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ITER fuel cycle options and optimizations
Fusion Engineering and Design 16 (1991) 11-22 North-Holland
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ITER fuel cycle options and optimizations P.J. D i n n e r a a n d H. Y o s h i d a b a The NET Team, Max-Planck-lnstitut ]fir Plasmaphysik, W-8046 Garching, Germany b Japan Atomic Energy Research Institute, Tokaimura, lbaraki, Japan
Inter-connected elements are required to carry out fuelling of the torus, torus vacuum pumping, processing of (exhaust) fuel recovery of tritium from the breeder blanket and test sectors, and several "common" processes such as hydrogen isotope separation, storage and management of fuel gases, and treatment of solid, liquid and gaseous tritiated wastes. Results of the International Tokamak Experimental Reactor - Conceptual Design Activity (ITER-CDA) design efforts are summarized, and technical and organizational proposals made to facilitate design optimization during the next phase of the Project. 1. Introduction During the International Tokamak Experimental Reactor - Engineering Design Activity (ITER-EDA), the process of Fuel Cycle design which began in the Conceptual Design Activity (CDA) will be continued, but with increased emphasis on producing an integrated design which balances design criteria while continuing to meet all Project requirements. This will require that such criteria are well understood by all levels of the project, and an engineering management structure exists to ensure orderly evolution of the design and cohsistency with interfacing systems and proposed operating scenarios. With these assumptions, it is possible to define the steps in preparing an optimized Fuel Cycle design. In the early stages of the EDA, it is reasonable to expect that a complete review of all CDA options will be conducted. In addition to the main approaches used in the CDA, this review should include an assessment of novel concepts and approaches to design implementation, some of which may have emerged during the period between the CDA and EDA. In this paper the main design options for ITER Fuel Cycle are reviewed, and a number of novel approaches and recent developments are discussed. Steps envisaged for the design "optimization" process are presented. 2. Fuel cycle design concepts and innovations The main elements of the ITER Fuel Cycle (FC) are shown in fig. 1, and key design parameters are listed in table 1. During the CDA phase, ITER FC
conceptual design was performed using well-advanced technologies to determine if: (1) Acceptable process options exist for all essential FC functions. (2) Subsystem concept design requirements are correct, since FC systems involve many "recycle" loops. (3) Impacts of the FC on plant arrangement, safety, and cost are acceptable. (4) Specifications for ITER FC R & D have been correctly formulated, and R & D priorities are consistent with the design process and schedule. The Fuel Cycle designs evolved in the ITER-CDA are described in ref. [1]. R & D progress involving many of the key processes is summarized in ref. [2]. In the following discussion, options used in the ITER FC design integration exercise are presented first. These are either based on proven technology or subject to detailed R & D investigations in ITER home organizations. Additional proposals to improve on concepts used for design integration in the CDA have arisen since the C D A work was performed. These developments are also introduced in this paper, with reference to specific advantages they may offer. 2.1. Fuelling
A combination of gas puffing and pellet injection past the scrape-off layer is required for fuelling. Fastvalve concepts for gas-puffing are being demonstrated on current-generation machines, and appear to be scaleable to ITER. However, further attention must be paid to increasing flow rate, reducing time-constants of valve and gas distribution circuit, and demonstrating
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12
P.J. Dinner, H. Yoshida / ITER fuel cycle
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tritium, neutron and gamma radiation, and electromagnetic field compatibility of the gas-puffing valves. For fuelling during ramp-up, pellet injectors capable of producing a limited number ( < 100/pulse) of pellets at velocities in excess of 2 k m / s will be provided to produce peaked profiles. Pellet injectors for continuous fuelling beyond the scrape-off layer (1-2 k m / s ) are also needed. For CDA-phase design integration an assembly consisting of two single-stage light gas guns and a two-stage light gas gun, all employing hydrogen propellent were used (fig. 2). The figure illustrates the compact arrangement necessitated by limited space and the requirement for port-sharing with neutron diagnostics. Both fuelling and plasma chamber vacuum pumping (section 2.2) have been designed for remote maintenance. The addition of a centrifugal pellet launcher to the basic design would permit shallow pellet injection without the high pro-
tium propellant loads associated with gas guns. The centrifuge would require the development of very high DT extrusion rates, since only pellets a few millimeters in diameter could be accelerated by a centrifuge. Deeper fuel injection without high propellant loads might be achieved using either rail guns or compact toroids. Neither approach has yet been tested as a fuelling method for a large tokamak. 2.2. Plasma chamber t,acuum p u m p i n g
Plasma chamber vacuum pumping options for burn and dwell pumping include both compound cryopumps (CCP) and turbopumps. Argon-spray CCPs (fig. 3) have been used in the initial design integration effort, since the design for this pump is well advanced [3], and At-spray efficiencies confirmed in recent measurements [4]. Cryopumps are arranged in 8 stations of
P.J. Dinner, H. Yoshida / ITER fuel cycle Table 1 ITER F C
-
13
Key design parameters
Fuelling and exhaust Fuelling and exhaust rate (mol/h) Effective pumping speed at torus during burn (m3/s) Down-pipe conductance assumed (m3/s) Operating pump speed (m3/s) Ultimate pressure in torus (mbar) Installed TMP capacity (ma/s) for torus conditioning
35-75 350-700 1000-1500 1000-1500 4x10 -7 120
Fuel (exhaust) processing Total flow (mbar l/s) Impurity in burn-time exhaust (mbar l/s)
H2
He Low-Z High-Z Glow discharge in He, flow rate (mbar I/s) He impurity conc. (mbar l/s) Low-Z
220-470 <5 <18 <10 <1 300 0.3
Blanket tritium processing Tritium production rate (g/full-power-day) Ceramic breeder hydrogen swamping ratio ( H / T ) Maximum H + T recovered as oxide (%-water) Nominal production of tritiated water (mol/d)
152 100 10 < 25
Common processes Max. flows to Isotopic Separation (mol/h) protium deuterium tritium Max. tritium conc. in fuel (%) Max. T in H z/HD exhaust (%) Effluent water detritiation rate (kg/h) Nominal tritium concentration of water to be detritiated (Ci/kg) Discharge concentration (Ci/kg) Plant volumes requiring air-detritiation (m3) Air He Graphite dust processing rate (kg/d) Max. local tritium inventory design target (g)
three (fig. 4). For initial pump-down and torus conditioning, TMPs are required (8 pumps of > 15 m3/s capacity, integrated into the cryopumping stations). Substantial progress has been made in cryosorption panel development and testing for He pumping. Thin cryosorbent layers bonded inorganically to a metal substrate are being tested for resistence to repeated temperature cycling, and the ability of the cryosorbent to pump He in the presence of DT and impurities has
<1×104 < 450 < 50 < 80 l x l 0 -7 < 150 <0.1 <1×10 -5 <2x10 s <5x104 1-10 180
been confirmed [4,5]. This would permit pump designs employing a single 4 K pumping surface, which implies higher specific pumping speeds in a simpler, more compact design. This innovation has already been incorporated into CCPs for JET [6]. An all-turbopump option would result in a single torus pumping system with potential advantages, especially lower overall tritium inventories (table 2), if development programs meeting ITER requirements are
P.J. Dinner, 14. Yoshida / 1TER fuel cycle
14
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I. Moderate velocity injector 2. Moderate velocity injector 3. High velocity injector (HVI) 4. HVI pump tube 5. HVI propellant reservoir 6. HVI propellant compressor 7. Propellant gas manifolds 8. Fuel gas manifolds 9. High pressure cage 10. Low vacuum chamber 11. Mid vacuum chamber 12. High vacuum chamber 13. Torus isolation valves, breaks, bellows 14. Structural shielding wall 15. Modular shielding 16. Normetex 1300 m3/h scroll pump 17. Normetex control cabinets 18. Turbopumps, 5000 L / s 19. Turbopumps, 500 L / s 20. Rotatable shielding drums & motors 21. Flow-limiting fast valves 22. Rotatable targets 23. Backing pumps 24. Instrumentation cabinets
1000 I
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25. Airlock 26. Crane, 10 ton capacity 27. Crane beam 28. Crane beam gearmotors 29. Crane rails 30. Crane rail columns 31. Manipulator turret 32. Telescoping boom 33. Manipulator & camera 34. Dual replaceable end effectors 35. Manipulator bridge 36. Manipulator bridge gearmotors 36. Manipulator bridge rails 37. Manipulator bridge rails 38. ITER cryostat 39. ITER vacuum vessel port no. 14 40. Metal room liner 41. In-port shielding
Fig. 2. ITER pellet injector assembly consisting of one two-stage and two single-stage hydrogen propellent guns.
successful. A 25 m ~ / s oil-lubricated T M P has b e e n tested (fig. 5), and efforts are underway to fit it with magnetic bearings.
R o u g h i n g and backing could be provided for the cryo-pumping options by available (Normetex-type) scroll p u m p s backed by metal bellows pumps. The
P.J. Dinner, H. Yoshida / ITER fuel cycle
15
VACUUHAND THERHAL BARRIER ARGONNANIFOLO
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16
P.J. Dinner, 14. Yoshida /ITER]uel cycle
Table 2 Comparison of compound cryopump and turbopump issues Feature
Single flow TMP
Ar spray CCP
Operating mode Tritium inventory
Cyclical High but isolated from "downstream"
Tritium processing
Continuous Low (not isolated) Pumps He/DT Easy
Scaleability He capacity (GDC)
Best to lower S~f Unlimited
Space requirement
Compact
Remote handling
Compact Bearing cleaning? Rotor heating Sensitive Components'? Component level
Magnetic field Radiation Experimental data avail. Dust ingress
Water ingress
Sudden venting Loss of power Seismic impact Prob. of meeting ITER perf. spec.
Mag. bearing fails'? Emerg. bearing Overheats? Bearing tolerance'? Insulation Tolerance Thrust -= tonnes Pump survival'? Emerg. bearing touchdown + ?? Pump survival? Questioned (EC)
complete pump-set consists of 6 of each of the 600 and 1200 m3/s pumps occupying 300 m 2 of floor space, and has a relatively high tritium inventory (50 g for the complete set). Improved pump designs are needed to reduce this inventory while simultaneously improving pumping speed in the 10-1-10 -2 mbar range. Introduction of heavy gas in the pump appears to significantly improve pumping performance for hydrogens [7] and would be compatible with the Ar spray CCP concept [3]. Cryotransfer pumps [8], offer compact means of achieving He compression and Ar separation for the Ar-spray pumps, and are therefore included in the reference configuration. It may also be possible to use these in "cryodiffusion pump" [9] mode to provide compact backing and He separation in an alI-TMP option or to connect them directly as primary pumps.
Complicated (Ar Separation) Best to higher S~< Limited (Need some TMPs) Not compact: pump cryogen distribution Not compact Valve seal R&R No effort Few sensitive components Component level Valve seal fails? Pump regenerates (undamaged) Pump regenerates (undamaged) Pump regenerates (undamaged) Survival expected Questioned (J)
"Cryodiffusion" pumping, which depends on DT molecule collision with He to compress the latter. would be particularly interesting if higher gas flow rates are required to alleviate divertor sputtering. Since the pump would not contain cryosorbents, DT could be "flash" regenerated, permitting lower average operating inventories. Turbo-pump backing requires improvement to the throughput achievable with existing mechanical backing pumps at low-pressure. A low-inventory, tritium compatible, reciprocating pump with 200 m~/h throughput at 13 mbar is under development in Japan as part of a complete 25 m3/s TMP train. A scaled~up molecular drag pump (Holweg stage) has been proposed as an alternative to Normetex pumps in the intermediate pressure range (10 3-10- 1 mbar).
P.J. Dinner, 1t. Yoshida / I T E R fuel cycle
17
2.3. Fuel processing
Fig. 5. 25 m3/s oil-lubricated turbomolecular pump. The pump will eventuallybe tested with magneticbearings.
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Fuel processing removes impurities from highly tritiated gas streams. It requires numerous interconnections of components to provide multiple processing pathways to permit an optimum response to multiple sources and variations in operating conditions (fig. 6). Impurities can be separated from the hydrogen streams using permeation, molecular-sieve cryosorption, cryocondensation or gettering techniques. Tritium may be recovered from the impurity stream by high-temperature isotopic exchange or oxidation and cold-trapping of the impurities in the stream, followed by reduction employing catalyst-beds, electrolysis, or oxidation of metal (e.g. iron). Neutral Beam and Pellet Injector gas impurities will be extracted via the molecular-sieve path, as this path is most suitable to high gas flows with lower tritium concentrations. Another approach, which avoids inventories of tritiated water, combines a permeation membrane with a catalyst to "crack" hydrocarbons and reduce water vapour via the water-gas-shift reaction [10]. High-temperature isotopic exchange (HITEX) uses exchange reactions between tritium in impurities and a swamping stream of protium over a hot platinum wire [11]. Electrolysis cells suitable for reducing water with very high tritium concentrations have been developed and are undergoing testing [12,13]. Experiments are underway to optimize molecularsieve impurity trapping. This involves selection of sieve type, which may include specially modified zeolites
IMPURITY PROCESSING Catalytic Cracking
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Oxidation ColdTrap
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I NeutralBeam Pumpoff Blanketsect. Impurity
Fig. 6. Design options for fuel purification showing multiple feed sources, two options for impurity removal and multiple options for tritium recoveryfrom the impurities.
18
P.,L Dinner, H. Yoshida / ITER fuel eycle
[14], able to retain impurities while minimizing DT cosorption. Permeator designs which increase low-pressure-side conductance, and thereby decrease both hydrogen carry-over with the impurity stream and backing pump size are undergoing tests [10]. Experiments are underway to determine if tritiated water from oxidation of plasma impurities may be reduced on regenerable metal beds [15]. This would provide a low tritium-inventory, low-waste alternative to electrolysis. HITEX also offers potentially low-waste, low-inventory processing of plasma exhaust, but experimental results are yet to be reported.
strongly on the purge gas flow required, the H / T swamping ratio, and the concentration of water vapour generated in the breeder or occurring as a consequence of coolant leakage. Ceramic breeder designs may involve two He sweep gas streams: one from the breeder zone and a second from the Be multiplier zone. Due to the uncertainty in requirements, three tritium recovery options were considered for the ceramic breeder and multiplier purge gas streams. The option used for CDA design integration studies (fig. 7) involves cryosorption on molecular sieve of thc tritium produced in the blanket, along with the swamping hydrogen and any impurities present. The impurities are subsequently separated and tritium recovered using processes similar to those found in the plasma exhaust fuel processing. This option avoids generation of a large HTO inventory by recovering hydrogen isotopes in elemental form. The process uses generally proven concepts, although not yet demonstrated on large-scale. A process based on pressure-swing adsorp-
2.4. Blanket tritium recol,~ery
Blanket tritium recovery options depend on the reference blanket option. D-ceramic, LiPb and aqueous lithium salt have all been considered as candidates for the ITER "driver" blanket. The preferred recovery option for the ceramic breeder blanket depends
]
H"2 Recycle
Product T2
He
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Fig. 7. Process for tritium recovery from ceramic breeder employing large cryogenic tool-sieve beds which remove all hydrogens and impurities from the purge gas flow.
P.J. Dinner, H. Yoshida / ITER fuel cycle
tion may provide a lower-inventory alternative for extracting hydrogen isotopes from the He gas stream, but is still at the early experimental stage [16]. If a large fraction of the ceramic breeder blanket hydrogen and tritium is recovered in oxide form, it is proposed to oxidize the entire hydrogen flow and process it using Liquid Phase Catalytic Exchange (LPCE) and Vapour Phase Catalytic Exchange (VPCE). All elements of this process have been demonstrated in conjunction with Heavy Water Reactor facilities in France and Canada. Most elements of this process will likely be required in any event to process tritiated water wastes. For Aqueous Salt Blanket tritium recovery, pre-concentration using water distillation is envisaged, followed by Vapour Phase Catalytic Exchange. This process is proven on industrial scale. For the solid LiPb eutectic blanket, melting of the eutectic during shutdown and subsequent vacuum extraction with purification using a hot metal trap followed by recovery on metal getter beds is envisaged. Test sector tritium recovery will be carried out as part of test sector design, as tritium recovery is in some cases a part of the blanket test. One of the above driver-blanket processes could be used as a "default" scheme for recovery for any test sector, if required. For the ceramic breeder, design optimization of the tritium recovery will require minimizing the hydrogen swamping ratio and fraction of tritium recovered as oxide. If LPCE + VPCE is used, it is possible to combine the blanket tritium recovery and tritiated wastewater processing. However, optimization of this process requires demonstration of LPCE under high tritium conditions over long periods. For the LiPb eutectic, an efficient tritium permeation barrier is required. Introduction of the driver blanket after operation of test sectors would also permit the blanket tritium recovery system design to be based on operational experience with scaleable test rigs under realistic conditions. 2.5. Auxilliaries and common processes
Tritiated water processing required to detritiate water collected in the facility to concentrations suitable for re-use or environmental discharge can be accomplished by isotopic enrichment and isotopic exchange by either distillation of water (DW), combined electrolysis catalytic-exchange (CECE), or a combination of Water Distillation and Liquid Phase Catalytic Exchange (LPCE). Both CECE and D W processes enrich the tritium (and deuterium) concentrations of the water prior to feeding it to isotopic separation. For water
19
to be re-used in the plant, detritiation to a much less stringent level is required, and D W will be the process of choice. Tritiated water must be filtered and purified by ion-exchange, and possibly e v a p o r a t i o n / condensation before enrichment. All of the preceding processes have been demonstrated at industrial scale. Optimization will essentially involve minimization of tritiated aqueou s source terms. Tritiated atmosphere processing uses proven catalytic oxidation and drier technology. For high flow rates of inert atmospheres, direct adsorption of hydrogens and impurities using modified zeolites [17] would provide an alternative solution which avoids production of tritiated waste-water, and facilitates re-use of the inert gas. To cope with a loss-of cooling accident accompanied by major tritium release, a passive "filter vent" system [18] is undergoing detailed study to provide for pressure relief and tritium confinement. Solid Waste Tritium Recovery will involve heating, isotopic exchange, and vacuum degassing of components such as metal components exposed directly to the plasma. Processes for efficient tritium recovery from graphite and steel require laboratory demonstration and scale-up. Isotopic Separation is based on cryogenic distillation of hydrogen isotopes, which has been demonstrated at a scale sufficient for ITER [19,20], but without the variation in feed composition and flow rates likely to characterize ITER operation. All hydrogen feeds may be processed in a single train, or multiple trains for different feed compositions, depending on the outcome of experiments and process modelling to determine robustness and tritium inventories. Low-inventory column packings are being sought to permit tritium inventory reduction. As for tritiated water source terms, the flow rates of protium and deuterium must be carefully scrutinized during the EDA. A tradeoff is required between ISS feed-flow and acceptable tritium contamination of H 2 and D 2 feeds re-used in pellet injector and neutral beams respectively. Fuel storage will use metal hydride beds with a maximum capacity of 200 g of tritium per bed for long-term storage of tritium. LaNi 5_ r M n r , ZrNi, ZrCo and uranium can be used as getter materials depending on process requirements. A storage system able to accomodate multiple, independent, storage modules is being pursued as a means to minimize inventory releasable under single process failure, while resulting in a relatively compact storage system [21]. For large quantities of fuel gases containing relatively low tritium concentrations ( < 1%), metal tanks will be used for storage. The number and capacity of these tanks must
20
P.J. Dinner, H. Yoshida / ITER fuel cycle
be optimized via dynamic analysis of the fuel cycle operation, and careful attention to system interfaces to minimize tritiated gas hold-up. With the exception of components connected directly to the torus (Fuelling, Primary Exhaust, and the main loop of Blanket Tritium Recovery), FC components are located away from the machine, in an area dedicated to components containing tritium (fig. 8). This area has separate access and ventilation system to prevent the spread of contamination, and simplify emissions control under normal and upset conditions.
3. Design optimization steps for the EDA While the CDA demonstrated that feasible design options exist for all essential FC elements, only limited optimization of the FC design was possible during the CDA Phase. An "optimized" design for the Fuel Cycle is tied to the overall process of ITER design integration and can only emerge towards the end of the EDA Phase. Furthermore, it will depend on component development priorities and schedules followed by the ITER Parties. Detailed long-range R & D plans were
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Fig. 8. Tritium building - vertical cut illustrating several of the large tritium-handling components, in particular for isotope separation, water, and atmosphere detritiation.
P.J. Dinner, H. Yoshida / ITER fuel cycle prepared as part of the ITER Fuel Cycle CDA activities. Extended testing of components, particularly those for which operation in tritium is forseen, will only be available in the period 1995-98. Development of an optimum design for ITER Fuel Cycle systems requires that maximum benefit be obtained from expenditures in the accompanying R & D program. Criteria for FC R & D resource allocation should recognize potential contributions to: - Safety, in particular those features which could improve "passive safety", primarily T inventory minimization. - Provision of essential technical information for discussion with regulatory bodies to advance to the ITER construction phase on schedule. - Technological risk minimization, i.e. reducing the level of uncertainty posed by engineering solutions. - Reduced development cost and time. This would permit development resources to be focussed on the most important design feasibility issues. - Reduced capital cost (for overall project as well as fuel cycle). A number of organizational measures should be taken to facilitate design optimization at the start of the EDA Phase. These include: - Identification of R & D tasks for the testing of reference design concepts. Design variations and alternatives should be tested in the same loops, and under similar conditions. - Establishment of a computerized data base for approved design requirements which can be remotely accessed by all parties. - Identification and support of common software tools for all aspects of fuel cycle design. For example, for vacuum pumping, process flow-sheeting, drawing and document transfer. - Establishment of a computerized "Tritium Manual" data base for defining standard designs of components which form the building blocks of systems. Applications, cost, experience and other data should be "tagged" to the appropriate component.
4. Conclusion
The ITER CDA established that feasible design concepts existed for all ITER FC elements. New design concepts continue to evolve. During the EDA, all FC design concepts must be reviewed according to projectwide criteria to select the preferred processes and ensure these are adequately covered by the R & D Programs of the Parties. FC Design optimization can
21
therefore be expected to occur as part of the overall ITER design integration and optimization process. A number of organizational measures can be taken to facilitate FC design optimization, such as standardization of component testing, and development of common data bases and analysis programs for design.
5. Acknowledgements
The authors would like to acknowledge the efforts of all their ITER colleagues and the design and R & D staffs of the home organizations associated with Fuel Cycle definition during the ITER CDA.
References
[1] ITER Fuel Cycle Technical Report, Prepared by the ITER Fuel Cycle Design Unit, ITER Documentation Series, No. 31, IAEA, 1-6 October 1991. [2] R.D. Penzhorn, J. Anderson, R, Haange, B. Hircq, A. Meikle and Y. Naruse, Technology and component development for a closed tritium cycle, ISFNT-2, Karlsruhe 2-7 June, 1991, these Proceedings, Part A; Fusion Engrg. Des. 16 (1991) 141-157. [3] D.K. Murdoch and J.C. Boissin, The influence of argon cryotrapping agent on ITER fuel cycle design, ISFNT-2, Karlsruhe 2-7 June, 1991, these Proceedings, Part C; Fusion Engrg. Des. 18 (1991) 85-90. [4] D. Perinic, A. Mack, J.-C. Boissin and D.K. Murdoch, Experimental investigations of helium cryotrapping by argon frost, ISFNT-2, Karlsruhe 2-7 June, 1991. [5] C.R. Walthers, E.M. Jenkins, T.H. Batzer, D.W. Sedgley, S. Konishi, S. Ohira and Y. Naruse, Continued studies of co-pumping of deuterium and helium on a single, 4K activated charcoal panel, 9th Topical Meeting on Technology of Fusion Energy, Oak Brook, Illinois, USA, October 7-11, 1990. [6] W. Obert, personal communication, (February 27, 1990). [7] U. Berndt, T.L. Le, E. Kirste, M. Glugla and R.-D. Penzhorn, Performance characteristics of large scroll pumps, ISFNT-2, Karlsruhe 2-7 June, 1991, these Proceedings, Part C; Fusion Engrg. Des. 18 (1991) 73-77. [8] D. Murdoch, J.C. Boissin, A. Conrad and D. Perinic, Forepumping concept for NET torus exhaust, 16th SOFT, London, UK, 3-7 September 1990 (Elsevier, Amsterdam, 1991). [9] P.J. Dinner and D.K. Murdoch, Integrated plasma exhaust cryotransfer and processing for NET, Fusion Engrg. Des. 10 (1989) 209-215. [10] M. Glugla and R.-D. Penzhorn, Catalytic plasma exhaust clean-up for NET II/ITER, 16th SOFT, London, UK, 3-7 September 1990 (Elsevier, Amsterdam, 1991).
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t:J. Dinner, H. Yoshida / ITER /uel cycle
[11] A. Busigin, S.K. Sood and K,M. Kalyanam, A high temperature isotopic exchange process for recovering tritium from fusion fuel impurities, ISFNT-2, Karlsruhe 2 7 June, 1991, these Proceedings, Part C; Fusion Engrg. Des. 18 (1991) 49-52. [12] A. Rahier, R. Cornelissen, A. Bruggeman and P. de Regge, Fusion Technol. (1988) 1365. [13] P. Giroux, Th. Pelletier, Ph. Boucquey and J.F. Bressieux, Electrolysis cell for highly tritiated water, 16th SOFT, London, UK, 3-7 September 1990 (Elsevier, Amsterdam, 1991 ), [14] A. Facchini, C. Malara, G. Pierini, 1. Ricapito, B. Spelta and A. Viola, Adsorption constants and kinetic coefficients evaluation for H z / D 2 separation on Na-, Ca-, Ni-mordenites, 6th SOFT London, UK, 3-7 September 1990 (Elsevier, Amsterdam, 1991). [15] G. Bonizzoni, C. Carretti, M. Colombo and F. Ghezzi, Tritiated water reduction on metal bed - technology task TCP 1.3, Status Report, Istituto Di Fisica Del Plasma, Associazione C.N.R. - EURATOM, December 1990. [16] B. Keefer, B. Bora, M. Chew, M. Rump and O. Kveton,
[17]
[18]
[19]
[2/I]
[21]
A low inventory adsorptive process for tritium extraction and purification, 16th SOFT, London, UK, 3-7 September 1990. G. Vassallo, Reducing atmosphere processings on mordenite, task TCP 3-2, Association technology programme for NET, 1989-1991, Revision 1, January 1991. S.J. Piet and S.J. Brereton, Fusion radioactivity confinement and application to postulated ITER accidents, 1NEL Report EGG-FSP-94705, March 1991. R.B. Davidson, P. von flatten, M. Schaub and D. Ulrich, Commissioning and first operating experience at Darlington tritium removal facility, Fusion Technol. 14 No. 2 (1988) 472-479. T. Yamanishi, H. Yoshida, S. Hirata, T. Naito, Y. Narusc, R.H. Sherman, J.R. Bartlit, K.M. Gruetzmacher and J.L. Anderson, Single column and two column H - D - T distillation experiments at TSTA, ibid, 489-494. W. Brainy, B. Hircq, D. Leger, P. Muguet and M. Peyrat, Tritiated hydrogen gas storage study for the NET Project, 16th SOFT, London, UK. 3 7 September 19911 (Elsevier, Amsterdam, 1991).
11
ITER fuel cycle options and optimizations P.J. D i n n e r a a n d H. Y o s h i d a b a The NET Team, Max-Planck-lnstitut ]fir Plasmaphysik, W-8046 Garching, Germany b Japan Atomic Energy Research Institute, Tokaimura, lbaraki, Japan
Inter-connected elements are required to carry out fuelling of the torus, torus vacuum pumping, processing of (exhaust) fuel recovery of tritium from the breeder blanket and test sectors, and several "common" processes such as hydrogen isotope separation, storage and management of fuel gases, and treatment of solid, liquid and gaseous tritiated wastes. Results of the International Tokamak Experimental Reactor - Conceptual Design Activity (ITER-CDA) design efforts are summarized, and technical and organizational proposals made to facilitate design optimization during the next phase of the Project. 1. Introduction During the International Tokamak Experimental Reactor - Engineering Design Activity (ITER-EDA), the process of Fuel Cycle design which began in the Conceptual Design Activity (CDA) will be continued, but with increased emphasis on producing an integrated design which balances design criteria while continuing to meet all Project requirements. This will require that such criteria are well understood by all levels of the project, and an engineering management structure exists to ensure orderly evolution of the design and cohsistency with interfacing systems and proposed operating scenarios. With these assumptions, it is possible to define the steps in preparing an optimized Fuel Cycle design. In the early stages of the EDA, it is reasonable to expect that a complete review of all CDA options will be conducted. In addition to the main approaches used in the CDA, this review should include an assessment of novel concepts and approaches to design implementation, some of which may have emerged during the period between the CDA and EDA. In this paper the main design options for ITER Fuel Cycle are reviewed, and a number of novel approaches and recent developments are discussed. Steps envisaged for the design "optimization" process are presented. 2. Fuel cycle design concepts and innovations The main elements of the ITER Fuel Cycle (FC) are shown in fig. 1, and key design parameters are listed in table 1. During the CDA phase, ITER FC
conceptual design was performed using well-advanced technologies to determine if: (1) Acceptable process options exist for all essential FC functions. (2) Subsystem concept design requirements are correct, since FC systems involve many "recycle" loops. (3) Impacts of the FC on plant arrangement, safety, and cost are acceptable. (4) Specifications for ITER FC R & D have been correctly formulated, and R & D priorities are consistent with the design process and schedule. The Fuel Cycle designs evolved in the ITER-CDA are described in ref. [1]. R & D progress involving many of the key processes is summarized in ref. [2]. In the following discussion, options used in the ITER FC design integration exercise are presented first. These are either based on proven technology or subject to detailed R & D investigations in ITER home organizations. Additional proposals to improve on concepts used for design integration in the CDA have arisen since the C D A work was performed. These developments are also introduced in this paper, with reference to specific advantages they may offer. 2.1. Fuelling
A combination of gas puffing and pellet injection past the scrape-off layer is required for fuelling. Fastvalve concepts for gas-puffing are being demonstrated on current-generation machines, and appear to be scaleable to ITER. However, further attention must be paid to increasing flow rate, reducing time-constants of valve and gas distribution circuit, and demonstrating
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12
P.J. Dinner, H. Yoshida / ITER fuel cycle
Hydrogen Discharge ,,a
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tritium, neutron and gamma radiation, and electromagnetic field compatibility of the gas-puffing valves. For fuelling during ramp-up, pellet injectors capable of producing a limited number ( < 100/pulse) of pellets at velocities in excess of 2 k m / s will be provided to produce peaked profiles. Pellet injectors for continuous fuelling beyond the scrape-off layer (1-2 k m / s ) are also needed. For CDA-phase design integration an assembly consisting of two single-stage light gas guns and a two-stage light gas gun, all employing hydrogen propellent were used (fig. 2). The figure illustrates the compact arrangement necessitated by limited space and the requirement for port-sharing with neutron diagnostics. Both fuelling and plasma chamber vacuum pumping (section 2.2) have been designed for remote maintenance. The addition of a centrifugal pellet launcher to the basic design would permit shallow pellet injection without the high pro-
tium propellant loads associated with gas guns. The centrifuge would require the development of very high DT extrusion rates, since only pellets a few millimeters in diameter could be accelerated by a centrifuge. Deeper fuel injection without high propellant loads might be achieved using either rail guns or compact toroids. Neither approach has yet been tested as a fuelling method for a large tokamak. 2.2. Plasma chamber t,acuum p u m p i n g
Plasma chamber vacuum pumping options for burn and dwell pumping include both compound cryopumps (CCP) and turbopumps. Argon-spray CCPs (fig. 3) have been used in the initial design integration effort, since the design for this pump is well advanced [3], and At-spray efficiencies confirmed in recent measurements [4]. Cryopumps are arranged in 8 stations of
P.J. Dinner, H. Yoshida / ITER fuel cycle Table 1 ITER F C
-
13
Key design parameters
Fuelling and exhaust Fuelling and exhaust rate (mol/h) Effective pumping speed at torus during burn (m3/s) Down-pipe conductance assumed (m3/s) Operating pump speed (m3/s) Ultimate pressure in torus (mbar) Installed TMP capacity (ma/s) for torus conditioning
35-75 350-700 1000-1500 1000-1500 4x10 -7 120
Fuel (exhaust) processing Total flow (mbar l/s) Impurity in burn-time exhaust (mbar l/s)
H2
He Low-Z High-Z Glow discharge in He, flow rate (mbar I/s) He impurity conc. (mbar l/s) Low-Z
220-470 <5 <18 <10 <1 300 0.3
Blanket tritium processing Tritium production rate (g/full-power-day) Ceramic breeder hydrogen swamping ratio ( H / T ) Maximum H + T recovered as oxide (%-water) Nominal production of tritiated water (mol/d)
152 100 10 < 25
Common processes Max. flows to Isotopic Separation (mol/h) protium deuterium tritium Max. tritium conc. in fuel (%) Max. T in H z/HD exhaust (%) Effluent water detritiation rate (kg/h) Nominal tritium concentration of water to be detritiated (Ci/kg) Discharge concentration (Ci/kg) Plant volumes requiring air-detritiation (m3) Air He Graphite dust processing rate (kg/d) Max. local tritium inventory design target (g)
three (fig. 4). For initial pump-down and torus conditioning, TMPs are required (8 pumps of > 15 m3/s capacity, integrated into the cryopumping stations). Substantial progress has been made in cryosorption panel development and testing for He pumping. Thin cryosorbent layers bonded inorganically to a metal substrate are being tested for resistence to repeated temperature cycling, and the ability of the cryosorbent to pump He in the presence of DT and impurities has
<1×104 < 450 < 50 < 80 l x l 0 -7 < 150 <0.1 <1×10 -5 <2x10 s <5x104 1-10 180
been confirmed [4,5]. This would permit pump designs employing a single 4 K pumping surface, which implies higher specific pumping speeds in a simpler, more compact design. This innovation has already been incorporated into CCPs for JET [6]. An all-turbopump option would result in a single torus pumping system with potential advantages, especially lower overall tritium inventories (table 2), if development programs meeting ITER requirements are
P.J. Dinner, 14. Yoshida / 1TER fuel cycle
14
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I. Moderate velocity injector 2. Moderate velocity injector 3. High velocity injector (HVI) 4. HVI pump tube 5. HVI propellant reservoir 6. HVI propellant compressor 7. Propellant gas manifolds 8. Fuel gas manifolds 9. High pressure cage 10. Low vacuum chamber 11. Mid vacuum chamber 12. High vacuum chamber 13. Torus isolation valves, breaks, bellows 14. Structural shielding wall 15. Modular shielding 16. Normetex 1300 m3/h scroll pump 17. Normetex control cabinets 18. Turbopumps, 5000 L / s 19. Turbopumps, 500 L / s 20. Rotatable shielding drums & motors 21. Flow-limiting fast valves 22. Rotatable targets 23. Backing pumps 24. Instrumentation cabinets
1000 I
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Fig. 2. ITER pellet injector assembly consisting of one two-stage and two single-stage hydrogen propellent guns.
successful. A 25 m ~ / s oil-lubricated T M P has b e e n tested (fig. 5), and efforts are underway to fit it with magnetic bearings.
R o u g h i n g and backing could be provided for the cryo-pumping options by available (Normetex-type) scroll p u m p s backed by metal bellows pumps. The
P.J. Dinner, H. Yoshida / ITER fuel cycle
15
VACUUHAND THERHAL BARRIER ARGONNANIFOLO
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16
P.J. Dinner, 14. Yoshida /ITER]uel cycle
Table 2 Comparison of compound cryopump and turbopump issues Feature
Single flow TMP
Ar spray CCP
Operating mode Tritium inventory
Cyclical High but isolated from "downstream"
Tritium processing
Continuous Low (not isolated) Pumps He/DT Easy
Scaleability He capacity (GDC)
Best to lower S~f Unlimited
Space requirement
Compact
Remote handling
Compact Bearing cleaning? Rotor heating Sensitive Components'? Component level
Magnetic field Radiation Experimental data avail. Dust ingress
Water ingress
Sudden venting Loss of power Seismic impact Prob. of meeting ITER perf. spec.
Mag. bearing fails'? Emerg. bearing Overheats? Bearing tolerance'? Insulation Tolerance Thrust -= tonnes Pump survival'? Emerg. bearing touchdown + ?? Pump survival? Questioned (EC)
complete pump-set consists of 6 of each of the 600 and 1200 m3/s pumps occupying 300 m 2 of floor space, and has a relatively high tritium inventory (50 g for the complete set). Improved pump designs are needed to reduce this inventory while simultaneously improving pumping speed in the 10-1-10 -2 mbar range. Introduction of heavy gas in the pump appears to significantly improve pumping performance for hydrogens [7] and would be compatible with the Ar spray CCP concept [3]. Cryotransfer pumps [8], offer compact means of achieving He compression and Ar separation for the Ar-spray pumps, and are therefore included in the reference configuration. It may also be possible to use these in "cryodiffusion pump" [9] mode to provide compact backing and He separation in an alI-TMP option or to connect them directly as primary pumps.
Complicated (Ar Separation) Best to higher S~< Limited (Need some TMPs) Not compact: pump cryogen distribution Not compact Valve seal R&R No effort Few sensitive components Component level Valve seal fails? Pump regenerates (undamaged) Pump regenerates (undamaged) Pump regenerates (undamaged) Survival expected Questioned (J)
"Cryodiffusion" pumping, which depends on DT molecule collision with He to compress the latter. would be particularly interesting if higher gas flow rates are required to alleviate divertor sputtering. Since the pump would not contain cryosorbents, DT could be "flash" regenerated, permitting lower average operating inventories. Turbo-pump backing requires improvement to the throughput achievable with existing mechanical backing pumps at low-pressure. A low-inventory, tritium compatible, reciprocating pump with 200 m~/h throughput at 13 mbar is under development in Japan as part of a complete 25 m3/s TMP train. A scaled~up molecular drag pump (Holweg stage) has been proposed as an alternative to Normetex pumps in the intermediate pressure range (10 3-10- 1 mbar).
P.J. Dinner, 1t. Yoshida / I T E R fuel cycle
17
2.3. Fuel processing
Fig. 5. 25 m3/s oil-lubricated turbomolecular pump. The pump will eventuallybe tested with magneticbearings.
FEEDSOURCE ~
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1
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.....~ , ~
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Fuel processing removes impurities from highly tritiated gas streams. It requires numerous interconnections of components to provide multiple processing pathways to permit an optimum response to multiple sources and variations in operating conditions (fig. 6). Impurities can be separated from the hydrogen streams using permeation, molecular-sieve cryosorption, cryocondensation or gettering techniques. Tritium may be recovered from the impurity stream by high-temperature isotopic exchange or oxidation and cold-trapping of the impurities in the stream, followed by reduction employing catalyst-beds, electrolysis, or oxidation of metal (e.g. iron). Neutral Beam and Pellet Injector gas impurities will be extracted via the molecular-sieve path, as this path is most suitable to high gas flows with lower tritium concentrations. Another approach, which avoids inventories of tritiated water, combines a permeation membrane with a catalyst to "crack" hydrocarbons and reduce water vapour via the water-gas-shift reaction [10]. High-temperature isotopic exchange (HITEX) uses exchange reactions between tritium in impurities and a swamping stream of protium over a hot platinum wire [11]. Electrolysis cells suitable for reducing water with very high tritium concentrations have been developed and are undergoing testing [12,13]. Experiments are underway to optimize molecularsieve impurity trapping. This involves selection of sieve type, which may include specially modified zeolites
IMPURITY PROCESSING Catalytic Cracking
~
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~
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I NeutralBeam Pumpoff Blanketsect. Impurity
Fig. 6. Design options for fuel purification showing multiple feed sources, two options for impurity removal and multiple options for tritium recoveryfrom the impurities.
18
P.,L Dinner, H. Yoshida / ITER fuel eycle
[14], able to retain impurities while minimizing DT cosorption. Permeator designs which increase low-pressure-side conductance, and thereby decrease both hydrogen carry-over with the impurity stream and backing pump size are undergoing tests [10]. Experiments are underway to determine if tritiated water from oxidation of plasma impurities may be reduced on regenerable metal beds [15]. This would provide a low tritium-inventory, low-waste alternative to electrolysis. HITEX also offers potentially low-waste, low-inventory processing of plasma exhaust, but experimental results are yet to be reported.
strongly on the purge gas flow required, the H / T swamping ratio, and the concentration of water vapour generated in the breeder or occurring as a consequence of coolant leakage. Ceramic breeder designs may involve two He sweep gas streams: one from the breeder zone and a second from the Be multiplier zone. Due to the uncertainty in requirements, three tritium recovery options were considered for the ceramic breeder and multiplier purge gas streams. The option used for CDA design integration studies (fig. 7) involves cryosorption on molecular sieve of thc tritium produced in the blanket, along with the swamping hydrogen and any impurities present. The impurities are subsequently separated and tritium recovered using processes similar to those found in the plasma exhaust fuel processing. This option avoids generation of a large HTO inventory by recovering hydrogen isotopes in elemental form. The process uses generally proven concepts, although not yet demonstrated on large-scale. A process based on pressure-swing adsorp-
2.4. Blanket tritium recol,~ery
Blanket tritium recovery options depend on the reference blanket option. D-ceramic, LiPb and aqueous lithium salt have all been considered as candidates for the ITER "driver" blanket. The preferred recovery option for the ceramic breeder blanket depends
]
H"2 Recycle
Product T2
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J Li20
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Fig. 7. Process for tritium recovery from ceramic breeder employing large cryogenic tool-sieve beds which remove all hydrogens and impurities from the purge gas flow.
P.J. Dinner, H. Yoshida / ITER fuel cycle
tion may provide a lower-inventory alternative for extracting hydrogen isotopes from the He gas stream, but is still at the early experimental stage [16]. If a large fraction of the ceramic breeder blanket hydrogen and tritium is recovered in oxide form, it is proposed to oxidize the entire hydrogen flow and process it using Liquid Phase Catalytic Exchange (LPCE) and Vapour Phase Catalytic Exchange (VPCE). All elements of this process have been demonstrated in conjunction with Heavy Water Reactor facilities in France and Canada. Most elements of this process will likely be required in any event to process tritiated water wastes. For Aqueous Salt Blanket tritium recovery, pre-concentration using water distillation is envisaged, followed by Vapour Phase Catalytic Exchange. This process is proven on industrial scale. For the solid LiPb eutectic blanket, melting of the eutectic during shutdown and subsequent vacuum extraction with purification using a hot metal trap followed by recovery on metal getter beds is envisaged. Test sector tritium recovery will be carried out as part of test sector design, as tritium recovery is in some cases a part of the blanket test. One of the above driver-blanket processes could be used as a "default" scheme for recovery for any test sector, if required. For the ceramic breeder, design optimization of the tritium recovery will require minimizing the hydrogen swamping ratio and fraction of tritium recovered as oxide. If LPCE + VPCE is used, it is possible to combine the blanket tritium recovery and tritiated wastewater processing. However, optimization of this process requires demonstration of LPCE under high tritium conditions over long periods. For the LiPb eutectic, an efficient tritium permeation barrier is required. Introduction of the driver blanket after operation of test sectors would also permit the blanket tritium recovery system design to be based on operational experience with scaleable test rigs under realistic conditions. 2.5. Auxilliaries and common processes
Tritiated water processing required to detritiate water collected in the facility to concentrations suitable for re-use or environmental discharge can be accomplished by isotopic enrichment and isotopic exchange by either distillation of water (DW), combined electrolysis catalytic-exchange (CECE), or a combination of Water Distillation and Liquid Phase Catalytic Exchange (LPCE). Both CECE and D W processes enrich the tritium (and deuterium) concentrations of the water prior to feeding it to isotopic separation. For water
19
to be re-used in the plant, detritiation to a much less stringent level is required, and D W will be the process of choice. Tritiated water must be filtered and purified by ion-exchange, and possibly e v a p o r a t i o n / condensation before enrichment. All of the preceding processes have been demonstrated at industrial scale. Optimization will essentially involve minimization of tritiated aqueou s source terms. Tritiated atmosphere processing uses proven catalytic oxidation and drier technology. For high flow rates of inert atmospheres, direct adsorption of hydrogens and impurities using modified zeolites [17] would provide an alternative solution which avoids production of tritiated waste-water, and facilitates re-use of the inert gas. To cope with a loss-of cooling accident accompanied by major tritium release, a passive "filter vent" system [18] is undergoing detailed study to provide for pressure relief and tritium confinement. Solid Waste Tritium Recovery will involve heating, isotopic exchange, and vacuum degassing of components such as metal components exposed directly to the plasma. Processes for efficient tritium recovery from graphite and steel require laboratory demonstration and scale-up. Isotopic Separation is based on cryogenic distillation of hydrogen isotopes, which has been demonstrated at a scale sufficient for ITER [19,20], but without the variation in feed composition and flow rates likely to characterize ITER operation. All hydrogen feeds may be processed in a single train, or multiple trains for different feed compositions, depending on the outcome of experiments and process modelling to determine robustness and tritium inventories. Low-inventory column packings are being sought to permit tritium inventory reduction. As for tritiated water source terms, the flow rates of protium and deuterium must be carefully scrutinized during the EDA. A tradeoff is required between ISS feed-flow and acceptable tritium contamination of H 2 and D 2 feeds re-used in pellet injector and neutral beams respectively. Fuel storage will use metal hydride beds with a maximum capacity of 200 g of tritium per bed for long-term storage of tritium. LaNi 5_ r M n r , ZrNi, ZrCo and uranium can be used as getter materials depending on process requirements. A storage system able to accomodate multiple, independent, storage modules is being pursued as a means to minimize inventory releasable under single process failure, while resulting in a relatively compact storage system [21]. For large quantities of fuel gases containing relatively low tritium concentrations ( < 1%), metal tanks will be used for storage. The number and capacity of these tanks must
20
P.J. Dinner, H. Yoshida / ITER fuel cycle
be optimized via dynamic analysis of the fuel cycle operation, and careful attention to system interfaces to minimize tritiated gas hold-up. With the exception of components connected directly to the torus (Fuelling, Primary Exhaust, and the main loop of Blanket Tritium Recovery), FC components are located away from the machine, in an area dedicated to components containing tritium (fig. 8). This area has separate access and ventilation system to prevent the spread of contamination, and simplify emissions control under normal and upset conditions.
3. Design optimization steps for the EDA While the CDA demonstrated that feasible design options exist for all essential FC elements, only limited optimization of the FC design was possible during the CDA Phase. An "optimized" design for the Fuel Cycle is tied to the overall process of ITER design integration and can only emerge towards the end of the EDA Phase. Furthermore, it will depend on component development priorities and schedules followed by the ITER Parties. Detailed long-range R & D plans were
AT I ~ i ~ P ~ II[ TRI TI ATION
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Fig. 8. Tritium building - vertical cut illustrating several of the large tritium-handling components, in particular for isotope separation, water, and atmosphere detritiation.
P.J. Dinner, H. Yoshida / ITER fuel cycle prepared as part of the ITER Fuel Cycle CDA activities. Extended testing of components, particularly those for which operation in tritium is forseen, will only be available in the period 1995-98. Development of an optimum design for ITER Fuel Cycle systems requires that maximum benefit be obtained from expenditures in the accompanying R & D program. Criteria for FC R & D resource allocation should recognize potential contributions to: - Safety, in particular those features which could improve "passive safety", primarily T inventory minimization. - Provision of essential technical information for discussion with regulatory bodies to advance to the ITER construction phase on schedule. - Technological risk minimization, i.e. reducing the level of uncertainty posed by engineering solutions. - Reduced development cost and time. This would permit development resources to be focussed on the most important design feasibility issues. - Reduced capital cost (for overall project as well as fuel cycle). A number of organizational measures should be taken to facilitate design optimization at the start of the EDA Phase. These include: - Identification of R & D tasks for the testing of reference design concepts. Design variations and alternatives should be tested in the same loops, and under similar conditions. - Establishment of a computerized data base for approved design requirements which can be remotely accessed by all parties. - Identification and support of common software tools for all aspects of fuel cycle design. For example, for vacuum pumping, process flow-sheeting, drawing and document transfer. - Establishment of a computerized "Tritium Manual" data base for defining standard designs of components which form the building blocks of systems. Applications, cost, experience and other data should be "tagged" to the appropriate component.
4. Conclusion
The ITER CDA established that feasible design concepts existed for all ITER FC elements. New design concepts continue to evolve. During the EDA, all FC design concepts must be reviewed according to projectwide criteria to select the preferred processes and ensure these are adequately covered by the R & D Programs of the Parties. FC Design optimization can
21
therefore be expected to occur as part of the overall ITER design integration and optimization process. A number of organizational measures can be taken to facilitate FC design optimization, such as standardization of component testing, and development of common data bases and analysis programs for design.
5. Acknowledgements
The authors would like to acknowledge the efforts of all their ITER colleagues and the design and R & D staffs of the home organizations associated with Fuel Cycle definition during the ITER CDA.
References
[1] ITER Fuel Cycle Technical Report, Prepared by the ITER Fuel Cycle Design Unit, ITER Documentation Series, No. 31, IAEA, 1-6 October 1991. [2] R.D. Penzhorn, J. Anderson, R, Haange, B. Hircq, A. Meikle and Y. Naruse, Technology and component development for a closed tritium cycle, ISFNT-2, Karlsruhe 2-7 June, 1991, these Proceedings, Part A; Fusion Engrg. Des. 16 (1991) 141-157. [3] D.K. Murdoch and J.C. Boissin, The influence of argon cryotrapping agent on ITER fuel cycle design, ISFNT-2, Karlsruhe 2-7 June, 1991, these Proceedings, Part C; Fusion Engrg. Des. 18 (1991) 85-90. [4] D. Perinic, A. Mack, J.-C. Boissin and D.K. Murdoch, Experimental investigations of helium cryotrapping by argon frost, ISFNT-2, Karlsruhe 2-7 June, 1991. [5] C.R. Walthers, E.M. Jenkins, T.H. Batzer, D.W. Sedgley, S. Konishi, S. Ohira and Y. Naruse, Continued studies of co-pumping of deuterium and helium on a single, 4K activated charcoal panel, 9th Topical Meeting on Technology of Fusion Energy, Oak Brook, Illinois, USA, October 7-11, 1990. [6] W. Obert, personal communication, (February 27, 1990). [7] U. Berndt, T.L. Le, E. Kirste, M. Glugla and R.-D. Penzhorn, Performance characteristics of large scroll pumps, ISFNT-2, Karlsruhe 2-7 June, 1991, these Proceedings, Part C; Fusion Engrg. Des. 18 (1991) 73-77. [8] D. Murdoch, J.C. Boissin, A. Conrad and D. Perinic, Forepumping concept for NET torus exhaust, 16th SOFT, London, UK, 3-7 September 1990 (Elsevier, Amsterdam, 1991). [9] P.J. Dinner and D.K. Murdoch, Integrated plasma exhaust cryotransfer and processing for NET, Fusion Engrg. Des. 10 (1989) 209-215. [10] M. Glugla and R.-D. Penzhorn, Catalytic plasma exhaust clean-up for NET II/ITER, 16th SOFT, London, UK, 3-7 September 1990 (Elsevier, Amsterdam, 1991).
22
t:J. Dinner, H. Yoshida / ITER /uel cycle
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