Improvement of the Inertial Fusion Energy Reactor Performance by Means of UF₄ and ThF₄

REFERENCES

• R. W. CONN, F. KANTROWITZ, and W. F. VOGELSANG “Hybrids for Direct Enrichment and Self-Protected Fissile Fuel Production,” Nucl. Technol., 49, 458 (1980).
   Web of Science ®Google Scholar
• E. GREENSPAN, “Fusion-Fission Hybrid Reactors,” Adv. Nucl. Sci. Technol., 16, 289 (1984).
   Google Scholar
• A. KUMAR and S. SAHN, “(Deuterium-Deuterium) Driven Experimental Hybrid Blankets and Their Neutronic Analysis,” Fusion Technol., 6, 225 (1984).
   Web of Science ®Google Scholar
• R. W. MOIR et al., “Design of Helium-Cooled Molten-Salt Fusion Breeder,” Fusion Technol., 8, 465 (1985).
   Web of Science ®Google Scholar
• J. S. WATSON et al., “Cost of Processing Fuel from a Molten Salt, Fusion/Fission, Hybrid Reactor Blanket,” Fusion Technol., 8, 2113 (1985).
   Google Scholar
• S. ŞAHiN and M. AL-ESHAIKH, “Fission Power Flattening in Hybrid Reactors Using Mixed Fuel,” Fusion Technol., 12, 395 (1987).
   Google Scholar
• D. H. BERWALD et al., “Updated Reference Design of a Liquid-Metal-Cooled Tandem Mirror Fusion Breeder,” Fusion Technol., 12, 30 (1987).
   Web of Science ®Google Scholar
• S. ŞAHiN, E. BALTACIOGLU, and H. YAPICI, “Rejuvenation of LWR Spent Fuel in a Catalyzed Fusion Hybrid Blanket,” Kerntechnik, 59, 243 (1994).
   Web of Science ®Google Scholar
• A. HATAYAMA et al., “Basic Requirements for a 1000 MW(electric) Class Tokamak Fusion-Fission Hybrid Reactor and Its Blanket Concept,” Fusion Technol., 26, 27 (1994).
   Google Scholar
• S. ÜNALAN, “Rejuvenation of CANDU Spent Fuel in (D, T) Driven Hybrid Reactors,” Fusion Technol., 33, 398 (1998).
   Google Scholar
• S. ÜNALAN, “Rejuvenation of LWR Spent Fuel in (D, T) Driven Hybrids Reactors,” Fusion Eng. Des., 38, 393 (1998).
   Google Scholar
• S. ÜNALAN and S. O. AKANSU, “Improvement of Neutronic Performance of the Hybrid Reactor Rejuvenating Spent Fuels Using Various Moderators,” Fusion Technol., 34, 109 (1998).
   Google Scholar
• P. A. HOUSE, “HYLIFE-II Reactor Chamber Mechanical Design,” Fusion Technol., 21, 1487 (1992).
   Web of Science ®Google Scholar
• R. W. MOIR, “HYLIFE-II Inertial Fusion Energy Power Plant Design,” Fusion Technol., 21, 1475 (1992).
   Google Scholar
• R. W. MOIR et al., “HYLIFE-II: A Molten Salt Inertial Fusion Energy Power Plant Design—Final Report,” Fusion Technol., 25, 5 (1994).
   Web of Science ®Google Scholar
• R. W. MOIR, “Improvements to the HYLIFE-II Inertial Fusion Power Plant Design,” Fusion Technol., 26, 1169 (1994).
   Web of Science ®Google Scholar
• P. A. HOUSE, “HYLIFE-II Reactor Chamber Design Refinements,” Fusion Technol., 26, 1178 (1994).
   Google Scholar
• J. D. LEE, “Waste Disposal Assessment of HYLIFE-II Structure,” Fusion Technol., 26, 74 (1994).
   Web of Science ®Google Scholar
• R. W. MOIR, “The High-Yield Lithium-Injection Fusion-Energy HYLIFE-II Inertial Fusion Energy (IFE) Power Plant Concept and Implications for IFE,” Phys. Plasmas, 6, 2447 (1995).
   Google Scholar
• S. ŞAHLN et al., “Radiation Damage in Liquid-Protected First Wall Materials for IFE-Reactors,” Fusion Technol., 30, part 2A, 1027 (1996).
   Google Scholar
• C. J. CALL and R. W. MOIR, “A Novel Fusion Power Concept Based on Molten-Salt Technology: PACER Revisited,” Nucl. Sci. Eng., 104, 364 (1990).
   Web of Science ®Google Scholar
• J. R. ENGEL et al., “Conceptual Design Characteristics of a Denatured Molten-Salt Reactor with Once-Through Fueling,” ORNL/TM-7207, Oak Ridge National Laboratory (1980).
   Google Scholar
• W. W. ENGLE, Jr., “ANISN, A One-Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering,” K1693, Oak Ridge National Laboratory (1970).
   Google Scholar
• R. E. MacFARLANE, “MATXS10, 30-Neutron, 12-Group Photon Cross Sections from ENDF/B-VI with TRANSX-2,” Los Alamos National Laboratory (1994).
   Google Scholar
• T. A. AL-KUSAYER, S. SAHLN, and A. DRIRA, “CLAW-IV, Coupled 30 Neutrons, 12 Gamma Ray Group Cross Sections with Retrieval Programs for Radiation Transport Calculations,” in “RSIC Newsletter,” p. 4, Oak Ridge National Laboratory Radiation Shielding Information Center (May 1988).
   Google Scholar
• R. J. BARRETT and R. E. MacFARLANE, “CLAW, Coupled 30 Neutrons, 12 Gamma-Ray Group Cross Sections for Neutron Transport Calculations,” LA-7808-MS, Los Alamos Scientific Laboratory (Apr. 1979).
   Google Scholar
• Group Constants for Nuclear Reactor Calculations, I. I. BONDARENKO, Ed., Consultants Bureau, New York (1964).
   Google Scholar
• S. SAHIN, H. YAPICI, and S. ÜNALAN, “ERDEMLI, A Computer Program to Process ANISN Output Data,” Gazi University (1991).
   Google Scholar
• S. ŞAHIN and J. LIGOU, “The Effect of the Spontaneous Fission of Plutonium-240 on Energy Release in a Nuclear Explosive,” Nucl. Technol., 50, 88 (1980).
   Web of Science ®Google Scholar