SCALE Publications
Reactor and Nuclear Systems Division
Design, Safety and Simulation Integration Group
B. T. Rearden and R. A. Lefebvre, Getting Started with VIBE as a DICE Plug-in Module, ORNL/TM-2010/60, Oak Ridge National Laboratory, Oak Ridge, Tenn., August 2010.
B. T. Rearden, "Verification Methods for the SCALE Code System," Proc. Verification and Validation for Nuclear Systems Analysis Workshop II, North Myrtle Beach, SC, May 24-28, 2010.
S. M. Bowman and I. C. Gauld, OrigenArp Primer: How to Perform Isotopic Depletion and Decay Calculations with SCALE/ORIGEN, ORNL/TM-2010/43, Oak Ridge National Laboratory, Oak Ridge, Tenn., April 2010.
M. L. Williams, S. M. Bowman, and C. V. Parks, “Plans for Future SCALE Development Beyond Version 6.0,” The American Nuclear Society and the European Nuclear Society 2007 International Conference on Making the Renaissance Real, November 11–15, 2007, Washington, D.C. Trans. Am. Nucl. Soc. 97, 606–607 (2007).
S. M. Bowman, “Overview of the SCALE Code System,” The American Nuclear Society and the European Nuclear Society 2007 International Conference on Making the Renaissance Real, November 11–15, 2007, Washington, D.C. Trans. Am. Nucl. Soc. 97, 589–591 (2007).
Stephen M. Bowman, Bradley T. Rearden, and James E. Horwedel, “GeeWiz Integrated Visualization Interface for SCALE 5.1,” p. 12–16 in Proceedings of the 8th International Conference on Nuclear Criticality Safety (ICNC 2007), Vol. II, May 28–June 1, 2007, St. Petersburg, Russia.
S. Goluoglu and M. L. Williams, "Modeling Doubly Heterogeneous Systems in SCALE," Trans. Am. Nucl. Soc. 93, 963-965 (2005).
S. M. Bowman, "Overview of Advances in SCALE Development," Trans. Am. Nucl. Soc. 92, 747-748 (2005).
Stephen M. Bowman and James E. Horwedel, "GeeWiz: Integrated User Interface for SCALE," Trans. Am. Nucl. Soc. 92, 767-769 (2005).
SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluations, ORNL/TM-2005/39, Version 5.1, Vols. I-III, November 2006. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-732.
S. M. Bowman, D. F. Hollenbach, M. D. DeHart, B. T. Rearden, I. C. Gauld, and S. Goluoglu, "SCALE 5: Powerful New Criticality Safety Analysis Tools," pp. 26-32 in Proc. of The 7th International Conference on Nuclear Criticality Safety (ICNC2003), October 20-24, 2003, Tokai-mura, Japan.
SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, Vols. I-III, NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2/R6), May 2000. Available from Radiation Shielding Information Center at Oak Ridge National Laboratory as CCC-545.
C. V. Parks, "Application of SCALE to Analysis of Spent Fuel Casks," Vol. 2, pp. 385-393 in Proceedings of the IAEA International Symposium on the Packaging and Transport of Radioactive Materials (PATRAM '86), IAEA-SM-286/62P, June 16-20, 1986, Davos, Switzerland.
SCALE Validation
G. Radulescu, I. C. Gauld, and G. Ilas, SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions, ORNL/TM-2010/44, Oak Ridge National Laboratory, Oak Ridge, Tenn., March 2010.
G. Ilas, I. C. Gauld, F. C. Difilippo, and M. B. Emmett, Analysis of Experimental
Data for High Burnup PWR Spent Fuel Isotopic Validation—Calvert Cliffs, Takahama, and Three Mile
Island Reactors, NUREG/CR-6968 (ORNL/TM-2008/071), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., February 2010.
G. Ilas, I. C. Gauld, and B. D. Murphy, Analysis of Experimental
Data for High Burnup PWR Spent Fuel Isotopic Validation—ARIANE and REBUS Programs (UO2 Fuel), NUREG/CR-6969 (ORNL/TM-2008/072), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., February 2010.
B. D. Murphy and I. C. Gauld, Spent Fuel Decay Heat Measurements Performed at the Swedish Central Interim Storage Facility, NUREG/CR-6971 (ORNL/TM-2008/016), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., February 2010.
I. C. Gauld, G. Ilas, B. D. Murphy, and C. F. Weber, Validation of SCALE 5 Decay Heat Predictions for
LWR Spent Nuclear Fuel, NUREG/CR-6972 (ORNL/TM-2008/015), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., February 2010.
I. C. Gauld and B. D. Murphy, Technical Basis for a Proposed Expansion of Regulatory Guide 3.54—Decay Heat Generation in an Independent Spent Fuel Storage Installation, NUREG/CR-6999 (ORNL/TM-2007/231), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., February 2010.
M. D. DeHart and S. M. Bowman, “Improved radiochemical assay analyses using TRITON depletion sequences in SCALE,” Proc. of International Atomic Energy Agency Technical Meeting “Advances in Applications of Burnup Credit to Enhance Spent Fuel Transportation, Storage, Reprocessing and Disposition,” August 29–September 2, 2005, London, United Kingdom. IAEA-TECDOC-CD-1547, Session 2, pg. 99–108 (May 2007).
Bradley T. Rearden, “Criticality Code Validation Exercises with TSUNAMI,” p. 84–88 in Proceedings of the 8th International Conference on Nuclear Criticality Safety (ICNC 2007), Vol. I, May 28–June 1, 2007, St. Petersburg, Russia.
Tyler Sumner and Sedat Goluoglu, “Verification of KENO V.A and KENO-VI Using Analytical Benchmarks,” p. 361–363 in Proceedings of the 8th International Conference on Nuclear Criticality Safety (ICNC 2007), Vol. I, May 28–June 1, 2007, St. Petersburg, Russia.
I. C. Gauld, “Validation of ORIGEN-S Decay Heat Predictions for LOCA Analysis,” C183.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
G. Ilas and I. C. Gauld, “Analysis of Decay Heat Measurements for BWR Fuel Assemblies,” Trans. Am. Nucl. Soc., 94, 385–387 (2006).
D. F. Hollenbach and P. B. Fox, "Benchmark Analysis of the SCALE 5 Versions of the KENO-VI and CENTRM Codes," dfz-2.pdf in Proc. of 2005 NCSD Topical Meeting – Integrating Criticality Safety into the Resurgence of Nuclear Power, September 19–22, 2005, Knoxville, Tennessee, on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2005).
P. B. Fox and D. F. Hollenbach, KENO-VI Validation, ORNL/TM-2004/60, Oak Ridge National Laboratory, Oak Ridge, Tenn., May 2005.
D. F. Hollenbach and P. B. Fox, CENTRM Validation, ORNL/TM-2004/66, Oak Ridge National Laboratory, Oak Ridge, Tenn., May 2005.
S. Goluoglu and C. M. Hopper, Assessment of Degree of Applicability of Benchmarks for Gadolinium Using KENO V.a and the 238-Group SCALE Cross-section Library, ORNL/TM-2003/106, UT-Battelle, LLC, Oak Ridge National Laboratory, December 2003.
S. Goluoglu, K. R. Elam, B. T. Rearden, B. L. Broadhead, C. M. Hopper, and C. V. Parks, “Validation of the 10B Capture Reaction in Nuclear Fuel Casks with Sensitivity Analysis,” Trans. Am. Nucl. Soc. 89, 134–135 (2003).
K. R. Elam and B. T. Rearden, “Use of Sensitivity and Uncertainty Analysis to Select Benchmark Experiments for the Validation of Computer Codes and Data,” Nucl. Sci. and Eng. 145, 196 - 212 (2003).
I. C. Gauld, MOX Cross-Section Libraries for ORIGEN-ARP, ORNL/TM-2003/2,
UT-Battelle, LLC, Oak Ridge National Laboratory, July 2003.
S. Goluoglu, C. M. Hopper, and B. T. Rearden, "Extended Interpretation of Sensitivity Data for Benchmark Areas of Applicability," Trans. Am. Nucl. Soc. 88, 77-79 (2003).
Z. Zhong, T. Downar, and M. DeHart, "Benchmarking the U.S. NRC Neutronic Codes
NEWT and PARCS with the VENUS-2 MOX Critical Experiments ," 097.pdf in
Proc. of Nuclear Mathematical and Computational Sciences: A Century
in Review, A Century Anew, April 6-11, 2003, Gatlinburg, Tennessee.
C. E. Sanders and I. C. Gauld, Isotopic Analysis of High-Burnup PWR Spent Fuel Samples From the Takahama-3 Reactor, NUREG/CR-6798, (ORNL/TM-2001/259), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, January 2003.
M. B. Emmett, Calculational Benchmark Problems for VVER-1000 Mixed Oxide Fuel Cycle, ORNL/TM-1999/207, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, March 2000.S. M. Bowman, M. B. Emmett, and W. D. Jordan, "SCALE Criticality Safety Verification and Validation Package," Trans. Am. Nucl. Soc., 78, 160-162 (1998).
O. W. Hermann and M. D. DeHart, Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel, ORNL/TM-13315, Lockheed Martin Energy Research Corp.,Oak Ridge National Laboratory, September 1998.
B. L. Broadhead, M. B. Emmett, and J. S. Tang, Guide to Verification and Validation of the SCALE-4 Radiation Shielding Software, NUREG/CR-6484 (ORNL/TM-13277), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, December 1996
M. B. Emmett and W. C. Jordan, Guide to Verification and Validation of the SCALE-4 Criticality Safety Software, NUREG/CR-6483 (ORNL/TM-12834), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, November 1996.
S. M. Bowman, R. Q. Wright, M. D. DeHart, C. V. Parks, and L. M. Petrie, "Recent Validation Experience with Multigroup Cross-Section Libraries and SCALE," ICNC '95 Fifth International Conference on Nuclear Criticality Safety, September 17-21, 1995, Albuquerque, New Mexico.
Douglas E. Peplow and Lester M. Petrie, Jr., "Criticality Accident Alarm System Modeling with SCALE," 200725.pdf in Proc. of International Conference on Mathematics, Computational Methods, and Reactor Physics (M&C 2009), May 3–7, 2009, Saratoga Springs, New York.
S. M. Bowman, KENO-VI Primer: A Primer for Criticality Calculations with SCALE/KENO-VI Using GeeWiz, ORNL/TM-2008/069, UT-Battelle, LLC, Oak Ridge National Laboratory, September 2008.
S. Goluoglu, "Performance of the New Continuous Energy Capability in KENO V.a," Trans. Am. Nucl. Soc. 99, 407-408 (2008).
Sedat Goluoglu, Michael E. Dunn, Lester M. Petrie, and Tyler S. Sunmer, "Development and validation of the new continuous-energy capability in the criticality safety code in KENO," FP138.pdf in Proc. of PHYSOR’08 International Conference on the Physics of Reactors “Nuclear Power: A Sustainable Resource,” September 14–19, 2008, Interlaken, Switzerland.
S. Goluoglu, S. M. Bowman, and M. E. Dunn, "KENO Monte Carlo Code Capabilities," The American Nuclear Society and the European Nuclear Society 2007 International Conference on Making the Renaissance Real, November 11–15, 2007, Washington, D.C. Trans. Am. Nucl. Soc. 97, 592–594 (2007).
Stephen M. Bowman, Mark D. DeHart, Michael E. Dunn, Sedat Goluoglu, James E. Horwedel, Lester M. Petrie, Jr., Bradley T. Rearden, and Mark L. Williams, “New Criticality Safety Analysis Capabilities in SCALE 5.1,” p. 403–407 in Proceedings of the 8th International Conference on Nuclear Criticality Safety (ICNC 2007), Vol. I, May 28–June 1, 2007, St. Petersburg, Russia.
S. Goluoglu, M. E. Dunn, N. M. Greene, L. M. Petrie, and D. F. Hollenbach, “Generation and Testing of the Continuous-Energy Cross-Section Library for Use with Continuous-Energy Versions of KENO,” p. 364–366 in Proceedings of the 8th International Conference on Nuclear Criticality Safety (ICNC 2007), Vol. I, May 28–June 1, 2007, St. Petersburg, Russia.
Tyler Sumner and Sedat Goluoglu, “Verification of KENO V.A and KENO-VI Using Analytical Benchmarks,” p. 361–363 in Proceedings of the 8th International Conference on Nuclear Criticality Safety (ICNC 2007), Vol. I, May 28–June 1, 2007, St. Petersburg, Russia.
Aaron M. Fleckenstein and Bradley T. Rearden, "Multigroup Cross Section and Cross Section Covariance Data Visualization with Javapeno," Trans. Am. Nucl. Soc., 95, 292-295 (2006).
D. F. Hollenbach, L. M. Petrie, and S. M. Bowman, “Advances in the KENO-VI Geometry Package,” abstract submitted to ANS 2006 Winter Meeting “Ensuring the Future in Times of Change: Nonproliferation and Security,” November 12–16, 2006, Albuquerque, New Mexico. Trans. Am. Nucl. Soc., 95, 296–298 (2006).
J. E. Horwedel, S. M. Bowman, and D. F. Hollenbach, “New Capabilities to Calculate Volumes of SCALE/KENO-VI Geometry Models,” Trans. Am. Nucl. Soc., 95, 287–289 (2006).
R. D. Busch and S. M. Bowman, KENO V.a Primer: A Primer for Criticality Calculations with SCALE/KENO V.a Using GeeWiz, ORNL/TM-2005/135, UT-Battelle, LLC, Oak Ridge National Laboratory, December 2005. [Export Controlled Document published on CD-ROM and distributed by RSICC.]
D. F. Hollenbach and M. E. Dunn, "Status and Preliminary Testing of Continuous-Energy KENO V.a and KENO-VI Results," dfz-1.pdf in Proc. of 2005 NCSD Topical Meeting – Integrating Criticality Safety into the Resurgence of Nuclear Power, September 19–22, 2005, Knoxville, Tennessee, on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2005).
D. F. Hollenbach and P. B. Fox, "Benchmark Analysis of the SCALE 5 Versions of the KENO-VI and CENTRM Codes," dfz-2.pdf in Proc. of 2005 NCSD Topical Meeting – Integrating Criticality Safety into the Resurgence of Nuclear Power, September 19–22, 2005, Knoxville, Tennessee, on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2005).
S. M. Bowman, B. T. Rearden, and J. E. Horwedel, “ Complete User Visualization Interface for KENO,” stevebowman-1.pdf in Proc. of 2005 NCSD Topical Meeting – Integrating Criticality Safety into the Resurgence of Nuclear Power, September 19–22, 2005, Knoxville, Tennessee, on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2005).
D. F. Hollenbach and M. E. Dunn, "Continuous-Energy Version of the SCALE Control Modules for Use with Continuous-Energy KENO V.a and KENO-VI," Trans. Am. Nucl. Soc. 92, 749-750 (2005).
M. E. Dunn , P. B. Fox, N. M. Greene, L. M. Petrie, “ENDF/B-VI Library Generation and Testing for the SCALE Code System,” Trans. Am. Nucl. Soc., 92, 758–759 (June 2005).
P. B. Fox and D. F. Hollenbach, KENO-VI Validation, ORNL/TM-2004/60, Oak Ridge National Laboratory, Oak Ridge, Tenn., May 2005.
S. M. Bowman, B. T. Rearden, and J. E. Horwedel, “Integrated Interactive Visualization for KENO,” usr-stevebowman-2-paper.pdf in Proc. of The Monte Carlo 2005 Topical Meeting, The Monte Carlo Method: Versatility Unbounded in a Dynamic Computing World, April 17–21, 2005, Chattanooga, Tennessee, on CD-ROM, American Nuclear Society, La Grange Park, Illinois (2005).
M. E. Dunn, N. M. Greene, D. F. Hollenbach, and L. M. Petrie, "Monte Carlo Methods Development for a Continuous-Energy Version of KENO," usr-dunnme-1-paper.pdf in Proc. of The Monte Carlo 2005 Topical Meeting, The Monte Carlo Method: Versatility Unbounded in a Dynamic Computing World, April 17–21, 2005, Chattanooga, Tennessee; on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2005).
B. T. Rearden, A. M. Fleckenstein, and K. C. Whitney, “Development of HTML Formatted Output for SCALE,” Trans. Am. Nucl. Soc. 91, 605-607 (2004).
S. Goluoglu and C. M. Hopper, “Impact of Benchmarks on Potential MOX Throughput,” Trans. Am. Nucl. Soc., 91, 585-587 (2004).
M. E. Dunn, D. F. Hollenbach, N. M. Greene, and L. M. Petrie, “Point KENO V.a: A Continuous-Energy Monte Carlo Code for Transport Applications,” in Proc. of PHYSOR-2004 The Physics of Fuel Cycles and Advanced Nuclear Systems Global Developments, April 25-29, 2004, Chicago, Illinois.
C. F. Weber and C. M. Hopper, "Modeling Actinide Solution Densities with the Pitzer Method," Trans. Am. Nucl. Soc. 89, 123-124 (2003).
M. E. Dunn, N. M. Greene, and L. M. Petrie, "Continuous-energy Version of KENO V.a for Criticality Safety Applications," pp. 21-28 in Proc. of the 7th International Conference on Nuclear Criticality Safety (ICNC2003)," October 20-24, 2003, Tokai-mura, Japan.
S. M. Bowman, and J. E. Horwedel, "New SCALE Graphical Interface for Criticality Safety," pp. 118-124 in Proc. of the 7th International Conference on Nuclear Criticality Safety (ICNC2003), October 20-24, 2003, Tokai-mura, Japan.
S. M. Bowman, D. F. Hollenbach, M. D. DeHart, B. T. Rearden, I. C. Gauld, and S. Goluoglu, "SCALE 5: Powerful New Criticality Safety Analysis Tools," pp. 26-32 in Proc. of The 7th International Conference on Nuclear Criticality Safety (ICNC2003), October 20-24, 2003, Tokai-mura, Japan.
R. D. Busch and S. M. Bowman, "The KENO V.a Primer," Trans. Am. Nucl. Soc. 88, 80-81 (2003).
S. Goluoglu, C. M. Hopper, and B. T. Rearden, "Extended Interpretation of Sensitivity Data for Benchmark Areas of Applicability," Trans. Am. Nucl. Soc. 88, 77-79 (2003).
Y. Karni, D. Regev, E. Greenspan, S. Goluoglu, L. M. Petrie, and C. M. Hopper, "On the SMORES Capability for Minimum Critical Mass Determination," Trans. Am. Nucl. Soc. 88, 82-83 (2003).
R. D. Busch and S. M. Bowman, KENO V.a Primer: A Primer for Criticality
Calculations with SCALE/KENO V.a Using CSPAN for Input, ORNL/TM-2002/155,
UT-Battelle, LLC, Oak Ridge National Laboratory, January 2003.
D. F. Hollenbach and P. B. Fox, Neutronics Benchmarks of Mixed-Oxide Fuels using the SCALE/CENTRM Sequence, ORNL/TM-1999/299, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, February 2000.
S. M. Bowman, M. B. Emmett, and W. D. Jordan, "SCALE Criticality Safety Verification and Validation Package," Trans. Am. Nucl. Soc., 78, 160-162 (1998).
M. B. Emmett and W. C. Jordan, Guide to Verification and Validation of the SCALE-4 Criticality Safety Software, NUREG/CR-6483 (ORNL/TM-12834), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, November 1996.
SCALE Depletion/Source Terms/Decay Heat
Mark D. DeHart, Ian C. Gauld, and Kenya Suyama, “Three-dimensional depletion analysis of the axial end of a Takahama fuel rod,” FP243.pdf in Proc. of PHYSOR’08 International Conference on the Physics of Reactors “Nuclear Power: A Sustainable Resource,” September 14–19, 2008, Interlaken, Switzerland.
Germina Ilas, Brian D. Murphy, and Ian C. Gauld, “Overview of ORIGEN-ARP and its Applications to VVER RBMK,” The American Nuclear Society and the European Nuclear Society 2007 International Conference on Making the Renaissance Real, November 11–15, 2007, Washington, D.C. Trans. Am. Nucl. Soc. 97, 601–603 (2007).
Germina Ilas, Brian Murphy, and Ian Gauld, "Overview of ORIGEN-ARP and its Applications to VVER and RBMK," presented at The American Nuclear Society and the European Nuclear Society 2007 International Conference on Making the Renaissance Real, November 11–15, 2007, Washington, D.C.
M. D. DeHart and S. M. Bowman, “Improved radiochemical assay analyses using TRITON depletion sequences in SCALE,” Proc. of International Atomic Energy Agency Technical Meeting “Advances in Applications of Burnup Credit to Enhance Spent Fuel Transportation, Storage, Reprocessing and Disposition,” August 29–September 2, 2005, London, United Kingdom. IAEA-TECDOC-CD-1547, Session 2, pg. 99–108 (May 2007).
Germina Ilas, Brian D. Murphy, and Ian C. Gauld, “VVER and RBMK Cross Section Libraries for ORIGEN-ARP,” p. 413–417 in Proceedings of the 8th International Conference on Nuclear Criticality Safety (ICNC 2007), Vol. II, May 28–June 1, 2007, St. Petersburg, Russia.
B. D. Murphy, ORIGEN-ARP Cross-Section Libraries for the RBMK-1000 System, ORNL/TM-2006/139, UT-Battelle, LLC, Oak Ridge National Laboratory, November 2006.
G. Ilas, I. C. Gauld, and V. J. Jodoin, “LWR Cross Section Libraries for ORIGEN-ARP in SCALE 5.1,” Trans Am. Nucl. Soc., 95, 706 (2006).
C. F. Weber and B. L. Broadhead, “Inverse Depletion/Decay Analysis Using the SCALE Code System,” Trans. Am. Nucl. Soc., 95, 248–249 (2006).
Ian Gauld, “Validation of ORIGEN-S Decay Heat Predictions for LOCA Analysis,” C183.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
I. C. Gauld, S. M. Bowman and B. D. Murphy, "Application of ORIGEN to Spent Fuel Safeguards and Non-Proliferation," in Proc.of INMM 47th Annual Meeting, July 16–20, 2006, Nashville, Tennessee.
Germina Ilas and Ian C. Gauld, “Analysis of Decay Heat Measurements for BWR Fuel Assemblies,” Trans. Am. Nucl. Soc., 94, 385–387 (2006).
J. J. Klingensmith and I. C. Gauld, "ORIGEN-S Gamma Decay Spectra Characterization and Benchmarking," Trans. Am. Nucl. Soc., 92, 33-34 (2006).
S. M. Bowman, M. D. DeHart, and L. M. Petrie, “ Integrated KENO Monte Carlo Transport for 3-D Depletion with SCALE,” usr-stevebowman-1-paper.pdf in Proc. of The Monte Carlo 2005 Topical Meeting, The Monte Carlo Method: Versatility Unbounded in a Dynamic Computing World, April 17–21, 2005, Chattanooga, Tennessee, on CD-ROM, American Nuclear Society, La Grange Park, Illinois (2005).
I. C. Gauld, "Automated Depletion Analysis of PBMR Fuel Using SCALE," Trans. Am. Nucl. Soc., 91, 673-674 (2004).
B. D. Murphy and I. C. Gauld, "Spent-Fuel Decay Heat Investigations for BWR Assemblies Using Both One- and Two-Dimensional Model Simulations," Trans. Am. Nucl. Soc., 91, 670-672 (2004).
M. D. DeHart and L. M. Petrie, "Integrated KENO V.a Monte Carlo Transport for Multidimensional Depletion Within SCALE," Trans. Am. Nucl. Soc., 91, 667-669 (2004).
I. C. Gauld and B. D. Murphy,
Updates to the ORIGEN-S Data Libraries Using ENDF/B-VI, FENDL-2.0,
M. D. DeHart and L. M. Petrie, “A Radioisotpe Depletion Method Using Monte Carlo Transport with Variance Reduction and Error Propagation,” in Proc. of PHYSOR 2004 - The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments, April 25–29, 2004, Chicago, Illinois, on CD-ROM, American Nuclear Society, La Grange Park, Illinois (2004).
B. D. Murphy, ORIGEN-ARP Cross-Section Libraries for Magnox, Advanced Gas-Cooled, and VVER Reactor Designs, ORNL/TM-2003/263, UT-Battelle, LLC, Oak Ridge National Laboratory, February 2004.
I. C. Gauld, P. Chare, and R. C. Clark, "Development of ORIGEN-ARP Methods and Data for LEU and MOX Safeguards Applications," 0135.pdf in Proc of the 44th Annual Institute of Nuclear Materials Management (INMM) Annual Meeting, July 13–17, 2003, Phoenix, Arizona.
I. C. Gauld, MOX Cross-Section Libraries for ORIGEN-ARP, ORNL/TM-2003/2, UT-Battelle, LLC, Oak Ridge National Laboratory, July 2003.
C. E. Sanders and I. C. Gauld, Isotopic Analysis of High-Burnup PWR Spent Fuel Samples From the Takahama-3 Reactor, NUREG/CR-6798, (ORNL/TM-2001/259), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, January 2003.
I. C. Gauld, E. F. Shores, and R. T. Perry, "New Neutron Source Algorithms in the ORIGEN-S Code," in Proc. of the ANS 12th Biennial RPSD Topical Meeting, April 14-18, 2002, Santa Fe, New Mexico.
O. W. Hermann, Benchmark of SCALE (SAS2H) Isotopic Predictions of Depletion Analyses for San Onofre PWR MOX Fuel, ORNL/TM-1999/326, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, February 2000.
O. W. Hermann, P. R. Daniel, and J. C. Ryman, ORIGEN-S Decay Data Library and Half-Life Uncertainties, ORNL/TM-13624, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, September 1998.
O. W. Hermann and M. D. DeHart, Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel, ORNL/TM-13315, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, September 1998.M. D. DeHart and O. W. Hermann, An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel, ORNL/TM-13317, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, September 1996.
O. W. Hermann, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analyses, ORNL/TM-12667, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, March 1995.C. V. Parks, O. W. Hermann, and B. L. Broadhead, "The SCALE Analysis Sequence for LWR Fuel Depletion," pp. 10.2 3.1 - 10.2 3.14 in Proc. of ANS/ENS International Topical Meeting, April 28-May 1, 1991, Pittsburgh, Pennsylvania.
B. Duchemin and C. Nordborg, DECAY HEAT CALCULATION An International Nuclear Code Comparison, Nuclear Energy Agency Report NEACRP-319"L," NEANDC-275"U," 1989.
SCALE Shielding
D. Wiarda, M. E. Dunn, D. E. Peplow, T. M. Miller, and H. Akkurt, Development and Testing of ENDF/B-VI.8 and ENDF/B-VII.0 Coupled Neutron-Gamma Libraries for SCALE 6, NUREG/CR-6990 (ORNL/TM-2008/047), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., February 2009.
John C. Wagner, Edward D. Blakeman, and Douglas E. Peplow, "Forward-Weighted CADIS Method for Variance Reduction of Monte Carlo Calculations of Distributions and Multiple Localized Quantities," 203271.pdf in Proc. of International Conference on Mathematics, Computational Methods, and Reactor Physics (M&C 2009), May 3–7, 2009, Saratoga Springs, New York.
Douglas E. Peplow and Lester M. Petrie, Jr., "Criticality Accident Alarm System Modeling with SCALE," 200725.pdf in Proc. of International Conference on Mathematics, Computational Methods, and Reactor Physics (M&C 2009), May 3–7, 2009, Saratoga Springs, New York.
Douglas E. Peplow, Edward D. Blakeman, and John C. Wagner, “Advanced Variance Reduction Strategies for Optimizing Mesh Tallies in MAVRIC,” The American Nuclear Society and the European Nuclear Society 2007 International Conference on Making the Renaissance Real, November 11–15, 2007, Washington, D.C. Trans. Am. Nucl. Soc. 97, 595–597 (2007).
J. C. Wagner, E. D. Blakeman, and D. E. Peplow, “Forward-Weighted CADIS Method for Global Variance Reduction,” The American Nuclear Society and the European Nuclear Society 2007 International Conference on Making the Renaissance Real, November 11–15, 2007, Washington, D.C. Trans. Am. Nucl. Soc. 97, 630–633 (2007).
M. L. Williams and S. Goluoglu, “Sensitivity Analysis for Coupled Neutron-Gamma Calculations,” American Nuclear Society 2007 Annual Meeting “It’s All About the People: The Future of Nuclear,” June 24-28, 2007, Boston, Massachusetts. Trans. Am. Nucl. Soc., 96, p. 533–534 (2007).
D. E. Peplow, S. M. Bowman, J. E. Horwedel, and J. C. Wagner, “Monaco/MAVRIC: Computational Resources for Radiation Protection and Shielding in SCALE,” Trans. Am. Nucl. Soc., 95, 669–671 (2006).
D. Ilas and J. C. Wagner, "--A SCALE Sequence for 3-D Discrete Ordinates Calculations," Trans. Am. Nucl. Soc., 95, 667-668 (2006).
D. E. Peplow and J. C. Wagner, "Automated Variance Reduction for SCALE Shielding Calculations," in Proc. of ANS 14th Biennial Topical Meeting of the Radiation Protection and Shielding Division, pp. 556-558, Carlsbad, New Mexico, April 2-6, 2006.
M. B. Emmett and J. C. Wagner, “MONACO : A New 3-D Monte Carlo Shielding Code for SCALE,” Trans. Am. Nucl. Soc., 91, 701-703 (2004).
M. B. Emmett, S. M. Bowman, and B. L. Broadhead, "SCALE Radiation Shielding Verification and Validation Package," Vol. 1, pp. 91-97 in Proceedings of the 1998 ANS Radiation Protection and Shielding Division Topical Conference on Technologies for the New Century, April 19-23, 1998, Nashville, Tennessee.
B. L. Broadhead, M. B. Emmett, and J. S. Tang, Guide to Verification and Validation of the SCALE-4 Radiation Shielding Software, NUREG/CR-6484 (ORNL/TM-13277), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, December 1996.
S. M. Bowman, "SCALE-PC Shielding Analysis Sequences," ANS 1996 Radiation Protection and Shielding Division Topical Meeting, April 21-25, 1996, Cape Cod, Massachusetts.
Matthew A. Jessee, Mark L. Williams, and Mark D. DeHart, “Development of Generalized Perturbation Theory Capability within the SCALE Code Package,” 201120.pdf in Proc. of International Conference on Mathematics, Computational Methods, and Reactor Physics (M&C 2009), May 3–7, 2009, Saratoga Springs, New York.
Brad Rearden and
Don Mueller, "TSUNAMI Methods for Validation,
Gap Analysis and
Experiment Design Verification," presented at the Validation
for Nuclear Systems
Analysis Workshop,
July 24, 2008, Idaho Falls, Idaho.
M. L. Williams and B. T. Rearden, “SCALE-6 Sensitivity/Uncertainty Methods and Covariance Data,” Nuclear Data Sheets, Vol. 109, Issue 12, 2796–2800 (December 2008), Special Issue on Workshop on Neutron Cross Section Covariances, June 24-28, 2008, Port Jefferson, USA.
Bradley T. Rearden, “TSUNAMI Sensitivity and Uncertainty Analysis Capabilities in SCALE 5.1,” The American Nuclear Society and the European Nuclear Society 2007 International Conference on Making the Renaissance Real, November 11–15, 2007, Washington, D.C. Trans. Am. Nucl. Soc. 97, 604–605 (2007) .
M. L. Williams, B. L. Broadhead, M. E. Dunn, and B. T. Rearden, “Approximate Techniques for Representing Nuclear Data Uncertainties,”p. 744-752 in Proc. of the Eighth International Topical Meeting on Nuclear Applications and Utilization of Accelerators (ACCAPP '07) , July 30–August 2, 2007, Pocatello, Idaho.
Bradley T. Rearden and Mark L. Williams, “Overview of the SCALE TSUNAMI Sensitivity and Uncertainty Analysis Tools,” American Nuclear Society 2007 Annual Meeting “It’s All About the People: The Future of Nuclear,” June 24-28, 2007, Boston, Massachusetts. Trans. Am. Nucl. Soc , 96, p.535–537 (2007).
M. L. Williams and S. Goluoglu, “Sensitivity Analysis for Coupled Neutron-Gamma Calculations,” American Nuclear Society 2007 Annual Meeting “It’s All About the People: The Future of Nuclear,” June 24-28, 2007, Boston, Massachusetts. Trans. Am. Nucl. Soc , 96, p.533–534 (2007).
D. E. Mueller and J. C. Wagner, “Application of sensitivity/uncertainty methods to burnup credit validation,” Proc. of International Atomic Energy Agency Technical Meeting “Advances in Applications of Burnup Credit to Enhance Spent Fuel Transportation, Storage, Reprocessing and Disposition,” August 29–September 2, 2005, London, United Kingdom. IAEA-TECDOC-CD-1547, Session 2.2, pg. 183–195 (May 2007).
Bradley T. Rearden and Mark L. Williams, “Eigenvalue Contribution Estimator for Sensitivity Calculations with TSUNAMI-3D,” p. 408–412 in Proceedings of the 8th International Conference on Nuclear Criticality Safety (ICNC 2007), Vol. I, May 28–June 1, 2007, St. Petersburg, Russia.
B. T. Rearden, “Criticality Code Validation Exercises with TSUNAMI,” p. 84–88 in Proceedings of the 8th International Conference on Nuclear Criticality Safety (ICNC 2007), Vol. I, May 28–June 1, 2007, St. Petersburg, Russia.
B. T. Rearden and J. E. Horwedel, "Automatic Differentiation with Code Coupling and Applications to SCALE Modules," mcsna01317full.pdf in Proc. of Joint International Topical Meeting on Mathematics & Computation and Supercomputing in Nuclear Applications (M&C + SNA2007), April 15–19, 2007, Monterey, California, on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2007).
M. L. Williams, "Sensitivity and Uncertainty Analysis for Eigenvalue-Difference Responses," Nucl. Sci. Eng., Vol. 155, No. 1, pp. 18-36 (January 2007).
J. E. Horwedel, "Automatic Differentiation to Couple SCALE Modules Using GRESS 90--Part I: Methodology," Trans. Am. Nucl. Soc., 95, 699-701 (2006).
B. T. Rearden and J. E. Horwedel, "Automatic Differentiation to Couple SCALE Modules Using GRESS 90--Part II: Application," Trans. Am. Nucl. Soc., 95, 702-705 (2006).
B. T. Rearden, "A Criticality Code Validation Exercise for a LEU Lattice," Trans. Am. Nucl. Soc., 95, 381-386 (2006).
B. L. Broadhead, C. M. Hopper, and J. J. Wagschal, "Sensitivity of Adjusted Responses to Parameter and Response Uncertainties," B033.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
M. L. Williams, J. C. Gehin, and K. T. Clarno, “
Sensitivity Analysis of Reactivity Responses Using One-Dimensional Discrete Ordinates and Three-Dimensional Monte Carlo Methods,” C135.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
B. T. Rearden, M. L. Williams, and J. E. Horwedel , "Advances in the TSUNAMI Sensitivity and Uncertainty Analysis Codes Beyond SCALE 5," Trans. Am. Nucl. Soc. 92, 760-762 (2005).
D. E. Mueller and B. T. Rearden, "Sensitivity Coefficient Generation for a Burnup Credit Cask Model using TSUNAMI-3D," reardenb-2.pdf in Proc. of 2005 NCSD Topical Meeting, September 19-22, 2005, Knoxville, Tennessee.
B. T. Rearden, W. J. Anderson, and G. A. Harms, "Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel," Nucl. Technol. 151, 133-158 ( August 2005).
I. C. Gauld and D. E. Mueller, Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations, ORNL/TM-2005/48, Oak Ridge National Laboratory, Oak Ridge, Tenn., August 2005.
D. E. Mueller and G. A. Harms, "Using the SCALE 5 TSUNAMI-3D Sequence in Critical Experiment Design," Trans. Am. Nucl. Soc. 93, 263-266 (2005).
B. T. Rearden, “Improvements in KENO V.a to Support TSUNAMI-3D Sensitivity Calculations,” usr-rearden-1-paper.pdf in Proc. of The Monte Carlo 2005 Topical Meeting, The Monte Carlo Method: Versatility Unbounded in a Dynamic Computing World, April 17–21, 2005, Chattanooga, Tennessee, on CD-ROM, American Nuclear Society, La Grange Park, Illinois (2005).
B. T. Rearden, C. M. Hopper, and K. R. Elam, "TSUNAMI Analysis of the Applicability of Proposed Experiments to Reactor-Grade and Weapons-Grade Mixed Oxide Systems," presented at the International Symposium NUCEF2005, Tokai, Japan, February 9-10, 2005.
B. L. Broadhead, B. T. Rearden, C. M. Hopper, J. J. Wagschal, and C. V. Parks, “Sensitivity- and Uncertainty-Based Criticality Safety Validation Techniques,” Nucl. Sci. Eng . 146, 340–366 (2004).
B. T. Rearden, "Perturbation Theory Eignvalue Sensitivity Analysis with Monte Carlo Techniques," Nucl. Sci. Eng., 146, 367-382 (2004).
B. T. Rearden, W. J. Anderson, and G. A. Harms, "Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel," to be published in the Nuclear Technology journal.
S. Goluoglu, Sensitivity Analysis Applied to the Validation of the 10B Capture Reaction in Nuclear Fuel Casks, ORNL/TM-2004/48, UT-Battelle, LLC, Oak Ridge National Laboratory, February 2004.
S. Goluoglu and C. M. Hopper, Assessment of Degree of Applicability of Benchmarks for Gadolinium Using KENO V.a and the 238-Group SCALE Cross-section Library, ORNL/TM-2003/106, UT-Battelle, LLC, Oak Ridge National Laboratory, December 2003.
B. T. Rearden, C. M. Hopper, K. R. Elam, S. Goluoglu, and C. V. Parks, Applications of the TSUNAMI Sensitivity and Uncertainty Analysis Methodology,” pp. 61-66 in Proc. of The 7th International Conference on Nuclear Criticality Safety (ICNC2003)”, October 20-24, 2003, Tokai-mura, Japan.
S. Goluoglu, K. R. Elam, B. T. Rearden, B. L. Broadhead, C. M. Hopper, and C. V. Parks, “Validation of the 10B Capture Reaction in Nuclear Fuel Casks with Sensitivity Analysis,” Trans. Am. Nucl. Soc. 89, 134–135 (2003).
K. R. Elam and B. T. Rearden, “Use of Sensitivity and Uncertainty Analysis to Select Benchmark Experiments for the Validation of Computer Codes and Data,” Nucl. Sci. and Eng. 145, 196 - 212 (2003).
S. Goluoglu, C. M. Hopper, and B. T. Rearden, "Extended
Interpretation of Sensitivity Data for Benchmark Areas of Applicability," Trans. Am. Nucl. Soc. 88, 77-79 (2003).
SCALE Reactor Physics
Mark L. Williams and Germina Ilas, “ENDF/B-VII Nuclear Data Libraries for SCALE 6,” usr-xmw-1-paper.pdf, Advances in Nuclear Fuel Management IV (ANFM 2009),Hilton Head Island, South Carolina, USA, April 12–15, 2009, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2009).
Mark D. DeHart and Megan L. Pritchard, “Validation of SCALE and the TRITON Depletion Sequences for Gas Reactor Analysis,” Trans. Am. Nuc. Soc., 99, 683–685 (November 2008).
Mark D. DeHart, Ian C. Gauld, and Kenya Suyama, “Three-dimensional depletion analysis of the axial end of a Takahama fuel rod,” FP243.pdf in Proc. of PHYSOR’08 International Conference on the Physics of Reactors “Nuclear Power: A Sustainable Resource,” September 14–19, 2008, Interlaken, Switzerland.
Mark D. DeHart, “High-Fidelity Lattice Physics Capabilities of the SCALE Code System Using TRITON,” The American Nuclear Society and the European Nuclear Society 2007 International Conference on Making the Renaissance Real, November 11–15, 2007, Washington, D.C. Trans. Am. Nucl. Soc. 97, 598–600 (2007).
Germina Ilas, Brian D. Murphy, and Ian C. Gauld, “VVER and RBMK Cross Section Libraries for ORIGEN-ARP,” p. 413–417 in Proceedings of the 8th International Conference on Nuclear Criticality Safety (ICNC 2007), Vol. II, May 28–June 1, 2007, St. Petersburg, Russia.
M. D. DeHart, S. Goluoglu, and M. E. Dunn, “KENO Continuous Energy Calculations for a Suite of Computational Benchmarks for the Doppler Reactivity Defect,” mcsna01087full.pdf in Proc. of Joint International Topical Meeting on Mathematics & Computation and Supercomputing in Nuclear Applications (M&C + SNA2007), April 15–19, 2007, Monterey, California, on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2007).
M. D. DeHart, I. C. Gauld, and M. L. Williams, “High-Fidelity Lattice Physics Capabilities of the SCALE Code System Using TRITON,” mcsna05008full.1.pdf in Proc. of Joint International Topical Meeting on Mathematics & Computation and Supercomputing in Nuclear Applications (M&C + SNA2007), April 15–19, 2007, Monterey, California, on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2007).
M. D. DeHart, “Simplification of Multi-Group Cross-Section Processing for Large Depletion Calculations in TRITON,” mcsna01086full.pdf in Proc. of Joint International Topical Meeting on Mathematics & Computation and Supercomputing in Nuclear Applications (M&C + SNA2007), April 15–19, 2007, Monterey, California, on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2007).
Mark L. Williams, "Sensitivity and Uncertainty Analysis for Eigenvalue-Difference Responses," Nucl. Sci. Eng., Vol. 155, No. 1, pp. 18-36 (January 2007).
B. D. Murphy, ORIGEN-ARP Cross-Section Libraries for the RBMK-1000 System, ORNL/TM-2006/139, UT-Battelle, LLC, Oak Ridge National Laboratory, November 2006.
G. Ilas, I. C. Gauld, and V. J. Jodoin, “LWR Cross Section Libraries for ORIGEN-ARP in SCALE 5.1,” Trans Am. Nucl. Soc., 95, 706 (2006).
S. Goluoglu, "Analysis of a Computational Benchmark for a High-Temperature Reactor Using SCALE," C022.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
Kevin T. Clarno and Jess C. Gehin, "Physics Analysis of the LS-VHTR: Salt Coolant and Fuel Block Design," D045.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
S. M. Bowman and D. F. Gill, “Validation of Standardized Computer Analyses for Licensing Evaluation/TRITON Two-Dimensional and Three-Dimensional Models for Light Water Reactor Fuel,” B153.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
Mark D. DeHart, “Advancements in Generalized-Geometry Discrete Ordinates Transport for Lattice Physics Calculations,” A154.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
Mark D. DeHart, “Lattice Physics Capabilities of the SCALE Code System Using TRITON,” A121.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
Mark L. Williams, Jess C. Gehin, and Kevin T. Clarno, “Sensitivity Analysis of Reactivity Responses Using One-Dimensional Discrete Ordinates and Three-Dimensional Monte Carlo Methods,” C135.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
Zhaopeng Zhong, Thomas J. Downar, Mark D. DeHart, and Mark L. Williams, “Parallelization of the SCALE Continuous-Energy Resonance Processing Module GEMINEWTRN,” Trans. Am. Nucl. Soc., 94, 473–475 (2006).
Zhaopeng Zhong and Mark D. DeHart, “Coarse–Mesh Finite–Difference Acceleration in the NEWT Generalized-Geometry Lattice Physics Package,” Trans. Am. Nucl. Soc., 94, 471–472 (2006).
Mark D. DeHart , Jacopo Sacchari and David Diamond, “Assessment of TRITON and PARCS for Full-Core MOX Fuel Calculations,” Trans. Am. Nucl. Soc., 92, 763-766 (June 2005).
Z. Zhong, T. Downar, and M. DeHart, Implementation of a Two-Level Coarse-Mesh Finite-Difference Accelerator in the NEWT Transport Code, ORNL/TM-2004/162, UT-Battelle, LLC, Oak Ridge National Laboratory, June 2005.
F. C. Difilippo, "Analysis of VHTRs with the SCALE System," Trans. Am. Nucl. Soc., 91, 763-765 (2004).
M. D. DeHart and L. M. Petrie, "Integrated KENO V.a Monte Carlo Transport for Multidimensional Depletion Within SCALE," Trans. Am. Nucl. Soc., 91, 667-669 (2004).
M. D. DeHart, Z. Zhong, and T. J. Downar, "TRITON: An Advanced Lattice Code for MOX Fuel Calculations," in Proc. of American Nuclear Society, Advances in Nuclear Fuel Management III, October 5-8, 2003, Hilton Head Island, South Carolina.
Z. Zhong, T. Downar, and M. DeHart, "Benchmarking the U.S. NRC Neutronics Codes NEWT and PARCS with the VENUS-2 MOX Critical Experiments," in Proc. of Nuclear Mathematical and Computational Sciences: A Century in Review, A New Century Anew, April 6-11, 2003, Gatlinburg, Tennessee.
B. B. Bevard, J. C. Wagner, C. V. Parks, and M. Aissa, Review of Information for Spent Nuclear Fuel Burnup Confirmation, NUREG/CR-6998, prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., December 2009.
Jeremy A. Roberts and Donald E. Mueller, "Designing Critical Experiments in Support of Full Burnup Credit," Trans. Am. Nucl. Soc. 99, 391-393 (2008).
J. C. Wagner, Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask, NUREG/CR-6955 (ORNL/TM-2004/52), U. S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, January 2008.
G. Radulescu, D. E. Mueller, and J. C. Wagner, Sensitivity and Uncertainty Analysis of Commercial Reactor Criticals for Burnup Credit, NUREG/CR-6951 (ORNL/TM-2006/87), U. S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, July 2007.
M. D. DeHart and S. M. Bowman, “Improved radiochemical assay analyses using TRITON depletion sequences in SCALE,” Proc. of International Atomic Energy Agency Technical Meeting “Advances in Applications of Burnup Credit to Enhance Spent Fuel Transportation, Storage, Reprocessing and Disposition,” August 29–September 2, 2005, London, United Kingdom. IAEA-TECDOC-CD-1547, Session 2, pg. 99–108 (May 2007).
D. E. Mueller and J. C. Wagner, “Application of sensitivity/uncertainty methods to burnup credit validation,” Proc. of International Atomic Energy Agency Technical Meeting “Advances in Applications of Burnup Credit to Enhance Spent Fuel Transportation, Storage, Reprocessing and Disposition,” August 29–September 2, 2005, London, United Kingdom. IAEA-TECDOC-CD-1547, Session 2.2, pg. 183–195 (May 2007).
C. V. Parks, J. C. Wagner, D. E. Mueller, and I. C. Gauld, “Full Burnup Credit in Transport and Storage Casks: Benefits and Implementation,” RadWaste Solutions, Vol. 14, No. 2, March/April 2007, pp. 32–41.
Susan N. Williams and Donald E. Mueller, "Survey of Operating Parameters for Use in Burnup Credit Calculations," Trans. Am. Nucl. Soc., 95, 269-273 (2006).
D. E. Mueller and B. T. Rearden, "Sensitivity Coefficient Generation for a Burnup Credit Cask Model using TSUNAMI-3D," reardenb-2.pdf in Proc. of 2005 NCSD Topical Meeting, September 19-22, 2005, Knoxville, Tennessee.
I. C. Gauld and D. E. Mueller, Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations, ORNL/TM-2005/48, Oak Ridge National Laboratory, Oak Ridge, Tenn., August 2005.
C. V. Parks and J. C. Wagner, “Status of Burnup Credit for Transport of SNF in the United States,” Paper #154 in Proc. of The 14th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2004), September 20–24, 2004, Berlin, Germany (2005).
C. V. Parks and J. C. Wagner, "Current Status and Potential Benefits of Burnup for Spent Fuel Transportation," pp. 233-240 in Proc. of the 14th Pacific Basin Nuclear Conference, March 21-25, 2004, Honolulu, Hawaii.
J. C. Wagner, "Impact of Soluble Boron Modeling for PWR Burnup Credit Criticality Safety Analyses," Trans. Am. Nucl. Soc., 89, 120-122 (2003).
J. C. Wagner, "Evaluation of Burnup Credit for Accommodating PWR Spent Nuclear Fuel in High-Capacity Cask Designs," pp. 684-689 in Proc. of the 7th International Conference on Nuclear Criticality Safety (ICNC2003), October 20-24, 2003, Tokai-mura, Japan.
I. C. Gauld, Strategies for Application of Isotopic Uncertainties in Burnup Credit, NUREG/CR-6811 (ORNL/TM-2001/257), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, June 2003.
C. V. Parks and C. J. Withee, "Recommendations for PWR Storage and Transportation Casks That Use Burnup Credit," pp.500-507 in Proc. of the 10th International High-Level Radioactive Waste Management (IHLRWM) Conference, "Progress Through Cooperation," March 30-April 2, 2003, Las Vegas, Nevada.
J. C. Wagner and C. E. Sanders, Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs, NUREG/CR-6800 (ORNL/TM-2002/6), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, March 2003.
J. C. Wagner, M. D. DeHart, and C. V. Parks, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses, NUREG/CR-6801 (ORNL/TM-2001/273), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, March 2003.
J. C. Wagner and C. V. Parks, Recommendations on the Credit for Cooling Time in PWR Burnup Credit Analyses, NUREG/CR-6781 (ORNL/TM-2001/272), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, January 2003.
C. E. Sanders and I. C. Gauld, Isotopic Analysis of High-Burnup PWR Spent Fuel Samples From the Takahama-3 Reactor, NUREG/CR-6798,
(ORNL/TM-2001/259), U.S. Nuclear Regulatory Commission, Oak Ridge
National Laboratory, January 2003.
J. C. Wagner and C. V. Parks, Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit, NUREG/CR-6761 (ORNL/TM-2000/373), U. S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, March 2002.
C. E. Sanders and J. C. Wagner, Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit, NUREG/CR-6760 (ORNL/TM-2000-321), U. S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, March 2002.
I. C. Gauld and S. M. Bowman, STARBUCS: A Prototypic SCALE Control Module for Automated Criticality Safety Analyses Using Burnup Credit, NUREG/CR-6748 (ORNL/TM-2001/33), U. S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, October 2001.
J. C. Wagner, Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit, NUREG/CR-6747 (ORNL/TM-2000/306), U. S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, October 2001.
I. C. Gauld, SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor
Critical Configurations, ORNL/TM-1999/247,
Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, March
2000.
J. C. Wagner and M. D. DeHart, Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations, ORNL/TM-1999/246, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, March 2000.
C. V. Parks, M. D. DeHart, and J. C. Wagner, Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel, NUREG/CR-6665 (ORNL/TM-1999/303), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, February 2000.
M. D. DeHart, Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term Disposal Criticality Safety, ORNL/TM-1999/99, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, August 1999.
H. R. Dyer and C. V. Parks, Recommendations for Preparing the Criticality Safety Evaluation of Transportation Packages, NUREG/CR-5661 (ORNL/TM-11936), U. S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, April 1997.
S. M. Bowman and T. Suto, SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 5-North Anna Unit 1 Cycle 5 , ORNL/TM-12294/V5, Lockheed Martin Energy Research Systems, Inc., Oak Ridge National Laboratory, October 1996.
M. D. DeHart, M. C. Brady, and C. V. Parks, OECD/NEA Burnup Credit Calculational
Criticality Benchmark Phase I-B Results, NEA/NSC/DOC(96)-06 (ORNL-6901),
June 1996.
Cross-Section Processing Methods
S. Goluoglu and M. L. Williams, "Modeling Doubly Heterogeneous Systems in SCALE," Trans. Am. Nucl. Soc. 93, 963-965 (2005).
D. F. Hollenbach and P. B. Fox, CENTRM Validation, ORNL/TM-2004/66, Oak Ridge National Laboratory, Oak Ridge, Tenn., May 2005.
M. E. Dunn, N. M. Greene, and L. M. Petrie, "Status of ORNL Cross-Section Processing and Library Generation Capabilities," presented at the International Workshop on Nuclear Data Needs for Generation IV Systems, April 5-7, 2005, Antwerp, Belgium
L. C. Leal, M. E. Dunn, K. H. Guber, H. Derrien, and R. O. Sayer, "Cross-Section Measurements and Evaluations Effort at ORNL," presented at the International Workshop on Nuclear Data Needs for Generation IV Systems, April 5-7, 2005, Antwerp, Belgium
.
M. L. Williams, S. Goluoglu, and L. M. Petrie, "Recent Enhancements to the SCALE 5 Resonance Self-Shielding Methodology," Trans. Am. Nucl. Soc. 92, 751-753 (2005).
Z. Zhong, T. J. Downar, M. D. DeHart, and M. L. Williams, “Continuous-Energy Multidimensional SN Transport for Problem-Dependent Resonance Self-Shielding Calculations,” Trans. Am. Nucl. Soc., 92, 754-757 (June 2005).
M. E. Dunn , P. B. Fox, N. M. Greene, L. M. Petrie, “ENDF/B-VI Library Generation and Testing for the SCALE Code System,” Trans. Am. Nucl. Soc., 92, 758–759 (June 2005).
M. E. Dunn and L. C. Leal, "Calculating Probability Tables for the Unresolved-Resonance
Region Using Monte Carlo Methods," in Proc. of International
Conference on the New Frontiers of Nuclear Technology: Reactor Physics,
Safety and High Performance Computing (PHYSOR 2002), October 7-10, 2002,
Seoul, Korea (October 2002). Also published in Nucl. Sci. Eng. (2003).
General Criticality
D. E. Mueller, B. T. Rearden, and D. F. Hollenbach, Application of the SCALE TSUNAMI Tools for the Validation of Criticality Safety Calculations Involving 233U, ORNL/TM-2008/196, Oak Ridge National Laboratory, Oak Ridge, Tenn., January 2009.
Donald E. Mueller, "Evaluation of the HTC Critical Experiment Data for Spent Nuclear Fuel," Trans. Am. Nucl. Soc. 98, 219-222 (2008).
Kirill Raskach and Calvin M. Hopper, “Statistical Analysis of PST Types of Experiments Relative to Examining ‘Safety Applications,” p. 64–68 in Proceedings of the 8th International Conference on Nuclear Criticality Safety (ICNC 2007), Vol. II, May 28–June 1, 2007, St. Petersburg, Russia.
S. Goluoglu and C. M. Hopper, "Application of Validation Methodologies for a Generic Validation Problem," sedat-2.pdf in Proc. of 2005 NCSD Topical Meeting – Integrating Criticality Safety into the Resurgence of Nuclear Power, September 19–22, 2005, Knoxville, Tennessee, on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2005).
B. T. Rearden, W. J. Anderson, and G. A. Harms, "Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel," Nucl. Technol. 151, 133-158 ( August 2005).
J. J. Wagschal and C. M. Hopper, "Determination of Consistent Benchmarks Used for Nuclear Criticality Safety Analysis Applications," Trans. Am. Nucl. Soc. 93, 257-259 (2005).
J. T. Thomas, R. M. Westfall, and C. M. Hopper, “History of the Oak Ridge Critical Experiments Program,” Trans. Am. Nucl. Soc. 92, 470-471 (2005).
C. M. Hopper, “Guide for Nuclear Criticality Safety in the Storage of Fissile Materials,” Trans. Am. Nucl. Soc. 91, 642-643 (2004).
C. V. Parks and C. M. Hopper, “Technical Basis for Proposed Fissile Exemption Criteria for Transport Packages,” presented at the 14th International Symposium on the Packaging and Transportation of Radioactive Materials, September 20-24, 2004, Berlin, Germany.
K. R. Elam, Criticality Safety Study of UF6 and UO2F2 in 8-in. -Diameter Piping,ORNL/TM-2003/239, UT-Battelle, LLC, Oak Ridge National Laboratory, October 2003.
K. R. Elam, J. C. Wagner, and C. V. Parks, Effects of Fuel Failure on Criticality Safety and Radiation Dose of Spent Fuel Casks, NUREG/CR-6835 (ORNL/TM-2002/255), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, September 2003.
B. L. Broadhead, C. M. Hopper, R. L. Childs, and C. V. Parks, Sensitivity
and Uncertainty Analyses Applied to Criticality Safety Validation, Volume 1:
Methods Development, NUREG/CR-6655, Vol. 1
(ORNL/TM-13692/V1), U.S. Nuclear Regulatory Commission, Oak Ridge
National Laboratory, November 1999.
D. F. Hollenbach, L. M. Petrie, and H. L. Dodds, "Vectorization Methods Development for a New Version of the KENO V.a Criticality Safety Code," Nucl. Sci. Eng. 116, 147-164 (1994).
General Depletion/Source
Terms/Decay Heat
Germina Ilas and Ian C. Gauld, “Analysis of Decay Heat Measurements for BWR Fuel Assemblies,” Trans. Am. Nucl. Soc., 94, 385–387 (2006).
B. D. Murphy, Calculating Fork Detector Response from Spent-Fuel Inventories Using Monte-Carlo Techniques, ORNL/TM-2004/310, UT-Battelle, LLC, Oak Ridge National Laboratory, May 2005.
General Shielding
S. W. Mosher, T. M. Evans, T. M. Miller, and J. C. Wagner, "Efficient Transport Simulations of Difficult Detection Problems Using ADVANTG," accepted for publication in Proc. IEEE Nuclear Science Symposium, Orlando, FL, October 25-31, 2009.
D. Wiarda, M. E. Dunn, D. E. Peplow, T. M. Miller, and H. Akkurt, Development and Testing of ENDF/B-VI.8 and ENDF/B-VII.0 Coupled Neutron-Gamma Libraries for SCALE 6, NUREG/CR-6990 (ORNL/TM-2008/047), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., February 2009.
B. L. Broadhead and J. C. Wagner, “Effective Biasing Schemes for Duct Streaming Problems,” presented at the 10th International Conference on Radiation Shielding, Radiation Protection Dosimetry, (ICRS-10), May 9-14, 2004, Funchal, Portugal.
K. R. Elam, J. C. Wagner, and C. V. Parks, Effects of Fuel Failure on Criticality Safety and Radiation Dose of Spent Fuel Casks, NUREG/CR-6835 (ORNL/TM-2002/255), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, September 2003.
Herschel P. Smith and J. C. Wagner, "A Case Study in Manual
and Automated Monte Carlo Variance Reduction with a Deep Penetration Reactor
Shielding Problem," in Proc. of Nuclear Mathematical
and Computational Sciences: A Century in Review, A Century Anew, April
6-11, 2003, Gatlinburg, Tennessee.
B. L. Broadhead, Recommendations for Shielding Evaluations for Transport and Storage Packages, NUREG/CR-6802 (ORNL/TM-2002/31), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, May 2003.
A. Haghighat and J. C. Wagner, "Monte Carlo Variance Reduction with Deterministic Importance Functions," Progress in Nuclear Energy 42(1), 25-53, January 2003.
B. L. Broadhead, "Shielding Analyses: The Rabbit vs the Turtle?" Vol. 1,
p. 322 in Proc. of Radiation Protection & Shielding 1996 Topical
Meeting, Advances and Applications in Radiation Protection and Shielding
(1996).
General Reactor Physics
Tom Greifenkamp, Kevin Clarno, and Jess Gehin, "Effect of Fuel Temperature on Eigenvalue Calculations," in Proc. of the 2008 American Nuclear Society National Student Conference “Expanding the Nuclear Family,”
February 28 – March 1, 2008, Texas A&M University,
College Station, Texas.
Kevin Clarno, "NEWTRNX: Massively Parallel Neutron Transport for Multi-Physics Nuclear Reactor Simulations," presented February 26, 2008, at the University of Texas Graduate Seminar, Austin, Texas.
Kevin Clarno, Valmor de Almeida, and Mark Williams, "High Performance Computing Neutronics for Coupled-Physics Simulations," presented to the Nuclear Science and Technology Division Advisory Committee, Oak Ridge National Laboratory, November 7, 2007.
Kevin Clarno, Valmor de Almeida, Ed d’Azevedo, Cassiano de Oliveira, and Steven Hamilton, “GNES-R: Global Nuclear Energy Simulator for Reactors Task 1: High-Fidelity Neutron Transport,” D025.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
R. J. Ellis, J. C. Gehin, and R. T. Primm, III, “Cross Section Generation and Physics Modeling in a Feasibility Study of the Conversion of the High Flux Isotope Reactor Core to use Low-Enriched Uranium Fuel,” B021.pdf in Proc. of PHYSOR–2006, American Nuclear Society Topical Meeting on Reactor Physics: Advances in Nuclear Analysis and Simulation, September 10–14, 2006, Vancouver, British Columbia, Canada.
C. W. Forsberg, D. T. Ingersoll, P. F. Peterson, H. Zhao, J. E. Cahalan, T. Taiwo, J. A. Enneking, R. A. Kochendarfer, and P. E. MacDonald, Refueling Options and Considerations for Liquid-Salt-Cooled Very High-Temperature Reactors, ORNL/TM-2006/92, UT-Battelle, LLC, Oak Ridge National Laboratory, June 2006.
B. T. Rearden, W. J. Anderson, and G. A. Harms, "Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel," Nucl. Technol. 151, 133-158 ( August 2005).
SCALE GUIs & Visualization
Aaron M. Fleckenstein and Bradley T. Rearden, "Extensible SCALE Intelligent Text Editor—ExSITE," Trans Am. Nucl. Soc., 98, 223-226 (2008).
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