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Radial build between helical coil and plasma in the Large Helical Device
Fusion Engineering and Design 20 (1993) 73-77 North-Holland
73
Radial build between helical coil and plasma in the Large Helical Device N. O h y a b u , K. Y a m a z a k i , H a n t a o JI, S. I m a g a w a , H. K a n e k o , S. M o r i m o t o , N. N o d a , T. Satow, J. Y a m a m o t o , O. M o t o j i m a a n d the L H D Design G r o u p National Institute for Fusion Science, Chikusa-ku, Nagoya 464-01, Japan
The Large Helical Device (LHD) is a heliotron/torsatron-type confinement device (B - 4 T, R = 3.9 m) equipped with a helical divertor. In the LHD configuration, the plasma region is shifted inwards approximately a third of the plasma minor radius relative to the center of the two pairing helical coils, thus making the distance between the coil center and the edge plasma small, 328 mm on the small major radius side of the torus. Within this space, many components must be installed, such as the superconducting helical coil (163), its coil can (45), the thermal shield (30), the vacuum gap (35), the plasma vacuum vessel (15), the first wall (25) (the numbers in the parentheses are allocated radial space in mm for the respective components). Effective edge plasma control by the divertor requires a space of more than 15 mm between the plasma and the first wall. To meet the above space requirement, the thickness of the helical coil is designed to be as small as possible, yet the coil current density and maximum field strength are within the limits of reliable coil operation.
1. Introduction The National Institute for Fusion science (NIFS) has committed to construct a large heliotron/torsatron type (l = 2) helical device, called the Large Helical Device (LHD) [1,3]. The objective of the project is to demonstrate the attractiveness of the helical system at reactor-relevant plasma parameters. The device parameters are: R (the major radius)= 3.9 m, a (the minor radius) - 0.6 m, B (the magnetic field) = 4 T, m (the toroidal mode number of the helical coil) = 10, Yc (the pitch of the helical coil)= 1.25. The inherent advantage of helical devices over tokamaks is that steady-state operation can be realized since plasma current is not required for confinement of the helical plasma. In order to demonstrate this advantage clearly, we have decided to adopt superconducting coils for the LHD magnets. Steady-state plasma operations (>30 min) are planned in the LHD experiments. With superconducting coils, a steady-state magnetic confinement configuration can be maintained in a steady state for the helical device, but it does not guarantee steady-state plasma operation. The outward heat flow Correspondence to: Dr. N. Ohyabu, National Institute for Fusion Science, Chikusa-ku, Nagoya 464-01, Japan.
must be removed safely at the plasma boundary and the impurity contamination should be controlled for successful steady-state plasma operation. We install a helical divertor to handle these edge problems. The divertor is a key ingredient of the LHD device [4], which is also expected to enhance the energy confinement of the discharge. A gap of - 1 5 mm between the plasma and the vessel wall or limiter is necessary for more than 90% of the particle and heat fluxes from the core plasma region to be guided into the divertor chamber, while these fluxes are handled in a controllable fashion. The radial build between the helical coil center and the plasma is determined by the requirement of maintaining the gap mentioned above at B = 4 T. Section 2 deals with physical requirements on the radial build. Section 3 discusses how these requirements are accommodated in the device design, particularly the superconducting coil design.
2. Physical requirements on the radial build Figure 1 shows the magnetic configurations of standard LHD operation at the two poloidal planes with = 0° and 18°. The whole plasma is shifted inwards
0 9 2 0 - 3 7 9 6 / 9 3 / $ 0 6 . 0 0 © 1 9 9 3 - Elsevier Science Publishers B.V. All rights reserved
74
N. Ohyabu et al. / Radial build between helical coil and plasma in L H D
Fig. 1. Poloidal cross-sectional view of the LHD configuration at two toroidal angles (q5 = 0 °, q~ = 18°).
Torus Cen
Coil Can (45ram) Thermal
Insulation
Gap
(~65mm) Vacuum V e s s e l (15mm) First
Wall (~25mm)
642 955 nter
Radius
9 7 5
Divertor Plate
/
Major R a d i u s
3 g 0 0
> ( u n i t : mm) Fig. 2. Radial build at 6 = 0° on the small major radius side o f t h e torus.
75
N. Ohyabu et al. / Radial build between helical coil and plasma in LHD '
I~ = 0 °
'
'
'1
I >.
/L j ""
= 4.5 °
i"
:~
/,
,
'
i.~o ' ' q~-7.0"
:iclli. i:.i, °f,
ayer
Helical
Coil
-o.m
-o.,
.o.,
-o.~
o.o
Fig. 3. Proximity of the scrape-off layer plasma to the helical coil and the first wall.
76
N. Ohyabu et al. / Radial build between helical coil and plasma in L H D
relative to the center of the two pairing helical coils approximately by a third of the plasma minor radius. Consequently, on the small major radius side of the t o m s , the space between the coil center and the edge of the plasma region is very tight at the toroidal location d' = 0 °, where the plasma shape is vertically elongated. In this space, many components must be installed, such as the coil, the coil can, the vacuum gap, the thermal shield, the vacuum vessel and the first wall, as shown in fig. 2. As mentioned in section 1, the physical requirement on the radial build is that the plasma in the scrape-off layer, which is connected to the divertor chamber through magnetic field lines, should not touch the first wall for the divertor to function. Figure 3 shows the proximity of the scrapeoff layer plasma to the first wall under the helical coils. The gap of 15 mm is needed to avoid wall-plasma interaction. The space for the first wall and its associated components is determined by the heat flux on the wall. We are planning a steady-state operation with 3 M W input power and the first wall is designed to handle a steady-state radiative heat load of 1.5 W / c m z. This requires at least a radial space of 20ram. The thickness of the vacuum vessel wall needed to be 15 mm because it should withstand atmospheric pressure load as well as the magnetic pressure load during disruption of the plasma current (we do not intend to drive the plasma current, but it is predicted that the bootstrap current driven by the pressure gradient flows through the plasma). As shown in fig. 2, the radial distance between the coil center and the edge of the plasma is 328 mm and thus the distance between the coil center and the outer vacuum vessel is 273 mm. The superconducting coil and its associated components must be installed in this radial space.
3. Tight-space requirement on the superconducting coil design The current density in the helical coil package is the most critical factor for meeting the space requirement. During the course of the conceptional design study of the L H D device, we had attempted to increase the coil current density as high as possible. But the coil integrity, the superconducting coil stability must be maintained. After detailed engineering considerations and evaluation of the coil system, the coil current density was determined to be 4 0 A / m m - ' for 3 T operation (phase I) and 53 A / r a m 2 for 4 T operation (phase II). With such current densities, the full thickness of the coil is 326 ram. The gap must include the displacement of the position of the components. The coils and coil cans shrink by 10mm in the radial direction when the t e m p e r a t u r e is decreased from room temperature to helium temperature. When current is driven through the coils, displacement of the coils (a few mm) due to the force on the coil occurs. The vacuum vessel temperature changes due to the radiative power loss from the plasma or the vessel baking from room temperature to 100°C, allowing a radial vessel displacement of 5 mm. The thermal shield between the coils (helium temperature) and the plasma vacuum vessel wall (100°C) requires a radial space of 30 ram. Table 1 summarizes the radial displacement of each component at various conditions. The gap between the coil can and the thermal shield is 35 mm at the desigt~ specification, but it becomes minimum, 20 mm, during vacuum test of the vacuum vessel. Accordingly, the accuracy in manufacturing and assembly of the coil and the plasma vessel must be 2 0 m m to meet the
Table 1 Vacuum gap betwecn thermal shield and helical can at various conditions ~ Conditions
After assembly
Vacuum test of V.V.
Design reference
HFC only
Normal op. (low V.V. temperature)
Disruption
Normal op. (high V.V. temperature)
Pressure in V.V. Temperature of V.V. Pressure in cryostat Temperature of HFC
A R A R
V R A R
V R V C
V R V C
V R V C
V R V C
V H V C
Vacuum gap between coil can and shield (mm)
25
20
35
30
40
35
45
~'V.V.: vacuum vessel, HFC: helical field coil, A: atmosphere, V: vacuum, R: room temperature, H: 100°C, C: helium temperature.
N. Ohyabu et al. / Radial build between helical coil and plasma in LHD
requirements on the radial build, while the accuracy for the coil is less than 2 mm to minimize the error field of the confining magnetic configuration. This is quite a challenging engineering task which we must accomplish.
References [1] A. liyoshi, M. Fujiwara, O. Motojima, N. Ohyabu and K. Yamazaki, Design study for the large helical device, Fusion Technol. 17 (1990) 169-187.
77
[2] O, Motojima et al., Engineering design study of superconducting large helical device, in: Plasma Physics and Controlled Nuclear Fusion Research 1990 (Proc. 13th Int. Conf. Washington, 1990) IAEA-CN-53/G-1-5, 1991. [3] K. Yamazaki et al., Physics studies on helical confinement configurations with / = 2 continuous coil systems, in: Plasma Physics and Controlled Nuclear Fusion Research 1990 (Proc. 13th Int. Conf. Washington, 1990), Vol. 2 (IAEA, Vienna, 1991) p. 709. [4] N, Ohyabu et al., Helical divertor in the Large Helical Device, to appear in the proceedings of the PSI conference (Monterey, CA) (March 1992).
73
Radial build between helical coil and plasma in the Large Helical Device N. O h y a b u , K. Y a m a z a k i , H a n t a o JI, S. I m a g a w a , H. K a n e k o , S. M o r i m o t o , N. N o d a , T. Satow, J. Y a m a m o t o , O. M o t o j i m a a n d the L H D Design G r o u p National Institute for Fusion Science, Chikusa-ku, Nagoya 464-01, Japan
The Large Helical Device (LHD) is a heliotron/torsatron-type confinement device (B - 4 T, R = 3.9 m) equipped with a helical divertor. In the LHD configuration, the plasma region is shifted inwards approximately a third of the plasma minor radius relative to the center of the two pairing helical coils, thus making the distance between the coil center and the edge plasma small, 328 mm on the small major radius side of the torus. Within this space, many components must be installed, such as the superconducting helical coil (163), its coil can (45), the thermal shield (30), the vacuum gap (35), the plasma vacuum vessel (15), the first wall (25) (the numbers in the parentheses are allocated radial space in mm for the respective components). Effective edge plasma control by the divertor requires a space of more than 15 mm between the plasma and the first wall. To meet the above space requirement, the thickness of the helical coil is designed to be as small as possible, yet the coil current density and maximum field strength are within the limits of reliable coil operation.
1. Introduction The National Institute for Fusion science (NIFS) has committed to construct a large heliotron/torsatron type (l = 2) helical device, called the Large Helical Device (LHD) [1,3]. The objective of the project is to demonstrate the attractiveness of the helical system at reactor-relevant plasma parameters. The device parameters are: R (the major radius)= 3.9 m, a (the minor radius) - 0.6 m, B (the magnetic field) = 4 T, m (the toroidal mode number of the helical coil) = 10, Yc (the pitch of the helical coil)= 1.25. The inherent advantage of helical devices over tokamaks is that steady-state operation can be realized since plasma current is not required for confinement of the helical plasma. In order to demonstrate this advantage clearly, we have decided to adopt superconducting coils for the LHD magnets. Steady-state plasma operations (>30 min) are planned in the LHD experiments. With superconducting coils, a steady-state magnetic confinement configuration can be maintained in a steady state for the helical device, but it does not guarantee steady-state plasma operation. The outward heat flow Correspondence to: Dr. N. Ohyabu, National Institute for Fusion Science, Chikusa-ku, Nagoya 464-01, Japan.
must be removed safely at the plasma boundary and the impurity contamination should be controlled for successful steady-state plasma operation. We install a helical divertor to handle these edge problems. The divertor is a key ingredient of the LHD device [4], which is also expected to enhance the energy confinement of the discharge. A gap of - 1 5 mm between the plasma and the vessel wall or limiter is necessary for more than 90% of the particle and heat fluxes from the core plasma region to be guided into the divertor chamber, while these fluxes are handled in a controllable fashion. The radial build between the helical coil center and the plasma is determined by the requirement of maintaining the gap mentioned above at B = 4 T. Section 2 deals with physical requirements on the radial build. Section 3 discusses how these requirements are accommodated in the device design, particularly the superconducting coil design.
2. Physical requirements on the radial build Figure 1 shows the magnetic configurations of standard LHD operation at the two poloidal planes with = 0° and 18°. The whole plasma is shifted inwards
0 9 2 0 - 3 7 9 6 / 9 3 / $ 0 6 . 0 0 © 1 9 9 3 - Elsevier Science Publishers B.V. All rights reserved
74
N. Ohyabu et al. / Radial build between helical coil and plasma in L H D
Fig. 1. Poloidal cross-sectional view of the LHD configuration at two toroidal angles (q5 = 0 °, q~ = 18°).
Torus Cen
Coil Can (45ram) Thermal
Insulation
Gap
(~65mm) Vacuum V e s s e l (15mm) First
Wall (~25mm)
642 955 nter
Radius
9 7 5
Divertor Plate
/
Major R a d i u s
3 g 0 0
> ( u n i t : mm) Fig. 2. Radial build at 6 = 0° on the small major radius side o f t h e torus.
75
N. Ohyabu et al. / Radial build between helical coil and plasma in LHD '
I~ = 0 °
'
'
'1
I >.
/L j ""
= 4.5 °
i"
:~
/,
,
'
i.~o ' ' q~-7.0"
:iclli. i:.i, °f,
ayer
Helical
Coil
-o.m
-o.,
.o.,
-o.~
o.o
Fig. 3. Proximity of the scrape-off layer plasma to the helical coil and the first wall.
76
N. Ohyabu et al. / Radial build between helical coil and plasma in L H D
relative to the center of the two pairing helical coils approximately by a third of the plasma minor radius. Consequently, on the small major radius side of the t o m s , the space between the coil center and the edge of the plasma region is very tight at the toroidal location d' = 0 °, where the plasma shape is vertically elongated. In this space, many components must be installed, such as the coil, the coil can, the vacuum gap, the thermal shield, the vacuum vessel and the first wall, as shown in fig. 2. As mentioned in section 1, the physical requirement on the radial build is that the plasma in the scrape-off layer, which is connected to the divertor chamber through magnetic field lines, should not touch the first wall for the divertor to function. Figure 3 shows the proximity of the scrapeoff layer plasma to the first wall under the helical coils. The gap of 15 mm is needed to avoid wall-plasma interaction. The space for the first wall and its associated components is determined by the heat flux on the wall. We are planning a steady-state operation with 3 M W input power and the first wall is designed to handle a steady-state radiative heat load of 1.5 W / c m z. This requires at least a radial space of 20ram. The thickness of the vacuum vessel wall needed to be 15 mm because it should withstand atmospheric pressure load as well as the magnetic pressure load during disruption of the plasma current (we do not intend to drive the plasma current, but it is predicted that the bootstrap current driven by the pressure gradient flows through the plasma). As shown in fig. 2, the radial distance between the coil center and the edge of the plasma is 328 mm and thus the distance between the coil center and the outer vacuum vessel is 273 mm. The superconducting coil and its associated components must be installed in this radial space.
3. Tight-space requirement on the superconducting coil design The current density in the helical coil package is the most critical factor for meeting the space requirement. During the course of the conceptional design study of the L H D device, we had attempted to increase the coil current density as high as possible. But the coil integrity, the superconducting coil stability must be maintained. After detailed engineering considerations and evaluation of the coil system, the coil current density was determined to be 4 0 A / m m - ' for 3 T operation (phase I) and 53 A / r a m 2 for 4 T operation (phase II). With such current densities, the full thickness of the coil is 326 ram. The gap must include the displacement of the position of the components. The coils and coil cans shrink by 10mm in the radial direction when the t e m p e r a t u r e is decreased from room temperature to helium temperature. When current is driven through the coils, displacement of the coils (a few mm) due to the force on the coil occurs. The vacuum vessel temperature changes due to the radiative power loss from the plasma or the vessel baking from room temperature to 100°C, allowing a radial vessel displacement of 5 mm. The thermal shield between the coils (helium temperature) and the plasma vacuum vessel wall (100°C) requires a radial space of 30 ram. Table 1 summarizes the radial displacement of each component at various conditions. The gap between the coil can and the thermal shield is 35 mm at the desigt~ specification, but it becomes minimum, 20 mm, during vacuum test of the vacuum vessel. Accordingly, the accuracy in manufacturing and assembly of the coil and the plasma vessel must be 2 0 m m to meet the
Table 1 Vacuum gap betwecn thermal shield and helical can at various conditions ~ Conditions
After assembly
Vacuum test of V.V.
Design reference
HFC only
Normal op. (low V.V. temperature)
Disruption
Normal op. (high V.V. temperature)
Pressure in V.V. Temperature of V.V. Pressure in cryostat Temperature of HFC
A R A R
V R A R
V R V C
V R V C
V R V C
V R V C
V H V C
Vacuum gap between coil can and shield (mm)
25
20
35
30
40
35
45
~'V.V.: vacuum vessel, HFC: helical field coil, A: atmosphere, V: vacuum, R: room temperature, H: 100°C, C: helium temperature.
N. Ohyabu et al. / Radial build between helical coil and plasma in LHD
requirements on the radial build, while the accuracy for the coil is less than 2 mm to minimize the error field of the confining magnetic configuration. This is quite a challenging engineering task which we must accomplish.
References [1] A. liyoshi, M. Fujiwara, O. Motojima, N. Ohyabu and K. Yamazaki, Design study for the large helical device, Fusion Technol. 17 (1990) 169-187.
77
[2] O, Motojima et al., Engineering design study of superconducting large helical device, in: Plasma Physics and Controlled Nuclear Fusion Research 1990 (Proc. 13th Int. Conf. Washington, 1990) IAEA-CN-53/G-1-5, 1991. [3] K. Yamazaki et al., Physics studies on helical confinement configurations with / = 2 continuous coil systems, in: Plasma Physics and Controlled Nuclear Fusion Research 1990 (Proc. 13th Int. Conf. Washington, 1990), Vol. 2 (IAEA, Vienna, 1991) p. 709. [4] N, Ohyabu et al., Helical divertor in the Large Helical Device, to appear in the proceedings of the PSI conference (Monterey, CA) (March 1992).