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Liquid-metal-cooled divertor for ARIES
Fusion Engineering and Design Fusion
Engineering
and Design
Liquid-metal-cooled
29 (1995) 98- 104
divertor for ARIES
E. Muraviev Russian Scientific
Center Kurchutm
Institute,
11025 North
Torrey Pines Roud, La Jollu, CA 920.77. USA
Abstract The key features of the proposed divertor design concept based on the specific liquid metal (LM) coolant properties are as follows: the requirement of the vacuum tightness of the divertor cooling tract is dismissed; the pressurized coolant ducts can be separated from the plasma-facing structure (PFS) elements which are subjected to the thermal loads, and with this feature the PFS can be replaced independently, without disturbing the cooling system. The divertor targets of the type found in free LM-flow-cooled cells have been adjusted to the double-null gaseous divertor geometry. The performed design study included the following major issues: a materials review; in-reactor thermal hydraulics; divertor target thermal-stress analysis; LM loop evaluation; safety considerations, including Ga activation, accident analysis and estimate of the dose due to tritium release.
1. Introduction This paper represents an overview of the design study of a divertor system with liquid metal (LM) coolant (gallium) related to the ARIES project. The work has been conducted by a group of specialists from the Institute of Nuclear Fusion of the Russian Scientific Center Kurchatov Institute [I]. The divertor design has always presented a challenge for physicists and engineers. The major concerns identified at the very early stages of tokamak reactor design studies were high thermal and particle loads resulting in thermal stresses, fast erosion and short lifetime of the divertor targets In this connection attractive ideas of using LM to form a stress-free and erosion-insensitive working surface of the divertor target have appeared. These concepts have been investigated for many years in different countries, though the major work, including experimental studies on MHD test facilities and in a tokamak, was probably done in Russia [2,3]. In partic0920-3796/95/$09.50 &:: 1995 Elsevier SSDIO920-3796(94)00100-6
Science S.A. All rights
ular in these studies gallium was named as the most promising LM for use in the divertor environment. Nevertheless the major concern with respect to the open LM surface devices in the tokamak plasma chamber remained the problem of plasma impurity influx that might result from the plasma contact with LM. At the same time the divertor problem was recognized as the most crucial one for the practical development of a fusion reactor. This fact was confirmed by the conceptual design of the International Thermonuclear Reactor (ITER) [4]. In this context the advantages of LM coolants in genera1 and Ga in particular seemed to be very important for the specific divertor environment and worth trying to employ. Taking into account difficulties with experimental proving of the free LM surface divertor target concept, a new divertor concept where the LM (Ga) is to be used only as a coolant has been proposed by the author together with Chuyanov et al. [5]. In this case the plasma-facing surface (PFS) can be made of the pertinent solid material. Luckily reserved
E. Murucie~
/ Fusion Engineering
enough all the presently considered candidate materials for the PFS of the divertor plate, namely graphite, tungsten and beryllium, turn out to be the best materials compatible with Ga, which is generally known as a highly corrosive LM (especially above 300°C). The corrosion rate in Ga at 400°C is 10mh g mm2 hh’ for W and 10e4 g mm2 h-’ for Be while it is 7.1 g mm7 h-’ for Fe and 1100 g rnmr h-i for Cu [6]. This new concept has been applied in a design study for the ARIES project.
2. Key features of the concept The key features of the proposed divertor design concept based on the specific LM coolant properties are as follows. (a) The requirement of the vacuum tightness of the divertor cooling tract is dismissed. (b) The pressurized coolant ducts can be separated from the PFS elements which are subjected to the thermal loads, and with this feature the PFS can be replaced independently, without disturbing the cooling system; this is achieved using free LM jets sprayed on the back of the PFS elements, free LM film cooling and free LM draining under the action of gravity; obviously this point is more suitable for a single-null bottom divertor. The first basic feature of the proposed concept gives us the following major advantages against water or gas cooling options: non-sensitivity to a small leakage in the divertor target cooling tract within the vacuum chamber and therefore elimination of the possibility of a whole type of internal LOVA that might have the strongest impact on the reactor reliability and availability; elimination of the possibility of another whole type of accident with the vacuum chamber pressurizing due to internal cooling tract rupture; easy replacement of the whole divertor target module without welding or brazing operations. The second basic feature provides additional important advantages: the feasibility of PFS element replacement separately from the cooling system, without disturbing the whole module; a longer PFS element fatigue life in the absence of a brazed joint with the heat sink structure; a longer operating life of the pressurized coolant feed pipes due to absence of thermal stresses. Besides these, which are specifically for comparison with an ITER-like water cooled divertor, such advantages should be emphasized as: no coolant interaction with hot plasma-facing material in the case of an ingress-of-coolant event and no hydrogen production; no
and Design 29 (1995)
98- 104
99
need for cooling system protection from run-away electrons (due to the LM’s ability to absorb large amounts of energy without producing high pressure); no critical heat flux limits. For the ARIES project the described system had to be embedded into a completed and self-sufficient reactor design affording no significant changes in it. Though the concept of a gaseous divertor accepted in the ARIES project assumed very low and uniform thermal loads to the divertor plates and no sputtering, some strong doubts about its viability still exist. So for this study it was supposed expedient to try to apply the specific advantage of the new divertor concept, namely its tolerance to higher heat flux and shortened erosion lifetime of the plasma-facing components, to ARIES conditions assuming that in the gaseous divertor some non-uniformity of thermal load might occur. Specifically the target value of the peak heat flux of 15 MW mm2 has been established. As the structural material for the divertor system the same vanadium alloy has been accepted as for the blanket design of ARIES-II. A protective (and renewable in situ) Be coating of the PFS can be considered in principle. Should Be utilization be assumed, the PFS elements with free LM flow cooling could be made entirely of Be. An electroinsulating coating is envisaged on all metallic surfaces in contact with LM. It is of selfformed, self-healing oxide film type, lo-20 urn thick, which basically does not affect heat transfer. Its feasibility is assumed on the basis of experience of developing such films on vanadium alloys in lead and lithium [7,8], as well as on existing technology of Pb-Bi coolant used in fission reactors [9].
3. Design description In the developed design the overall configuration of the LM-cooled divertor target system is practically identical to that of the reference gaseous divertor (Fig. 1). Both the top and bottom divertor incorporate LM feeding pipelines, targets and drain ducts. Each divertor structure is split into separate modules. The total number of modules in a divertor is 32. One module incorporates five targets of two different types: for low heat flux (LHF) and high heat flux (HHF). To provide a continuous toroidal divertor target belt there are two types of modules arranged in pairs: one of rectangular and the other of trapezium shape in plane. The module is fixed on a tube frame attached to the inner and outer blanket walls. Each module has its own system of feeding and
100
E. Muraviev 1 Fusion Engineering and Design 29 (1995) 98-104
Fig. 1. Vertical cross-section of the top divertor: 1, inboard LHF target; 2, inboard HHF target; 3, central LHF target; 4, outboard HHF target; 5, outboard LHF target; 6, supporting tubular structure; 7, LM feeding pipes; 8, drain pipes to the drain channels in the blanket.
draining pipelines. LM draining from the top divertor is carried out by using part of the blanket ducts. In the bottom divertor special draining pipelines are envisaged that consist of arrays of tubes with comparatively small diameter and smooth bending. The non-pressurized draining ducts inside the vacuum chamber have collective joints allowing easy connections and disconnections without welding or brazing operations. Small LM leakages from these joints are removed into the same draining duct. The maximum heat flux from the plasma is expected within a narrow belt of about 60-80 mm width near the separatrix. The HHF target design has been developed as a set of inclined plates (troughs) with a plasma-facing wall of thickness 1 mm, length 250 mm (established with respect to possible oscillations of the separatrix) and width 80 mm, with the side stiffening walls 5 mm high. The troughs are fixed cantilevered by means of legs formed as some higher parts of the side walls (Fig. 2). The holder is combined with the manifold tube that has a set of orifices from which LM jets are ejected on the cooled trough surfaces. After hitting the plates LM jets form a thin film that flows down due to its own inertia as well as under the action of gravity. At the end of the trough the film enters a flattened receiver manifold from which the LM flows into the draining ducts. The LHF targets are made in the form of curved panels consisting of a set of flat wedged cooling tubes encased with clearances into a protective shell also filled
Fig. 2. HHF target design: 1, troughs; 2, inlet manifold; 4, orifices.
3
fixing
legs; 3, LM
2
Fig. 3. LHF panel target cross-section: I, pane1 case front wall; 2, panel case back wall; 3, flattened cooling tube; 4, distancing ribs. with LM (Fig. 3). On the outer surface of the cooling tubes there are segmented longitudinal ribs. The protective shell walls are joined with the cooling tubes by diffusion welding and then the entire stack is curved by pressing in the vertical plane according to the required target configuration. Since the toroidal width of a single target is only about 1 m it does not need to be curved in this direction, thus a complicated technology for shaping double-curved panels is not necessary. The pressurized duct formed by the cooling tubes is separated from the low pressure LM between them within the protective shell. In the case of a cooling tube rupture the LM can flow through the intertube space to the drain system without spraying into plasma. The entire LM divertor system is split into 16 separate loops according to the number of toroidal field
E. Muraviev
Table 1 Summary
/ Fusion
Engineering
and Design
29 (1995)
98- 104
101
of the design parameters
Total thermal power removed by LM divertor system (MW) Total cooled surface area (m*) Number of divertor target modules in the top/bottom divertor Total LM coolant flow rate (m3 SK’) Average LM coolant temperature rise (‘C) Required pressure at the reactor inlet (top/bottom) (MPa) Inboard and central targets Outboard targets LM inventory inside reactor (m’) Top divertor Bottom divertor, pressurized pipelines and targets Bottom divertor, drain system Total LM inventory in the divertor system ( m3) Total required pumping power (MW) Estimated maximum tritium inventory in the LM loop (g)
316 278 32/32 1.3 150 1.0/1.2 0.7/1.0 6.79 1.36 2.96 2.41 28.2 2.63 16-30 _.
coils
and
the
between
them.
bottom
divertor
subsystems:
vacuum Each and
chamber
LM loop incorporates
access serves the
ports
for
both
following
located top
and
major
LM clean-up and conditioning system; heat tritium extraction-processing system; main and auxiliary pumps; LM dump tank and
exchangers; circulation storage. Since the different cooling lines of the top and bottom divertors require different inlet pressures to circulate LM, the pump system of each loop is equipped with several pumps connected in series to permit intermediate coolant bleeding. The design parameters are summarised in Table 1.
4. In-reactor thermal hydraulics In the open cooling cell of the HHF target the impinging LM jets form a film flow. Since the jet orifices make a row in the toroidal direction the film flows in the quasi-coplanar (close to coplanar) magnetic field with the major toroidal component B, lying in the film plane and perpendicular to the velocity vector. The strong coplanar magnetic field should stimulate fast formation of a uniform film flow with a transition length of about the distance between the jets (the orifices pitch). This portion of the target is screened from the heat flux by the LHF panel (Fig. 1). For the selected jet diameter above 2 mm and 10 mm pitch the minimum film thickness of 0.6 mm can be expected. This was considered acceptable for the comparatively small cooling cells (only 80 mm wide). Fast thin LM films with almost exactly such a geometry in a coplanar
magnetic field have been obtained by Evtushenko et al. in experiments, including “flow on ceiling” [lo]. The coplanar magnetic field interacts rather weakly with averaged flow and in this respect is similar to the longitudinal magnetic field. It suppresses turbulent pulsations and thus can reduce the heat transfer coefficients. Therefore in the absence of experimental data on heat transfer in a coplanar magnetic field an empirical relation suggested for LM flow in longitudinal magnetic field by Genin and Sviridov [ 1I] has been used. The pressure drops within the reactor magnetic field have been calculated for the detailed geometry of the LM-feeding pipelines in a quasi-one-dimensional approximation of MHD flow. This means splitting long segments of the duct into pieces for which the quasiuniform flow approach is applicable. The total pressure drop for a line was found as the sum of the pressure losses in each of these duct pieces and the local pressure drops in various bends and joints (detailed model description and calculation results are presented in [ 11). Ten pressurized feeding pipelines (five lines for five targets within the module of each of two divertors) were assessed with respect to LM inventory minimization at the selected pressure loss levels. Besides this four drain lines located within the bottom divertor chamber were calculated: the inboard blanket drain line receiving the total coolant flow from the inboard LHF and HHF targets of the top divertor; the inboard HHF target drain line of the bottom divertor; the outboard HHF target drain line of the bottom divertor; the outboard blanket drain line receiving the total coolant flow from the central and outboard LHF and HHF targets of the top divertor.
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The minimum LM inventory in this case was found under the condition p = 0 (where p is the absolute pressure in the LM) at any point along the line.
5. Thermal stress analysis Temperature and stress distributions were calculated for the outboard HHF target, where the thermal load was supposed to be a factor of 1.5 higher than that for the inboard one. For this analysis temperature-dependent properties of the vanadium alloy have been used. All calculations were carried out by means of a finite element code. The maximum temperature of about 740°C occurred in the heat flux peak area on the PFS of the cell. The maximum stresses cr,,, (von Mises) and oi = 0, - c3 are localized in the same area. The maximum stress value g,,, is about 216 MPa and 0, = 224 MPa. For the mentioned temperature of 740°C the yield stress is about 330 MPa and the ultimate tensile strength is 560 MPa. According to American Society for Mechanical Engineers code the limit value S,,, is the least of the following quantities: ( 1/3)fl, (average ultimate tensile strength) and (2/3)00,, (average yield stress) at the corresponding temperature. In our case S, = 187 MPa. For the stress calculations based on a linear elastic material model the equivalent stresses g,,, and cri should be lower than 1.2& = 225 MPa. This condition is fulfilled in our case. Similar results were obtained for the inboard, central and outboard LHF panel targets. In this case the analysis was conducted in two steps: first, a simplified three-dimensional model was employed to determine the location of maximum stresses and then more detailed two-dimensional temperature-stress analysis was carried out for critical regions. The stress values were checked both for the target case walls and for the cooling tubes.
6. Safety considerations Gallium activation characteristics have been calculated (by Kashirskij) to obtain activation afterheat data for accident analysis. From this viewpoint the most interesting parameters were maximum values of specific activation energy release during reactor operation and afterheat dynamics. Neutron flux with 46 energy groups was used in activation calculations. The flux data were available for three points in the middle plane of the ARIES-II blanket: at the first wall of the outboard
and Design 29 (1995) 98- 104
blanket; and at 10 and 20 cm inside the blanket. The activation cross-section library and nuclear physical data were based on ACTL-82 [ 121 and other sources. In particular, the decay energy values were calculated with data from Refs. [13,14]. It was found that the energy release in gallium is rather high during operation and about 3 days after shut-down. At the moment of shut-down it equals about 4.2 x 10” MeV s-’ cm-3. So if there are any zones with large amounts of stagnant gallium staying permanently under exposure during all the operation time it may result in a serious problem with the afterheat. In flowing gallium coolant the average energy release at shut-down is considerably lower and amounts to about 6.8 x IO’* MeV s-’ cme3. ;’ rays produced by Ga activation products have rather high energy and, since the divertor target structure is relatively thin, only a small fraction of energy released in the material will be actually transformed into afterheat in situ. An estimated transformation factor is approximately 0.1 at O.S- 1.O cm of total target thickness. From the viewpoint of accident analysis the afterheat in Ga should be a matter for consideration only within the first 10 days after shut-down. For radioactive waste disposal it does not present a greater problem than V alloy. The following accidents have been selected for firstpriority consideration: ( 1) LOFA in two types of targets (HHF and LHF) (1 .l) LOFA resulting from an external event, two versions: (a) coolant stays inside cooling tubes (b) coolant drained from cooling tubes (1.2) LOFA resulting from an internal event (a) plugging up of one or a few cooling tubes in an LHF target (b) plugging up of one or a few jet orifices in an HHF target (2) loss-of-coolant accident (LOCA) (2.1) LOCA resulting from external events (2.2) spilling of liquid gallium outside the reactor On the basis of the obtained results it was concluded that neither LOFAs nor LOCAs in LHF targets would be dangerous if the reactor were shut down (the fusion reaction terminated) within the first 10 s after the accident began and the first wall cooling system kept working. In this case the LHF target temperature would not exceed 500°C. LOFA with 10 s reactor shut-down time and no heat removal would lead to a temperature rise of up to 1000°C within the first 5 h after the accident began. LOFA calculations taking into account various scenarios of the heat flux decrease during plasma shut-
E. Murmiev
/ Fusion Engineering
down have demonstrated this to be an important issue that must be subject to further analyses. Spilling of the LM coolant would not lead to any appreciable temperature rise if the thickness of the spilled coolant layer is about 1 cm, but there may be major consequences if it is about 10 cm. Some preliminary estimates of tritium doses were performed for accidents with Ga spilling from the divertor target module during maintenance and in the case of coolant loop rupture during reactor operation. For the first case tritium release into the reactor hall was estimated as 0.2-4 g depending on the efficiency of the Ga-cleaning system during previous operation. For the second case the release value can be as high as 2-40 g. As a tritium dose characteristic a dose equivalent at 1 km distance from the release point was used. An upper limit estimate under most conservative assumptions (including no containment credit) gave an accident dose of 85 mSv (the accident dose limit is 100 mSv).
cud Design 29 (1995)
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103
gen adsorption properties in the vacuum chamber environment; technology for Ga purification from PFS erosion products and tritium. It was hoped that the design solutions suggested in this study would allow the corresponding tasks to be specified more exactly.
Acknowledgments The author, on behalf of the entire working team, wishes to acknowledge the initiative and friendly support from our American colleagues that made this work possible on a contractual basis with General Atomics, San Diego, CA, USA: Dr. K. Schultz, Mr. S. Ethridge and Dr. C. Wong, as well as useful cooperation from Dr. L. Elguebaly and Dr. C. Bathke.
References
[II E.V. Muraviev, V.S. Petrov, P.V. Romanov, L.N. Topil7. Conclusions The performed design study was the first attempt at a complete divertor system analysis based on a new LMcooled divertor target concept. It seems that such major objectives of this work as to insure higher permissible thermal loads to the divertor targets (up to 15 MW m-‘) and to demonstrate the feasibility of a divertor design with easy maintenance and high duty cycle potential have been successfully achieved. The impact from the divertor system on the tokamak reactor can be characterized by the following major points: the necessity of using some part (though not large) of the blanket cooling ducts for draining LM from the top divertor; the need for rooms directly adjacent to the tokamak reactor for placing the divertor LM loop equipment to minimize gallium inventory. On the whole the feasibility of the proposed concept will depend on successful solution of a number of problems that require certain research and development efforts. The most important of these are the following: the technology of thin electroinsulating films or coatings formation on metallic walls in contact with Ga; stability and self-healing of the electroinsulating films under neutron irradiation; a data base on corrosion and mass transfer under the conditions of a non-isothermal gallium loop with a working section in a magnetic field; a data base for the LM heat transfer in MHD film flow; a data base for MHD jet flow-wall interaction; a data base for hydrogen isotope solubility in Ga and hydro-
ski, A.V. Kashirski et al., Liquid metal cooled divertor for ARIES, Technical Report of General Atomics GAA21755, General Atomics, San Diego, CA, 1994. [21B.G. Karasev, O.A. Lielausis, E.V. Muraviev and A.V. Tananaev, Liquid metals in fusion reactor with magnetic confinement, in Fusion Reactor Design and Technology 1986, International Atomic Energy Agency, Vienna, 1987. A.F. Kolesnichenko, S.V. Mirnov, [31 V.N. Dem’yanenko, E.V. Muraviev et al.. Liquid metal limiter of a tokamak: task formulation and first results, Fiz. Plazmy l4(5) ( 1988) 628-632. ITER Documentation [41 ITER plasma facing components, Series 30, International Atomic Energy Agency, Vienna, 1991. [5] V.A. Chuyanov, A.V. Klishchenko, E.V. Muraviev and V.S. Petrov, New concept of ITER divertor. presented at ITER Topical Meeting, Garching, June 1992. [6] S.P. Jatsenko, Gallium: Interaction with Metals, Nauka, Moscow. 1974. [7] Conceptual design of a DEMO reactor, Preliminary Report, Russian Scientific Center Kurchatov Institute, La Jolla, CA, 1993 (in Russian). [8] D.K. Sze, R.F. Mattas, A.B. Hull, B. Picologlou and D.L. Smith, MHD considerations for a self-cooled lithium blanket, presented at the 10th Topical Meeting on the Technology of Fusion Energy, Boston, MA, June 1992. [9] A.V. Beznosov. B.F. Gromov, E.V. Muraviev, V.V. Orlov, Yu.1. Orlov et al., Heavy liquid metal coolants on the basis of lead in cooling system of fusion reactor with magnetic plasma confinement, At. Energ. 71(4) (1991) 506-51 I. [IO] I.A. Evtushenko, I.R. Kirillov, Yu.M. Krivchenkov, T.V. Mazul, V.L. Matveev et al., Divertor target with liquid
104
E. Muraviev 1 Fusion Engineering and Design 29 (1995) 98- 104
metal jet impinging upon a solid surface, presented at ISFNT-3, Los Angeles, CA, 27 June- I July 1994, Paper DP2-16. [ 1l] L.C. Genin and V.G. Sviridov, Heat transfer and temperature fields in the thermal entry length in liquid metal flow in longitudinal magnetic field, Magn. Gidrodin. (2) (1983) (in Russian). [ 121 H.D. Lemmel, ACTL-82, The LLNL Neutron Activation
Cross-Section Library of 1982 (Summary of Contents), International Atomic Energy Agency Nuclear Data Section 55, March 1983. [ 131 O.F. Nemets and Ju.V. Gofman, Spravochnik po jadernoj fizike, Naukova Dumka, Kiev, 1975 (in Russian). [I41 N.G. Gusev and P.P. Dmitriev, Kvantovoe izluchenie radioaktivnykh nuklidov, Atomizdat, Moscow, 1977 (in Russian).
Engineering
and Design
Liquid-metal-cooled
29 (1995) 98- 104
divertor for ARIES
E. Muraviev Russian Scientific
Center Kurchutm
Institute,
11025 North
Torrey Pines Roud, La Jollu, CA 920.77. USA
Abstract The key features of the proposed divertor design concept based on the specific liquid metal (LM) coolant properties are as follows: the requirement of the vacuum tightness of the divertor cooling tract is dismissed; the pressurized coolant ducts can be separated from the plasma-facing structure (PFS) elements which are subjected to the thermal loads, and with this feature the PFS can be replaced independently, without disturbing the cooling system. The divertor targets of the type found in free LM-flow-cooled cells have been adjusted to the double-null gaseous divertor geometry. The performed design study included the following major issues: a materials review; in-reactor thermal hydraulics; divertor target thermal-stress analysis; LM loop evaluation; safety considerations, including Ga activation, accident analysis and estimate of the dose due to tritium release.
1. Introduction This paper represents an overview of the design study of a divertor system with liquid metal (LM) coolant (gallium) related to the ARIES project. The work has been conducted by a group of specialists from the Institute of Nuclear Fusion of the Russian Scientific Center Kurchatov Institute [I]. The divertor design has always presented a challenge for physicists and engineers. The major concerns identified at the very early stages of tokamak reactor design studies were high thermal and particle loads resulting in thermal stresses, fast erosion and short lifetime of the divertor targets In this connection attractive ideas of using LM to form a stress-free and erosion-insensitive working surface of the divertor target have appeared. These concepts have been investigated for many years in different countries, though the major work, including experimental studies on MHD test facilities and in a tokamak, was probably done in Russia [2,3]. In partic0920-3796/95/$09.50 &:: 1995 Elsevier SSDIO920-3796(94)00100-6
Science S.A. All rights
ular in these studies gallium was named as the most promising LM for use in the divertor environment. Nevertheless the major concern with respect to the open LM surface devices in the tokamak plasma chamber remained the problem of plasma impurity influx that might result from the plasma contact with LM. At the same time the divertor problem was recognized as the most crucial one for the practical development of a fusion reactor. This fact was confirmed by the conceptual design of the International Thermonuclear Reactor (ITER) [4]. In this context the advantages of LM coolants in genera1 and Ga in particular seemed to be very important for the specific divertor environment and worth trying to employ. Taking into account difficulties with experimental proving of the free LM surface divertor target concept, a new divertor concept where the LM (Ga) is to be used only as a coolant has been proposed by the author together with Chuyanov et al. [5]. In this case the plasma-facing surface (PFS) can be made of the pertinent solid material. Luckily reserved
E. Murucie~
/ Fusion Engineering
enough all the presently considered candidate materials for the PFS of the divertor plate, namely graphite, tungsten and beryllium, turn out to be the best materials compatible with Ga, which is generally known as a highly corrosive LM (especially above 300°C). The corrosion rate in Ga at 400°C is 10mh g mm2 hh’ for W and 10e4 g mm2 h-’ for Be while it is 7.1 g mm7 h-’ for Fe and 1100 g rnmr h-i for Cu [6]. This new concept has been applied in a design study for the ARIES project.
2. Key features of the concept The key features of the proposed divertor design concept based on the specific LM coolant properties are as follows. (a) The requirement of the vacuum tightness of the divertor cooling tract is dismissed. (b) The pressurized coolant ducts can be separated from the PFS elements which are subjected to the thermal loads, and with this feature the PFS can be replaced independently, without disturbing the cooling system; this is achieved using free LM jets sprayed on the back of the PFS elements, free LM film cooling and free LM draining under the action of gravity; obviously this point is more suitable for a single-null bottom divertor. The first basic feature of the proposed concept gives us the following major advantages against water or gas cooling options: non-sensitivity to a small leakage in the divertor target cooling tract within the vacuum chamber and therefore elimination of the possibility of a whole type of internal LOVA that might have the strongest impact on the reactor reliability and availability; elimination of the possibility of another whole type of accident with the vacuum chamber pressurizing due to internal cooling tract rupture; easy replacement of the whole divertor target module without welding or brazing operations. The second basic feature provides additional important advantages: the feasibility of PFS element replacement separately from the cooling system, without disturbing the whole module; a longer PFS element fatigue life in the absence of a brazed joint with the heat sink structure; a longer operating life of the pressurized coolant feed pipes due to absence of thermal stresses. Besides these, which are specifically for comparison with an ITER-like water cooled divertor, such advantages should be emphasized as: no coolant interaction with hot plasma-facing material in the case of an ingress-of-coolant event and no hydrogen production; no
and Design 29 (1995)
98- 104
99
need for cooling system protection from run-away electrons (due to the LM’s ability to absorb large amounts of energy without producing high pressure); no critical heat flux limits. For the ARIES project the described system had to be embedded into a completed and self-sufficient reactor design affording no significant changes in it. Though the concept of a gaseous divertor accepted in the ARIES project assumed very low and uniform thermal loads to the divertor plates and no sputtering, some strong doubts about its viability still exist. So for this study it was supposed expedient to try to apply the specific advantage of the new divertor concept, namely its tolerance to higher heat flux and shortened erosion lifetime of the plasma-facing components, to ARIES conditions assuming that in the gaseous divertor some non-uniformity of thermal load might occur. Specifically the target value of the peak heat flux of 15 MW mm2 has been established. As the structural material for the divertor system the same vanadium alloy has been accepted as for the blanket design of ARIES-II. A protective (and renewable in situ) Be coating of the PFS can be considered in principle. Should Be utilization be assumed, the PFS elements with free LM flow cooling could be made entirely of Be. An electroinsulating coating is envisaged on all metallic surfaces in contact with LM. It is of selfformed, self-healing oxide film type, lo-20 urn thick, which basically does not affect heat transfer. Its feasibility is assumed on the basis of experience of developing such films on vanadium alloys in lead and lithium [7,8], as well as on existing technology of Pb-Bi coolant used in fission reactors [9].
3. Design description In the developed design the overall configuration of the LM-cooled divertor target system is practically identical to that of the reference gaseous divertor (Fig. 1). Both the top and bottom divertor incorporate LM feeding pipelines, targets and drain ducts. Each divertor structure is split into separate modules. The total number of modules in a divertor is 32. One module incorporates five targets of two different types: for low heat flux (LHF) and high heat flux (HHF). To provide a continuous toroidal divertor target belt there are two types of modules arranged in pairs: one of rectangular and the other of trapezium shape in plane. The module is fixed on a tube frame attached to the inner and outer blanket walls. Each module has its own system of feeding and
100
E. Muraviev 1 Fusion Engineering and Design 29 (1995) 98-104
Fig. 1. Vertical cross-section of the top divertor: 1, inboard LHF target; 2, inboard HHF target; 3, central LHF target; 4, outboard HHF target; 5, outboard LHF target; 6, supporting tubular structure; 7, LM feeding pipes; 8, drain pipes to the drain channels in the blanket.
draining pipelines. LM draining from the top divertor is carried out by using part of the blanket ducts. In the bottom divertor special draining pipelines are envisaged that consist of arrays of tubes with comparatively small diameter and smooth bending. The non-pressurized draining ducts inside the vacuum chamber have collective joints allowing easy connections and disconnections without welding or brazing operations. Small LM leakages from these joints are removed into the same draining duct. The maximum heat flux from the plasma is expected within a narrow belt of about 60-80 mm width near the separatrix. The HHF target design has been developed as a set of inclined plates (troughs) with a plasma-facing wall of thickness 1 mm, length 250 mm (established with respect to possible oscillations of the separatrix) and width 80 mm, with the side stiffening walls 5 mm high. The troughs are fixed cantilevered by means of legs formed as some higher parts of the side walls (Fig. 2). The holder is combined with the manifold tube that has a set of orifices from which LM jets are ejected on the cooled trough surfaces. After hitting the plates LM jets form a thin film that flows down due to its own inertia as well as under the action of gravity. At the end of the trough the film enters a flattened receiver manifold from which the LM flows into the draining ducts. The LHF targets are made in the form of curved panels consisting of a set of flat wedged cooling tubes encased with clearances into a protective shell also filled
Fig. 2. HHF target design: 1, troughs; 2, inlet manifold; 4, orifices.
3
fixing
legs; 3, LM
2
Fig. 3. LHF panel target cross-section: I, pane1 case front wall; 2, panel case back wall; 3, flattened cooling tube; 4, distancing ribs. with LM (Fig. 3). On the outer surface of the cooling tubes there are segmented longitudinal ribs. The protective shell walls are joined with the cooling tubes by diffusion welding and then the entire stack is curved by pressing in the vertical plane according to the required target configuration. Since the toroidal width of a single target is only about 1 m it does not need to be curved in this direction, thus a complicated technology for shaping double-curved panels is not necessary. The pressurized duct formed by the cooling tubes is separated from the low pressure LM between them within the protective shell. In the case of a cooling tube rupture the LM can flow through the intertube space to the drain system without spraying into plasma. The entire LM divertor system is split into 16 separate loops according to the number of toroidal field
E. Muraviev
Table 1 Summary
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Engineering
and Design
29 (1995)
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101
of the design parameters
Total thermal power removed by LM divertor system (MW) Total cooled surface area (m*) Number of divertor target modules in the top/bottom divertor Total LM coolant flow rate (m3 SK’) Average LM coolant temperature rise (‘C) Required pressure at the reactor inlet (top/bottom) (MPa) Inboard and central targets Outboard targets LM inventory inside reactor (m’) Top divertor Bottom divertor, pressurized pipelines and targets Bottom divertor, drain system Total LM inventory in the divertor system ( m3) Total required pumping power (MW) Estimated maximum tritium inventory in the LM loop (g)
316 278 32/32 1.3 150 1.0/1.2 0.7/1.0 6.79 1.36 2.96 2.41 28.2 2.63 16-30 _.
coils
and
the
between
them.
bottom
divertor
subsystems:
vacuum Each and
chamber
LM loop incorporates
access serves the
ports
for
both
following
located top
and
major
LM clean-up and conditioning system; heat tritium extraction-processing system; main and auxiliary pumps; LM dump tank and
exchangers; circulation storage. Since the different cooling lines of the top and bottom divertors require different inlet pressures to circulate LM, the pump system of each loop is equipped with several pumps connected in series to permit intermediate coolant bleeding. The design parameters are summarised in Table 1.
4. In-reactor thermal hydraulics In the open cooling cell of the HHF target the impinging LM jets form a film flow. Since the jet orifices make a row in the toroidal direction the film flows in the quasi-coplanar (close to coplanar) magnetic field with the major toroidal component B, lying in the film plane and perpendicular to the velocity vector. The strong coplanar magnetic field should stimulate fast formation of a uniform film flow with a transition length of about the distance between the jets (the orifices pitch). This portion of the target is screened from the heat flux by the LHF panel (Fig. 1). For the selected jet diameter above 2 mm and 10 mm pitch the minimum film thickness of 0.6 mm can be expected. This was considered acceptable for the comparatively small cooling cells (only 80 mm wide). Fast thin LM films with almost exactly such a geometry in a coplanar
magnetic field have been obtained by Evtushenko et al. in experiments, including “flow on ceiling” [lo]. The coplanar magnetic field interacts rather weakly with averaged flow and in this respect is similar to the longitudinal magnetic field. It suppresses turbulent pulsations and thus can reduce the heat transfer coefficients. Therefore in the absence of experimental data on heat transfer in a coplanar magnetic field an empirical relation suggested for LM flow in longitudinal magnetic field by Genin and Sviridov [ 1I] has been used. The pressure drops within the reactor magnetic field have been calculated for the detailed geometry of the LM-feeding pipelines in a quasi-one-dimensional approximation of MHD flow. This means splitting long segments of the duct into pieces for which the quasiuniform flow approach is applicable. The total pressure drop for a line was found as the sum of the pressure losses in each of these duct pieces and the local pressure drops in various bends and joints (detailed model description and calculation results are presented in [ 11). Ten pressurized feeding pipelines (five lines for five targets within the module of each of two divertors) were assessed with respect to LM inventory minimization at the selected pressure loss levels. Besides this four drain lines located within the bottom divertor chamber were calculated: the inboard blanket drain line receiving the total coolant flow from the inboard LHF and HHF targets of the top divertor; the inboard HHF target drain line of the bottom divertor; the outboard HHF target drain line of the bottom divertor; the outboard blanket drain line receiving the total coolant flow from the central and outboard LHF and HHF targets of the top divertor.
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The minimum LM inventory in this case was found under the condition p = 0 (where p is the absolute pressure in the LM) at any point along the line.
5. Thermal stress analysis Temperature and stress distributions were calculated for the outboard HHF target, where the thermal load was supposed to be a factor of 1.5 higher than that for the inboard one. For this analysis temperature-dependent properties of the vanadium alloy have been used. All calculations were carried out by means of a finite element code. The maximum temperature of about 740°C occurred in the heat flux peak area on the PFS of the cell. The maximum stresses cr,,, (von Mises) and oi = 0, - c3 are localized in the same area. The maximum stress value g,,, is about 216 MPa and 0, = 224 MPa. For the mentioned temperature of 740°C the yield stress is about 330 MPa and the ultimate tensile strength is 560 MPa. According to American Society for Mechanical Engineers code the limit value S,,, is the least of the following quantities: ( 1/3)fl, (average ultimate tensile strength) and (2/3)00,, (average yield stress) at the corresponding temperature. In our case S, = 187 MPa. For the stress calculations based on a linear elastic material model the equivalent stresses g,,, and cri should be lower than 1.2& = 225 MPa. This condition is fulfilled in our case. Similar results were obtained for the inboard, central and outboard LHF panel targets. In this case the analysis was conducted in two steps: first, a simplified three-dimensional model was employed to determine the location of maximum stresses and then more detailed two-dimensional temperature-stress analysis was carried out for critical regions. The stress values were checked both for the target case walls and for the cooling tubes.
6. Safety considerations Gallium activation characteristics have been calculated (by Kashirskij) to obtain activation afterheat data for accident analysis. From this viewpoint the most interesting parameters were maximum values of specific activation energy release during reactor operation and afterheat dynamics. Neutron flux with 46 energy groups was used in activation calculations. The flux data were available for three points in the middle plane of the ARIES-II blanket: at the first wall of the outboard
and Design 29 (1995) 98- 104
blanket; and at 10 and 20 cm inside the blanket. The activation cross-section library and nuclear physical data were based on ACTL-82 [ 121 and other sources. In particular, the decay energy values were calculated with data from Refs. [13,14]. It was found that the energy release in gallium is rather high during operation and about 3 days after shut-down. At the moment of shut-down it equals about 4.2 x 10” MeV s-’ cm-3. So if there are any zones with large amounts of stagnant gallium staying permanently under exposure during all the operation time it may result in a serious problem with the afterheat. In flowing gallium coolant the average energy release at shut-down is considerably lower and amounts to about 6.8 x IO’* MeV s-’ cme3. ;’ rays produced by Ga activation products have rather high energy and, since the divertor target structure is relatively thin, only a small fraction of energy released in the material will be actually transformed into afterheat in situ. An estimated transformation factor is approximately 0.1 at O.S- 1.O cm of total target thickness. From the viewpoint of accident analysis the afterheat in Ga should be a matter for consideration only within the first 10 days after shut-down. For radioactive waste disposal it does not present a greater problem than V alloy. The following accidents have been selected for firstpriority consideration: ( 1) LOFA in two types of targets (HHF and LHF) (1 .l) LOFA resulting from an external event, two versions: (a) coolant stays inside cooling tubes (b) coolant drained from cooling tubes (1.2) LOFA resulting from an internal event (a) plugging up of one or a few cooling tubes in an LHF target (b) plugging up of one or a few jet orifices in an HHF target (2) loss-of-coolant accident (LOCA) (2.1) LOCA resulting from external events (2.2) spilling of liquid gallium outside the reactor On the basis of the obtained results it was concluded that neither LOFAs nor LOCAs in LHF targets would be dangerous if the reactor were shut down (the fusion reaction terminated) within the first 10 s after the accident began and the first wall cooling system kept working. In this case the LHF target temperature would not exceed 500°C. LOFA with 10 s reactor shut-down time and no heat removal would lead to a temperature rise of up to 1000°C within the first 5 h after the accident began. LOFA calculations taking into account various scenarios of the heat flux decrease during plasma shut-
E. Murmiev
/ Fusion Engineering
down have demonstrated this to be an important issue that must be subject to further analyses. Spilling of the LM coolant would not lead to any appreciable temperature rise if the thickness of the spilled coolant layer is about 1 cm, but there may be major consequences if it is about 10 cm. Some preliminary estimates of tritium doses were performed for accidents with Ga spilling from the divertor target module during maintenance and in the case of coolant loop rupture during reactor operation. For the first case tritium release into the reactor hall was estimated as 0.2-4 g depending on the efficiency of the Ga-cleaning system during previous operation. For the second case the release value can be as high as 2-40 g. As a tritium dose characteristic a dose equivalent at 1 km distance from the release point was used. An upper limit estimate under most conservative assumptions (including no containment credit) gave an accident dose of 85 mSv (the accident dose limit is 100 mSv).
cud Design 29 (1995)
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103
gen adsorption properties in the vacuum chamber environment; technology for Ga purification from PFS erosion products and tritium. It was hoped that the design solutions suggested in this study would allow the corresponding tasks to be specified more exactly.
Acknowledgments The author, on behalf of the entire working team, wishes to acknowledge the initiative and friendly support from our American colleagues that made this work possible on a contractual basis with General Atomics, San Diego, CA, USA: Dr. K. Schultz, Mr. S. Ethridge and Dr. C. Wong, as well as useful cooperation from Dr. L. Elguebaly and Dr. C. Bathke.
References
[II E.V. Muraviev, V.S. Petrov, P.V. Romanov, L.N. Topil7. Conclusions The performed design study was the first attempt at a complete divertor system analysis based on a new LMcooled divertor target concept. It seems that such major objectives of this work as to insure higher permissible thermal loads to the divertor targets (up to 15 MW m-‘) and to demonstrate the feasibility of a divertor design with easy maintenance and high duty cycle potential have been successfully achieved. The impact from the divertor system on the tokamak reactor can be characterized by the following major points: the necessity of using some part (though not large) of the blanket cooling ducts for draining LM from the top divertor; the need for rooms directly adjacent to the tokamak reactor for placing the divertor LM loop equipment to minimize gallium inventory. On the whole the feasibility of the proposed concept will depend on successful solution of a number of problems that require certain research and development efforts. The most important of these are the following: the technology of thin electroinsulating films or coatings formation on metallic walls in contact with Ga; stability and self-healing of the electroinsulating films under neutron irradiation; a data base on corrosion and mass transfer under the conditions of a non-isothermal gallium loop with a working section in a magnetic field; a data base for the LM heat transfer in MHD film flow; a data base for MHD jet flow-wall interaction; a data base for hydrogen isotope solubility in Ga and hydro-
ski, A.V. Kashirski et al., Liquid metal cooled divertor for ARIES, Technical Report of General Atomics GAA21755, General Atomics, San Diego, CA, 1994. [21B.G. Karasev, O.A. Lielausis, E.V. Muraviev and A.V. Tananaev, Liquid metals in fusion reactor with magnetic confinement, in Fusion Reactor Design and Technology 1986, International Atomic Energy Agency, Vienna, 1987. A.F. Kolesnichenko, S.V. Mirnov, [31 V.N. Dem’yanenko, E.V. Muraviev et al.. Liquid metal limiter of a tokamak: task formulation and first results, Fiz. Plazmy l4(5) ( 1988) 628-632. ITER Documentation [41 ITER plasma facing components, Series 30, International Atomic Energy Agency, Vienna, 1991. [5] V.A. Chuyanov, A.V. Klishchenko, E.V. Muraviev and V.S. Petrov, New concept of ITER divertor. presented at ITER Topical Meeting, Garching, June 1992. [6] S.P. Jatsenko, Gallium: Interaction with Metals, Nauka, Moscow. 1974. [7] Conceptual design of a DEMO reactor, Preliminary Report, Russian Scientific Center Kurchatov Institute, La Jolla, CA, 1993 (in Russian). [8] D.K. Sze, R.F. Mattas, A.B. Hull, B. Picologlou and D.L. Smith, MHD considerations for a self-cooled lithium blanket, presented at the 10th Topical Meeting on the Technology of Fusion Energy, Boston, MA, June 1992. [9] A.V. Beznosov. B.F. Gromov, E.V. Muraviev, V.V. Orlov, Yu.1. Orlov et al., Heavy liquid metal coolants on the basis of lead in cooling system of fusion reactor with magnetic plasma confinement, At. Energ. 71(4) (1991) 506-51 I. [IO] I.A. Evtushenko, I.R. Kirillov, Yu.M. Krivchenkov, T.V. Mazul, V.L. Matveev et al., Divertor target with liquid
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E. Muraviev 1 Fusion Engineering and Design 29 (1995) 98- 104
metal jet impinging upon a solid surface, presented at ISFNT-3, Los Angeles, CA, 27 June- I July 1994, Paper DP2-16. [ 1l] L.C. Genin and V.G. Sviridov, Heat transfer and temperature fields in the thermal entry length in liquid metal flow in longitudinal magnetic field, Magn. Gidrodin. (2) (1983) (in Russian). [ 121 H.D. Lemmel, ACTL-82, The LLNL Neutron Activation
Cross-Section Library of 1982 (Summary of Contents), International Atomic Energy Agency Nuclear Data Section 55, March 1983. [ 131 O.F. Nemets and Ju.V. Gofman, Spravochnik po jadernoj fizike, Naukova Dumka, Kiev, 1975 (in Russian). [I41 N.G. Gusev and P.P. Dmitriev, Kvantovoe izluchenie radioaktivnykh nuklidov, Atomizdat, Moscow, 1977 (in Russian).