A. "Sagara

T." Tanaka

National Institute for Fusion Science (NIFS-Repository)

§2. Long-Life Flibe Blanket Design for the LHD-Type Reactor FFHR Sagara, A., Tanaka, T. Aiming at system integration for D-T fusion demo-reactors on the basis of physics and engineering results established in the LHD project, much progress has been made in design studies, including R&D works on engineering issues in the LHD-type heliotron power reactor FFHR [1]. In those studies the coil pitch parameter y of continuous helical winding has been adjusted beneficially to reduce the magnetic hoop force while expanding the blanket space, and a self-cooled liquid blanket using molten salt flibe (BeF,-LiF) has been proposed, due to its advantages of low MHD pressure loss, low reactivity with air, low pressure operation, and low tritium solubility. In case of a liquid blanket, since the breeder liquid can be continuously circulated and refreshed during the reactor operation, the lifetime of blanket is essentially limited by the total displacement damage and He production in structural materials under irradiation by fusion neutrons. There are many candidates for structural materials such as reduced activation ferritic steel (RAF), vanadium alloy, and SiCISiC composite. In case of RAF, which has a very mature material database and is chemically compatible with f1ibe, the design limit is about 15MWalm' (about 120dpa). This means that the lifetime is 10 years under 1.5MW/m' as adopted so far in FFHR designs, and replacement is needed three times in . the reactor life of 30 years. Therefore, if the effective wall loading could be reduced by a factor of 3, then no replacement would be required. This concept was proposed about 30 years ago as ISSEC (Internal Spectral Shifter and Energy Converter) by employing thick carbon shields as armor tiles on the blanket wall [2]. In this concept, therefore, the breeder blanket radioactive waste is largely reduced, while the carbon armors, low-level waste with no y-ray,have to be replaced due to neutron damage. In that ISSEC study, however, the TBR was below 1.05 even with the Be neutron multiplier in C and 90% enriched liquid Li, and there was no practical means for actively cooling the carbon tiles below about 2000°C to avoid high carbon vapor pressure. Figure I shows the new proposal of our STB (Spectral-shifter and Tritium breeder Blanket) of flibe in the limited thickness of about 1 m, where the Be layer in C and the flibe zone are optimized. Comparing with our original blanket [1] without ISSEC, the MCNP-4C calculations using IENDL3.2 nuclear data show that in this STB the flux of fast neutrons (> O.IMeV) at the first wall of ILF-l (RAF) is reduced to about 1/3 the original flux as shown in Fig.2. At the same time the local TBR of 1.17 is kept and the fast neutron f1uence to SC magnets is one order reduced to 5xlo" nlm'in 30 years, which is sufficient to keep TclTco >0.9 for Nb,Sn, for instance. According to a preliminary thermal analysis, the surface temperature of the carbon armor is about 1,600°C under conditions of total volumetric heat deposition of IMW/m', an effective thermal conductivity of 180WlmlK for C-Be-C bonded armors of 16cm thickness and heat transfer coefficient of 6,OOOW/m'/K for the mechartical contact using a graphite sheet[3] between the armor and the first wall of JLF-1. These conditions are not yet fully available, due to dimensional changes and degradation of thermal properties of carbon under neutron irradiations. Those are key R&D issues. As for the neutron multiplier, the use of higher melting point Be, such as Bel, Ti, BeO or Beze, and the optimization of the nuclear properties are under investigation. Fortunately the tritium inventory trapped in carbon is negligible at such high temperatures. Tiles First wall ip'~a , , Self-cooled Radiation shield T breeder & j500C Thermal shield Vacuum vessel & T boundary Fig.1 The radial build of typical STB of Flibe. Fig.2 Neutron energy spectrum at the first wall with and without carbon armors as STB. Reference 1) A.Sagara et al., 17th IAEA Fusion Energy Conf., Yokohama, FJP-3 (1998). 2) R.W. Conn, G.L. Kulchinski et aI., Nucl. Technology 26 (1975) 125. 3) Y.Kubota et al., Fusion Eng. Des. 56-57 (2001) 205. 241 

Long-Life Flibe Blanket Design for the LHD-Type Reactor FFHR

Abstract

Similar works

Full text

Available Versions

National Institute for Fusion Science (NIFS-Repository)