The use of water in a fusion power core

Water has both advantages and disadvantages as a coolant in conceptual designs of future fusion power plants. In the United States, water has not been chosen as a fusion power core coolant for decades. Researchers in other countries continue to adopt water in their designs, in some cases as the leading or sole candidate. In this article, we summarize the technical challenges resulting from the choice of water coolant and the differences in approach and assumptions that lead to different design decisions amongst researchers in this field.

[1]  Takashi Tsukada,et al.  Stress corrosion cracking susceptibility of a reduced-activation martensitic steel F82H , 2009 .

[2]  L. Giancarli,et al.  A conceptual study of commercial fusion power plants. Final report of the European Fusion Power Plant Conceptual Study (PPCS) , 2005 .

[3]  Minami Yoda,et al.  The ARIES Advanced and Conservative Tokamak Power Plant Study , 2014 .

[4]  Steven J. Zinkle,et al.  Materials RD for a timely DEMO: Key findings and recommendations of the EU Roadmap Materials Assessment Group , 2014 .

[5]  M. Richou,et al.  Design of a water cooled monoblock divertor for DEMO using Eurofer as structural material , 2014 .

[6]  S. Ide,et al.  Compact DEMO, SlimCS: design progress and issues , 2009 .

[7]  C. Tomastik,et al.  Oxidation of beryllium—a scanning Auger investigation , 2005 .

[8]  Kikuchi Mitsuru,et al.  Blanket and divertor design for the Steady State Tokamak Reactor (SSTR) , 1991 .

[9]  Hiroshi Kawamura,et al.  Surface reaction of titanium beryllide with water vapor , 2006 .

[10]  David A. Petti,et al.  Steam-chemical reactivity for irradiated beryllium , 1998 .

[11]  Brad J. Merrill,et al.  Updated Modeling of Postulated Accident Scenarios in ITER , 2009 .

[12]  Steven J. Zinkle,et al.  Irradiation performance of stainless steels for ITER application , 1996 .

[13]  Frederic Escourbiac,et al.  Overview and status of ITER internal components , 2014 .

[14]  Michael L. Corradini,et al.  Lithium alloy chemical reactivity with reactor materials: current state of knowledge , 1991 .

[15]  K. Kleefeldt Thermal-hydraulic model of the helium-cooled pebble bed test blanket module for ITER FEAT , 2003 .

[16]  Mohamed A. Abdou,et al.  Physics and technology conditions for attaining tritium self-sufficiency for the DT fuel cycle , 2006 .

[17]  David A. Petti,et al.  The safety implications of tokamak dust size and surface area , 1998 .

[18]  M. L. Corradini,et al.  Liquid metal chemical reaction safety in fusion facilities , 1987 .

[19]  Hiroshi Kawamura,et al.  Stability of titanium beryllide under water vapor , 2004 .

[20]  C. Kessel,et al.  The ARIES Advanced and Conservative Tokamak ( ACT ) Power Plant Study , 2008 .

[21]  David A. Petti,et al.  On the mechanisms associated with the chemical reactivity of Be in steam , 2000 .

[22]  T. Ihli,et al.  Development of a helium-cooled divertor: Material choice and technological studies , 2007 .

[23]  Minami Yoda,et al.  The ARIES-CS Compact Stellarator Fusion Power Plant , 2008 .

[24]  Pierre Cortes,et al.  ITER safety and licensing update , 2012 .

[25]  Brad J. Merrill,et al.  Reaction of porous beryllium in steam , 1992 .

[26]  R. J. Pawelko,et al.  Steam chemical reactivity of Be pebbles and Be powder , 2000 .

[27]  A. Sagara,et al.  Helical reactor design FFHR-d1 and c1 for steady-state DEMO , 2014 .

[28]  M. A. Fütterer,et al.  Development of the EU water-cooled Pb-17Li blanket , 1998 .

[29]  B. Snow,et al.  Criteria for practical fusion power systems: Report from the EPRI fusion panel , 1994 .

[30]  R. C. Weast CRC Handbook of Chemistry and Physics , 1973 .

[31]  Steven J. Zinkle,et al.  Evaluation of copper alloys for fusion reactor divertor and first wall components , 1996 .

[32]  L. C. Alves,et al.  Structural and Oxidation Studies of Titanium Beryllides , 2007 .

[33]  J. K. Bienlein,et al.  The half-life of N16 , 1964 .

[34]  R. J. Pawelko,et al.  Chemical reactivity and mobilization of beryllium exposed to steam , 1997 .

[35]  Brad J. Merrill,et al.  A preliminary assessment of beryllium dust oxidation during a wet bypass accident in a fusion reactor , 2009 .

[36]  M. Gadomska,et al.  Overview of EU DEMO design and R&D activities , 2014 .

[37]  Farrokh Najmabadi,et al.  Spherical torus concept as power plants—the ARIES-ST study , 2003 .

[38]  C. E. Kessel Fusion Nuclear Science Pathways Assessment , 2012 .

[39]  Mikio Enoeda,et al.  Overview of the ITER TBM Program , 2012 .

[40]  Brad J. Merrill,et al.  Implications of beryllium : steam interactions in fusion reactors , 1992 .

[41]  Giacomo Aiello,et al.  Development of the Helium Cooled Lithium Lead blanket for DEMO , 2014 .

[42]  M. A. Fütterer,et al.  A water cooled lithium-lead blanket without tritium permeation barriers: feasibility and economical analysis , 2001 .

[43]  Kenji Tobita,et al.  DEMO design activities in the broader approach under Japan/EU collaboration , 2014 .

[44]  Yves Poitevin,et al.  Demo blanket technology R&D results in EU , 2002 .

[45]  C. San Marchi,et al.  4.16 – Tritium Barriers and Tritium Diffusion in Fusion Reactors , 2012 .

[46]  Chien-chang Lin,et al.  TRITIUM RELEASE FROM NUCLEAR POWER PLANTS IN TAIWAN , 2003, Health physics.

[47]  K. Kim,et al.  A preliminary conceptual design study of blanket for Korean DEMO Reactor (K-DEMO) , 2013, 2013 IEEE 25th Symposium on Fusion Engineering (SOFE).

[48]  L. L. Lao,et al.  The ARIES-AT advanced tokamak, Advanced technology fusion power plant , 2006 .

[49]  P. Sardain,et al.  Power plant conceptual studies in Europe , 2007 .