Carbon Materials for Advanced Technologies Episode 12 doc

420 50 100 250 200 250 300 3 50 Mean Thermal Conductivity (Wlm-K) Fig. 15. Weight loss as a function of mean thermal conductivity of graphite. 5 Tritium Retention in Graphite In the previous section the interaction of the plasma particle flux with the surface of graphite was discussed. However, the fate of the implanted particles (most importantly deuterium and tritium) following their impact with the graphite surface is also an important issue, and is seen by some as the major impediment to graphite's use as a PFM [SI. Quantifkation of the problem, and determination of possible mitigating steps, is complicated by experimental data which can vary by orders of magnitude [59-661 as reviewed by Wilson [67]. The physical process involved in the retention of hydrogen, as it corresponds to graphte PFMs, is fairly well understood. The energetic hydrogen isotopes are implanted to depths of less than a micron in the PFM surface. Once implanted, the hydrogen ions are either trapped, re-emitted, or diffused through the bulk graphite. At temperatures less than 100°C [68-721 the majority of ions are trapped near the end of their range. These trapped ions are not in solution in the graphite, but are 42 1 held [73] in the highly defective structure. The amount of hydrogen isotope whch can be accommodated is largely dependent on implantation temperature [72, 741 and to a lesser extent by implantation depth [70,71]. The total retained isotopic H can reach as much as 0.4-0.5 WC in the implanted layer at room temperature [68, 71, 751. As the mount of implanted hydrogen increases toward its saturation value, a larger fraction of ions are released from the graphite surface. None of these reemitted atoms become trapped in unsaturated regions. For intermediate and high temperatures (>250"C) diffusion of hydrogen in the graphite lattice occurs. This in-lattice difision most llkely occurs along internal surfaces, such as micro-pores and micro-cracks, while transgranular diffusion has been seen above 750°C [76, 771. This bulk diffusion, along with the associated trapping of hydrogen at defect sites, has been studied widely with quite variable results. This variation is shown in Fig. 16 where the temperature dependence of the hydrogen dffusion coefficient is shown for several carbon and graphite materials. 1 - Atsumi et al. (ref. 60) 5 - Rohrig et al. 64) 8 - Tanabe and Watanabe (66) 0.4 0.6 0.8 1 1.2 1.4 lo3 / T (K) Fig. 16. Hydrogen diffusion coefficient as a function of inverse temperature. 422 * 0 Unirradiated * 7 NeutmnIrradiated It would be expected that the diffusion of hydrogen through graphite would be highly dependent on the graphite microstructure, which may explain the wide range of the data of Fig. 16. In any event, the transport of hydrogen through the bulk graphite and associated solubility limits, can significantly increase the hydrogen inventory. The effect of the perfection of graphitic structure on the solubility of hydrogen is shown by Atsumi's data [78] in Fig. 17 which indicates that the more defect-free, highly-graphitized material has a lower solubility limit. Further evidence for the role of structural perfection comes from the observation that material which has been disordered by neutron irradiation has significantly higher solubility for hydrogen [78,79]. Graphific Perfection (%) Fig. 17. Hydrogen solubility as a function of graphitic perfection. The effect of atomic cllsplacements on the hydrogen retention of graphite was first shown by Wampler using 6 MeV ion beams (801. Wampler used four types of intermediate and high-quality graphites and irradiated with a high energy carbon beam at room temperature, followed by exposure to deuterium gas. Wampler's results indicated that the residual deuterium concentration increased by more than a factor of 30 to 600 appm for displacement doses appropriate to ITER. However, for reasons that are not yet clear, neutron irradiated high-quality CFCs retain 423 significantly less tritium than would be expected from the earlier work. This was reported by Atsumi [78] and is clearly shown by the recent work of Causey [SI] (Fig. 18). Causey irradiated high thermal conductivity MKC-1PH unidirectional composite and FMI-222 3D composite at -150°C (a particularly damaging irradiation temperature regime) to a range of displacement doses up to 1 dpa. As is seen in Fig. 18 the tritium retention is greater than one order of magnitude less than expected from earlier work on GraphNOL-N3M [82]. FML222CFC 1000 100 10 1 0.001 0.01 0.1 1 10 Radiation Damage (dpa) Fig. 18. Tritium retention as a function of neutron damage in graphite and graphite composite. The primary concern related to fuel retention in the PFC is the inventory of hydrogen adsorbed into the graphite and subsequent release of near surface hydrogen (due to sputtering) as plasma discharge begins. The hydrogen sputtered from the wall oversupplies the plasma edge with fuel, causing instabilities and making plasma control problematic. Tritium inventory concerns are generally safety related, but can have significant economic consequences because of the high cost of tritium. The potential release to the environment in an accident situation has limited the allowed inventory in TFTR, and may have si@icant consequences 424 for the sighting of the ITER. It has been estimated [58] that as much as -1.5 kg of tritium would reside in the graphite PFM of ITER, corresponding to an additional fuel cost of 1.5 to 3 million dollars. A source of trapped hydrogen which has not been discussed to this point, and which may dominate the tritium inventory in ITER-like machines, is the "co- deposited layer" [58,83]. This layer is formed by the simultaneous deposition of carbon, which is eroded from the first wall, and hydrogen. Thick layers of carbon redeposited to low erosion areas are common, and have been seen in every large tokamak utilizing graphite PFMs. As this layer grows, the hydrogen contained therein cannot be liberated by surface sputtering and becomes permanently trapped. This problem is unique to graphite and will require continual surface conditioning to minimize the total inventory of trapped species. 6 Summary and Conclusions Carbon and graphite materials have enjoyed considerable success as plasma-facing materials in current tokamaks because of their low atomic number, high thermal shock resistance, and favorable properties. However, their use is not without problems and their application in next generation fusion energy devices is by no means certain. Significant amongst the issues for carbon and graphite PFMs are: neutron irradiation damage, which degrades the thermal conductivity and causes increased PFC surface temperatures; physical sputtering? chemical erosion, and radiation enhanced sublimation? which results in surface material loss to the plasma, and redeposition of carbon; and tritium inventory, which poses both a safety problem and an economic impediment to the use of graphite. The high-heat loads and surface temperature that occur after plasma disruptions are also problematic for carbons. However, the same high temperatures make the use of Be, which has a significantly lower melting temperature, very unlikely. Next generation machines will impose increasingly greater thermal loads on their PFCs. High thermal conductivity CFC materials may offer a solution to the high- heat loads, but further research is needed to overcome the problems noted above and to assure the place of carbon materials in future fusion power reactors. 7 Acknowledgments Research sponsored by the U.S. Department of Energy under contract DE-ACOS- 960R22464 with Lockheed Martin Energy Research Corporation at Oak Ridge National Laboratory. 425 8 References 1. 2. 3. 4. 5. 6. 7. 8. 9. M. Akiba and H. Madarame, J. Nucl. Mat., 212-215,90-96 (1994). R.D. Watson, et al., High Heat Flux Testing of CITFirst Wall Tiles, 1988. G.H. Kinchin and R.S. Pease, Rep. Phys. Prog. Phys., 18 (1) (1955). J.W.H. Simmons, Radiation Damage in Graphite, Pergamon Press, (1965 ). C.R. Kennedy. InExtendedAbstracts for 14th Bienniel Conference on Carbon, 1979, Ervine, California. B.T. Kelly, Physics of Graphite, Applied Science Publishers (1981). G.B. Engle, Carbon, 12,291-306 (1974). T.D. BurchelI and W.P. Eatherly,J. Nucl. 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BURCHELL Metals and Ceramics Division Oak Ridge National Laboratory Oak Ridge, Tennessee 37831-6088, USA 1 The Role of Carbon Materials in Fission Reactors I. I Nuclear fission - basic concepts Nuclear (fission) reactors produce useful thermal energy from the fission (or disintegration) of isotopes such as 92U235, g2UB3, and 94Pu239. Fission of a heavy element, with release of energy and further neutrons, is usually initiated by an impinging neutron. The fission of 92V35 may be written: 92V35 + ,,n' - 92TJ236* - A + B + n + energy. The two fission fragments "A" and "B" will vary over a range of mass numbers fiom about 74 to 160, there being a whole range of possible reactions. Although an integral number of neutrons is emitted for any one fission, the average yield per fission (when all possible methods of fission are considered) is about 2.5 neutrons. The neutrons emitted by the fission reactions can be described by a Maxwellian distribution in energy, with a mean value of about 2 MeV. The total amount of energy given up per atom fissioned is of the order of 200 MeV, which is distributed approximately as indicated in Table 1. Table 1. Energy distribution from the fission of a UZ3' atom [I] Nuclear Process Energy (MeV) Kinetic energy of fission fragments Kinetic energy of fission neutrons 170 6 5 5 Instantaneous y-rays 6 Radioactive decay of fission fragments, p energy Radioactive decay of fission fragments, y energy [...]... U.S.A E 1.75 1119.6 15/13 2 0124 60160 1501135 3.514.5 Germany M 1.78 9.919.2 15/14 2 6123 2 67163 125 4.714.9 IG-110 Jzlpm I 1.75 10 25 34 71 124 1138 413.6 TSX U.S.A E 1.7 1413.8 2517 1I4 GR-280 Russia E I 72 6.515 7.616 3 4124 103189 3.214.9 GR2 -125 Russia E 1.85 121 8.5 1518 59159 1601100 3.915.2 Pitch-coke ASR- 1RS aE-extruded,M-molded, I-isostatic pressing; to the forming axis 437 1.4 Historical... graphite i 5 Q\ Table 4 Physical and mechanical properties of some common nuclear graphites[l l-191 Forming Grade Bulk Density glcm' StrengthbMPa Elastic Modulusb GPa Tensile Bend Comp Thermal Cond? WlmK CTEb 1OdK-' Source Method" PGA U.K E 1.74 121 5 17111 19 /12 2 7127 2001109 0. 912. 8 SM2-24 U.K M 1.7 818.5 12 19 47 U.K./Fr E 1.8 13/10 25117 32/26 70163 1301135 413.8 IM 1-24 U.K M 1.81 11 27.5 23 70... properties of some nuclei [l] Element Hydrogen Deuterium Helium Lithium Beryllium Carbon Oxygen Uranium z 1 2 4 7 9 12 16 23 8 4 1.000 0.725 0.425 0.268 0.208 0.158 0 .120 0.0084 n 18 25 43 67 87 114 150 2150 The parameter 5 gives a good indication of a material’s moderating ability, but it is not entirely dependable For example, as shown in Table 2, hydrogen is a good moderator based on its high 5... coal-tar pitch The binder plasticizes the filler coke particles so that they can be formed Commonly used forming processes include extrusion, molding, and isostatic pressing The binder phase is carbonized during the subsequent baking operation (- 1000°C) Frequently, engineering graphites are pitch impregnated to densify the carbon artifact, followed by rebaking Useful increases in density and strength are... the Hanford (U.S.A.) site in 1943 The mission of the Oak Ridge and Hanford reactors was the production of weapons grade U and Pu under the auspices of the U.S Government's Manhattan Project It is worth noting that the first irradiated fuel was discharged from the Hanford B reactor less than two years after the historic demonstration of a self-sustainingnuclear reaction in CP- 1 [23] The early Hanford... cladding Therefore, considerable effort was expended in selecting appropriate materials for the PIPPA design The moderator graphite, Pile Grade A (PGA), was manufactured from a particularly pure coke, thus reducing its neutron capture cross section substantially relative to the graphites used in earlier experimental reactors such as BEPO The choice of fuel canning materials was limited to those with... brick The methane is added to the carbon dioxide coolant as a radiolytic corrosion inhibitor (see Section 4) 444 Fig 4 A typical advanced gas-cooled reactor graphite core (Hinkley Point B under construction) [I 11 The graphite core is located within a steel envelope called the gas baffle, which provides for reentrant cooling of the graphite structure (Fig 5) Cooled carbon dioxide is drawn from the... Publishing, Dartford, U.K., with permission Four steam generators, each consisting of three separate factory assembled u i s nt, are positioned in the annulus between the gas baffle and the inner wall of the pressure vessel After passing down through a steam generator, the cooled carbon dioxide gas discharges into one of the quadrants of the circulator annulus which forms the entry plenum for eight 5.2... with initial criticality being attained on January 31, 1974 The major performance parameters of Fort St Vrain are given in Table 9 The fuel was of the Triso particle type (see Section 5 ) with kernels of fissile uranium dicarbide (93% enriched) or fertile thorium dicarbide [40] TabIe 9 The major performance parameters of the Fort St Vrain HTGR [29,38-401 Parameter Coolant Pressure Core inlet temperature... The fuel particles were formed into a fuel compact (Section 5.3) and sealed into the fuel channels The core was divided into 37 regions, each containing 7 columns, except for the 6 regions at the core periphery which each contained 5 columns The center column fuel elements and top reflector additionally contained three control rod channels, two for the operational rods and one for the B,C reserve shutdown . 10. T.D. Burchell and T. Oh, Materials Properties Data for Fusion Reactor Plasma Facing Carbon- Carbon Materials. Nuclear Fusion, 1994 5(Suppl.), 77 128 . 11. H.H. Yoshikawa, et al.,. GPa Tensile Bend Comp. WlmK 1 OdK-' Forming PGA U.K. E 1.74 121 5 1711 1 19 /12 2 7127 2001109 0. 912. 8 __ SM2-24 U.K. M 1.7 818.5 12 19 47 Pitch-coke U.K./Fr. E 1.8 13/10 25117. 11. H.H. Yoshikawa, et al., Radiation Damage in Reactor Materials, 1963. 12. M. Eto, et al., J: Nucl. Mat., 212- 215, 122 3 -122 7 (1994), 13. L. Ahlf, et aZ.,J: Nucl. Mat., 171, 31