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The first in-vessel cryopump for EAST divertor experiment
Fusion Engineering and Design 85 (2010) 1508–1512
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
The first in-vessel cryopump for EAST divertor experiment Q.S. Hu ∗ , D.M. Yao, G.N. Luo, H. Xie, C.S. Xu, J.G. Li, X.M. Wang Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, China
a r t i c l e
i n f o
Article history: Available online 7 May 2010 Keywords: Cryopump Divertor Tokamak Long pulse
a b s t r a c t To operate divertor experiment, an in-vessel cryopump was installed on the EAST tokamak in 2008. It can limit gas impurity recycling from divertor region into core plasma area, and provide plasma density control with toroidally distributed high pumping speed. In this paper, the designing and manufacturing is basically described. Most parts are manufactured in ASIPP, except for some procedures such as laser cutting, plasma-spray coating, and pipe annealing. For this first in-vessel cryopump, liquid helium and nitrogen supplying system is upgraded. Functional tests for this cryopump show a good radiation shield and pumping capability. A campaign utilizing this device for divertor physics research has been successful. © 2010 Elsevier B.V. All rights reserved.
1. Introduction The EAST tokamak is a superconducting magnetically confined fusion experiment device in China. Its primary purpose is to explore a way to operate a long stable pulse discharge, which is pivotal for establishing a commercialized fusion device in the future. As required by fusion technology development, an elongated confined plasma configuration is expected. To comply with this requirement, EAST has an up-down symmetrical divertor construction with two divertors be integrated. Each integrated divertor is toroidally continued and made up of inner target blanket, dome and outer passive blanket. Electromagnetic control allows the plasma profile to be elongated as double null or single null configuration with more effective neutral particles’ venting and less erosion of target surface (Fig. 1). For a long pulse discharge that may reach up to 1000 s for EAST, high speed pumping of neutral particles in divertor region becomes more and more crucial. To prevent large accumulations of impurity gas from diffusing back into core plasma region, a toroidally high pumping speed evacuation system must be set up and the impurity gas be pumped away rapidly from divertor region to maintain the discharge balance between gas puffing in and venting out. If an appropriate balance has been set up, plasma discharge disruption may be avoided. However, the long mechanical duct from divertor region to any outside pumping device seriously decreases any effective pumping speed and capability. So only an in-vessel cryopump can solve this problem. With the proper design and construction, it can effectively supply almost all its pumping capability to evacuate gas impurities from divertor region and away from core plasma region.
∗ Corresponding author. Tel.: +86 551 5595011; fax: +86 551 5591310. E-mail address: [email protected] (Q.S. Hu). 0920-3796/$ – see front matter © 2010 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2010.04.015
Eventually EAST plans to install 4 sets of in-vessel cryopump for divertor experiment, two for the lower divertor and two for the upper one. In 2008, the first set was constructed and installed under the lower divertor passive target blanket partially for the tightened engineering scheme. This first location was chosen to provide EAST with the same lower single null configuration as what ITER will have. Experiments in this configuration will help to explore more relevant information essential to ITER research. The other three sets are to be expected in next 5 years. For this initial cryopump, we evaluated every step from overall design to installation in vacuum vessel to avoid any risk. After installation in 2008, we made the first time functional test only at the end of the third EAST experimental campaign to minimize the effects of a possible failure on the whole campaign. The planned performance of the system was confirmed with no leakage, anticipated cooling temperature and required pumping speed. In the next campaign for EAST experiment in 2009, this device was utilized for divertor physics research. These recent results have shown that the in-vessel cryopump has an impact on divertor plasma behavior and thus the interaction between plasma and target blanket surface. The relevant divertor physics experiment results with this in-vessel cryopump will be presented in other articles.
2. Designs and manufacture The initial design scheme was evaluated several times in regard to EAST specialization. The primary projected pumping speed of each in-vessel cryopump is 15,000 L/s. Several schemes for implementing the pumping can be considered. The first scheme is to set up several individually cooled plates distributed evenly along the divertor torus region. The second is to set up a torus continuously cooled by liquid helium, which will supply the evenly distributed
Q.S. Hu et al. / Fusion Engineering and Design 85 (2010) 1508–1512
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Fig. 2. A sample of copper coating belt array.
Fig. 1. The cross-section of the divertor and the torus section.
pumping capability around the divertor region. The third is to set up 16 sets of commercialized refrigerated cryopumps installed in the top or underside windows of the tokamak. Individually cooled plates require several ports be used for liquid helium or liquid nitrogen supplying. And cryogenic liquid distributor also requires several feed lines. It makes engineering very difficult. At the same time, the pumping speed is not toroidally symmetric and this makes plasma condition more difficult to analyze. So we give up the first scheme. Even though we can put commercialized cryopumps on top windows or underside windows of EAST, their pumping efficiency is not good enough for divertor neutral particles’ evacuation. Long distance between port window and pump itself limits gas conduction and heavily limits pumping effective speed. So we employ the second scheme similar to that in DIII-D [1]. With an overall condensation surface in form of torus, the in-vessel cryopump also minimizes the required volume in the confined divertor space. The in-vessel cryopump comprises two kinds of subassembly: torus and feed line. Torus pumping capability mainly comes from the cooled surface of liquid helium pipe. For this first set, the surface area of liquid helium pipe for the torus is about 1 m2 , yielding a theoretical pumping speed for deuterium of about 79,000, 110,000 L/s for hydrogen, 37,000 L/s for water vapor and 24,000 L/s for carbon dioxide. At the same time, the outward area of nitrogen cooling shield is about 4.5 m2 . Its pumping speed for water vapor is about 25,000 L/s and about 16,000 L/s for carbon dioxide. The torus section center is located radially 2115 mm from tokamak axis and 900 mm below the equatorial plane. When the torus was installed, continuous measurement was carefully kept to make sure that the torus was in correct position and there was no bending force from installation. The nitrogen shield was cut to have windows and a series of slots through which welded or brazed metal jointing are filled to connect the shield and the nitrogen pipe. All
these slots or windows were laser cut quickly and effectively. To improve the shield cooling speed and uniformity, an array of copper belts of about 0.4 mm thickness were designed and were coated on the outer surface of the shield using plasma-spray coating (Fig. 2). The feed line in the lower port pipeline of the tokamak connects the torus in vacuum vessel to a cryogenic coolant distributor outside the tokamak. It transmits liquid helium and liquid nitrogen to torus that function as divertor pumping drive. Since the cryostat shield has a large volume to protect a large number of superconducting magnetic field coil (16 sets of toroidal field coil sets, 14 sets of poloidal field coil sets, and a central solenoid set), the port pipeline is relatively long and can easily give extra heat load to liquid helium or nitrogen. Accordingly, the feed line is surrounded by a polished tube to keep the temperature of liquid helium coolant be as lower as possible (Fig. 3). The torus is designed as an assembly of 8 units. Each unit is made up of a helium pipe, a helium pipe support, 2 nitrogen pipes, an inner radiation shield, an outer radiation shield, a warm particle shield and a unit support. Each unit is assembled outside the toka-
Fig. 3. The feed line and the torus pumping system.
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Fig. 4. A section of the radiation shield.
mak. After each unit is preinstalled in vacuum vessel, we connected each helium pipe to its neighboring one, then the two nitrogen pipes, forming one torus assembly. This modular design and fabrication is very flexible to comply with any adjustment of inner components for tokamak improvement. When the torus position is required to be modified, or if there is any interference with inner components, corrections can be made rapidly. Located in the center of the torus, the helium pipe and its metallic surface can be cooled to about 5 K. The inner and outer nitrogen shield can maintain at about 80 K. During tokamak discharges, the divertor pumping drive comes mainly from the cooled helium pipe, while the nitrogen cooling shield acts to condense most water vapor and carbon dioxide gas, reducing the heat load on the helium pipe. Surrounding the cryopump torus, the warm particle shield protects the nitrogen shield from higher ambient heat load or hot gas convection [2]. The warm particle shield is not actively cooled or warmed, but its surface is polished and can reflect most heat load from the vacuum vessel and the blanket structure. During the experiment, the warm particle shield’s temperature is subject to its surroundings and is roughly same as the vacuum vessel baking temperature of about 310 K. These torus shields are helpful in cooling down and maintaining the helium pipe at about 5 K (Fig. 4). To increase heat conductance impedance, a spring support structure is utilized to support the helium pipe inside the torus unit. This spring support increases the conduction length and thus conduction resistance between the 80 K shield and the 5 K helium pipe. Between the nitrogen shield and warm particle shield, there are several ceramic supports with very small contact area to nitrogen shield. This method provides acceptable support while maintaining low thermal conduction from warm shield to nitrogen shield. The support for torus unit is a kind of plate spring [3]. This makes it more adaptable to shrinking of cooled torus. To avoid electrical accident, an electrical break is designed for the overall in-vessel cryopump to separate it from the plasma vacuum chamber and the cryostat shield chamber. The coolant distributor is also isolated from the cryostat shield chamber and the plasma vacuum chamber. The insulation limit is 10 kV. This eliminates the accidental current heating load to coolant or electrical strike to personnel. To reduce impact from disruption halo currents and to minimize the distortion of the cooled torus, most of the components were fabricated from the nickel chromium alloy named Inconel 625. It has relatively high electrical impedance and lower thermal expansion ratio. For this first set of divertor pump on EAST, a number of manufacturing challenges needed to be overcome. The main processes were bending and cutting shield pipes, welding cooling pipes, plasma
Fig. 5. A leakage test for a torus pumping unit.
spraying of copper coating, annealing of bended pipes, leak detecting of welded pipes, and the fabrication of all parts. Bending is an important process to start. The torus is assembled with some coaxial pipes and the coaxial tolerance between them is very small. Since the curvature of some pipes may change after welding process, experiments were done to make sure that all bended parts were in coaxial tolerance after being installed in their final position. For each liquid nitrogen radiation shield, there is a bent pipe that is attached by welding. Before installation, all bended or welded parts are annealed to eliminate the strain force produced by machining. Before installation in vacuum vessel, all pipes of a torus unit must be leak detected to satisfy an ultrahigh vacuum criterion. This test focuses on the welded spots to search for metal cracks by welding. After all torus units were welded into one ring in vacuum vessel, a leakage test was done again to make sure that all welded spots were well sealed. Once the lower passive blanket was installed, we checked all diagnostic windows that opened on the passive blanket and block off torus preventing it from direct exposure to divertor plasma. In this way we can ensure that neutral particles in divertor region can transmit through circular throat gap evenly to torus condensing surface (Figs. 5 and 6).
Fig. 6. A torus unit being installed in vacuum chamber.
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Most parts are manufactured by ASIPP, except for some procedures such as laser cutting, plasma-spray coating and pipe annealing. The manufacturing experience should be helpful to establishing the future subsequent divertor pumping units.
3. Function test and divertor experiment During the third campaign of EAST experiment in August 2008, a series of test were made. The first test was a functional test of the in-vessel cryopump during plasma operation. For these tests the conditional parameters such as rate of outgas, baking temperature of vacuum vessel, heat sink temperature, vacuum changes in discharge’s duration were kept stable. In plasma discharge condition, the impurity gas is made up of helium, hydrogen, deuterium, carbon dioxide, carbon monoxide, etc. The pumping speed for this mixture gas in the course of plasma discharge is about 26,000 L/s. Once we get this data, we stopped the discharge experiment. After a while, with the in-vessel cryopump working and the vacuum of the tokamak vessel decreased to balanced pressure of about 1E−5 Pa, we began to puff only deuterium in typical puffing speed of EAST during plasma discharge at 416 Pa L/s and find its pumping speed for deuterium is about 75,600 L/s. The deuterium gas puffing into the vacuum vessel lasts for 100 min. The purpose is to check the pumping capability for a long pulse discharge up to 1000 s, one of the physics goals for EAST tokamak. The condensation ability criterion is whether this torus cryopump can condense impurity gas effectively for 1000 s long pulse discharge on EAST. As presently planned, EAST shall inject several times more puffed gas in the future. Tests using 6 times the previous mass of gas by 6000 s at average puffing speed of 416 Pa L/s, gave condensational ability for deuterium for about 2.5 × 106 Pa L, which is sufficient to fit a long pulse experiment on EAST. When nitrogen shield and helium pipe has been cooled down to its minimal temperature point, the inlet temperature of liquid nitrogen pipe is about 78.5 K, while inlet liquid helium temperature is about 7 K. When we began to puff deuterium gas at the average rate of 416 Pa L/s, the pressure curve of the tokamak’ vessel increases rapidly to 6E−3 Pa and then decrease to 4.7E−3 Pa. After 100 min of steady rate puffing, we shut the puffing valve, and the pressure of the vessel decreased steeply because of the high pumping speed supplied by cooled torus surface. However, because of the condensed gas layer, the cooling metallic surface’s temperature may rise a little and the pumping ability drops incrementally. The vacuum vessel pressure then stabilizes at a somewhat higher level. When the puffing is shut, the outlet temperature of liquid nitrogen pipe remains at about 80.5 K, while the outlet temperature of liquid helium pipe is 7.75 K. The differential temperature rise between of the helium pipe’s inlet and outlet is about 0.58 K while the change of the nitrogen pipe is negligible. After the gas puffing, the balanced pressure is 1.8E−4 Pa, which is tolerant to discharge pressure. However, after a saturation puffing for 100 min, the torus shade cryopump loses part of its condensation ability due to the buildup of frost on the cooling surface. When we start several other types of pumping devices, the vessel vacuum is drop to 2E−5 Pa, still higher than the pressure of 1E−5 Pa (Fig. 7). During the test or discharge experiment, liquid helium consumption seems to be stable at about 3 g/s. Nitrogen shield can be cooled down from normal room temperature to 80 K in 2 h. The helium pipe can be cooled down to 7 K in 3.5 h. The test shows agreement with the expected coolant consumption rate and cooling down speed. During discharge experiment, thermal losses seen to be negligible because the thermometer at outlet of the pipe indicates no obvious changes in contrast to that in idle operation. The average inlet pressure of liquid helium is 1.2 bar. The pressure drop is about 0.08 bar. For more efficient cooling in the future, cryo-
Fig. 7. A test curve of vacuum vessel pressure changes and can be used to determine the condensation saturation.
genic valve should be controlled and modulated at different cooling temperature. During the fourth campaign of EAST experiment in 2009, this in-vessel cryopump was used for divertor physics experiment. Double null and single null plasma configurations were researched with this divertor pumping in operation. Impurity gas influence to plasma was also researched. The initial results are encouraging. Impurity gas in plasma has been minimized, indicating a longer and more stable plasma discharge. Beside its utilizing for physics experiment, operation rules for the cryopump is also been explored. Because the buildup of the frost layer on cooling surface may impede effective condensation pumping, a regular regeneration should be operated periodically to warm up and then cool down its surface to release trapped neutral particles and maintain its condensational pumping capability. During the experiment in 2009, this cryopump has been regenerated several times, using two kinds of operation. The first is named partial regeneration in which the helium pipe warmed only to above 20 K. This kind of operation needs only half an hour and can be done conveniently during plasma discharge’s interval. The second named complete regeneration is to warm the whole system including torus and feed line up to normal room temperature and then to cool down to their lowest temperature. This can evacuate all impurity gases including water vapor and other kinds of frost. A complete regeneration needs about 8 h. So it is preferred to operating in idle interval. It is estimated that a complete regeneration may be needed every about 400 shots on EAST. Even though the in-vessel cryopump is very effective in pumping deuterium and hydrogen, analysis of mass spectrum shows that it cannot pump some kinds of deuterated gases effectively. When the liquid helium cooling surface’s temperature is a little above 5 K, condensation of helium gas for the main vacuum vessel is not very effective. These gases should be pumped out mainly by high vacuum turbo-molecular pumps or multi-dragging pumps. 4. Summary The first in-vessel cryopump for divertor experiment on EAST has been established. It has a compact structure and minimized heat load from outside and locates under lower outer passive target blanket, which makes it possible to try lower single null plasma configuration that similar to ITER with divertor pumping facility. Modular design and fabrication has improved its flexibility to comply with adjustment of adjacent components. Leakage detection
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has ensured the qualified parts to be used for the next process. The pumping capability seems to be good for a long pulse divertor plasma discharge. The pumping speed is consistent with what is expected. The coolant consuming speed seems in reasonable level and radiation shield seems to be in good function. With this first in-vessel cryopump, more exciting divertor plasma phenomena can be expected in the near future. Acknowledgments This work is partly supported by the ASIPP-GA Cooperation Program in 2005. The Authors acknowledge helpful discussions with
Mr. M.A. Mahdavi and host reception from Mr. P.M. Anderson and helpful language reviewed by Mr. Paul Thomas. References [1] E.E. Reis, et al., 19th Symposium on Fusion Technology, Lisbon, Portugal, 1996. [2] E.E. Reis, et al., 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, New Mexico, 1999. [3] A.S. Bozek, et al., 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, New Mexico, 1999.
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
The first in-vessel cryopump for EAST divertor experiment Q.S. Hu ∗ , D.M. Yao, G.N. Luo, H. Xie, C.S. Xu, J.G. Li, X.M. Wang Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, China
a r t i c l e
i n f o
Article history: Available online 7 May 2010 Keywords: Cryopump Divertor Tokamak Long pulse
a b s t r a c t To operate divertor experiment, an in-vessel cryopump was installed on the EAST tokamak in 2008. It can limit gas impurity recycling from divertor region into core plasma area, and provide plasma density control with toroidally distributed high pumping speed. In this paper, the designing and manufacturing is basically described. Most parts are manufactured in ASIPP, except for some procedures such as laser cutting, plasma-spray coating, and pipe annealing. For this first in-vessel cryopump, liquid helium and nitrogen supplying system is upgraded. Functional tests for this cryopump show a good radiation shield and pumping capability. A campaign utilizing this device for divertor physics research has been successful. © 2010 Elsevier B.V. All rights reserved.
1. Introduction The EAST tokamak is a superconducting magnetically confined fusion experiment device in China. Its primary purpose is to explore a way to operate a long stable pulse discharge, which is pivotal for establishing a commercialized fusion device in the future. As required by fusion technology development, an elongated confined plasma configuration is expected. To comply with this requirement, EAST has an up-down symmetrical divertor construction with two divertors be integrated. Each integrated divertor is toroidally continued and made up of inner target blanket, dome and outer passive blanket. Electromagnetic control allows the plasma profile to be elongated as double null or single null configuration with more effective neutral particles’ venting and less erosion of target surface (Fig. 1). For a long pulse discharge that may reach up to 1000 s for EAST, high speed pumping of neutral particles in divertor region becomes more and more crucial. To prevent large accumulations of impurity gas from diffusing back into core plasma region, a toroidally high pumping speed evacuation system must be set up and the impurity gas be pumped away rapidly from divertor region to maintain the discharge balance between gas puffing in and venting out. If an appropriate balance has been set up, plasma discharge disruption may be avoided. However, the long mechanical duct from divertor region to any outside pumping device seriously decreases any effective pumping speed and capability. So only an in-vessel cryopump can solve this problem. With the proper design and construction, it can effectively supply almost all its pumping capability to evacuate gas impurities from divertor region and away from core plasma region.
∗ Corresponding author. Tel.: +86 551 5595011; fax: +86 551 5591310. E-mail address: [email protected] (Q.S. Hu). 0920-3796/$ – see front matter © 2010 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2010.04.015
Eventually EAST plans to install 4 sets of in-vessel cryopump for divertor experiment, two for the lower divertor and two for the upper one. In 2008, the first set was constructed and installed under the lower divertor passive target blanket partially for the tightened engineering scheme. This first location was chosen to provide EAST with the same lower single null configuration as what ITER will have. Experiments in this configuration will help to explore more relevant information essential to ITER research. The other three sets are to be expected in next 5 years. For this initial cryopump, we evaluated every step from overall design to installation in vacuum vessel to avoid any risk. After installation in 2008, we made the first time functional test only at the end of the third EAST experimental campaign to minimize the effects of a possible failure on the whole campaign. The planned performance of the system was confirmed with no leakage, anticipated cooling temperature and required pumping speed. In the next campaign for EAST experiment in 2009, this device was utilized for divertor physics research. These recent results have shown that the in-vessel cryopump has an impact on divertor plasma behavior and thus the interaction between plasma and target blanket surface. The relevant divertor physics experiment results with this in-vessel cryopump will be presented in other articles.
2. Designs and manufacture The initial design scheme was evaluated several times in regard to EAST specialization. The primary projected pumping speed of each in-vessel cryopump is 15,000 L/s. Several schemes for implementing the pumping can be considered. The first scheme is to set up several individually cooled plates distributed evenly along the divertor torus region. The second is to set up a torus continuously cooled by liquid helium, which will supply the evenly distributed
Q.S. Hu et al. / Fusion Engineering and Design 85 (2010) 1508–1512
1509
Fig. 2. A sample of copper coating belt array.
Fig. 1. The cross-section of the divertor and the torus section.
pumping capability around the divertor region. The third is to set up 16 sets of commercialized refrigerated cryopumps installed in the top or underside windows of the tokamak. Individually cooled plates require several ports be used for liquid helium or liquid nitrogen supplying. And cryogenic liquid distributor also requires several feed lines. It makes engineering very difficult. At the same time, the pumping speed is not toroidally symmetric and this makes plasma condition more difficult to analyze. So we give up the first scheme. Even though we can put commercialized cryopumps on top windows or underside windows of EAST, their pumping efficiency is not good enough for divertor neutral particles’ evacuation. Long distance between port window and pump itself limits gas conduction and heavily limits pumping effective speed. So we employ the second scheme similar to that in DIII-D [1]. With an overall condensation surface in form of torus, the in-vessel cryopump also minimizes the required volume in the confined divertor space. The in-vessel cryopump comprises two kinds of subassembly: torus and feed line. Torus pumping capability mainly comes from the cooled surface of liquid helium pipe. For this first set, the surface area of liquid helium pipe for the torus is about 1 m2 , yielding a theoretical pumping speed for deuterium of about 79,000, 110,000 L/s for hydrogen, 37,000 L/s for water vapor and 24,000 L/s for carbon dioxide. At the same time, the outward area of nitrogen cooling shield is about 4.5 m2 . Its pumping speed for water vapor is about 25,000 L/s and about 16,000 L/s for carbon dioxide. The torus section center is located radially 2115 mm from tokamak axis and 900 mm below the equatorial plane. When the torus was installed, continuous measurement was carefully kept to make sure that the torus was in correct position and there was no bending force from installation. The nitrogen shield was cut to have windows and a series of slots through which welded or brazed metal jointing are filled to connect the shield and the nitrogen pipe. All
these slots or windows were laser cut quickly and effectively. To improve the shield cooling speed and uniformity, an array of copper belts of about 0.4 mm thickness were designed and were coated on the outer surface of the shield using plasma-spray coating (Fig. 2). The feed line in the lower port pipeline of the tokamak connects the torus in vacuum vessel to a cryogenic coolant distributor outside the tokamak. It transmits liquid helium and liquid nitrogen to torus that function as divertor pumping drive. Since the cryostat shield has a large volume to protect a large number of superconducting magnetic field coil (16 sets of toroidal field coil sets, 14 sets of poloidal field coil sets, and a central solenoid set), the port pipeline is relatively long and can easily give extra heat load to liquid helium or nitrogen. Accordingly, the feed line is surrounded by a polished tube to keep the temperature of liquid helium coolant be as lower as possible (Fig. 3). The torus is designed as an assembly of 8 units. Each unit is made up of a helium pipe, a helium pipe support, 2 nitrogen pipes, an inner radiation shield, an outer radiation shield, a warm particle shield and a unit support. Each unit is assembled outside the toka-
Fig. 3. The feed line and the torus pumping system.
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Fig. 4. A section of the radiation shield.
mak. After each unit is preinstalled in vacuum vessel, we connected each helium pipe to its neighboring one, then the two nitrogen pipes, forming one torus assembly. This modular design and fabrication is very flexible to comply with any adjustment of inner components for tokamak improvement. When the torus position is required to be modified, or if there is any interference with inner components, corrections can be made rapidly. Located in the center of the torus, the helium pipe and its metallic surface can be cooled to about 5 K. The inner and outer nitrogen shield can maintain at about 80 K. During tokamak discharges, the divertor pumping drive comes mainly from the cooled helium pipe, while the nitrogen cooling shield acts to condense most water vapor and carbon dioxide gas, reducing the heat load on the helium pipe. Surrounding the cryopump torus, the warm particle shield protects the nitrogen shield from higher ambient heat load or hot gas convection [2]. The warm particle shield is not actively cooled or warmed, but its surface is polished and can reflect most heat load from the vacuum vessel and the blanket structure. During the experiment, the warm particle shield’s temperature is subject to its surroundings and is roughly same as the vacuum vessel baking temperature of about 310 K. These torus shields are helpful in cooling down and maintaining the helium pipe at about 5 K (Fig. 4). To increase heat conductance impedance, a spring support structure is utilized to support the helium pipe inside the torus unit. This spring support increases the conduction length and thus conduction resistance between the 80 K shield and the 5 K helium pipe. Between the nitrogen shield and warm particle shield, there are several ceramic supports with very small contact area to nitrogen shield. This method provides acceptable support while maintaining low thermal conduction from warm shield to nitrogen shield. The support for torus unit is a kind of plate spring [3]. This makes it more adaptable to shrinking of cooled torus. To avoid electrical accident, an electrical break is designed for the overall in-vessel cryopump to separate it from the plasma vacuum chamber and the cryostat shield chamber. The coolant distributor is also isolated from the cryostat shield chamber and the plasma vacuum chamber. The insulation limit is 10 kV. This eliminates the accidental current heating load to coolant or electrical strike to personnel. To reduce impact from disruption halo currents and to minimize the distortion of the cooled torus, most of the components were fabricated from the nickel chromium alloy named Inconel 625. It has relatively high electrical impedance and lower thermal expansion ratio. For this first set of divertor pump on EAST, a number of manufacturing challenges needed to be overcome. The main processes were bending and cutting shield pipes, welding cooling pipes, plasma
Fig. 5. A leakage test for a torus pumping unit.
spraying of copper coating, annealing of bended pipes, leak detecting of welded pipes, and the fabrication of all parts. Bending is an important process to start. The torus is assembled with some coaxial pipes and the coaxial tolerance between them is very small. Since the curvature of some pipes may change after welding process, experiments were done to make sure that all bended parts were in coaxial tolerance after being installed in their final position. For each liquid nitrogen radiation shield, there is a bent pipe that is attached by welding. Before installation, all bended or welded parts are annealed to eliminate the strain force produced by machining. Before installation in vacuum vessel, all pipes of a torus unit must be leak detected to satisfy an ultrahigh vacuum criterion. This test focuses on the welded spots to search for metal cracks by welding. After all torus units were welded into one ring in vacuum vessel, a leakage test was done again to make sure that all welded spots were well sealed. Once the lower passive blanket was installed, we checked all diagnostic windows that opened on the passive blanket and block off torus preventing it from direct exposure to divertor plasma. In this way we can ensure that neutral particles in divertor region can transmit through circular throat gap evenly to torus condensing surface (Figs. 5 and 6).
Fig. 6. A torus unit being installed in vacuum chamber.
Q.S. Hu et al. / Fusion Engineering and Design 85 (2010) 1508–1512
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Most parts are manufactured by ASIPP, except for some procedures such as laser cutting, plasma-spray coating and pipe annealing. The manufacturing experience should be helpful to establishing the future subsequent divertor pumping units.
3. Function test and divertor experiment During the third campaign of EAST experiment in August 2008, a series of test were made. The first test was a functional test of the in-vessel cryopump during plasma operation. For these tests the conditional parameters such as rate of outgas, baking temperature of vacuum vessel, heat sink temperature, vacuum changes in discharge’s duration were kept stable. In plasma discharge condition, the impurity gas is made up of helium, hydrogen, deuterium, carbon dioxide, carbon monoxide, etc. The pumping speed for this mixture gas in the course of plasma discharge is about 26,000 L/s. Once we get this data, we stopped the discharge experiment. After a while, with the in-vessel cryopump working and the vacuum of the tokamak vessel decreased to balanced pressure of about 1E−5 Pa, we began to puff only deuterium in typical puffing speed of EAST during plasma discharge at 416 Pa L/s and find its pumping speed for deuterium is about 75,600 L/s. The deuterium gas puffing into the vacuum vessel lasts for 100 min. The purpose is to check the pumping capability for a long pulse discharge up to 1000 s, one of the physics goals for EAST tokamak. The condensation ability criterion is whether this torus cryopump can condense impurity gas effectively for 1000 s long pulse discharge on EAST. As presently planned, EAST shall inject several times more puffed gas in the future. Tests using 6 times the previous mass of gas by 6000 s at average puffing speed of 416 Pa L/s, gave condensational ability for deuterium for about 2.5 × 106 Pa L, which is sufficient to fit a long pulse experiment on EAST. When nitrogen shield and helium pipe has been cooled down to its minimal temperature point, the inlet temperature of liquid nitrogen pipe is about 78.5 K, while inlet liquid helium temperature is about 7 K. When we began to puff deuterium gas at the average rate of 416 Pa L/s, the pressure curve of the tokamak’ vessel increases rapidly to 6E−3 Pa and then decrease to 4.7E−3 Pa. After 100 min of steady rate puffing, we shut the puffing valve, and the pressure of the vessel decreased steeply because of the high pumping speed supplied by cooled torus surface. However, because of the condensed gas layer, the cooling metallic surface’s temperature may rise a little and the pumping ability drops incrementally. The vacuum vessel pressure then stabilizes at a somewhat higher level. When the puffing is shut, the outlet temperature of liquid nitrogen pipe remains at about 80.5 K, while the outlet temperature of liquid helium pipe is 7.75 K. The differential temperature rise between of the helium pipe’s inlet and outlet is about 0.58 K while the change of the nitrogen pipe is negligible. After the gas puffing, the balanced pressure is 1.8E−4 Pa, which is tolerant to discharge pressure. However, after a saturation puffing for 100 min, the torus shade cryopump loses part of its condensation ability due to the buildup of frost on the cooling surface. When we start several other types of pumping devices, the vessel vacuum is drop to 2E−5 Pa, still higher than the pressure of 1E−5 Pa (Fig. 7). During the test or discharge experiment, liquid helium consumption seems to be stable at about 3 g/s. Nitrogen shield can be cooled down from normal room temperature to 80 K in 2 h. The helium pipe can be cooled down to 7 K in 3.5 h. The test shows agreement with the expected coolant consumption rate and cooling down speed. During discharge experiment, thermal losses seen to be negligible because the thermometer at outlet of the pipe indicates no obvious changes in contrast to that in idle operation. The average inlet pressure of liquid helium is 1.2 bar. The pressure drop is about 0.08 bar. For more efficient cooling in the future, cryo-
Fig. 7. A test curve of vacuum vessel pressure changes and can be used to determine the condensation saturation.
genic valve should be controlled and modulated at different cooling temperature. During the fourth campaign of EAST experiment in 2009, this in-vessel cryopump was used for divertor physics experiment. Double null and single null plasma configurations were researched with this divertor pumping in operation. Impurity gas influence to plasma was also researched. The initial results are encouraging. Impurity gas in plasma has been minimized, indicating a longer and more stable plasma discharge. Beside its utilizing for physics experiment, operation rules for the cryopump is also been explored. Because the buildup of the frost layer on cooling surface may impede effective condensation pumping, a regular regeneration should be operated periodically to warm up and then cool down its surface to release trapped neutral particles and maintain its condensational pumping capability. During the experiment in 2009, this cryopump has been regenerated several times, using two kinds of operation. The first is named partial regeneration in which the helium pipe warmed only to above 20 K. This kind of operation needs only half an hour and can be done conveniently during plasma discharge’s interval. The second named complete regeneration is to warm the whole system including torus and feed line up to normal room temperature and then to cool down to their lowest temperature. This can evacuate all impurity gases including water vapor and other kinds of frost. A complete regeneration needs about 8 h. So it is preferred to operating in idle interval. It is estimated that a complete regeneration may be needed every about 400 shots on EAST. Even though the in-vessel cryopump is very effective in pumping deuterium and hydrogen, analysis of mass spectrum shows that it cannot pump some kinds of deuterated gases effectively. When the liquid helium cooling surface’s temperature is a little above 5 K, condensation of helium gas for the main vacuum vessel is not very effective. These gases should be pumped out mainly by high vacuum turbo-molecular pumps or multi-dragging pumps. 4. Summary The first in-vessel cryopump for divertor experiment on EAST has been established. It has a compact structure and minimized heat load from outside and locates under lower outer passive target blanket, which makes it possible to try lower single null plasma configuration that similar to ITER with divertor pumping facility. Modular design and fabrication has improved its flexibility to comply with adjustment of adjacent components. Leakage detection
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has ensured the qualified parts to be used for the next process. The pumping capability seems to be good for a long pulse divertor plasma discharge. The pumping speed is consistent with what is expected. The coolant consuming speed seems in reasonable level and radiation shield seems to be in good function. With this first in-vessel cryopump, more exciting divertor plasma phenomena can be expected in the near future. Acknowledgments This work is partly supported by the ASIPP-GA Cooperation Program in 2005. The Authors acknowledge helpful discussions with
Mr. M.A. Mahdavi and host reception from Mr. P.M. Anderson and helpful language reviewed by Mr. Paul Thomas. References [1] E.E. Reis, et al., 19th Symposium on Fusion Technology, Lisbon, Portugal, 1996. [2] E.E. Reis, et al., 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, New Mexico, 1999. [3] A.S. Bozek, et al., 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, New Mexico, 1999.