Radiation Shielding Using DUO2 in Nonmetallic Matrices
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E. Hopf, Conceptual Design Report for a Transportable DUCRETE™ Spent Fuel Storage Cask System, INEL-95/0167 (Idaho Falls, Idaho: Idaho National Engineering and Environmental Laboratory, April 1995). 10 F. P. Powell, Comparative Economics for DUCRETE™ Spent Fuel Storage Cask Handling, Transportation, and Capital Requirements, INEL-95/0166 (Idaho Falls, Idaho: Idaho National Engineering and Environmental Laboratory, March 1995). 11 P. A. Lessing, Development of “DUCRETE™,” INEL-94/0029 (Idaho Falls, Idaho: Idaho National Engineering and Environmental Laboratory, March 1995).
2 H. R. : Sandia National Laboratories, 1994). 3 F. E. Kosinski, “Review of Commercial Uranium Processing Capability,” Letter Report to C. R. S. DOE, EM50, TDC-100, Rev. 1, Technics Development Corporation, October 29, 1993. 4 W. J. Quapp and P. A. S. Patent No. 5,786,611, “Radiation Shielding Composition,” July 28, 1998. 5 W. J. Quapp and P. A. S. Patent No. 6,166,390, “Radiation Shielding Composition,” Dec. 26, 2000. 6 J. W. Sterbentz, Shielding Evaluation of a Depleted Uranium Heavy Aggregate Concrete for Spent Fuel Storage Casks, INEL-94/0092 (Idaho Falls, Idaho: Idaho National Engineering and Environmental Laboratory, October 1994).
18 R. Diersch and A. : Gesellschaft für Nuklear-Behälter [GNB] January 1998). 19 J. Tang, Shielding Characteristics of Various Materials on PWR Waste Packages, BBAC00000001717-0210-00008, Rev. 00 (Yucca Mountain Project Office, February 1998). 20 W. J. , March 1999. 21 W. J. , August 29–September 2, 1999. 22 N. Seagle, Duke Engineering Services, “DUCRETE HLW Storage System Feasibility,” unpublished report, February 1998. 23 ASTM C 289–94, “Standard Test Method for Potential Alkali-Silica Reactivity of Aggregate” (Chemical Method).