Oak Ridge National Laboratory

Nuclear Engineering Applications Section
1993

  Author(s) / Report Title / Publication Number / Date
J. S. Tang, Report of Foreign Travel, ORNL/FTR-4604, 1993.
M. B. Emmett, Report of Foreign Travel, ORNL/FTR-4603, 1993.
R. M. Westfall and J. F. Alexander,  The 1992 Annual Criticality Review Committee Appraisal, ORNL/CF-93/9(ES), Martin Marietta Energy Systems, Oak Ridge National Laboratory, 1993.
W. C. Jordan and H. R. Dyer, Nuclear Criticality Safety Assessment of the Proposed CFC Replacement Coolants, ORNL/TM-12475, Martin Marietta Energy Systems, Oak Ridge National Laboratory, December 1993.
E. C. Beahm, R. A. Lorenz, and C. F. Weber, "Iodine Evolution and pH Control," 1993 Winter Meeting American Nuclear Society, November 14-18, 1993, San Francisco, California.   Trans. Am. Nucl. Soc., 69, 387 (1993).
M. B. Emmett, "Validation and Verification of the ORNL Monte Carlo Codes for Nuclear Safety Analyses," 1993 Winter Meeting American Nuclear Society, November 14-18, 1993, San Francisco, California.  Trans. Am. Nucl. Soc.,  69,  199-200 (1993).
C. M. Hopper, "Nuclear Criticality Safety Specialist Training and Qualification Programs," 1993 Winter Meeting American Nuclear Society, November 14-18, 1993, San Francisco, California.   Trans. Am. Nucl. Soc., 69,  255-256 (1993).
T. S. Kress, E. C. Beahm, C. F. Weber, and G. W. Parker, "Fission Product Transport Behavior 2," Nucl. Technol., 101, 262-269 (1993).
D. F. Hollenbach, L. M. Petrie, and N. F. Landers, "KENO-VI: A Monte Carlo Criticality Program with Generalized Quadratic Geometry," pp. 58-63 in Proc. 1993 Topical Meeting on Physics and Methods in Criticality Safety ,September 19-23, 1993, Nashville, Tennessee.
S. M. Bowman, R. Q. Wright, H. Taniuchi, and M. D. DeHart, "Validation of SCALE-4 Criticality Sequences Using ENDF/B-V Data," in Proc. 1993 Topical Meeting on Physics and Methods in Criticality Safety, September 19-23, 1993, Nashville, Tennessee.
D. T. Ingersoll, J. E. White, R. Q. Wright, and R. W. Roussin, "Generation and Testing of an ENDF/B-VI Multigroup Cross-Section Library for LWR Shielding Applications,"  Proc. 8th ASTM-EURATOM Symp. on Reactor Dosimetry, August 29-September 3, 1993, Vail, Colorado.
D. T. Ingersoll, J. E. White, R. Q. Wright, H. T. Hunter, C. O. Slater, N. M. Greene, and R. E. MacFarlane, "Production and Testing of the Vitamin B6 Fine-Group and the Bugle-93 Broad-Group Neutron/Photon Cross-Section Libraries Derived from ENDF/B-VI Nuclear Data," Proc. 8th ASTM-Euratom Symp. on Reactor Dosimetry, August 29-September 3, 1993, Vail, Colorado.
W. A. Rhoades and R. L. Childs, "Application of TORT Three-Dimensional Neutron/Photon Transport Calculations," Proc. 8th ASTM-EURATOM Symp. of Reactor Dosimetry, August 29-September 3, 1993, Vail, Colorado.
C. V. Parks,  "The SCALE Criticality Safety Analysis Sequences: Status and Future Directions 2," 1993 Annual Meeting American Nuclear Society, June 20-24, 1993, San Diego, California . Trans. Am. Nucl. Soc.,  68(A), 237-238 (1993).
C. F. Weber and E. C. Beahm,  "Iodine Transport in a Severe Accident at the High Flux Isotope Reactor 2," 1993 Annual Meeting American Nuclear Society, June 20-24, 1993, San Diego, California .  Trans. Am. Nucl. Soc., 68(A), 275 (1993).
S. M. Bowman, O. W. Hermann, and M. C. Brady, "Burnup Credit Validation of SCALE-4 Using Light Water Reactor Criticals," 1993 Annual Meeting American Nuclear Society, June 20-24, 1993, San Diego, California .  Trans. Am. Nucl. Soc., 68(A), 243 (1993).
S. M. Bowman and H. Taniuchi, "Burnup Credit Validation of SCALE-4 Using Mixed-Oxide Critical Experiments," 1993 Annual Meeting American Nuclear Society, June 20-24, 1993, San Diego, California .Trans. Am. Nucl. Soc., 68(A), 241 (1993).
R. Q. Wright, "Revised Evaluations for ENDF/B-VI Revision 2," 1993 Annual Meeting American Nuclear Society, June 20-24, 1993, San Diego, California. Trans. Am. Nucl. Soc., 68(A), 468 (1993).
C. V. Parks,  Report of Foreign Travel, ORNL/FTR-4636, June 1993.
B. L. Broadhead, C. V. Parks, J. S. Tang, and H. Taniuchi, "Shielding Benchmark Calculations of Selected Spent Fuel Storage Cask Experiments," Vol. 2, pp. 1377-1382 in  Proc. of Fourth Annual Inter. Conf. on High Level Radioactive Waste Management, April 26-30, 1993, Las Vegas, Nevada.
M. B. Emmett,  "Status of the MORSE Multigroup Monte Carlo Radiation Transport Code," Proc. Seminar on Advanced Monte Carlo Computer Programs for Radiation Transport,  April 27-29, 1993, Saclay, France.
J. S. Tang and B. L. Broadhead,  "Development and Application of the Automated Monte Carlo Biasing Procedure in SAS4," Proc. Advanced Monte Carlo Computer Programs for Radiation Transport, OECD-NEA, April 27-29, 1993, Saclay, France.
W. C. Jordan,  Validation of SCALE 4.0 - CSAS25 Module and the 27-Group ENDF/B-IV Cross-Section Library for Low-Enriched Uranium Systems, ORNL/CSD/TM-287, Martin Marietta Energy Systems, Oak Ridge National Laboratory, February 1993.
W. C. Jordan,  Calculational Criticality Analyses of 10- and 20-MW UF6Freezer/Sublimer Vessels, ORNL/CSD/TM-288, Martin Marietta Energy Systems, Oak Ridge National Laboratory, February 1993.
R. L. Childs and W. A. Rhoades,  Theoretical Basis of the Linear Nodal and Linear Characteristics Methods in the TORT Computer Code, ORNL/TM-12246, Martin Marietta Energy Systems, Oak Ridge National Laboratory, January 1993.
S. M. Bowman, Criticality Safety Calculations for Region B of the Millstone Unit No. 2 Spent Fuel Pool, ORNL/M-2574, Martin Marietta Energy Systems, Oak Ridge National Laboratory, January 1993.


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