Tritium in Fusion: R&D in the EU

References

• D. MAISONNIER et al, “A conceptual study of commercial fusion power plants, Final report of the European fusion power plant conceptual study (PPCS)”, EFDA-RP-RE-5.0 (2005).
   Google Scholar
• P. SARDAIN et al, “The European power plant conceptual study: Helium-cooled lithium-lead reactor concept”, Fus. Eng. Des. 81 (2006) 2673-2678.
   Web of Science ®Google Scholar
• A. AIELLO et al, “Hydrogen isotopes permeability in Eurofer 97 martensitic steel”, Fus. Sci. Tech. 41 (2002) 872-876.
   Web of Science ®Google Scholar
• E. SERRA et al, “Influence of traps on the deuterium behavior in the low activation martensitic steels F82H and Batman”, J. Nucl. Mater. 245 (1997) 108-114.
   Web of Science ®Google Scholar
• G.A. ESTEBAN et al, Hydrogen transport and trapping in EUROFER’97, Proc. ICFRM-12, Santa Barbara, California, 2005.
   Google Scholar
• M.M.W. PEETERS et al: In-pile Tritium release behaviour of Lithium Metatitanate produced by Extrusion-spheroidisation-sintering process in EXOTIC-9/1 in the High Flux Reactor Petten, Proc SOFT Warsaw 2006.
   Google Scholar
• G. KIZANE et al, “Behaviour of tritium in breeding materials” Fus. Eng. Des. 75-79 (2005) 896-901.
   Web of Science ®Google Scholar
• F. REITER, “Solubility and diffusivity of hydrogen isotopes in liquid Pb-17Li”, Fus. Eng. Des. 14 (1991) 207-211.
   Web of Science ®Google Scholar
• A. AIELLO et al, “Determination of hydrogen solubility in lead lithium using sole device”, Fus. Eng. Des. 81 (2006) 639-644.
   Web of Science ®Google Scholar
• F. GABRIEL et al, “2D element modelling of tritium permeation for the HCLL DEMO blanket module”, Proc SOFT, Warsaw 2006.
   Google Scholar
• J. KONYS et al, “Corrosion behavior of EUROFER steel in flowing eutectic Pb-17Li alloy”, J. Nucl. Mater. 329-333 (2004) 1379-1383.
   Web of Science ®Google Scholar
• A. MÖSLANG et al, “Production and characterization of titanium beryllides for HIDOBE irradiation”, J. Nucl. Mater. 367-370 (2007) 1069-1072.
   Web of Science ®Google Scholar
• Fus. Sci. Tech. 41 (2002) 303-1174.
   Google Scholar
• D. MURDOCH et al, “ITER Design Review; tritium issues”, this volume
   Google Scholar
• Y. POITEVIN et al, “The Test Blanket Modules project in Europe: From the strategy to the technical plan over next ten years”, Proc SOFT, Warsaw 2006.
   Google Scholar
• J-F. SALAVY et al, “Overview of the last progresses for the European test blanket modules projects”, Proc. SOFT, Warsaw 2006.
   Google Scholar
• M. Dalle DONNE et al, “European HCPB Test Blanket. ITER Design Description Document”, Dec. 1996, FZKA 5891, April 1997, FZK Karlsruhe.
   Google Scholar
• Y. KAWAMURA et al, “Extraction of hydrogen from water vapour by hydrogen pump using ceramic protonic conductor”, Fus. Sci. Tech. 45 (2004) 33-40.
   Web of Science ®Google Scholar
• N. BEKRIS et al, “Cold trapping of traces of tritiated water from the helium loops of a fusion breeder blanket”, Fus. Eng. Des. 58-59 (2001) 423-428.
   Web of Science ®Google Scholar
• S. BELOGLAZOV et al, Semi-technical cryogenic molecular sieve bed for the tritium extraction system of the TBM for ITER, Fus. Sci. Tech. 48 (2005) 662-665.
   Web of Science ®Google Scholar
• A. AIELLO et al, “TRIEX facility: an experimental loop to test tritium extraction systems from lead lithium” Proc SOFT, Warsaw 2006.
   Google Scholar
• G. DELL’ORCO et al, “HeFUS3-European Helium cooled blanket test facility for DEMO”, Proceedings of the 19th SOFT, Lisbon, Portugal, Fus. Tech. 1996, 1335-1338.
   Google Scholar
• B.E. GHIDERSA et al, “Helium Loop Karlsruhe (HELOKA): A valuable tool for testing and qualifying ITER components and their He cooling circuits”, Fus. Eng. Des. 81, (2006) 1471-1476.
   Web of Science ®Google Scholar
• M. GLUGLA et al, “Tritium tests with a Permcat reactor for isotopic swamping”, Proceedings of the 19th SOFT, Lisbon, Portugal, Fus. Tech. 1996, 1193.
   Google Scholar
• S.A. BIRDSELL et al, “Ultrahigh Tritium Decontamination of Simulated Fusion Fuel Exhaust using a 2-Stage Palladium Membrane Reactor”, Fus. Tech. 30 (1996) 905.
   Google Scholar
• M. NISHI et al, “Study on tritium accountancy in fusion DEMO plant at JAERI”, Fus. Eng. Des. 81 (2006) 745-751.
   Web of Science ®Google Scholar
• M. MAYER et al, “Further insight into the mechanism of hydrocarbon layer formation below the divertor of ASDEX Upgrade, Nucl. Fusion 46 (2006) 914-920.
   Web of Science ®Google Scholar
• H.G. ESSER et al, “Post mortem analysis of a JET quartz microbalance system», J. Nucl. Mat. 363-365 (2007) 146-151.
   Web of Science ®Google Scholar
• C.H. SKINNER et al, “Tritium removal from JET and THTR tiles by a scanning laser”, J.Nucl.Mater. 313-316 (2003) 499-503.
   Web of Science ®Google Scholar
• B. SCHWEER et al, “Laser desorption of deuterium retained in re-deposited carbon layers at TEXTOR and JET, J.Nucl.Mater. 337-339 (2005) 570-574.
   Web of Science ®Google Scholar
• K.K.J. GIBSON et al, “The removal of co-deposited hydrocarbon films from plasma facing components using high-power pulsed flaslamp irradiation”, J.Nucl.Mater. 337-339 (2005) 565-569.
   Web of Science ®Google Scholar
• C. GRISOLIA et al, “In-situ tokamak laser applications for detritiation and co-deposited layers studies”, J. Nucl. Mater. 363-365 (2007) 1138-1147.
   Web of Science ®Google Scholar
• IFMIF Comprehensive Design Report (CDR), printed January 2004.
   Google Scholar
• S.P. SIMAKOV et al, “Assessment of the 3H and 7Be generation in the IFMIF lithium loop”, J.Nucl. Mater. 329-333 (2004) 213-217.
   Web of Science ®Google Scholar
• A. C. BELL et al, “Detritiation processes needed for JET operations and their wider applicability”. Fus. Sci. Tech. 41 (2002) 626-631.
   Web of Science ®Google Scholar
• S. ROSANVALLON et al, “Tritium related studies within the JET fusion technology work programme”, Fus. Sci. Tech. 48 (2005) 268-273.
   Web of Science ®Google Scholar