Concept definition of vacuum system for loop operation of large helical device

Vacuum/volume

Pergamon PII: SOO42-207X(96)00111-X

47/numbers 6-Upages 1001 to 1004/1996 Copyright 0 1996 Elsevier Science Ltd Printed in Great Britain. All rights reserved 004%207X/36 $15.00+.00

Concept definition of vacuum system for loop operation of large helical device A Miyahara,” N Noda,b K Akaishi,* T Kawamura,b K-N Sato,b Y Murakami,” K Watanabe,d G Horikoshi” and G Tominaga,‘“School of Economics, Teikyo University, Hachioji 792-03, Japan, bNational institute for Fusion Science, Nagoya 464-O 1, Japan, “Naka Fusion Research Establishment, JAERI, lbaraki 31 I-01, Japan, dHydrogen Isotope Research Center, Toyama University, Toyama 930, Japan, “National Laboratory for High Energy Physics, Tukuba 305, Japan, ‘Faculty of Science, Toho University, Funahashi 274, Japan

As the Large Helical Device fLHD), which is under construction at the National Institute for Fusion Science, is installed in a super conducting magnet as confinement magnetic field, it is capable of operating in a steady state loop operation, if proper particle controls of both upstream (plasma region) and downstream (exhaust, impurity removal, H-isotope separation and refueling) are achieved. In the following, not only is the concept of the steady state loop operation described but also several hardware developments to meet this requirement are discussed. Copyright 0 1996 Elsevier Science Ltd. Key words: Fusion research,

helical device, particle balance, foop operation,

vacuum

system.

is the final target

Introduction

The Large Helical Device (LHD), which is under construction at the National Institute for Fusion Science, is installed in a super conducting magnet as the confining magnetic field, so that it is capable of maintaining steady state loop operation if the particle balance of the plasma core is achieved through the recovery of fuels downstream. This is an important research direction for the Helical System because it is not necessary to have a current drive mechanism such as is required for tokamak type machines like ITER,’ TRIAM’ and TORE SUPRA.3 The typical expected plasma and machine parameters of the Large Helical Device (LHD) are as illustrated in Table 1. For magnetic fusion devices, because fuel burning rate is rather low-about 5%-recycling of unburnt fuels is necessary to obtain fuel economy. This means that not only achievement of steady state operation but also the realization of a loop operation

for magnetic fusion machine toward reactor. However, if we consider the loop operation very seriously, downstream of the device must be carefully investigated. This subject has been discussed’ during CDA phase of ITER but not many machines are considering it yet. Although we do not intend to operate with D-T fuels in LHD, we can investigate and simulate the loop operation of magnetic fusion reactor if the proper additional heating device is adopted to keep the energy balance. The most important items for upstream plasma to realize the loop operation are particle unloading after confining time zP and exhaust through the divertor and refuelling by pellet injector. The LHD divertor has a helical configuration’ with the vacuum vessel being dumbell shaped and the magnetic configuration being topologically similar to the double-null shape found in tokamaks. This shape rotates poloidaily along the helical coils of I/m = Z/IO.” Buffle plates are set in order to separate the divertor chamber from the core plasma region. All the port holes are

Table 1. Typical plasma and machine parameters of LHD Machine parameters 3.9 m Major radius 0.5-0.6 m Average minor radius of plasma 210 m’ Volume of plasma vacuum vessel 3 T (on the axis) Magnetic field 300 m* Area of first wall IOm’ Wetted area of the divertor 30 m3 Volume of the divertor chamber

Expected plasma parameters 5 keV Typical plasma temperature 5 X IO”’ 117 1 Typical plasma density 200 ms Typical energy confining time 200 rn\ Typical particle confining time 2.5 m’ Typical plasma volume H, or I)_. Filling gas 5-10s Discharge time

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located at the divertor chamber and some of them are used for divertor pumping. The concept of a local island divertor has been proposed to improve the particle control efficiency.’ Magnetic island structure is utilized to concentrate most of the fluxes coming out of the last closed magnetic surfaces. Pumping at just one toroidal section is enough to exhaust a considerable part of the particle fluxes. Active pumping of the divertor is under discussion and several ideas are proposed. One idea is “Membrane” pumping proposed by Livshits et al.‘, and the other is “Ceramic” turbomolecular pump proposed by Y Murakami et ~1.~~” The former is using super permeability of metals such as Ni, Fe, Pd and Nb, through which hydrogen isotopes could be continuously pumped out to recovering lines, while for He and impurities they go directly to the exhaust systems. The latter scheme will be described in the next section. In the following, we will describe present design of the LHD vacuum system as the starting point of further improvement for the loop operation. Present status of vacuum exhaust system and future direction of improvement The present design of the vacuum exhaust system of LHD is such primarily to meet the requirements for the high performance operation of the machine, namely to reach sufficient nkTr, parameters with 10 s discharge time. Due to the shorter discharge time for the first step of LHD operation, the vacuum system is designed based on inertial pumping. In Figure 1, parameters of

m-2

I

170m3/h

I -3

I

ventstack airblower Figure 1. Block diagram for exhaust system of both vacuum and cryostat

the vacuum vessel and the cryostat together with block diagram for the exhaust system are given. Because LHD has a double vacuum chamber, where the outer chamber is a cryostat and the inner chamber is a plasma vacuum vessel, the vacuum exhaust system consists of a set of two individual pumping systems for both chambers. Usually the exhaust system for the cryostat is less of a problem, but for that of the plasma vacuum vessel, there are some technical restrictions to be solved. The first is that magnetically sensitive vacuum components must be installed sufficiently far from LHD, because of the requirement of very high accuracy for the magnetic configuration of LHD. Therefore, each pumping system must be connected to each chamber through a long duct of around 13 m in length. The second is that initial pumping down of the plasma vacuum vessel and the cryostat from the atmospheric pressure must begin at the same time, because the structural displacement of the vessel due to vacuum force has to be avoided. The third is that the maximum allowable baking temperature of the plasma vacuum vessel is limited to 100°C in order to avoid structural displacement of the vessel due to thermal expansion. As shown in Figure 1, the pumping system for the cryostat is connected to the vacuum vessel through a long vacuum duct (VD-1, 700 mm in diameter and consists of a roughing line (RL-1) and a high vacuum pumping line using two compound molecular pumps (CMP), while the pumping system for the plasma vessel is also connected to the vacuum vessel through a iong vacuum duct (VD-2) 1000 mm in diameter), and consists of a roughing line (RL-2), an ultrahigh vacuum pump line using two turbomolecular pumps (TMP) and an additional pump (CMP). This pump unit is set in parallel with TMP in the pumping line for pumping of high pressure filling gas such as He during glow discharge cleaning of the vessel. Two rotary pumps (RP2) are used in common for the roughing of the cryostat and the plasma vessel. The air blower is set in the backing line of pumping systems to do forced purge of explosive gas such as hydrogen. Of course this design is aimed at fulfilling the requirement of starting and carrying out the high performance experiment, so that it is not sufficient to realize the loop operation of LHD. For example, the active pumping system capable of unburnt fuel recovery is not considered in this design. Also separation of impurities and He ashes from unburnt fuels must be taken into account to achieve loop operation smoothly. Commercially available turbomolecular pumps with metal rotors are used in all presently available large fusion devices. However, it is difficult to use them in high magnetic fields because they are susceptible to magnetic fields. They have to be placed at a distance from the torus vacuum vessel where the stray magnetic field is lower than 0.01 Tesla, or to be shielded with a thick ferromagnetic body. Several years ago, during the time of ITER CDA, a ceramic turbo-molecular pump (TMP) and a new ceramic pump, named the turbo-viscous pump (TVP), were developed and tested in Japan for use in high magnetic fields.8,” The ceramic TVP works in a pressure range of 10-“-10’Pa and can act as a roughing pump or an auxiliary pump for high vacuum pumps. In both ceramic pumps, ceramic rotor assemblies are levitated by gas bearings and are driven by gas impulse turbines. The ceramic TMP uses a non-contact spiral-groove seal to isolate the vacuum side from the high-pressure side (gas bearing). They operate completely oil-free. The main specifications of the ceramic pumps of ceramic turbo-molecular pump and ceramic turbo-viscous pump are listed in Table 2.

A Miyahara

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Table 2. Main specifications

of the ceramic pumps Ceramic CT-500

TMP

Type of pump

Ceramic TVP CT-300H

Rotor material Rotor diameter

Silicon nitride 210 mm

Silicon nitride

Driving method Bearing Rotation speed

Gas impulse turbine Gas bearing 25,000 rpm

Operable Pumping

pressure

speed Ultimate pressure

range

150mm Gas impulse turbine Gas bearing

IO ‘-1 Pa

25.000 rpm loo)- lO”Pa

500 L s ’ (N2) lo-’ Pa

280 L min ~’ (N2) IO- Pa

Impurity removal from exhausted gases and recovery of unburnt fuels

It is desirable that the torus pumping system should be composed of oil-free vacuum pumps regardless of the location of the pumps. When the ceramic TMP and TVP are employed, a driving gas like helium must be mixed with exhausted gas. Therefore. the driving gas, in addition to the ordinary impurity gases such as HzO, CO and CH, must be removed from the downstream gas to recover unburnt fuel. The gas separation unit consists of two sets of adsorption chambers ofdifferent temperatures, presumably 77 K and 20 K, connected in series. In these chambers, the impurity gases and the fuels are adsorbed on molecular sieves separately. Helium passes through both chambers. One of the two parallel sets is in operation while the other is in regeneration. The regeneration temperatures will be 300 K for the ante-chamber and 77 K for the post-chamber. Another system is as proposed by Livshits et ~1.~Unburnt fuel pumping is through super permeative membrane while impurity and He ash will be pumped directly. If it is feasible, it is ideal because the system is very simple. Refuelling system by ice pellet injector In order to carry out the loop operation experiments, one of the most important and essential items to be studied is a continuous fueling method. such as a technique for continuous pellet fueling. In this sense, requirements for the pellet injector”.” in the LHD steady state experiment have been studied. After saturation of wall materials of the vacuum vessel, the particle flux should be equal to the outflow. If we turn off particle fueling, particles are balanced by the recycling with the wall. and a certain density profile will be obtained. However, we can never vary the density profile without an active fueling method. The condition of particle balance after wall saturation requires for the particle influx F;, to be as follows: Fp = ,I,db”t,,, here, 17, is the plasma density, dV the volume element, and T,, the particle confinement time, respectively. In the case of an LHD plasma, the typical value will be around I-IO Pa. m’,s. which is a fairly large value both for fueling and for pumping. The typical region of operation required for the pellet injection has been obtained by simulation study. Thus. if we choose the pellet size to be in the range from 1 to 3 mm, the repetition frequency of injection will be around l-100 Hz with the duration time of 0.5-I h.” In this sense. a new type of continuous freezing device with liquid and solid hydrogen tanks has been designed and constructed. which is planned to be composed in a centrifuge

system in the near future. A stable ice rod of I.2 mm in diameter has been produced in the preliminary experiment. Moreover, in order to achieve loop operation of particles in the LHD. pellets must be injected in steady state with controlled interval so that dN:dt = 0 will be fulfilled. It IS necessary to emphasize that the interval and size control of pellet injection is only the active element to reach the successful loop operation.

Tritium behaviors in the large helical device D-D experiments by the LHD will produce tritium amounting to about I2 Ci,lyear. Because its energy is as high as I MeV, the tritium produced will be hardly confined in the plasma. Most tritium wilt hit the first wall and the divertot- which will be made of carbon composite. This results In the formation of fairly stable T-C bonds. Namely, tritium is captured in the first wall and/or divertor, and gives rise to a problem of handling contaminated materiais as well as tritium release during the maintenance shut downs. To avoid unnecessary tritium exposure to personnel and into the environment, the captured tritium should be removed to a level as low as possible before the maintenance shuts down. To decontaminate relevant materials and;oreyuipment. it is essential to know in advance the tritium inventory distribution in the system. It is also important to establish reliable methods to decontaminate the materials and;or devices. It is expected that the tritium concentration in the exhaust gas from the vacuum line is at very low level. According to the experiences of JT--601; as well as TFTR. the tritium concentration will be around one tenth Bq, cm’ at the end of vacuum line.“-“ It LVH> routinely measured with use of ionization chambers. However. it requires a counting or ambient gas and consequently the tritium concentration in effluent gas was only measured at the end of the vacuum line so far for the similar D--D burning experiments. in addition, this concentration level is almost at the back ground level of ;I conventional

ionization

chamber

and

it is not easy

to measure

pre-

cisely the tritium concentration of this level. To look in more detail at the tritium behavior in the vacuum line. it is necessary to measure ill .sir;l the flow- of tritium t&n) the torus to the end of the vacuum line. This is possible with USC of‘suitable electron detectors. For this purpose. electron multipliers wil! be adequate. by which tritium [I rays as low ah 0. I cps could be measured.“’ ” With respect to chemical fbrm of tritium, ir is important to identify elemental (HT. DT) and water (HTO. DTO) species, because it is reported that wet tritium is about 75 000 times more hazardous than dry. For this purpose. an ionization chamber installed with a H?O permeator is developed and is now commercially available. This is planned to he used. With use of this device, chemical form identification should he possible by a single 1003

A Miyahara et al: Loop operation

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device. It does not require wet tritium trapping and liquid scintillation counting. Then the tritium monitoring at the end of the vacuum line should be possible in situ, shot by shot or day by day. Most of the tritium is expected to be captured by the first wall and/or divertor. Estimation of the surface tritium concentration has not been completed yet for the LHD. But according to the experiences of JET and TFTR, the surface concentrations were in the order of 1O’“-10” (17-l 70Bq/cm’) and of 10”-lO”T/cm ( IO?-104Bq/cm’), respectively. This, of course, depends on the experimental conditions, but it is noted that the surface concentration limit is 40Bq/cm* according to the Japanese regulation. Consequently, waste handling is unavoidable for those materials. The captured tritium was conventionally measured by dissolving the first wall material in acidic solution and by following liquid scintillation counting. However, surface tritium can be measured in sits with use of alternative devices. A device has been developed by Malinovsky,lY who proposed the use of a secondary electron multiplier, channeltron, by which secondary electrons produced by tritium /?-rays could be detected in situ. Another device for detecting bremsstrahlung X-rays is now under investigation by Matsuyama et aI.” Although, at present, this method is less sensitive in comparison with Malinovsky’s device, it can detect tritium presenting about 1000 times greater depth than the SEM method. Both of these devices can be applied to the LHD for in situ monitoring of surface tritium concentration in the torus.

Conclusions The loop operation with D-T fuels for the magnetic fusion reactor is a necessary step because the fuels are expensive and tritium has radio-activity. Moreover, the burn fraction is small and almost all fuels have to be recycled. This is quite different in comparison to the inertial fusion reactor, which has a high burn rate of over 90%. It means that the study of fuel recycling is a very important

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subject for future fusion reactors, but not many researchers have seriously considered this problem. For example, we do not know what kind of impurities will appear at the outlet of the roughing pump and in what amount they will appear. Also, the necessary times of impurities removal and ice pellet preparation must be measured in order to analyze the system stability. Preliminary but important data can be obtained by light hydrogen experiment, but more detail data will be gained from D-D discharge devices. In future we have to collect such data in order to establish a reasonable scenario for the loop operation of magnetic fusion reactors. References ‘S A Cohen et al., J Nucl Muter, 196-198, 50 (1992). ‘S Itoh rt ul., in Plasma Ph_vsicsand Controlled Nuclear Fusion Research 1990 (13th IAEA Conf), IAEA Vienna, 1, 733 (1991). ‘Equipe TORE SUPRA, in Plasma Physics and Controlled Nuclear Fusion Research 1992 (14th IAEA Co@, IAEA Vienna, 1.79 (1993). 4N Noda, Y Kubota, A Sagara rf al., Fusion Technology 1992, (Proc qf 17th SOFT, 1992, Rome), p 325-328. 5N Ohyabu ef al., J Nucl Mater, 220-222, 298 (1995). bA I Livshits et al., J Nucl Mater, 220-222, 259 (1995). ‘K Akaishi and T Kawamura, Vuclcum,41, 1552 (1990). ‘T Abe and Y Murakami et al., Vacuum, 41, 1992 (1990). ‘Y Murakami and T Abe et al., J Vuc Sci Technof, A9, 2053 (1991). ‘*H Suzuki and A I Livshits, Paper PSI-MoA-7, this conference. ” M Sakamoto er ul., Plasma Phys and Controlled Fusion, 33, 583 (199i). “K N Sato et al., 1994 International Conjkence on Plasma Physics, 1, 93 (1994).

“K N Sato et ul., Proc of’ 15th S)wzposium on Fusion Engineering, 1, 40 (1994). 14M Miya and N Toyoshima, Fusion Technol, 507,26 (1994). “H F Dylla el al., PPPL and Sandiu Report, PPPL-2523/SAND88-8212 (1988). “R A Kerst and M E Malinovsky, Proc Tritium Technol. in Fission, Fusion and Isotop Appl, A3, 346 (1985). “K Ichimura et al., Radioisotopes, 34, 83 (1985). “K Ichimura et al., Ann Report, TRC, Toyama University, 3, 17 (1993). “M E Malinovsky, Appr Phys Letters, 39, 509 (1980).

“‘M Matsuyama and K Watanabe, to be submitted.