversible) 3 1 day of operation; regeneration time< a few hours; impurity sorption capacity (irreversible) > 1 year; compatibility with thermal loads 2 10 kW; compatibility with high density (lo-’ mbar) discharge cleaning; immunity to desorption by fast atoms reflected by the neutralizer plates (the relevant desorption yields [ 14,151 are of the order of unity for monolayer coverage); and resistance to thermal stresses and magnetic forces. The design calls for 80 SAES getter modules, located -20 cm from the neutralizer plates, and operating at 300-350°C. During discharge cleaning after atmospheric exposure, they will be kept cold and passivated. However during routine discharge cleaning they will operate at 475°C to avoid hydrogen saturation and embrittlement, but will still getter impurities thus speeding up cleaning. 4. Impurity studies In spite of all the precautions taken, impurities will be generated, especially during reversal. To understand
334
C. Boucher et al. / Impurity
control in the Varennes
sources. The direct shielding factor of the scrape-off layer is sensitive to f,, and can be even more so if perpendicular diffusion is important. Results of this code coupled to the various impurity diagnostics will form the basis of the impurity studies in the Tokamak de Varennes.
References [I] P. Couture,
6 cm/dbv To
plate -
Fig. 2. Contour lines of the probability of ionization (arbitrary units) in the scrape-off. (Z = 1 lines for ionization from charge state Z=O to Z=l and Z=3 lines for Z=2 to Z=3.)
the transport of wall-generated impurities in the scrapeoff region, a 2-D Monte Carlo code is being developed. The code will be used for calculating the shielding efficiency of the scrape-off layer to injected impurities, their charge state distribution at the divertor plate, and for simulating laser blow-off experiments. The preliminary results obtained simulate the continuous injection of heavy impurity atoms (Fe) from a point source into the scrape-off region (see fig. 2). The plasma density distribution in the scrape-off layer was derived form a simple 1-D transport model assuming perpendicular Bohm diffusion [16], and including the effect of ionization by electrons [ 171. The temperature is taken as constant through the boundary layer. The injected neutral particles are randomly selected from a cosine distribution, with an energy spectrum similar to that measured for sputtered atoms [18]. A neutral particle is injected at a fixed point with velocity components uL and u ,, and is followed using a pseudocollision algorithm [19] until it reaches a collision point where it is ionized. The ion then moves parallel to the magnetic field with its guiding center having a parallel velocity equal to the flow velocity f,,C,, where f,, is the flow velocity factor and C, is the ion sound speed. ( f,, = 0.1 [ 111). The ion is followed with its constant velocity through succesive states of ionization. This simulation indicates that Fe ions of high charge state (Z = 5) are created in the scrape-off before reaching the neutralizer plate. For the case of an extended source representative of wall sputtering, the charge state distribution can be obtained by summing over a distribution of such point
[2] [3] [4] [S] [6]
[7] [8] [9]
[IO]
[ 1 l] [12]
[ 131 [l4] [l5]
[ 161
[17] [I81 [19]
[20] [21]
J. Geoffrion. B.C. Gregory. H.W.H. Van Andel, R.A. Bolton, P.B. Cumyn, B.L. Stansfield and B. Terreault, 9th Symp. on Engineering Problems of Fusion Research, Chicago (October 198 1). F. Wesner, H. Rapp and F. Schneider, in ref. 1. F. Schneider, in ref. 1. P. Couture, J. Geoffrion and B.C. Gregory, Internal Report TV RI-87f, (October 1981). I.P. Shkaforsky and M. Shoucri, J. Plasma Phys. 26. part 3 (1981). V. Fuchs. M. Shoucri. G. Thibaudeau, T.W. Johnston and I.P. Shkarofsky, 23rd Annual Meeting, Div. of Plasma Phys., APS (October 1981). TFR Group, J. Nucl. Mater. 99 (1981) 334. J. Winter, F. Waelbrook, B. Brandt, K.J. Dietz, 1. Ali-Khan and P. Wienhold, J. Nucl. Mater. 93/94 (1980) 812. Y. Sakamoto, Y. Ishibe, K. Yano, H. Oyama. Y. Tanaka, N. Fujisawa, Y. Matsuzaki, N. Suzuki, K. Maeno, T. Yamamoto and K. Yokokura. J. Nucl. Mater. 93/94 (1980) 333. C. Boucher, B. Terreault, Internal Report, INRS-Energie No. 149 (June 1981). M. Petravic, D. Post and D. Heifetz. J. Schmidt, Phys. Rev. Lett. 48 (1982) 316. M. Ali-Mahdavi, J.C. DeBoo, C.L. Hsieh, N. Ohyabu, R.D. Stambaugh and J.C. Wesley, Phys. Rev. Lett. 47 (1981) 1602. M. Keilhacker et al., IAEA-CN-38/01, p. 35 1. R. Bastasz and L.G. Haggmark, J. Nucl. Mater. 103 & 104 (1981) 499. A. Sagara, K. Akaishi, K. Kamada and A. Miyoharu. J. Nucl. Mater. 103 & 104 (1981) 357. M. Keilhacker, in: Tokamak Reactors for Breakeven, Int. School of Fusion Reactor Technology, Ettore Majorana Centre, (September 1976). W. Lotz, 2. Phys. 220 (1969) 466. R.B. Wright, M.J. Pellin, D.M. Gruen and C.E. Young, Nucl. Instr. and Meth. 170 (1980) 295. G.G. Lister, D.E. Post and R. Goldston, 3rd Symp. on Plasma Heating in Toroidal Devices, Grenoble (1976) p. 303. H. Towner, Princeton Plasma Physics Laboratory. private communication. W. Bauer, J. Nucl. Mater. 76 & 77 (1978) 3.
Journal
IMPURITY
CONTROL
IN THE VARENNES
of Nuclear
Materials I I1 & I I2 (19X.3) 332-334 North-Holland Publishing Company
TOKAMAK
C. BOUCHER ‘, P. COUTURE ‘, V. FUCHS ‘, J. GEOFFRION ‘, B.C. GREGORY T.W. JOHNSTON ‘, I.P. SHKAROFSKY ‘, M.M. SHOUCRI ‘, B.L. STANSFIELD B. TERREAULT ’ and J.C.A. VITAL1 ’
‘, ‘.
I INRS-Energie, Universit6 du Quibec. C.P. 1020. Vurennes. QuPbec. Canada. JOL 2PO .’ Institut de recherche d’HvdmQu6bec C.P. 1000, Varennes, QuPbec, Cunadu, JOL -7PO .I MBP Technologies In<,., C.P. 160. Ste-Anne-de-Bellecue. QuPbec, Canodu. H9X 3L5
1. Machine concept
ity line radiation).
The Tokamak de Varennes (TV) will be devoted to study quasisteady state machine technology during long pulses. It is designed for 30 s long periods of high duty cycle operation, repeatable every 15 min (“mode C” operation [I]). The plasma current is to be reversed within a short interval (variable from 8 to 40 ms). every 0.1 to 1.0 s (depending on flux consumption). Impurity control over the 30 s period is an essential objective of the program. and for this purpose the TV will have two symmetrically positioned poloidal divertor triplet coils [I]. The experience of Wesner et al. [2] on ASDEX showed the importance of initiating the plasma far from material walls to obtain clean plasmas. The TV has a large (0.2 m diameter) centrally located region of low poloidal field (G 2 X lop3 T) and a multipole start-up field provided by a carefully designed OH circuit and by the divertor triplets. A powerful positional feedback control system is also essential to avoid wall contact. It is designed to centre the plasma as much as possible during ramp-down and current reversal. Horizontal positional feedback control is provided by internal coils [I] having an instantaneous effect on the plasma. The advantages of feedforward [3] have been demonstrated on ASDEX and will be incorporated on the TV. For divertor operation, the equilibrium field has been carefully designed [4] so that the separatrix clears the divertor coil casing by 2 cm. The theory for the behavior of the plasma in the TV is being developed [5,6]. A I-D tokamak diffusion code, Vartok, which incorporates neo-classical ion and anomalous electron heat transport, and anomalous particle loss, is calibrated to describe normal tokamak operation during the current plateau (Alcator-scaling for the energy confinement time, ohmic heating, impur-
0022-3115/82/0000-0000/$02.75
Neutral particle dynamics are treated using FRANTIC [20]. During current cross-over, j = 0 and q = 00 surfaces develop within the plasma crosssection. In the simplest diffusion code this situation is approximated by using a constant form factor for the current so that j - 0 and q - cc globally over the crosssection as the current goes to zero. Hence, in the model, ohmic heating is underestimated while diffusion and drift losses are overestimated. With MHD equilibrium assumed, and with gas puffing, the diffusion issue is whether during cross-over the electron temperature will drop below the “radiation barrier” level. For instance it is calculated that with 10% oxygen impurity the radiated power loss exceeds the power input for 100 kA plasma current for T, between 20 and 58 eV [6] (see fig. 1). This would necessitate reignition at each cross-over. Our results so far indicate that such a catastrophic heat loss does not occur, the temperature staying above 50 eV. even at low values of plateau current (100 kA). provided
Fig. I. Radiation barrier (arrows) for 100 kA plasma current and 10% oxygen. Poh is the ohmic power and P, is the power loss (radiation +diffusion) dominated by radiation for T G 100 eV.
C. Boucher et al. / Impurity
that (e.g. mize end TV.
the impurity concentration remains low enough B 5% oxygen). It is therefore imperative to minithe amount of impurities in the plasma. To this thorough cleaning and gettering will be used in the
2. Discharge cleaning In order to reduce the amount of low Z impurities (C and 0) in the plasma, the liner of the TV will be discharge-cleaned with hydrogen plasmas, using ECR (2.45 GHz, 0.0875 T) CW discharge for “rinsing” and tokamak-type discharges for “scrubbing”. During “rinsing” the hydrogen ions and neutrals from the plasma react with the weakly bound carbon and oxygen of the surface, forming CH, and H,O. These may be evacuated by the pumping system provided that the electron temperature is low enough (T, < 4 eV). During “scrubbing”, higher energy H ions bombard the wall and reduce oxides of greater chemical stability or more deeply implanted. Because of the higher T,, the H,O molecules are mostly dissociated and the oxygen is readsorbed on the wall with relatively lower binding energy. Alternate rinsing and scrubbing will therefore result in thorough cleaning by increasing the power in the “scrubbing” plasma until it becomes a high power tokamak discharge. lnconel was chosen for the liner; nickel oxide has lower chemical stability than the chromium oxide present in stainless steel. In view of recent results obtained by other research groups [7,8], the liner will be maintained at 300°C during discharge cleaning. An ECR CW plasma has been compared to Taylor discharge cleaning in JFT-2 [9] where it was found that the two methods lead to equivalent cleaning effects. The advantages of an ECR plasma are its low operation cost (using magnetrons in the industrial heating band) and a better control of the plasma parameters. The power requirement to maintain an ECR plasma may be estimated as in ref. 9, using the power balance equation with the added assumption that the ions and electrons are lost to the walls by curvature drift [lo]. We get a requirement of - 25 kW for the TV for a filling pressure of = 10m3 mbar. The neutral and ion fluxes are estimated to be of the order of lOI and lOI cm-* s-’ respectively, showing that the walls would be cleaned uniformly, mainly by the neutrals. In addition, the outside of the liner and the vacuum vessel will be cleaned by glow discharge. This can be done with the liner hot (- 300°C) and the vessel and accessories (ports, etc.) at moderate temperature (- 1OO’C) to avoid con-
control in the Varennes
333
densation on cold surfaces, and to avoid arcs, as recommended by Winter et al. [8].
3. Gettering in the divertor region Strict control of the hydrogen density is also necessary for plasma sustainment during reversal [6]. In addition to fast gas puffing, we plan to achieve this through (1) quick but reduced wall recycling and (2) high thoughput pumping. Indeed, the hot (3OO’C) Inconel liner will speed up thermal recycling but it will contain at any instant less surface hydrogen subject to desorption than the room-temperature saturation dose of - 1 X 10” H/cm2 [21]. Concerning pumping, Princeton [ 1 l] and DOUBLET [ 121 groups have demonstrated a density enhancement in the divertor region, which allows high throughput at modest pumping speed. However, the effect depends nonlinearly on central plasma parameters and it is not certain that the proper conditions would be obtained in the TV. Flexibility then dictates that high speed pumping be available. Furthermore, ASDEX experience [ 131 has shown the importance of divertor gettering for impurity control. Out of the three possibilities: Ti surface gettering, cryopumping, and Zr/Al bulk gettering, the last appears at this stage to fulfill best all the requirements and constraints which follow: hydrogen throughput of 50 mbar l/s; pumpdown time from 1O-4 to 10e6 mbar of 1 s; pumping speed of lo5 l/s; hydrogen sorption capacity (reversible) 3 1 day of operation; regeneration time< a few hours; impurity sorption capacity (irreversible) > 1 year; compatibility with thermal loads 2 10 kW; compatibility with high density (lo-’ mbar) discharge cleaning; immunity to desorption by fast atoms reflected by the neutralizer plates (the relevant desorption yields [ 14,151 are of the order of unity for monolayer coverage); and resistance to thermal stresses and magnetic forces. The design calls for 80 SAES getter modules, located -20 cm from the neutralizer plates, and operating at 300-350°C. During discharge cleaning after atmospheric exposure, they will be kept cold and passivated. However during routine discharge cleaning they will operate at 475°C to avoid hydrogen saturation and embrittlement, but will still getter impurities thus speeding up cleaning. 4. Impurity studies In spite of all the precautions taken, impurities will be generated, especially during reversal. To understand
334
C. Boucher et al. / Impurity
control in the Varennes
sources. The direct shielding factor of the scrape-off layer is sensitive to f,, and can be even more so if perpendicular diffusion is important. Results of this code coupled to the various impurity diagnostics will form the basis of the impurity studies in the Tokamak de Varennes.
References [I] P. Couture,
6 cm/dbv To
plate -
Fig. 2. Contour lines of the probability of ionization (arbitrary units) in the scrape-off. (Z = 1 lines for ionization from charge state Z=O to Z=l and Z=3 lines for Z=2 to Z=3.)
the transport of wall-generated impurities in the scrapeoff region, a 2-D Monte Carlo code is being developed. The code will be used for calculating the shielding efficiency of the scrape-off layer to injected impurities, their charge state distribution at the divertor plate, and for simulating laser blow-off experiments. The preliminary results obtained simulate the continuous injection of heavy impurity atoms (Fe) from a point source into the scrape-off region (see fig. 2). The plasma density distribution in the scrape-off layer was derived form a simple 1-D transport model assuming perpendicular Bohm diffusion [16], and including the effect of ionization by electrons [ 171. The temperature is taken as constant through the boundary layer. The injected neutral particles are randomly selected from a cosine distribution, with an energy spectrum similar to that measured for sputtered atoms [18]. A neutral particle is injected at a fixed point with velocity components uL and u ,, and is followed using a pseudocollision algorithm [19] until it reaches a collision point where it is ionized. The ion then moves parallel to the magnetic field with its guiding center having a parallel velocity equal to the flow velocity f,,C,, where f,, is the flow velocity factor and C, is the ion sound speed. ( f,, = 0.1 [ 111). The ion is followed with its constant velocity through succesive states of ionization. This simulation indicates that Fe ions of high charge state (Z = 5) are created in the scrape-off before reaching the neutralizer plate. For the case of an extended source representative of wall sputtering, the charge state distribution can be obtained by summing over a distribution of such point
[2] [3] [4] [S] [6]
[7] [8] [9]
[IO]
[ 1 l] [12]
[ 131 [l4] [l5]
[ 161
[17] [I81 [19]
[20] [21]
J. Geoffrion. B.C. Gregory. H.W.H. Van Andel, R.A. Bolton, P.B. Cumyn, B.L. Stansfield and B. Terreault, 9th Symp. on Engineering Problems of Fusion Research, Chicago (October 198 1). F. Wesner, H. Rapp and F. Schneider, in ref. 1. F. Schneider, in ref. 1. P. Couture, J. Geoffrion and B.C. Gregory, Internal Report TV RI-87f, (October 1981). I.P. Shkaforsky and M. Shoucri, J. Plasma Phys. 26. part 3 (1981). V. Fuchs. M. Shoucri. G. Thibaudeau, T.W. Johnston and I.P. Shkarofsky, 23rd Annual Meeting, Div. of Plasma Phys., APS (October 1981). TFR Group, J. Nucl. Mater. 99 (1981) 334. J. Winter, F. Waelbrook, B. Brandt, K.J. Dietz, 1. Ali-Khan and P. Wienhold, J. Nucl. Mater. 93/94 (1980) 812. Y. Sakamoto, Y. Ishibe, K. Yano, H. Oyama. Y. Tanaka, N. Fujisawa, Y. Matsuzaki, N. Suzuki, K. Maeno, T. Yamamoto and K. Yokokura. J. Nucl. Mater. 93/94 (1980) 333. C. Boucher, B. Terreault, Internal Report, INRS-Energie No. 149 (June 1981). M. Petravic, D. Post and D. Heifetz. J. Schmidt, Phys. Rev. Lett. 48 (1982) 316. M. Ali-Mahdavi, J.C. DeBoo, C.L. Hsieh, N. Ohyabu, R.D. Stambaugh and J.C. Wesley, Phys. Rev. Lett. 47 (1981) 1602. M. Keilhacker et al., IAEA-CN-38/01, p. 35 1. R. Bastasz and L.G. Haggmark, J. Nucl. Mater. 103 & 104 (1981) 499. A. Sagara, K. Akaishi, K. Kamada and A. Miyoharu. J. Nucl. Mater. 103 & 104 (1981) 357. M. Keilhacker, in: Tokamak Reactors for Breakeven, Int. School of Fusion Reactor Technology, Ettore Majorana Centre, (September 1976). W. Lotz, 2. Phys. 220 (1969) 466. R.B. Wright, M.J. Pellin, D.M. Gruen and C.E. Young, Nucl. Instr. and Meth. 170 (1980) 295. G.G. Lister, D.E. Post and R. Goldston, 3rd Symp. on Plasma Heating in Toroidal Devices, Grenoble (1976) p. 303. H. Towner, Princeton Plasma Physics Laboratory. private communication. W. Bauer, J. Nucl. Mater. 76 & 77 (1978) 3.