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The Direct Internal Recycling concept to simplify the fuel cycle of a fusion power plant
Fusion Engineering and Design 88 (2013) 616–620
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
The Direct Internal Recycling concept to simplify the fuel cycle of a fusion power plant Christian Day ∗ , Thomas Giegerich Institute for Technical Physics, Karlsruhe Institute of Technology (KIT), 76021 Karlsruhe, Germany
h i g h l i g h t s • • • •
The fusion fuel cycle is presented and its functions are discussed. Tritium inventories are estimated for an early DEMO configuration. The Direct Internal Recycling concept to reduce tritium inventories is described. Concepts for its technical implementation are developed.
a r t i c l e
i n f o
Article history: Received 13 September 2012 Received in revised form 13 May 2013 Accepted 13 May 2013 Available online 14 June 2013 Keywords: Vacuum Fuel cycle Tritium Inventory Dir Demo
a b s t r a c t A new concept, the Direct Internal Recycling (DIR) concept, is proposed, which minimizes fuel cycle inventory by adding an additional short-cut between the pumped torus exhaust gas and the fuelling systems. The paper highlights quantitative modelling results derived from a simple fuel cycle spreadsheet which underline the potential benefits that can be achieved by implementation of the DIR concept into a fusion power plant. DIR requires a novel set-up of the torus exhaust pumping system, which replaces the batch-wise and cyclic operated cryogenic pumps by a continuous pumping solution and which offers at the same time an additional integral gas separation function. By that, hydrogen can be removed close to the divertor from all other gases and the main load to the fuel clean-up systems is a smaller, helium-rich gas stream. Candidate DIR relevant pump technology based on liquid metals (vapour diffusion and liquid ring pumps) and metal foils is discussed. © 2013 Karlsruhe Institute of Technology (KIT). Published by Elsevier B.V. All rights reserved.
1. Introduction Safe and efficient control and management of the fuel and fusion product streams are essential issues for the development of fusion energy. The challenges for the fuel cycle components are given by the need to handle large flows of tritium resulting in strong requirements on containment and accountancy. The chosen fuel cycle concept has two central implications to the feasibility of power plant operation, namely the issue of start-up inventory – which shall be small – and the question of processing times – which shall be short. ITER will in certain areas serve as a good basis for scale-up to a fusion power reactor, whereas in other areas new technologies have to be developed. This paper takes a functional point of view and starts from the schemes chosen for ITER, it presents the current state of technology and identifies the areas which are considered to require essential
∗ Corresponding author. Tel.: +49 72160822609. E-mail address: [email protected] (C. Day).
supporting R&D towards an optimized and technically ready system for a tokamak-based fusion reactor. A new concept, the Direct Internal Recycling (DIR) concept, is proposed, which minimizes fuel cycle inventory by adding an additional short-cut between the pumped torus exhaust gas and the fuelling systems. Various options for how to do that will be compared. 2. Elements of the fuel cycle of a fusion power plant Fig. 1 illustrates the basic scheme of the fusion fuel cycle of a demonstration power plant (DEMO). It comprises an inner loop and an outer loop. The inner part denotes the directly plasma related gas flows and includes the fuelling system and the vacuum pumping systems. The outer part covers the breeding blankets, the coolant purification and their tritium extraction and recovery systems. For ITER, tritium will be supplied from external sources; hence, the outer part is only established at a minimum level via test blanket modules that allow for initial studies in a fusion environment, but with negligible tritium production. Whereas the ITER inner fuel cycle part is functionally very similar to the one expected for a
0920-3796/$ – see front matter © 2013 Karlsruhe Institute of Technology (KIT). Published by Elsevier B.V. All rights reserved. http://dx.doi.org/10.1016/j.fusengdes.2013.05.026
C. Day, T. Giegerich / Fusion Engineering and Design 88 (2013) 616–620
617
Fig. 1. The fusion fuel cycle (generic, most simplified).
fusion power plant. The tritium plant with its main elements of fuel clean-up, isotope separation, storage and delivery is a key system for both loops. Principle considerations should start from the elementary particle control functions, a fusion fuel cycle has to provide [1,2]: (a) provision of the fuel to the plasma; (b) provision of fuel-type gases to the neutral beam injection systems (NBI); (c) provision of additional plasma control (ELM pacing, divertor de-/attachment conditions); (d) tritium accountancy and gas analysis measurement for tritium inventory determination; (e) fusion ash exhaust via divertor and vacuum pumping of exhaust gas from torus and NBI; (f) exhaust gas cleaning and processing as well as fuel recovery; (g) removal and recovery of tritium from the breeding blanket extraction system to achieve self-sufficiency. Especially the latter, which is not only to provide the fuel to be spent but also to compensate for decay and losses (e.g. by inventory build-up due to wall storage), and, last not least, to produce additional tritium to be used as start-up inventories for other fusion plants, is a functionality whose technical implementation for DEMO will be first-of-its-kind.
2. The power output is related to a fuel throughput which is a factor two higher (this does scale less than linearly as an increased burn-up fraction is expected). 3. The long pulse duration will result in the fuel cycle to be run in full steady-state operation [3]. Furthermore, in spite of the pulse duration and the large throughputs, the inventory is wished not to rise significantly beyond what is accepted for ITER. All technical solutions to be applied at DEMO must therefore be checked in terms of minimum processing time and minimum inventory build-up. The inventory issue is key for two reasons. First, there are safety and regulator requirements to comply with, second, if the inventory consumed by the steady-state process is significantly smaller than the limit, this leaves a lot more operational margin (e.g. for inventory buildup in walls etc.) and, thus, increased net electricity production to the grid.
2.2. Inventory and processing time estimations of the fuel cycle 2.1. Steps from ITER towards DEMO DEMO is a demonstration power plant which provides electric power to the grid. Different DEMO designs are currently discussed world-wide. The European roadmap has defined two DEMO configurations, an early DEMO-1 (producing several 100 MW electric output) with technology that is wherever possible extrapolated from ITER, and an advanced DEMO based on new technology and advanced plasma scenarios. In any case, DEMO will be a longpulse/steady-state fusion machine. The DEMO characteristics as described above require validated solutions in the following core areas, with significant extrapolation from ITER:
1. The tritium self-sufficiency asks for a working breeding blanket concept with a multiplication factor that makes up for the parts of the wall which are consumed by duct openings (around 10%).
This section discusses the various fuel cycle sub-systems from inventory point of view. To do that, a simple spreadsheet was elaborated based on MS-Excel macros which describes the complete fuel cycle as a block diagram with 30 sub-systems. The fuel cycle module is deliberately kept simple in structure so that it is reasonable to plug it into a system code. It is good enough to study basic dependencies among subsystems, to make sensitivity investigations and to delineate mathematical correlations between strongly coupled systems. The subsystems are connected via flows which are treated separately for the three hydrogen isotopes, helium and impurities, and the subsystems are balanced by mass-conservation equations. In a simplistic way the inventory of a certain species in a sub-system can be considered by an incoming flowrate and the residence time for this species [4]. From a system engineering point of view it becomes obvious that the inventories needed to fulfil a certain function are given firstly by the chosen process, and, secondly, by the specific component design. Usually, the process decides to a much larger extent than the detailed design of a component.
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C. Day, T. Giegerich / Fusion Engineering and Design 88 (2013) 616–620
Table 1 Major tritium inventory contributions for a DEMO-1 device based on extrapolation from ITER. Function and process
Residence time [s]
T-inventory [g-T]
Reference
Fuelling by pellet injection Torus exhaust primary pumping with cryopumps Torus exhaust rough pumping with cryopumps Blanket systems (generic) Isotope separation by cryogenic distillation (ISS) Water detritiation (WDS) Storage systems and fuel management
1200 600 1000 – 4800 3600 –
400 280 200 335 1700 (ISS + WDS)
[5] [6] [7] [8] [9]
The fuel cycle spreadsheet was successfully benchmarked against ITER and showed an inventory of 769 g-T (excluding any inventory in the vessel walls and any permanent inventory in the storage). This yielded a consistent set of the residence times for the individual sub-systems, which was then kept and used to extrapolate to an ITER-style DEMO by correspondingly increasing the flowrates. Table 1 lists the times used together with appropriate public references (ITER internal information was used on top) and gives estimates of the expected tritium inventories associated with the main subsystems. The numbers refer to the current DEMO-1 reference configuration with 1.94 GW fusion power, a plasma volume of 1527 m3 , operated at Greenwald line density (plasma current 16 MA, minor radius 2.25 m), a tritium breeding ratio of 1.1 and a burn-up fraction of 1%. The systems shown in the table contribute to 97% of the total inventory. The DEMO blanket concept is not chosen at the moment, and therefore not discussed in detail in this paper. The assumed inventory there is a generic number. Please note that we considered the cryogenic distillation based isotope separation system (ISS) in combination with the water detritiation system (WDS). An integrated design for this unit is currently under development for ITER. The value for the storage systems and fuel management does not consider permanent storage, but only mobile inventories in piping systems and buffers. It was calculated assuming pipe lengths, operation pressures and an average velocity of the gas streams of 5 m/s. Furthermore, we do not consider a tritium contribution from the neutral beams which we assume to be operated with deuterium. The table shows clearly that there are two main contributors, namely the ISS/WDS [10,11], and the cryogenic torus exhaust pumping system [12]. Both processes are basically batch-wise and, thus, directly resulting in large inventories and large processing times [13,14]. Hence, R&D has started since a few years to develop customized continuous processes for the tritium plant, based on inorganic membranes [15]. Already existing continuous processes like the PERMCAT in the tokamak exhaust processing unit add just in the per mille range to the total inventory. Whereas no real alternative was found until now for the cryopumping process, mainly because the relatively low achievable burn-up fraction results in very large integral flowrates (mainly unburnt fuel), which ask for very high pumping speeds that can be provided best by cryogenic pumps. However, in view of the large potential benefit, this issue was re-visited at KIT and the remaining part of the paper is introducing a novel concept which potentially simplifies the whole fuel cycle. 3. Functional re-design of the gas exhaust system The exhaust gas has to be extracted via the divertor system uniformly, hence, a number of identical pumps have to be connected via pumping ducts; for ITER there will be 6 pumps (recently reduced from 8) [16]. This configuration would approximately scale linearly for DEMO, the increase in flowrate would result in a corresponding increase of the number of pumps; eventually less pumps if the pump size can be increased, but there are technical limits with regard to valve sizes and forces.
45
The requested pumping speed (which translates into pump size) is given by the divertor neutral pressure (which is not assumed to change much for DEMO compared with ITER) [3] and the flow rate to be pumped. The helium ash content of the exhaust gas is small. The amount of gas that can be stored in a cryopump is not limited by the saturation capacity of the cryosorbent but by a safety limit given by the maximum explosion pressure in the assumed case of an oxy-hydrogen explosion safety event. This scales with the hydrogen start concentration and, thus, with the volume of the pump, eventually including any additionally connected volumes. When the inventory limit is achieved, the cryopump has to be regenerated. This process happens in a staggered manner for all cryopumps continuously during the pulse duration. For ITER, when the pulse length is shorter than the pump interval, this concept is feasible. But for a DEMO scenario with a many h pulse and a factor 10 higher duty factor, this represents a huge consumption of energy and adds significant complexity to the cryoplant. The ITER gas exhaust and cleaning system is based on the philosophy to transfer the full amount of exhaust gas via primary and rough vacuum pumps to the tritium plant, and there to separate the gas species in a defined way so as to produce a given fuel mixture again for re-injection into the torus. It would have a large positive impact on the size of the tritium plant, if the separation would be partly or completely performed in the pumping system, and, thus, a considerably smaller flowrate to be processed towards and in the tritium plant. The separated hydrogen species could potentially be lead to the fuelling systems directly. Obviously, this is working only, if the sharpness of separation is sufficiently high and if there is a feedback controlled fuel gas management which is able to produce the correct composition from the various source flows. The new concept, which we call ‘Direct Internal Recycling (DIR)’ is highlighted in Fig. 2. It provides a short-cut around the tritium plant. To quantify the benefit due to a fuel cycle with embedded DIR, further calculations with the fuel cycle flowsheet were conducted. Fig. 3 illustrates two calculation cases which are related to two different numbers for the burn-up fraction, namely 1 and 2.5%, but are otherwise based on identical assumptions which have been taken from the DEMO-1 configuration as given above. The found numbers for 3 GW (bottom of Fig. 3) do very well agree with the range 2.5–5.5 kg, as given in [8] for this case. The figure illustrates how the tritium inventory depends on fusion power and the DIR fraction, which is defined as the recycled percentage of the exhausted hydrogen. Fig. 3 shows that the fusion power is directly proportional to the burn-up ratio, hence, identical inventory numbers result for identical values of the product (power × burn-up ratio). This represents another feasibility check. Obviously, the larger the DIR ratio is chosen, the more the inventory will be decreased. At a DIR of unity, this reduction effect corresponds to about 40%. A parametric variation showed that the DIR figure of merit is only very weakly dependent on the burnup fraction. As already said above, a major part of the remaining inventory is due to the cryogenic distillation system. Fig. 4 is a 2D cut at a fusion power of 2 GW and a burn-up fraction of 1%. The plot presents how the DEMO tritium inventory (in
C. Day, T. Giegerich / Fusion Engineering and Design 88 (2013) 616–620
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Fig. 2. A simplified DEMO fuel cycle.
a dimensionless form related to the inventory of the ITER case) decreases with increasing DIR ratio. At a DIR ratio of zero, this denotes a reactor which adopts the ITER fuel cycle architecture and simply extrapolates, the DEMO inventory is roughly factor 3.75
higher than the one at ITER. With only increasing DIR ratio, the DEMO inventory can be potentially reduced down to double the ITER inventory (still at factor five higher fusion power, and assumed identical burn-up fraction). To be more correct, one has to take into account that smaller flow rates to the tritium plant allow for a reduction of plant size. If we estimate this by 50% of the tritium load, DEMO could be operated with an inventory very close to the one of ITER. 4. Potential technical implementation of the DIR concept Finally, all the studies above are only useful once a technical solution is found, which can provide the functionalities needed for a DIR loop. There are a number of ideas elaborated in the past 20 years of fusion research which enable a gas separation, all for a cryopump environment. A very simple approach is based on the different desorption patterns of helium and hydrogens during regeneration,
4
DEMO inventory related to ITER
2 GW(fus) DEMO (ITER-like)
3.5 3 DEMO with DIR
2.5 2 1.5 1
DEMO with DIR and resulting reduction of tritium plant size
ITER
0.5 0 0
0.2
0.4
0.6
0.8
1
DIR ratio Fig. 3. Calculated interrelation of tritium inventory, Fusion power and DIR ratio, for an assumed burn-up fraction of 1% (top) and 2.5% (bottom), respectively.
Fig. 4. Influence of DIR on tritium inventory (at a fusion power of 2 GW and a burnup fraction of 1%).
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resulting in a sequential release of the different species while heating up [17]. Under ITER-relevant conditions it was found that helium is released completely below 40 K, whereas hydrogens show a quantitative release only above 40 K [14]. This concept was successfully demonstrated, but it also turned out that additional R&D, mainly on cryosorbent characterization and desorption kinetics and its limitations, would have to be conducted to achieve a sufficiently sharp separation (with helium contents at sub per mille level in the hydrogen stream), which is an absolute pre-requisite to make the DIR concept attractive. Another concept is based on pumping the different species in different compartments of a cascaded cryopump and stagewise regeneration [18]. This concept involves shutters and valves and was hence at that time disfavoured against the ITER-style cryopump. Another idea for separation of helium from hydrogens is to exploit the viscous drag effect for helium at higher pressures passing through a cryopump which is otherwise pumping the hydrogens by condensation [19]. This principle is already used in the cryoviscous compressor (CVC) which ITER is employing as a forepump to the actual primary cryopump [7]. The second aspect is the continuous operation to reduce inventories. One idea, already developed to some level of maturity is the use of a snail pump which scrapes off condensed gas towards the pump exit and compresses helium at the end in a similar way as for the CVC [20]. More favourable would be a pump principle providing equally high speeds and working non-cryogenic. A systematic screening study was performed and it turned out that vapor diffusion pumps represent a viable alternative. To make these pumps tritium-compatible, the working fluid, which nowadays is synthetic oil, has to be replaced by a non-organic material. The use of liquid metal which offers an excellent compatibility with tritium is a promising option for which experience is reported in literature [21]. The liquid metal technology is especially interesting because it can also be employed for the forepump, e.g. if it would be based on the liquid ring pump principle, which is another heavily industrialized pump concept with broad experience (although for more conventional fluids only). This aspect is discussed in detail in a companion paper [22]. Ideally, the continuous and non-cryogenic pump technology would be combined with an appropriate continuous and noncryogenic separation function. A candidate technology is the use of superpermeable metal foils that provide a 100% sharp separation (only hydrogen atoms can pass the foil) and produce a significant pumping effect for hydrogen [23–25]. Playing with the active surface of a metal foil pump which can be regarded as a continuous mass transfer apparatus, the DIR ratio can be adjusted. Due to a potential isotopic effect in the separation by metal foils, the maximum achievable overall DIR ratio is 0.9. Principle considerations on such devices and intelligent technologies for the atomizers which are needed to produce hydrogen atoms at the appropriate energy are ongoing. In principle, this results in a variety of potential setups which are based on metal foils only (for the DIR recycle flow to the fuelling systems), on metal foil + diffusion pump + ring pump (for the remaining flow to the tritium plant), or on the diffusion pump + ring pump only (in dwell phases or burn phases with sufficiently high divertor neutral pressures). It must be noted that this concept critically relies on the availability of an adequate metal foil, the major challenges in this area are durability and long-term stability in the DEMO environment. KIT is currently having a detailed look on the fundamental physics aspects of the various DIR elements and is developing modelling approaches so that operational limits can be assessed. In parallel proof-of-principle tests are on the way [22] and reference pump configurations are being developed.
5. Concluding remarks The concept of Direct Internal Recycling has a high potential to significantly reduce the inventories and the processing times within the fuel cycle of a fusion power plant and thus act as an absolute condition for a feasible and economically attractive DEMO reactor concept. The figure of merit as discussed in an example case above is impressive. A new proposal has been made for candidate technology on which a technical solution to implement the DIR concept may build. If this technology will be confirmed in future R&D, the requirements and their complexity on the auxiliary systems (tritium plant, cryoplant) can be drastically reduced. Acknowledgments This work, partly supported by the European Communities under the contract of Association EURATOM/KIT was carried out within the framework of the European Fusion Development Agreement as part of the Power Plant Physics and Technology activities. The views and opinions expressed herein do not necessarily reflect those of the European Commission. The authors are grateful to Lukas Rudischhauser, University of Edinburgh, who dedicated his internship at KIT to the fuel cycle modelling. References [1] D. Babineau, S. Maruyama, R. Pearce, M. Glugla, L. Bo, B. Rogers, et al., Review of the ITER fuel cycle, in: IAEA Fusion Energy Conference, Daejeon, Korea, 2010 (Paper ITR/2-2). [2] W. Kuan, M.A. Abdou, Fusion Technology 35 (1999) 309–353. [3] G.H. Neilson, G. Federici, J. Li, D. Maisonnier, R. Wolf, Nuclear Fusion 52 (2012) 047001. [4] D. Baiquan, H. Jinhua, Fusion Engineering and Design 55 (2001) 359–364. [5] W. Kuan, M. Abdou, S. Willms, Fusion Technology 28 (1995) 664–671. [6] R.J. Pearce, A. Antipenkov, B. Boussier, St. Bryan, M. Dremel, B. Levesy, et al., The ITER divertor pumping system, this conference. [7] L.R. Baylor, S.J. Meitner, C. Barbier, S.K. Combs, R.C. Duckworth, T.G. Edgemon, et al., Cryogenic viscous compressor development and modeling for the ITER vacuum system, in: Proc. 24th Symp. Fusion Engineering, IEEE, USA, 2011. [8] T. Tanabe, Tritium fuel cycle in ITER and DEMO: issues in handling large amount of fuel, Journal of Nuclear Materials (2013) http://dx.doi.org/ 10.1016/j.jnucmat.2013.01.284 [9] I. Cristescu, Personal Communication, Karlsruhe Institute of Technology, 2012. [10] I. Cristescu, I.R. Cristescu, L. Doerr, M. Glugla, D. Murdoch, Fusion Science and Technology 52 (2007) 667–671. [11] R.G. Ana, I. Cristescu, L. Dörr, R. Michling, St. Welte, W. Wurster, Fusion Engineering and Design 84 (2009) 398–403. [12] Chr. Day, D. Murdoch, R. Pearce, Vacuum 83 (2009) 773–778. [13] I.R. Cristescu, I. Cristescu, L. Doerr, M. Glugla, D. Murdoch, Nuclear Fusion 47 (2007) S458–S463. [14] Chr. Day, A. Antipenkov, I.R. Cristescu, M. Dremel, G. Federici, H. Haas, et al., Fusion Engineering and Design 81 (2006) 777–784. [15] D. Demange, S. Stämmler, M. Kind, Fusion Engineering and Design 86 (2011) 2312–2316. [16] Chr. Day, H. Haas, St. Hanke, V. Hauer, X. Luo, M. Scannapiego, et al., Fusion Engineering and Design 86 (2011) 2188–2191. [17] Day Chr, Colloids and Surfaces A 187/188 (2001) 187–206. [18] A. Mack, D. Perinic, Fusion Engineering and Design 28 (1995) 318–323. [19] J.L. Hemmerich, E. Küssel, Journal of Vacuum Science and Technology 8 (1990) 141–144. [20] C.A. Foster, Journal of Vacuum Science and Technology 5 (1987) 2558–2562. [21] C.L. Volkers, V.P. Gede, Transfer operations with tritium – a review, in: Proc. 23rd Conf. on Remote Systems Technology, San Francisco, CA, USA, 1975. [22] Th. Giegerich, Chr. Day, Conceptuation of a continuously working vacuum pump train for fusion power plants, this conference. [23] R.K. Musyaev, A.A. Yukhimchuk, B.S. Lebedev, A.O. Busnyuk, M.E. Notkin, A.A. Samartsev, et al., Fusion Science and Technology 54 (2008) 523–525. [24] Y. Nakamura, N. Ohyabu, H. Suzuki, Y. Nakahara, A. Livshits, M. Notkin, et al., Fusion Engineering and Design 49/50 (2000) 899–904. [25] M.L. Zheludkevich, A.G. Gusakov, A.G. Voropaev, E.N. Kozyrski, S.A. Raspopov, A.A. Vecher, Journal of Membrane Science 320 (2008) 528–532.
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
The Direct Internal Recycling concept to simplify the fuel cycle of a fusion power plant Christian Day ∗ , Thomas Giegerich Institute for Technical Physics, Karlsruhe Institute of Technology (KIT), 76021 Karlsruhe, Germany
h i g h l i g h t s • • • •
The fusion fuel cycle is presented and its functions are discussed. Tritium inventories are estimated for an early DEMO configuration. The Direct Internal Recycling concept to reduce tritium inventories is described. Concepts for its technical implementation are developed.
a r t i c l e
i n f o
Article history: Received 13 September 2012 Received in revised form 13 May 2013 Accepted 13 May 2013 Available online 14 June 2013 Keywords: Vacuum Fuel cycle Tritium Inventory Dir Demo
a b s t r a c t A new concept, the Direct Internal Recycling (DIR) concept, is proposed, which minimizes fuel cycle inventory by adding an additional short-cut between the pumped torus exhaust gas and the fuelling systems. The paper highlights quantitative modelling results derived from a simple fuel cycle spreadsheet which underline the potential benefits that can be achieved by implementation of the DIR concept into a fusion power plant. DIR requires a novel set-up of the torus exhaust pumping system, which replaces the batch-wise and cyclic operated cryogenic pumps by a continuous pumping solution and which offers at the same time an additional integral gas separation function. By that, hydrogen can be removed close to the divertor from all other gases and the main load to the fuel clean-up systems is a smaller, helium-rich gas stream. Candidate DIR relevant pump technology based on liquid metals (vapour diffusion and liquid ring pumps) and metal foils is discussed. © 2013 Karlsruhe Institute of Technology (KIT). Published by Elsevier B.V. All rights reserved.
1. Introduction Safe and efficient control and management of the fuel and fusion product streams are essential issues for the development of fusion energy. The challenges for the fuel cycle components are given by the need to handle large flows of tritium resulting in strong requirements on containment and accountancy. The chosen fuel cycle concept has two central implications to the feasibility of power plant operation, namely the issue of start-up inventory – which shall be small – and the question of processing times – which shall be short. ITER will in certain areas serve as a good basis for scale-up to a fusion power reactor, whereas in other areas new technologies have to be developed. This paper takes a functional point of view and starts from the schemes chosen for ITER, it presents the current state of technology and identifies the areas which are considered to require essential
∗ Corresponding author. Tel.: +49 72160822609. E-mail address: [email protected] (C. Day).
supporting R&D towards an optimized and technically ready system for a tokamak-based fusion reactor. A new concept, the Direct Internal Recycling (DIR) concept, is proposed, which minimizes fuel cycle inventory by adding an additional short-cut between the pumped torus exhaust gas and the fuelling systems. Various options for how to do that will be compared. 2. Elements of the fuel cycle of a fusion power plant Fig. 1 illustrates the basic scheme of the fusion fuel cycle of a demonstration power plant (DEMO). It comprises an inner loop and an outer loop. The inner part denotes the directly plasma related gas flows and includes the fuelling system and the vacuum pumping systems. The outer part covers the breeding blankets, the coolant purification and their tritium extraction and recovery systems. For ITER, tritium will be supplied from external sources; hence, the outer part is only established at a minimum level via test blanket modules that allow for initial studies in a fusion environment, but with negligible tritium production. Whereas the ITER inner fuel cycle part is functionally very similar to the one expected for a
0920-3796/$ – see front matter © 2013 Karlsruhe Institute of Technology (KIT). Published by Elsevier B.V. All rights reserved. http://dx.doi.org/10.1016/j.fusengdes.2013.05.026
C. Day, T. Giegerich / Fusion Engineering and Design 88 (2013) 616–620
617
Fig. 1. The fusion fuel cycle (generic, most simplified).
fusion power plant. The tritium plant with its main elements of fuel clean-up, isotope separation, storage and delivery is a key system for both loops. Principle considerations should start from the elementary particle control functions, a fusion fuel cycle has to provide [1,2]: (a) provision of the fuel to the plasma; (b) provision of fuel-type gases to the neutral beam injection systems (NBI); (c) provision of additional plasma control (ELM pacing, divertor de-/attachment conditions); (d) tritium accountancy and gas analysis measurement for tritium inventory determination; (e) fusion ash exhaust via divertor and vacuum pumping of exhaust gas from torus and NBI; (f) exhaust gas cleaning and processing as well as fuel recovery; (g) removal and recovery of tritium from the breeding blanket extraction system to achieve self-sufficiency. Especially the latter, which is not only to provide the fuel to be spent but also to compensate for decay and losses (e.g. by inventory build-up due to wall storage), and, last not least, to produce additional tritium to be used as start-up inventories for other fusion plants, is a functionality whose technical implementation for DEMO will be first-of-its-kind.
2. The power output is related to a fuel throughput which is a factor two higher (this does scale less than linearly as an increased burn-up fraction is expected). 3. The long pulse duration will result in the fuel cycle to be run in full steady-state operation [3]. Furthermore, in spite of the pulse duration and the large throughputs, the inventory is wished not to rise significantly beyond what is accepted for ITER. All technical solutions to be applied at DEMO must therefore be checked in terms of minimum processing time and minimum inventory build-up. The inventory issue is key for two reasons. First, there are safety and regulator requirements to comply with, second, if the inventory consumed by the steady-state process is significantly smaller than the limit, this leaves a lot more operational margin (e.g. for inventory buildup in walls etc.) and, thus, increased net electricity production to the grid.
2.2. Inventory and processing time estimations of the fuel cycle 2.1. Steps from ITER towards DEMO DEMO is a demonstration power plant which provides electric power to the grid. Different DEMO designs are currently discussed world-wide. The European roadmap has defined two DEMO configurations, an early DEMO-1 (producing several 100 MW electric output) with technology that is wherever possible extrapolated from ITER, and an advanced DEMO based on new technology and advanced plasma scenarios. In any case, DEMO will be a longpulse/steady-state fusion machine. The DEMO characteristics as described above require validated solutions in the following core areas, with significant extrapolation from ITER:
1. The tritium self-sufficiency asks for a working breeding blanket concept with a multiplication factor that makes up for the parts of the wall which are consumed by duct openings (around 10%).
This section discusses the various fuel cycle sub-systems from inventory point of view. To do that, a simple spreadsheet was elaborated based on MS-Excel macros which describes the complete fuel cycle as a block diagram with 30 sub-systems. The fuel cycle module is deliberately kept simple in structure so that it is reasonable to plug it into a system code. It is good enough to study basic dependencies among subsystems, to make sensitivity investigations and to delineate mathematical correlations between strongly coupled systems. The subsystems are connected via flows which are treated separately for the three hydrogen isotopes, helium and impurities, and the subsystems are balanced by mass-conservation equations. In a simplistic way the inventory of a certain species in a sub-system can be considered by an incoming flowrate and the residence time for this species [4]. From a system engineering point of view it becomes obvious that the inventories needed to fulfil a certain function are given firstly by the chosen process, and, secondly, by the specific component design. Usually, the process decides to a much larger extent than the detailed design of a component.
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C. Day, T. Giegerich / Fusion Engineering and Design 88 (2013) 616–620
Table 1 Major tritium inventory contributions for a DEMO-1 device based on extrapolation from ITER. Function and process
Residence time [s]
T-inventory [g-T]
Reference
Fuelling by pellet injection Torus exhaust primary pumping with cryopumps Torus exhaust rough pumping with cryopumps Blanket systems (generic) Isotope separation by cryogenic distillation (ISS) Water detritiation (WDS) Storage systems and fuel management
1200 600 1000 – 4800 3600 –
400 280 200 335 1700 (ISS + WDS)
[5] [6] [7] [8] [9]
The fuel cycle spreadsheet was successfully benchmarked against ITER and showed an inventory of 769 g-T (excluding any inventory in the vessel walls and any permanent inventory in the storage). This yielded a consistent set of the residence times for the individual sub-systems, which was then kept and used to extrapolate to an ITER-style DEMO by correspondingly increasing the flowrates. Table 1 lists the times used together with appropriate public references (ITER internal information was used on top) and gives estimates of the expected tritium inventories associated with the main subsystems. The numbers refer to the current DEMO-1 reference configuration with 1.94 GW fusion power, a plasma volume of 1527 m3 , operated at Greenwald line density (plasma current 16 MA, minor radius 2.25 m), a tritium breeding ratio of 1.1 and a burn-up fraction of 1%. The systems shown in the table contribute to 97% of the total inventory. The DEMO blanket concept is not chosen at the moment, and therefore not discussed in detail in this paper. The assumed inventory there is a generic number. Please note that we considered the cryogenic distillation based isotope separation system (ISS) in combination with the water detritiation system (WDS). An integrated design for this unit is currently under development for ITER. The value for the storage systems and fuel management does not consider permanent storage, but only mobile inventories in piping systems and buffers. It was calculated assuming pipe lengths, operation pressures and an average velocity of the gas streams of 5 m/s. Furthermore, we do not consider a tritium contribution from the neutral beams which we assume to be operated with deuterium. The table shows clearly that there are two main contributors, namely the ISS/WDS [10,11], and the cryogenic torus exhaust pumping system [12]. Both processes are basically batch-wise and, thus, directly resulting in large inventories and large processing times [13,14]. Hence, R&D has started since a few years to develop customized continuous processes for the tritium plant, based on inorganic membranes [15]. Already existing continuous processes like the PERMCAT in the tokamak exhaust processing unit add just in the per mille range to the total inventory. Whereas no real alternative was found until now for the cryopumping process, mainly because the relatively low achievable burn-up fraction results in very large integral flowrates (mainly unburnt fuel), which ask for very high pumping speeds that can be provided best by cryogenic pumps. However, in view of the large potential benefit, this issue was re-visited at KIT and the remaining part of the paper is introducing a novel concept which potentially simplifies the whole fuel cycle. 3. Functional re-design of the gas exhaust system The exhaust gas has to be extracted via the divertor system uniformly, hence, a number of identical pumps have to be connected via pumping ducts; for ITER there will be 6 pumps (recently reduced from 8) [16]. This configuration would approximately scale linearly for DEMO, the increase in flowrate would result in a corresponding increase of the number of pumps; eventually less pumps if the pump size can be increased, but there are technical limits with regard to valve sizes and forces.
45
The requested pumping speed (which translates into pump size) is given by the divertor neutral pressure (which is not assumed to change much for DEMO compared with ITER) [3] and the flow rate to be pumped. The helium ash content of the exhaust gas is small. The amount of gas that can be stored in a cryopump is not limited by the saturation capacity of the cryosorbent but by a safety limit given by the maximum explosion pressure in the assumed case of an oxy-hydrogen explosion safety event. This scales with the hydrogen start concentration and, thus, with the volume of the pump, eventually including any additionally connected volumes. When the inventory limit is achieved, the cryopump has to be regenerated. This process happens in a staggered manner for all cryopumps continuously during the pulse duration. For ITER, when the pulse length is shorter than the pump interval, this concept is feasible. But for a DEMO scenario with a many h pulse and a factor 10 higher duty factor, this represents a huge consumption of energy and adds significant complexity to the cryoplant. The ITER gas exhaust and cleaning system is based on the philosophy to transfer the full amount of exhaust gas via primary and rough vacuum pumps to the tritium plant, and there to separate the gas species in a defined way so as to produce a given fuel mixture again for re-injection into the torus. It would have a large positive impact on the size of the tritium plant, if the separation would be partly or completely performed in the pumping system, and, thus, a considerably smaller flowrate to be processed towards and in the tritium plant. The separated hydrogen species could potentially be lead to the fuelling systems directly. Obviously, this is working only, if the sharpness of separation is sufficiently high and if there is a feedback controlled fuel gas management which is able to produce the correct composition from the various source flows. The new concept, which we call ‘Direct Internal Recycling (DIR)’ is highlighted in Fig. 2. It provides a short-cut around the tritium plant. To quantify the benefit due to a fuel cycle with embedded DIR, further calculations with the fuel cycle flowsheet were conducted. Fig. 3 illustrates two calculation cases which are related to two different numbers for the burn-up fraction, namely 1 and 2.5%, but are otherwise based on identical assumptions which have been taken from the DEMO-1 configuration as given above. The found numbers for 3 GW (bottom of Fig. 3) do very well agree with the range 2.5–5.5 kg, as given in [8] for this case. The figure illustrates how the tritium inventory depends on fusion power and the DIR fraction, which is defined as the recycled percentage of the exhausted hydrogen. Fig. 3 shows that the fusion power is directly proportional to the burn-up ratio, hence, identical inventory numbers result for identical values of the product (power × burn-up ratio). This represents another feasibility check. Obviously, the larger the DIR ratio is chosen, the more the inventory will be decreased. At a DIR of unity, this reduction effect corresponds to about 40%. A parametric variation showed that the DIR figure of merit is only very weakly dependent on the burnup fraction. As already said above, a major part of the remaining inventory is due to the cryogenic distillation system. Fig. 4 is a 2D cut at a fusion power of 2 GW and a burn-up fraction of 1%. The plot presents how the DEMO tritium inventory (in
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Fig. 2. A simplified DEMO fuel cycle.
a dimensionless form related to the inventory of the ITER case) decreases with increasing DIR ratio. At a DIR ratio of zero, this denotes a reactor which adopts the ITER fuel cycle architecture and simply extrapolates, the DEMO inventory is roughly factor 3.75
higher than the one at ITER. With only increasing DIR ratio, the DEMO inventory can be potentially reduced down to double the ITER inventory (still at factor five higher fusion power, and assumed identical burn-up fraction). To be more correct, one has to take into account that smaller flow rates to the tritium plant allow for a reduction of plant size. If we estimate this by 50% of the tritium load, DEMO could be operated with an inventory very close to the one of ITER. 4. Potential technical implementation of the DIR concept Finally, all the studies above are only useful once a technical solution is found, which can provide the functionalities needed for a DIR loop. There are a number of ideas elaborated in the past 20 years of fusion research which enable a gas separation, all for a cryopump environment. A very simple approach is based on the different desorption patterns of helium and hydrogens during regeneration,
4
DEMO inventory related to ITER
2 GW(fus) DEMO (ITER-like)
3.5 3 DEMO with DIR
2.5 2 1.5 1
DEMO with DIR and resulting reduction of tritium plant size
ITER
0.5 0 0
0.2
0.4
0.6
0.8
1
DIR ratio Fig. 3. Calculated interrelation of tritium inventory, Fusion power and DIR ratio, for an assumed burn-up fraction of 1% (top) and 2.5% (bottom), respectively.
Fig. 4. Influence of DIR on tritium inventory (at a fusion power of 2 GW and a burnup fraction of 1%).
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resulting in a sequential release of the different species while heating up [17]. Under ITER-relevant conditions it was found that helium is released completely below 40 K, whereas hydrogens show a quantitative release only above 40 K [14]. This concept was successfully demonstrated, but it also turned out that additional R&D, mainly on cryosorbent characterization and desorption kinetics and its limitations, would have to be conducted to achieve a sufficiently sharp separation (with helium contents at sub per mille level in the hydrogen stream), which is an absolute pre-requisite to make the DIR concept attractive. Another concept is based on pumping the different species in different compartments of a cascaded cryopump and stagewise regeneration [18]. This concept involves shutters and valves and was hence at that time disfavoured against the ITER-style cryopump. Another idea for separation of helium from hydrogens is to exploit the viscous drag effect for helium at higher pressures passing through a cryopump which is otherwise pumping the hydrogens by condensation [19]. This principle is already used in the cryoviscous compressor (CVC) which ITER is employing as a forepump to the actual primary cryopump [7]. The second aspect is the continuous operation to reduce inventories. One idea, already developed to some level of maturity is the use of a snail pump which scrapes off condensed gas towards the pump exit and compresses helium at the end in a similar way as for the CVC [20]. More favourable would be a pump principle providing equally high speeds and working non-cryogenic. A systematic screening study was performed and it turned out that vapor diffusion pumps represent a viable alternative. To make these pumps tritium-compatible, the working fluid, which nowadays is synthetic oil, has to be replaced by a non-organic material. The use of liquid metal which offers an excellent compatibility with tritium is a promising option for which experience is reported in literature [21]. The liquid metal technology is especially interesting because it can also be employed for the forepump, e.g. if it would be based on the liquid ring pump principle, which is another heavily industrialized pump concept with broad experience (although for more conventional fluids only). This aspect is discussed in detail in a companion paper [22]. Ideally, the continuous and non-cryogenic pump technology would be combined with an appropriate continuous and noncryogenic separation function. A candidate technology is the use of superpermeable metal foils that provide a 100% sharp separation (only hydrogen atoms can pass the foil) and produce a significant pumping effect for hydrogen [23–25]. Playing with the active surface of a metal foil pump which can be regarded as a continuous mass transfer apparatus, the DIR ratio can be adjusted. Due to a potential isotopic effect in the separation by metal foils, the maximum achievable overall DIR ratio is 0.9. Principle considerations on such devices and intelligent technologies for the atomizers which are needed to produce hydrogen atoms at the appropriate energy are ongoing. In principle, this results in a variety of potential setups which are based on metal foils only (for the DIR recycle flow to the fuelling systems), on metal foil + diffusion pump + ring pump (for the remaining flow to the tritium plant), or on the diffusion pump + ring pump only (in dwell phases or burn phases with sufficiently high divertor neutral pressures). It must be noted that this concept critically relies on the availability of an adequate metal foil, the major challenges in this area are durability and long-term stability in the DEMO environment. KIT is currently having a detailed look on the fundamental physics aspects of the various DIR elements and is developing modelling approaches so that operational limits can be assessed. In parallel proof-of-principle tests are on the way [22] and reference pump configurations are being developed.
5. Concluding remarks The concept of Direct Internal Recycling has a high potential to significantly reduce the inventories and the processing times within the fuel cycle of a fusion power plant and thus act as an absolute condition for a feasible and economically attractive DEMO reactor concept. The figure of merit as discussed in an example case above is impressive. A new proposal has been made for candidate technology on which a technical solution to implement the DIR concept may build. If this technology will be confirmed in future R&D, the requirements and their complexity on the auxiliary systems (tritium plant, cryoplant) can be drastically reduced. Acknowledgments This work, partly supported by the European Communities under the contract of Association EURATOM/KIT was carried out within the framework of the European Fusion Development Agreement as part of the Power Plant Physics and Technology activities. The views and opinions expressed herein do not necessarily reflect those of the European Commission. The authors are grateful to Lukas Rudischhauser, University of Edinburgh, who dedicated his internship at KIT to the fuel cycle modelling. References [1] D. Babineau, S. Maruyama, R. Pearce, M. Glugla, L. Bo, B. Rogers, et al., Review of the ITER fuel cycle, in: IAEA Fusion Energy Conference, Daejeon, Korea, 2010 (Paper ITR/2-2). [2] W. Kuan, M.A. Abdou, Fusion Technology 35 (1999) 309–353. [3] G.H. Neilson, G. Federici, J. Li, D. Maisonnier, R. Wolf, Nuclear Fusion 52 (2012) 047001. [4] D. Baiquan, H. Jinhua, Fusion Engineering and Design 55 (2001) 359–364. [5] W. Kuan, M. Abdou, S. Willms, Fusion Technology 28 (1995) 664–671. [6] R.J. Pearce, A. Antipenkov, B. Boussier, St. Bryan, M. Dremel, B. Levesy, et al., The ITER divertor pumping system, this conference. [7] L.R. Baylor, S.J. Meitner, C. Barbier, S.K. Combs, R.C. Duckworth, T.G. Edgemon, et al., Cryogenic viscous compressor development and modeling for the ITER vacuum system, in: Proc. 24th Symp. Fusion Engineering, IEEE, USA, 2011. [8] T. Tanabe, Tritium fuel cycle in ITER and DEMO: issues in handling large amount of fuel, Journal of Nuclear Materials (2013) http://dx.doi.org/ 10.1016/j.jnucmat.2013.01.284 [9] I. Cristescu, Personal Communication, Karlsruhe Institute of Technology, 2012. [10] I. Cristescu, I.R. Cristescu, L. Doerr, M. Glugla, D. Murdoch, Fusion Science and Technology 52 (2007) 667–671. [11] R.G. Ana, I. Cristescu, L. Dörr, R. Michling, St. Welte, W. Wurster, Fusion Engineering and Design 84 (2009) 398–403. [12] Chr. Day, D. Murdoch, R. Pearce, Vacuum 83 (2009) 773–778. [13] I.R. Cristescu, I. Cristescu, L. Doerr, M. Glugla, D. Murdoch, Nuclear Fusion 47 (2007) S458–S463. [14] Chr. Day, A. Antipenkov, I.R. Cristescu, M. Dremel, G. Federici, H. Haas, et al., Fusion Engineering and Design 81 (2006) 777–784. [15] D. Demange, S. Stämmler, M. Kind, Fusion Engineering and Design 86 (2011) 2312–2316. [16] Chr. Day, H. Haas, St. Hanke, V. Hauer, X. Luo, M. Scannapiego, et al., Fusion Engineering and Design 86 (2011) 2188–2191. [17] Day Chr, Colloids and Surfaces A 187/188 (2001) 187–206. [18] A. Mack, D. Perinic, Fusion Engineering and Design 28 (1995) 318–323. [19] J.L. Hemmerich, E. Küssel, Journal of Vacuum Science and Technology 8 (1990) 141–144. [20] C.A. Foster, Journal of Vacuum Science and Technology 5 (1987) 2558–2562. [21] C.L. Volkers, V.P. Gede, Transfer operations with tritium – a review, in: Proc. 23rd Conf. on Remote Systems Technology, San Francisco, CA, USA, 1975. [22] Th. Giegerich, Chr. Day, Conceptuation of a continuously working vacuum pump train for fusion power plants, this conference. [23] R.K. Musyaev, A.A. Yukhimchuk, B.S. Lebedev, A.O. Busnyuk, M.E. Notkin, A.A. Samartsev, et al., Fusion Science and Technology 54 (2008) 523–525. [24] Y. Nakamura, N. Ohyabu, H. Suzuki, Y. Nakahara, A. Livshits, M. Notkin, et al., Fusion Engineering and Design 49/50 (2000) 899–904. [25] M.L. Zheludkevich, A.G. Gusakov, A.G. Voropaev, E.N. Kozyrski, S.A. Raspopov, A.A. Vecher, Journal of Membrane Science 320 (2008) 528–532.