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RF TBMs for ITER tests
Fusion Engineering and Design 81 (2006) 425–432
RF TBMs for ITER tests I.R. Kirillov a,∗ , G.E. Shatalov b , YU.S. Strebkov c , the RF TBM Team a
b
D.V. Efremov Scientific Research Institute of Electrophysical Apparatus, 196641 St. Petersburg, Russia Russian Research Center, Kurchatov Institute, Kurchatov Square 1, 123182 Moscow, Russia c Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, 101000 Moscow, Russia Received 25 January 2005; received in revised form 25 April 2005; accepted 6 May 2005 Available online 27 December 2005
Abstract Following the DEMO design analysis, two test blanket modules (TBM) were chosen in the RF for the development and testing in ITER: ceramic helium-cooled TBM and lithium self-cooled TBM. In the first one, lithium containing ceramics is used for tritium breeding, helium is used as a coolant and purge gas for tritium extraction, beryllium—as a multiplier. Ferritic steel is a structure material. In the second one lithium is used as tritium breeder and a coolant, and vanadium alloy of V–Cr–Ti system as a structure material. Conceptual designs of both TBMs and ancillary systems for their tests in ITER, strategy of tests, key R&D issues for both concepts are summarized. An international collaboration in R&D, development and testing of TBMs is of great importance due to shortage of testing space in ITER and due to high cost of the program. © 2005 Elsevier B.V. All rights reserved. Keywords: Fusion; Blanket; ITER test module; Helium; Ceramics; Lithium
1. Introduction
and a commercial power reactor (CPR). The major objectives of a DEMO project are to demonstrate:
Fusion development strategy in the RF considers three major steps: an experimental thermonuclear reactor (ITER), a demonstration power reactor (DEMO)
• an ability to produce heat and electric power at the level of CPR; • tritium self sufficiency; • an availability ≥0.6; • reliability and increased operation safety of all reactor systems during time comparable to operation time of CPR (over 30 years);
∗ Corresponding author. Tel.: +7 812 464 4590; fax: +7 812 464 2069. E-mail address: [email protected] (I.R. Kirillov).
0920-3796/$ – see front matter © 2005 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2005.05.004
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• decreased accident consequences and waste disposal demands with respect to fission reactors. The design solutions and DEMO systems should be fully adequate for CPR operation, though some extrapolation is possible. Two blanket concepts developed for the RF DEMO reactor (CHC, ceramic He-cooled and Li self-cooled ones) are planned to be tested in ITER with the aim to demonstrate the integrated performance of “DEMOlike” blanket modules and associated systems in fusion environment and calibrate the design tools for DEMO blankets. These tests should be accompanied with materials testing in dedicated facilities with high neutron fluence. 2. Blanket concepts for DEMO reactor DEMO reactor studies started in the RF in 1992, first for the pulsed type reactor and later for steady state operation [1–4]. The main technical requirements for these reactors are the following: (1) total electric power around 1.5 GW; (2) the tritium breeding ratio (TBR) ≥1.05–1.1; (3) the structural materials should withstand the first wall fluence up to 15–20 MWa/m2 with the possible replacement of in-vessel components after 5–10 MWa/m2 .
The main characteristics of two DEMO blankets considered for steady state operation are presented in Table 1. Lithium-cooled blanket concept utilizes Li as a coolant and tritium breeder. Its potential advantages are the following: low pressure operation at high temperature required for effective electricity generation; attractive heat transfer and heat removal characteristics; good tritium breeding capability; outside-of-blanket tritium extraction with low tritium losses; immunity to radiation damage. V-alloy (type V–4% Cr–4% Ti) is considered as a structural material due to its good compatibility with Li, good mechanical properties at high temperatures, capability to accommodate high heat fluxes and high neutron fluence with low degradation of properties, low decay heat and low activation. Beryllium may be used in some design options as a neutron multiplier and tungsten carbide or titanium carbide as the reflector material. Li-cooled concept requires some electroinsulating material to be placed on the interface of Li/structure material to decrease induced currents and magnetohydrodynamic pressure drop resulting from the interaction of these currents with the tokamak magnetic fields. Some different concepts of selfhealing electroinsulating coatings, based on CaO,
Table 1 Main characteristics of two blanket concepts for the RF DEMO steady state tokamak reactor with 2.44 GW thermonuclear power Coolant
Lithium
Helium
Breeder Neutron fluence (MWa/m2 ) FW neutron load, average/maximum (MW/m2 ) FW heat flux, average/maximum (MW/m2 ) Blanket radial dimension (m) Module/section toroidal dimension (m) Module/section poloidal dimension (m) Channel length in poloidal direction (m) Coolant temperature, inlet/outlet (◦ C) Coolant pressure (MPa) Coolant maximum velocity (m/s) Structure material Structure material maximum temperature (◦ C) Multiplier Multiplier maximum temperature (◦ C) Reflector (shielding) material Reflector material maximum temperature (◦ C) Breeder maximum temperature (◦ C) Tritium breeding ratio
Lithium (50% 6 Li) 10–15 2.7/4.4 0.4/0.7 0.5 1.25–1.45 2.9–7.3 2.9–7.3 400/600 1.2 1.4–2.2 V–Cr–Ti ∼690 Be ∼700 WC 350 600 ∼1.09
Lithium orthosilicate (Li4 SiO4 ) 10
0.5–0.7 1.05–1.5 0.83 – 300/500 10 ∼110 Ferritic steel (10Cr9MoMn) ∼550 ∼650 – – ∼1000 ∼1.06
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AlN, Er2 O3 , Y2 O3 , or multi-layer structures are being considered. Li-cooled blanket together with Li-cooled divertor and NaK-cooled vacuum vessel ensures high safety potential and promises good technical characteristics for electric power production. The only concern is the high chemical activity of Li in case of accidents. This problem is planned to be solved by standard technical measures, used, for example, in fast breeder reactors: excluding possible Li–water contact, providing Li loop vaults/compartments with an inert (nitrogen/argon) atmosphere, a steel liner on the walls, passive systems of fire prevention/mitigation (double wall piping, collectors/trashes for spilled Li, etc.). CHC blanket utilizes helium as a coolant and helium + 1% hydrogen as a purge gas for tritium extraction from lithium containing ceramic (Li2 SiO4 ) with ferritic steel (10Cr9MoMn) as a structure material. Helium high temperature provides for high efficiency of thermal to electric power conversion. The combination of a ceramic breeder and beryllium multiplier with helium assures high safety. The main concerns are 6 Li
427
burn-out in ceramics during reactor exploitation and tritium confinement.
3. RF TBMs for ITER tests Two test blanket module (TBM) concepts corresponding to that of the DEMO blanket are being developed in the RF for in-ITER testing. 3.1. Li self-cooled TBM Basic modification of the TBM is shown in Fig. 1. Its design is governed by the ITER requirement on lithium volume to be <35 l due to safety reasons. This limitation is the main point, which differs this design from previously reported [5,6]. The TBM is divided into two submodules in toroidal direction, having the common liquid metal cooling and tritium extraction subsystems. Each submodule has a number of poloidal Li channels (3) with the flow U-turn at the top. The upward and downward flow parts of the channels are separated by Be-insert (5). WC- or TiC-insert (6) is
Fig. 1. Li self-cooled TBM: (a) poloidal cross section; (b) view from the Frame; (c) toroidal cross section. 1, FW; 2, double-walled Li pipe; 3, Li channels; 4, partitions; 5, Be multiplier; 6, WC primary shield; 7, gas drainage pipe; 8, Li inlet/outlet pipes; 9, TBM back wall; 10, cable connector; 11, key; 12, flexible cartridge; 13, gripped structure; 14, heat conducting inserts.
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located behind the downward flow duct. The first wall (FW-pos.1) is designed in the form of a box, moulded of 5 mm thick sheet, the box radial thickness is 55 mm. A honeycomb sandwich, consisting of the mutually perpendicular ribs, is inserted in the box. Vertical ribs (4) divide the duct in the toroidal direction. Be- and WCinserts are located on the horizontal ribs. The FW is connected to the back one at the rear side by welding along the perimeter. The supply and discharge branch pipes, transiting to the pipe lines (8) are welded to the back wall bottom, and the drain pipe line (7) is welded to its top to provide for Li-duct filling-up (gas cavity removal). All welded joints are covered with the safeguarding housing (2). The TBM structure is made of V–Cr–Ti alloy, and surfaces in contact with Li are separated from Li by an electroinsulating barrier. Volume of Li in the TBM’s ducts is 22 l (11 kg), mass of the TMB is 0.5 t. Several (4–5) variants of the TBM design are envisaged for different types of tests. The sensor’s placement for measurement of neutron fluxes, temperatures, deformations, displacements, magnetic field density will be provided.
The SS–water structure is placed at the back of the TBM to provide together with the Frame and Shield Plug the required dose rate for maintenance. The TBM is fastened to the Primary Shield Block with four flexible supports and three keys of the type used for ITER Shielding Blanket modules. The Primary Shield Block is fastened to the Frame with six bolts and three keys. 3.2. Ceramic He-cooled (CHC) TBM CHC TBM has the following overall dimensions: 1.72 m × 0.514 m × 0.7 m (poloidal × toroidal× radial). The TBM consists of the following main components: FW, casing, breeding zone, headers and pipelines. The detailed design of the TBM is presented in Fig. 2. The FW (2) is a steel panel with 36 rectangular (12 mm × 10 mm) poloidal coolant channels (3). The FW panel is welded to the TBM casing (1). The plasma facing surface of the FW is coated with the protective beryllium layer, if necessary. The TBM casing is fabricated from the 22 mm thickness bent sheet which has the rectangular cooling channels (7). These cooling channels are poloidally
Fig. 2. CHC TBM: (a) toroidal cross section; (b) poloidal cross section; (c) view from the Frame. 1, TBM casing; 2, first wall; 3, first wall cooling channel; 4, outer tube of breeding zone cooling channel; 5, inner tube of breeding zone cooling channel; 6, ceramic breeder; 7, casing cooling channel; 8, inlet header of breeding zone; 9, outlet header of breeding zone; 10, return branch-pipe of breeding zone; 11, supply branch-pipe of first wall; 12, purge gas inlet branch-pipe; 13, purge gas outlet branch-pipe; 14, inlet branch-pipe of casing coolant; 15, intermediate header; 16, bracket; 17, fit place for shear key; 18, stub-key for flexible support.
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directed. The space between the FW and the casing is used for the breeding zone arrangement. The breeder inside tube-concept is a basis for the breeding zone design. The breeding zone is a set of five breeding elements. Each element is a coaxial bent tube assembly. The annulus between coaxial tubes is used for the coolant circulation. The internal space of the inner tube (5) is used for breeder (6) arrangement. The coaxial tubes have the following dimensions: Ø56 mm × 1 mm (outer tube, pos. 4) and Ø38 mm × 1 mm (inner tube). The space between the FW, the back plate and outer tubes is used for multiplier arrangement. Helium cools the FW through 36 parallel poloidal channels and then the breeding zone through five parallel breeding elements. The TBM casing is cooled by helium of the main cooling system and is equipped by the separate branch-pipe of coolant supply. The breeding zone is also equipped with the tritium extraction system. The purge gas flows through the breeder in downward direction.
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4. TBMs’ ancillary systems The TBMs’ ancillary systems providing for heat rejection, tritium extraction, coolants purification are developed as well. A part of these systems for Li-cooled TBM with liquid lithium are placed inside the vacuum vessel port extension to minimize lithium amount and to exclude the possible lithium fire outside of the tokamak. The intermediate heat exchangers for Li TBM and the main part of He cooling system for CHC TBM are placed in the dedicated near-port transporter (see Fig. 3). Only He relief tank is placed in the tokamak cooling water system vault. This configuration of ancillary systems and poloidally oriented TBM designs provide for simultaneous testing of both RF TBMs in one horizontal equatorial port. Two variants are considered for tritium extraction from Li: a nonequlibrium molecular distillation [7] and a cold trap method [8]. Both methods require the corresponding R&D to develop an optimum technology.
Fig. 3. Ancillary equipment of Li-cooled and CHC TBMs of the RF in near-port transporter. 1, gas tanks; 2, organic coolant/organic coolant heat exchanger; 3, organic coolant pump; 4, organic coolant/water heat exchanger; 5, organic coolant dump tank; 6, box of electric supply system; 7, compressor; 8, He blower; 9, vacuum pump; 10, He blowers; 11, He heat-exchangers; 12, He regenerating heat-exchangers; 13, electrical heater.
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Table 2 Main parameters of the RF TBMs Parameters
Li self-cooled TBM
CHC TBM
TBM dimensions (radial × poloidal × toroidal) (m) Coolant channel length in poloidal direction (m) Coolant Breeder
0.105 × 1.72 × 0.514 ∼1.7 Lithium (90% 6 Li) Lithium (90% 6 Li)
0.7 × 1.72 × 0.514 ∼1.7 Helium Li4 SiO4
Coolant temperature (◦ C) Inlet Outlet Coolant pressure (MPa) Coolant velocity (maximum) (m/s) Structure material Structure material maximum temperature (◦ C) Neutron multiplier Neutron multiplier maximum temperature (◦ C) ◦ Breeder maximum temperature ( C) Primary shield material Shield material maximum temperature (◦ C)
250–450 350–550 0.5 0.45 V–4Cr–4Ti ∼600 Be ≤560 550 WC ∼650
300 500 10 100 Ferritic steel: Cr (8–10%), Mo (0.5–0.7%), V (0.2–0.35%) 520 Be 650 ∼1000 – –
In CHC TBM tritium is removed from Li containing ceramics by diffusion to purge gas. Then tritium is extracted with getters.
5. TBMs’ main parameters The detailed performance analysis of both TBMs was made [9] to get their main characteristics: neutronics, thermal hydraulics, thermal mechanics, electromagnetic, etc. Safety analysis was also performed to meet the ITER requirements. Some results of this analysis are presented in Table 2 for the FW neutron load 0.78 MW/m2 , FW heat flux 0.25 MW/m2 (design value).
6. Scenario of TBMs tests in ITER The relevance of the TBM designs to the DEMO blankets is achieved with the use of the same materials, the same main design features, with keeping the TBM’s main parameters close to the DEMO blankets in separate “act-like” TBM tests (temperatures of the main components, coolant temperature rise, coolant velocities, mechanical stresses, etc.). The main objectives of the TBM testing in ITER are the following:
• Validation of the codes for tritium generation and tritium permeation rates; for structure material, breeder and multiplier temperature calculations; coolant flow distribution and stress analysis. Validation of the neutronic codes and libraries used in ITER and DEMO analysis, especially for prediction of tritium generation rates, nuclear heat deposition, neutron multiplication and shielding efficiency. • Study of the tritium recovery process efficiency and temperature dependence of residual tritium inventory in the breeder, tritium permeation rate to coolant and permeation barriers efficiency. • Study of the electroinsulating barriers performance under integrated action of irradiation, high temperature, Li velocity flow pattern in high magnetic fields of the tokamak reactor, mechanical and electromagnetic loads. • Study of the breeder and beryllium multiplier temperature control efficiency. • Demonstration of the blanket ability to generate high grade heat for electricity production. • Validation of the structural integrity of the TBM under integrated action of thermal, mechanical and electromagnetic loads. • Validation of the low fluence irradiation effects studied in fission reactors spectrum. A possible scenario of the RF TBMs tests in ITER, supposing their simultaneous testing, is presented in
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Table 3 Possible scenario of TBMs tests in ITER-FEAT ITER-FEAT operation stages
Years from first plasma
Number of pulses
TBM types
H-plasma
1–3
–
Installation of 1–4 TBM-0 modules; replacement, ∼1 module per year
D-plasma (limited amount of T)
4
D–T-plasma with low duty cycles
5
750
Installation of 1–3 diagnostic modules TBM-1 and 2–3 modules TBM-2 to be replaced in 7–12 months
6–7
1000–1500 per year
8–10
2500–3000 per year
D–T-plasma with high duty cycles
TBM-3 (1–2 modules) to be replaced in 1–2 years
Note: TBM-0 (1–4 modules), modules for diagnostics and preliminary tests (the first 4 years); TBM-1 (1–3 modules), diagnostic modules (neutron flux characteristics); TBM-2 (2 modules), short term functioning tests of different “act-like” modules (fifth–seventh years of ITER operation); TBM-3 (1–2 modules), functioning tests.
Table 3. It may be modified with the integration of other countries-participants testing scenarios, if the number of the TBM lines is larger than the testing space available.
7. Key R&D issues, development plans For the Li self-cooled TBM the following issues are considered to be the key ones: • V-alloy–lithium compatibility. • Electroinsulating barriers development on the interface of structure material/lithium. • Tritium management (key issue for all TBMs). For the CHC TBM the key issues are the following: • Breeder (Li4 SiO4 ) development in pebble bed form and its characterization. • Porous Be development in pebble bed form and its characterization. • Structure material characterization. An extensive R&D was performed so far in all these directions, though additional efforts are needed prior to the TBM manufacturing. V-alloys of V–Cr–Ti system were developed and basic properties of unirradiated and irradiated alloys were studied; preliminary promising results on Valloy/lithium compatibility at temperatures up to 700 ◦ C were obtained [10,11]. Electroinsulating barriers development and testing give new hopes for solving the problem [12,13]. MHD and heat trans-
fer (HT) study provided good results for MHD/HT characteristics understanding in straight ducts, though additional data are needed for flow characterization in ducts of complex geometry and/or ducts with imperfect insulation on the wall/liquid metal interface [14]. The experimental fabricating technology of the CHC TBM materials has been developed and the breeding zone models for the in-pile testing have been manufactured [15]. The functional reactor facility (RITM-F) for tritium breeding models tests has been fabricated and put into operation [16,17]. The in-pile testing of breeding zone model for the CHC TBM in the conditions of permanent tritium extraction have been performed [18]. The RF plans to install two TBMs in ITER (Licooled and CHC ones) prior to the first ITER plasma. Before that, key R&D issues have to be solved, technological R&D, manufacturing and testing of the TBMs prototypes, manufacturing and qualification tests of the TBMs to be shipped to ITER site are planned to be performed.
8. Conclusion Two DEMO relevant TBMs are being developed in the RF with the goal to start their testing in ITER from the first day of its operation. The RF actively supports the idea of developing a coordinated test program in ITER through by-lateral and multi-lateral consultations in ITER Test Blanket
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Working Group with the partners proposing the same TBM lines. The joint R&D activity is also of great importance and different frameworks for this activity are being discussed, some work is planned under IEA agreement.
References [1] Yu.A. Sokolov, Overview of Russian DEMO plant study, Fusion Eng. Des. 29 (1995) 18–27. [2] Yu.A. Sokolov, I.V. Altovskij, A.A. Borisov, A.A. Grigor’yan, B.N. Kolbasov, D.K. Kurbatov, et al., Russian DEMO plant study, Fusion Eng. Des. 41 (1998) 525–533. [3] G. Shatalov, I. Kirillov, Yu. Sokolov, Yu. Strebkov, N. Vasiliev, RF DEMO Team, Russian DEMO-S reactor with continuos plasma burn, Fusion Eng. Des. 51–52 (2000) 289– 298. [4] I.R. Kirillov, RF DEMO team, Lithium-cooled blanket of RF DEMO reactor, Fusion Eng. Des. 49–50 (2000) 457– 465. [5] I.V. Vitkovsky, M.M. Golovanov, A.V. Divavin, I.R. Kirillov, A.V. Lipko, A.A. Malkov, et al., Neutronic, thermal-hydraulic and stress analysis of RF lithium-cooled test blanket module for ITER, Fusion Eng. Des. 49–50 (2000) 703–707. [6] I.R. Kirillov, RF TBM Team, Lithium self-cooled blanket test module for ITER, in: Proceedings of the Fourth International Conference on MagnetoHydroDynamic at Dawn of Third Millenium, Presqu’ile de Giens, France, 2000. v.1, 291– 296. [7] M.I. Solonin, Materials science problems of blankets in Russian concept of fusion reactor, J. Nucl. Mater. 258–263 (1998) 30–46. [8] D.K. Sze, R.F. Mattas, J. Anderson, R. Haanger, H. Yoshida, O. Kveton, Tritium recovery from lithium based on cold trap, Fusion Eng. Des. 28 (1995) 220–225.
[9] Design Description Document. Russian Test Blanket Modules, RDIPE/Efremov Institute, St. Petersburg, Moscow, 2005. [10] T. Muroga, T. Nagasaka, K. Abe, V.M. Chernov, H. Matsui, D.L. Smith, et al., Vanadium alloys—overview and recent results, J. Nucl. Mater. 307–311 (2002) 547–554. [11] V.M. Chernov, I.R. Kirillov, A.N. Konev, O.I. Eliseeva, Review of V-alloys and stainless steel corrosion in lithium, in: Presented at International Workshop on Tritium Management and Corrosion Activities for Liquid Breeder Blankets, Brasimone, Italy, September 27–29, 2004. [12] A. Suzuki, B. Pint, T. Muroga, T. Yoneoka, S. Tanaka, Corrosion behavior of oxide coatings in liquid lithium for V/Li blanket application, in: Proceedings of the 11th International Conference on Fusion Reactor Materials, Abstracts, Japan, December 7–12, 2003, p. 274. [13] I.V. Vitkovsky, A.V. Gorunov, V.I. Engelko, I.R. Kirillov, V.M. Khoroshikh, N.D. Kraev, et al., Development and testing of electroinsulating barriers for lithium-vanadium fusion blanket, Fusion Eng. Des. 61–62 (2002) 739–743. [14] I.R. Kirillov, C.B. Reed, L. Barleon, K. Miyazaki, Present understanding of MHD and heat transfer phenomena for liquid metal blankets, Fusion Eng. Des. 27 (1995) 553–569. [15] V. Kovalenko, D. Davydov, V. Kapyshev, L. Kiryak, A. Lopatkin, A. Mrachev, et al., Breeding zone models of DEMO ceramic helium-cooled blanket test module for reactor in-pile testing, Plasma Dev. Oper. 12 (2) (2004) 75–80. [16] V. Kovalenko, V. Demidov, A. Zyrianov, V. Kapyshev, L. Menkin, L. Milder, et al., Experimental channel and gas system of RITM-F facility for in-pile tests, in: Proceeding of the 10th International workshop on ceramic breeder blanket interactions, Karlsruhe, Germany, October 22–24, 2001. [17] V. Kovalenko, V. Demidov, V. Kapyshev, N. Paltusev, Yu. Strebkov, V. Tebus, et al., RITM-F facility, Plasma Dev. Oper. 12 (2) (2004) 81–88. [18] V. Kapychev, V. Gorokhov, D. Davydov, V. Demidov, Yu. Kazenov, V. Tebus, et al., Model and mockup of DEMO ceramic breeding zone for testing in IVV-2M reactor, Fusion Technol. 39 (1) (2001) 45–53.
RF TBMs for ITER tests I.R. Kirillov a,∗ , G.E. Shatalov b , YU.S. Strebkov c , the RF TBM Team a
b
D.V. Efremov Scientific Research Institute of Electrophysical Apparatus, 196641 St. Petersburg, Russia Russian Research Center, Kurchatov Institute, Kurchatov Square 1, 123182 Moscow, Russia c Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, 101000 Moscow, Russia Received 25 January 2005; received in revised form 25 April 2005; accepted 6 May 2005 Available online 27 December 2005
Abstract Following the DEMO design analysis, two test blanket modules (TBM) were chosen in the RF for the development and testing in ITER: ceramic helium-cooled TBM and lithium self-cooled TBM. In the first one, lithium containing ceramics is used for tritium breeding, helium is used as a coolant and purge gas for tritium extraction, beryllium—as a multiplier. Ferritic steel is a structure material. In the second one lithium is used as tritium breeder and a coolant, and vanadium alloy of V–Cr–Ti system as a structure material. Conceptual designs of both TBMs and ancillary systems for their tests in ITER, strategy of tests, key R&D issues for both concepts are summarized. An international collaboration in R&D, development and testing of TBMs is of great importance due to shortage of testing space in ITER and due to high cost of the program. © 2005 Elsevier B.V. All rights reserved. Keywords: Fusion; Blanket; ITER test module; Helium; Ceramics; Lithium
1. Introduction
and a commercial power reactor (CPR). The major objectives of a DEMO project are to demonstrate:
Fusion development strategy in the RF considers three major steps: an experimental thermonuclear reactor (ITER), a demonstration power reactor (DEMO)
• an ability to produce heat and electric power at the level of CPR; • tritium self sufficiency; • an availability ≥0.6; • reliability and increased operation safety of all reactor systems during time comparable to operation time of CPR (over 30 years);
∗ Corresponding author. Tel.: +7 812 464 4590; fax: +7 812 464 2069. E-mail address: [email protected] (I.R. Kirillov).
0920-3796/$ – see front matter © 2005 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2005.05.004
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• decreased accident consequences and waste disposal demands with respect to fission reactors. The design solutions and DEMO systems should be fully adequate for CPR operation, though some extrapolation is possible. Two blanket concepts developed for the RF DEMO reactor (CHC, ceramic He-cooled and Li self-cooled ones) are planned to be tested in ITER with the aim to demonstrate the integrated performance of “DEMOlike” blanket modules and associated systems in fusion environment and calibrate the design tools for DEMO blankets. These tests should be accompanied with materials testing in dedicated facilities with high neutron fluence. 2. Blanket concepts for DEMO reactor DEMO reactor studies started in the RF in 1992, first for the pulsed type reactor and later for steady state operation [1–4]. The main technical requirements for these reactors are the following: (1) total electric power around 1.5 GW; (2) the tritium breeding ratio (TBR) ≥1.05–1.1; (3) the structural materials should withstand the first wall fluence up to 15–20 MWa/m2 with the possible replacement of in-vessel components after 5–10 MWa/m2 .
The main characteristics of two DEMO blankets considered for steady state operation are presented in Table 1. Lithium-cooled blanket concept utilizes Li as a coolant and tritium breeder. Its potential advantages are the following: low pressure operation at high temperature required for effective electricity generation; attractive heat transfer and heat removal characteristics; good tritium breeding capability; outside-of-blanket tritium extraction with low tritium losses; immunity to radiation damage. V-alloy (type V–4% Cr–4% Ti) is considered as a structural material due to its good compatibility with Li, good mechanical properties at high temperatures, capability to accommodate high heat fluxes and high neutron fluence with low degradation of properties, low decay heat and low activation. Beryllium may be used in some design options as a neutron multiplier and tungsten carbide or titanium carbide as the reflector material. Li-cooled concept requires some electroinsulating material to be placed on the interface of Li/structure material to decrease induced currents and magnetohydrodynamic pressure drop resulting from the interaction of these currents with the tokamak magnetic fields. Some different concepts of selfhealing electroinsulating coatings, based on CaO,
Table 1 Main characteristics of two blanket concepts for the RF DEMO steady state tokamak reactor with 2.44 GW thermonuclear power Coolant
Lithium
Helium
Breeder Neutron fluence (MWa/m2 ) FW neutron load, average/maximum (MW/m2 ) FW heat flux, average/maximum (MW/m2 ) Blanket radial dimension (m) Module/section toroidal dimension (m) Module/section poloidal dimension (m) Channel length in poloidal direction (m) Coolant temperature, inlet/outlet (◦ C) Coolant pressure (MPa) Coolant maximum velocity (m/s) Structure material Structure material maximum temperature (◦ C) Multiplier Multiplier maximum temperature (◦ C) Reflector (shielding) material Reflector material maximum temperature (◦ C) Breeder maximum temperature (◦ C) Tritium breeding ratio
Lithium (50% 6 Li) 10–15 2.7/4.4 0.4/0.7 0.5 1.25–1.45 2.9–7.3 2.9–7.3 400/600 1.2 1.4–2.2 V–Cr–Ti ∼690 Be ∼700 WC 350 600 ∼1.09
Lithium orthosilicate (Li4 SiO4 ) 10
0.5–0.7 1.05–1.5 0.83 – 300/500 10 ∼110 Ferritic steel (10Cr9MoMn) ∼550 ∼650 – – ∼1000 ∼1.06
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AlN, Er2 O3 , Y2 O3 , or multi-layer structures are being considered. Li-cooled blanket together with Li-cooled divertor and NaK-cooled vacuum vessel ensures high safety potential and promises good technical characteristics for electric power production. The only concern is the high chemical activity of Li in case of accidents. This problem is planned to be solved by standard technical measures, used, for example, in fast breeder reactors: excluding possible Li–water contact, providing Li loop vaults/compartments with an inert (nitrogen/argon) atmosphere, a steel liner on the walls, passive systems of fire prevention/mitigation (double wall piping, collectors/trashes for spilled Li, etc.). CHC blanket utilizes helium as a coolant and helium + 1% hydrogen as a purge gas for tritium extraction from lithium containing ceramic (Li2 SiO4 ) with ferritic steel (10Cr9MoMn) as a structure material. Helium high temperature provides for high efficiency of thermal to electric power conversion. The combination of a ceramic breeder and beryllium multiplier with helium assures high safety. The main concerns are 6 Li
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burn-out in ceramics during reactor exploitation and tritium confinement.
3. RF TBMs for ITER tests Two test blanket module (TBM) concepts corresponding to that of the DEMO blanket are being developed in the RF for in-ITER testing. 3.1. Li self-cooled TBM Basic modification of the TBM is shown in Fig. 1. Its design is governed by the ITER requirement on lithium volume to be <35 l due to safety reasons. This limitation is the main point, which differs this design from previously reported [5,6]. The TBM is divided into two submodules in toroidal direction, having the common liquid metal cooling and tritium extraction subsystems. Each submodule has a number of poloidal Li channels (3) with the flow U-turn at the top. The upward and downward flow parts of the channels are separated by Be-insert (5). WC- or TiC-insert (6) is
Fig. 1. Li self-cooled TBM: (a) poloidal cross section; (b) view from the Frame; (c) toroidal cross section. 1, FW; 2, double-walled Li pipe; 3, Li channels; 4, partitions; 5, Be multiplier; 6, WC primary shield; 7, gas drainage pipe; 8, Li inlet/outlet pipes; 9, TBM back wall; 10, cable connector; 11, key; 12, flexible cartridge; 13, gripped structure; 14, heat conducting inserts.
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located behind the downward flow duct. The first wall (FW-pos.1) is designed in the form of a box, moulded of 5 mm thick sheet, the box radial thickness is 55 mm. A honeycomb sandwich, consisting of the mutually perpendicular ribs, is inserted in the box. Vertical ribs (4) divide the duct in the toroidal direction. Be- and WCinserts are located on the horizontal ribs. The FW is connected to the back one at the rear side by welding along the perimeter. The supply and discharge branch pipes, transiting to the pipe lines (8) are welded to the back wall bottom, and the drain pipe line (7) is welded to its top to provide for Li-duct filling-up (gas cavity removal). All welded joints are covered with the safeguarding housing (2). The TBM structure is made of V–Cr–Ti alloy, and surfaces in contact with Li are separated from Li by an electroinsulating barrier. Volume of Li in the TBM’s ducts is 22 l (11 kg), mass of the TMB is 0.5 t. Several (4–5) variants of the TBM design are envisaged for different types of tests. The sensor’s placement for measurement of neutron fluxes, temperatures, deformations, displacements, magnetic field density will be provided.
The SS–water structure is placed at the back of the TBM to provide together with the Frame and Shield Plug the required dose rate for maintenance. The TBM is fastened to the Primary Shield Block with four flexible supports and three keys of the type used for ITER Shielding Blanket modules. The Primary Shield Block is fastened to the Frame with six bolts and three keys. 3.2. Ceramic He-cooled (CHC) TBM CHC TBM has the following overall dimensions: 1.72 m × 0.514 m × 0.7 m (poloidal × toroidal× radial). The TBM consists of the following main components: FW, casing, breeding zone, headers and pipelines. The detailed design of the TBM is presented in Fig. 2. The FW (2) is a steel panel with 36 rectangular (12 mm × 10 mm) poloidal coolant channels (3). The FW panel is welded to the TBM casing (1). The plasma facing surface of the FW is coated with the protective beryllium layer, if necessary. The TBM casing is fabricated from the 22 mm thickness bent sheet which has the rectangular cooling channels (7). These cooling channels are poloidally
Fig. 2. CHC TBM: (a) toroidal cross section; (b) poloidal cross section; (c) view from the Frame. 1, TBM casing; 2, first wall; 3, first wall cooling channel; 4, outer tube of breeding zone cooling channel; 5, inner tube of breeding zone cooling channel; 6, ceramic breeder; 7, casing cooling channel; 8, inlet header of breeding zone; 9, outlet header of breeding zone; 10, return branch-pipe of breeding zone; 11, supply branch-pipe of first wall; 12, purge gas inlet branch-pipe; 13, purge gas outlet branch-pipe; 14, inlet branch-pipe of casing coolant; 15, intermediate header; 16, bracket; 17, fit place for shear key; 18, stub-key for flexible support.
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directed. The space between the FW and the casing is used for the breeding zone arrangement. The breeder inside tube-concept is a basis for the breeding zone design. The breeding zone is a set of five breeding elements. Each element is a coaxial bent tube assembly. The annulus between coaxial tubes is used for the coolant circulation. The internal space of the inner tube (5) is used for breeder (6) arrangement. The coaxial tubes have the following dimensions: Ø56 mm × 1 mm (outer tube, pos. 4) and Ø38 mm × 1 mm (inner tube). The space between the FW, the back plate and outer tubes is used for multiplier arrangement. Helium cools the FW through 36 parallel poloidal channels and then the breeding zone through five parallel breeding elements. The TBM casing is cooled by helium of the main cooling system and is equipped by the separate branch-pipe of coolant supply. The breeding zone is also equipped with the tritium extraction system. The purge gas flows through the breeder in downward direction.
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4. TBMs’ ancillary systems The TBMs’ ancillary systems providing for heat rejection, tritium extraction, coolants purification are developed as well. A part of these systems for Li-cooled TBM with liquid lithium are placed inside the vacuum vessel port extension to minimize lithium amount and to exclude the possible lithium fire outside of the tokamak. The intermediate heat exchangers for Li TBM and the main part of He cooling system for CHC TBM are placed in the dedicated near-port transporter (see Fig. 3). Only He relief tank is placed in the tokamak cooling water system vault. This configuration of ancillary systems and poloidally oriented TBM designs provide for simultaneous testing of both RF TBMs in one horizontal equatorial port. Two variants are considered for tritium extraction from Li: a nonequlibrium molecular distillation [7] and a cold trap method [8]. Both methods require the corresponding R&D to develop an optimum technology.
Fig. 3. Ancillary equipment of Li-cooled and CHC TBMs of the RF in near-port transporter. 1, gas tanks; 2, organic coolant/organic coolant heat exchanger; 3, organic coolant pump; 4, organic coolant/water heat exchanger; 5, organic coolant dump tank; 6, box of electric supply system; 7, compressor; 8, He blower; 9, vacuum pump; 10, He blowers; 11, He heat-exchangers; 12, He regenerating heat-exchangers; 13, electrical heater.
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Table 2 Main parameters of the RF TBMs Parameters
Li self-cooled TBM
CHC TBM
TBM dimensions (radial × poloidal × toroidal) (m) Coolant channel length in poloidal direction (m) Coolant Breeder
0.105 × 1.72 × 0.514 ∼1.7 Lithium (90% 6 Li) Lithium (90% 6 Li)
0.7 × 1.72 × 0.514 ∼1.7 Helium Li4 SiO4
Coolant temperature (◦ C) Inlet Outlet Coolant pressure (MPa) Coolant velocity (maximum) (m/s) Structure material Structure material maximum temperature (◦ C) Neutron multiplier Neutron multiplier maximum temperature (◦ C) ◦ Breeder maximum temperature ( C) Primary shield material Shield material maximum temperature (◦ C)
250–450 350–550 0.5 0.45 V–4Cr–4Ti ∼600 Be ≤560 550 WC ∼650
300 500 10 100 Ferritic steel: Cr (8–10%), Mo (0.5–0.7%), V (0.2–0.35%) 520 Be 650 ∼1000 – –
In CHC TBM tritium is removed from Li containing ceramics by diffusion to purge gas. Then tritium is extracted with getters.
5. TBMs’ main parameters The detailed performance analysis of both TBMs was made [9] to get their main characteristics: neutronics, thermal hydraulics, thermal mechanics, electromagnetic, etc. Safety analysis was also performed to meet the ITER requirements. Some results of this analysis are presented in Table 2 for the FW neutron load 0.78 MW/m2 , FW heat flux 0.25 MW/m2 (design value).
6. Scenario of TBMs tests in ITER The relevance of the TBM designs to the DEMO blankets is achieved with the use of the same materials, the same main design features, with keeping the TBM’s main parameters close to the DEMO blankets in separate “act-like” TBM tests (temperatures of the main components, coolant temperature rise, coolant velocities, mechanical stresses, etc.). The main objectives of the TBM testing in ITER are the following:
• Validation of the codes for tritium generation and tritium permeation rates; for structure material, breeder and multiplier temperature calculations; coolant flow distribution and stress analysis. Validation of the neutronic codes and libraries used in ITER and DEMO analysis, especially for prediction of tritium generation rates, nuclear heat deposition, neutron multiplication and shielding efficiency. • Study of the tritium recovery process efficiency and temperature dependence of residual tritium inventory in the breeder, tritium permeation rate to coolant and permeation barriers efficiency. • Study of the electroinsulating barriers performance under integrated action of irradiation, high temperature, Li velocity flow pattern in high magnetic fields of the tokamak reactor, mechanical and electromagnetic loads. • Study of the breeder and beryllium multiplier temperature control efficiency. • Demonstration of the blanket ability to generate high grade heat for electricity production. • Validation of the structural integrity of the TBM under integrated action of thermal, mechanical and electromagnetic loads. • Validation of the low fluence irradiation effects studied in fission reactors spectrum. A possible scenario of the RF TBMs tests in ITER, supposing their simultaneous testing, is presented in
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Table 3 Possible scenario of TBMs tests in ITER-FEAT ITER-FEAT operation stages
Years from first plasma
Number of pulses
TBM types
H-plasma
1–3
–
Installation of 1–4 TBM-0 modules; replacement, ∼1 module per year
D-plasma (limited amount of T)
4
D–T-plasma with low duty cycles
5
750
Installation of 1–3 diagnostic modules TBM-1 and 2–3 modules TBM-2 to be replaced in 7–12 months
6–7
1000–1500 per year
8–10
2500–3000 per year
D–T-plasma with high duty cycles
TBM-3 (1–2 modules) to be replaced in 1–2 years
Note: TBM-0 (1–4 modules), modules for diagnostics and preliminary tests (the first 4 years); TBM-1 (1–3 modules), diagnostic modules (neutron flux characteristics); TBM-2 (2 modules), short term functioning tests of different “act-like” modules (fifth–seventh years of ITER operation); TBM-3 (1–2 modules), functioning tests.
Table 3. It may be modified with the integration of other countries-participants testing scenarios, if the number of the TBM lines is larger than the testing space available.
7. Key R&D issues, development plans For the Li self-cooled TBM the following issues are considered to be the key ones: • V-alloy–lithium compatibility. • Electroinsulating barriers development on the interface of structure material/lithium. • Tritium management (key issue for all TBMs). For the CHC TBM the key issues are the following: • Breeder (Li4 SiO4 ) development in pebble bed form and its characterization. • Porous Be development in pebble bed form and its characterization. • Structure material characterization. An extensive R&D was performed so far in all these directions, though additional efforts are needed prior to the TBM manufacturing. V-alloys of V–Cr–Ti system were developed and basic properties of unirradiated and irradiated alloys were studied; preliminary promising results on Valloy/lithium compatibility at temperatures up to 700 ◦ C were obtained [10,11]. Electroinsulating barriers development and testing give new hopes for solving the problem [12,13]. MHD and heat trans-
fer (HT) study provided good results for MHD/HT characteristics understanding in straight ducts, though additional data are needed for flow characterization in ducts of complex geometry and/or ducts with imperfect insulation on the wall/liquid metal interface [14]. The experimental fabricating technology of the CHC TBM materials has been developed and the breeding zone models for the in-pile testing have been manufactured [15]. The functional reactor facility (RITM-F) for tritium breeding models tests has been fabricated and put into operation [16,17]. The in-pile testing of breeding zone model for the CHC TBM in the conditions of permanent tritium extraction have been performed [18]. The RF plans to install two TBMs in ITER (Licooled and CHC ones) prior to the first ITER plasma. Before that, key R&D issues have to be solved, technological R&D, manufacturing and testing of the TBMs prototypes, manufacturing and qualification tests of the TBMs to be shipped to ITER site are planned to be performed.
8. Conclusion Two DEMO relevant TBMs are being developed in the RF with the goal to start their testing in ITER from the first day of its operation. The RF actively supports the idea of developing a coordinated test program in ITER through by-lateral and multi-lateral consultations in ITER Test Blanket
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Working Group with the partners proposing the same TBM lines. The joint R&D activity is also of great importance and different frameworks for this activity are being discussed, some work is planned under IEA agreement.
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