Publications

SCALE Publications Related to:

SCALE General

SCALE Primers

Burnup Credit

SCALE Validation

SCALE Criticality

General Criticality

SCALE Shielding

General Shielding

Cross-Section Processing Methods

SCALE Depletion/Source Terms/Decay Heat

SCALE Sensitivity/Uncertainty

General Reactor Physics

SCALE Reactor Physics

SCALE GUIs & Visualization


SCALE General

SCALE: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge, Tennessee, June 2011. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-785.

S. M. Bowman, "SCALE 6: Comprehensive Nuclear Safety Analysis Code System," Nucl. Technol. 174(2), 126-148, May 2011.

I. C. Gauld, G. Radulescu, G. Ilas, B. D. Murphy, M. L. Williams, and D. Wiarda, "Isotopic Depletion and Decay Methods and Analysis Capabilities in SCALE," Nucl. Technol. 174(2), 169-195, May 2011.

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.

SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, ORNL/TM-2005/39, Version 6, Vols. I–III, Oak Ridge National Laboratory, Oak Ridge, Tennessee, January 2009. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-750.

M. L. Williams, S. M. Bowman, and C. V. Parks, "Plans for Future SCALE Development Beyond Version 6.0,"Trans. Am. Nucl. Soc. 97, 606-607 (2007).

S. M. Bowman, "Overview of the SCALE Code System,"Trans. Am. Nucl. Soc. 97, 589-591 (2007).

S. M. Bowman, B. T. Rearden, and J. 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, St. Petersburg, Russia, May 28-June 1, 2007.

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).

S. M. Bowman and J. 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), Tokai-mura, Japan, October 20-24, 2003.

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.

SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, Vols. I-III, NUREG/CR-0200, Rev. 5 (ORNL/NUREG/CSD-2/R5), March 1997. Available from Radiation Shielding Information Center at Oak Ridge National Laboratory as CCC-545.

SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, Vols. I-III, NUREG/CR-0200, Rev. 4 (ORNL/NUREG/CSD-2/R4), April 1995. Available from Radiation Shielding Information Center at Oak Ridge National Laboratory as CCC-545.

S. M. Bowman, C. V. Parks, and S. K. Martin, "Maintaining SCALE as a Reliable Computational System for Criticality Safety Analysis," Trans. Am. Nucl. Soc. 72, 198 (1995).

S. M. Bowman, OFFSCALE: A PC Input Processor for the SCALE Code System, The CSASIN Processor for the Criticality Sequences, NUREG/CR-6182, Vol. 1 (ORNL/TM-12663/V1), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., November 1994.

S. M. Bowman, OFFSCALE: A PC Input Processor for the SCALE Code System, The ORIGNATE Processor for ORIGEN-S, NUREG/CR-6182, Vol. 2 (ORNL/TM-12263/V2), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., November 1994.

C. V. Parks, "Overview of the SCALE-4 Code System," presented at Seminar on SCALE-4 and Related Modular Systems for the Evaluation of Nuclear Fuel Facilities and Package Design Featuring Criticality, Shielding, and Heat Transfer Capabilities, OECD NEA Data Bank, Saclay, France, September 17-19, 1991.

C. V. Parks, L. M. Petrie, J. Manneschmidt, and J. L. Bartley, "SCALE-4: An Updated Version of the Modular Code System for Performing Standardized Computer Analysis For Licensing Evaluation," in Proceedings of ANS/ENS International Topical Meeting on Advances in Mathematics, Computations, and Reactor Physics, Pittsburgh, Pennsylvania, April 28-May 1, 1991.

C. V. Parks, "SCALE-4: An Improved Computational System for Spent-Fuel Cask Analysis," Vol. 3, pp. 1545-1552 in Proceedings of the 9th International Symposium on the Packaging and Transportation of Radioactive Materials, Washington, D.C., June 11-16, 1989, CONF-890631.

C. V. Parks, Summary Description of the SCALE Modular Code System, NUREG/CR-5033 (ORNL/CSD/TM-252), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., December 1987.

C. V. Parks and R. M. Westfall, "History and Overview of the SCALE System," Workshop on the SCALE-3 Modular System, Saclay, France, June 24-27, 1986. Newsletter of the NEA Data Bank No. 33, pp. 5-19, October 1986.

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, Davos, Switzerland, June 16-20, 1986.

SCALE Validation

G. Ilas, D. Ilas, R. P. Kelly, and E. E. Sunny, Validation of SCALE for High Temperature Gas-Cooled Reactor Analysis, NUREG/CR-7107 (ORNL/TM-2011/161), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., July 2012.

B. T. Rearden and W. J. Marshall, "Examination of Validation Outlier Cases Using the Sensitivity and Uncertainty Analysis Tools of SCALE 6.1," Trans. Am. Nucl. Soc. 106, 461-464 (2012).

W. J. Marshall and B. T. Rearden, "Criticality Safety Validation of SCALE 6.1 with ENDF/B-VII.0 Libraries," Trans. Am. Nucl. Soc. 106, 456-460 (2012).

J. M. Scaglione, D. E. Mueller, J. C. Wagner, and W. J. Marshall, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (keff) Predictions, NUREG/CR-7109 (ORNL/TM-2011/514), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., April 2012.

G. Radulescu, I. C. Gauld, G. Ilas, and J. C. Wagner, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions, NUREG/CR-7108 (ORNL/TM-2011/509), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., April 2012.

D. Ilas, "SCALE Code Validation for Prismatic High-Temperature Gas-Cooled Reactors," PHYSOR 2012 – Advances in Reactor Physics – Linking Research, Industry, and Education, Knoxville, Tennessee, USA, April 15-20, 2012, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2012)

G. Ilas, I. C. Gauld, and G. Radulescu, "Validation of New Depletion Capabilities and ENDF/B-VII Data Libraries in SCALE," Annals of Nucl. Ener. 46, 43-55, August 2012.

W. J. Marshall and B. T. Rearden, Criticality Safety Validation of SCALE 6.1, ORNL/TM-2011/450, Oak Ridge National Laboratory, Oak Ridge, Tenn., November 2011.

U. Mertyurek, M. W. Francis, and I. C. Gauld, SCALE 5 Analysis of BWR Spent Nuclear Fuel Isotopic Compositions for Safety Studies, ORNL/TM-2010/286, Oak Ridge National Laboratory, Oak Ridge, Tenn., December 2010.

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," London, United Kingdom, August 29-September 2, 2005; IAEA-TECDOC-CD-1547, Session 2, pg. 99-108 (May 2007).

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, St. Petersburg, Russia, May 28-June 1, 2007.

T. Sumner and S. 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, St. Petersburg, Russia, May 28-June 1, 2007.

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, Vancouver, British Columbia, Canada, September 10-14, 2006.

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, Knoxville, Tennessee, September 19-22, 2005; 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, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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, Gatlinburg, Tennessee, April 6-11, 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), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., January 2003.

M. B. Emmett, Calculational Benchmark Problems for VVER-1000 Mixed Oxide Fuel Cycle, ORNL/TM-1999/207, Oak Ridge National Laboratory, Oak Ridge, Tenn., March 2000.

O. W. Hermann, Benchmark of SCALE (SAS2H) Isotopic Predictions of Depletion Analyses for San Onofre PWR MOX Fuel, ORNL/TM-1999/326, Oak Ridge National Laboratory, Oak Ridge, Tenn., February 2000.

D. F. Hollenbach and P. B. Fox, Neutronics Benchmarks of Mixed-Oxide Fuels using the SCALE/CENTRM Sequence, ORNL/TM-1999/299, Oak Ridge National Laboratory, Oak Ridge, Tenn. 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).

O. W. Hermann and M. D. DeHart, Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel, ORNL/TM-13315, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., 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), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., 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, Albuquerque, New Mexico, September 17-21, 1995.

B. L. Broadhead, J. S. Tang, R. L. Childs, C. V. Parks, and H. Taniuchi, Evaluation of Shielding Analysis Methods in Spent Fuel Cask Environments, EPRI TR-104329, Electric Power Research Institute, May 1995.

S. M. Bowman, M. D. DeHart, and C. V. Parks, "Validation of SCALE-4 for Burnup Credit Applications," Nucl. Tech. 110, 53 (1995).

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, Oak Ridge National Laboratory, Oak Ridge, Tenn., March 1995.

M. D. DeHart, O. W. Hermann, and C. V. Parks, "Validation of SCALE Depletion Methods for PWR Spent Fuel Isotopic Characterization," Trans. Am. Nucl. Soc. 73, 361-63 (1995).

O. W. Hermann, J. P. Renier, and C. V. Parks, Technical Support for a Proposed Decay Heat Guide Using SAS2/ORIGEN-S Data, NUREG/CR-5625 (ORNL-6698), prepared for the U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, Oak Ridge, Tenn., September 1994.

M. D. DeHart and S. M. Bowman, Validation of the SCALE Broad Structure 44-Group ENDF/B-V Cross-Section Library for Use in Criticality Safety Analyses, NUREG/CR-6102 (ORNL/TM-12460), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., September 1994.

S. M. Bowman and H. Taniuchi, "Burnup Credit Validation of SCALE-4 Using Mixed-Oxide Critical Experiments," Trans. Am. Nucl. Soc. 68(A), 241 (1993).

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.

S. M. Bowman, R. Q. Wright, H. Taniuchi, and M. D. DeHart, "Validation of SCALE-4 Criticality Sequences Using ENDF/B-V Data," 1993 Topical Meeting on Physics and Methods in Criticality Safety, Nashville, Tennessee, September 19-23, 1993.

M. B. Emmett, "Validation and Verification of the ORNL Monte Carlo Codes for Nuclear Safety Analyses," Proc. ANS 1993 Winter Meeting, San Francisco, California, November 14-19, 1993.

O. W. Hermann, M. C. Brady, and C. V. Parks, "Validation of Spent Fuel Isotopics Predicted by the SCALE-4 Depletion Sequence," Trans. Am. Nucl. Soc. 64, 147-149 (1991).

S. M. Bowman and C. V. Parks, "Validation of SCALE-4 for LWR Fuel in Transportation and Storage Cask Conditions," ANS Meeting, Washington, D.C., November 11-15, 1990.

W. C. Jordan, N. F. Landers, and L. M. Petrie, Validation of KENO V.a Comparison with Critical Experiments, ORNL/CSD/TM-238, Oak Ridge National Laboratory, Oak Ridge, Tenn., 1986.

R. M. Westfall and J. R. Knight, "SCALE System Cross Section Validation with Shipping-Cask Critical Experiments," Trans. Am. Nucl. Soc. 33, 368 (1979).

SCALE Criticality

A. M. Ibrahim, D. E. Peplow, K. B. Bekar, C. Celik, D. Ilas, J. M. Scaglione, and J. C. Wagner, "Acceleration of Monte Carlo Fission Source Convergence Using Deterministic Starting Sources for Used Nuclear Fuel Cask-Specific Criticality Analysis," Proceedings of the Nuclear Criticality Safety Division 2013, Wilmington, NC, September 29-October 3, 2013.

W. J. Marshall and B. T. Rearden, Criticality Safety Validation of SCALE 6.1, ORNL/TM-2011/450, Oak Ridge National Laboratory, Oak Ridge, Tenn., November 2011.

S. Goluoglu, L. M. Petrie, Jr., M. E. Dunn, D. F. Hollenbach, and B. T. Rearden, "Monte Carlo Criticality Methods and Analysis Capabilities in SCALE," Nucl. Technol. 174(2), 214-235, May 2011.

D. E. Peplow and L. 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), Saratoga Springs, New York, May 3-7, 2009.

S. M. Bowman, KENO-VI Primer: A Primer for Criticality Calculations with SCALE/KENO-VI Using GeeWiz, ORNL/TM-2008/069, Oak Ridge National Laboratory, Oak Ridge, Tenn., September 2008.

S. Goluoglu, "Performance of the New Continuous Energy Capability in KENO V.a," Trans. Am. Nucl. Soc. 99, 407-408 (2008).

S. Goluoglu, M. E. Dunn, L. M. Petrie, and T. 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," Interlaken, Switzerland, September 14-19, 2008.

S. Goluoglu, S. M. Bowman, and M. E. Dunn, "KENO Monte Carlo Code Capabilities," Trans. Am. Nucl. Soc. 97, 592-594 (2007).

S. M. Bowman, M. D. DeHart, M. E. Dunn, S. Goluoglu, J. E. Horwedel, L. M. Petrie, Jr., B. T. Rearden, and M. 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, St. Petersburg, Russia, May 28-June 1, 2007.

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, St. Petersburg, Russia, May 28-June 1, 2007.

T. Sumner and S. 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, St. Petersburg, Russia, May 28-June 1, 2007.

A. M. Fleckenstein and B. 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," 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, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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, Knoxville, Tennessee, September 19-22, 2005; 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, Knoxville, Tennessee, September 19-22, 2005; 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, Knoxville, Tennessee, September 19-22, 2005; 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, Chattanooga, Tennessee, April 17-21, 2005; 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, Chattanooga, Tennessee, April 17-21, 2005; 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, Chicago, Illinois, April 25-29, 2004.

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)," Tokai-mura, Japan, October 20-24, 2003.

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), Tokai-mura, Japan, October 20-24, 2003.

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), Tokai-mura, Japan, October 20-24, 2003.

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, Oak Ridge National Laboratory, Oak Ridge, Tenn., January 2003.

D. F. Hollenbach and P. B. Fox, Neutronics Benchmarks of Mixed-Oxide Fuels using the SCALE/CENTRM Sequence, ORNL/TM-1999/299, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., November 1996.

D. F. Hollenbach, L. M. Petrie, and N. F. Landers, KENO-VI: A General Quadratic Version of the KENO Program, ORNL/TM-13011, Oak Ridge National Laboratory, Oak Ridge, Tenn., 1996.

D. F. Hollenbach, S. M. Bowman, L. M. Petrie, and C. V. Parks, "New Enhancements to SCALE for Criticality Safety Analysis," presented at ICNC'95, Fifth International Conference on Nuclear Criticality Safety, Albuquerque, New Mexico, September 17-21, 1995.

C. V. Parks, R. Q. Wright, and W. C. Jordan, Adequacy of the 123-Group Cross-Section Library for Criticality Analyses of Water-Moderated Uranium Systems, NUREG/CR-6328 (ORNL/TM-12970), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., 1995.

S. M. Bowman, OFFSCALE: A PC Input Processor for the SCALE Code System, The CSASIN Processor for the Criticality Sequences, NUREG/CR-6182, Vol. 1 (ORNL/TM-12663/V1), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., November 1994.

M. D. DeHart and S. M. Bowman, Validation of the SCALE Broad Structure 44-Group ENDF/B-V Cross-Section Library for Use in Criticality Safety Analyses, NUREG/CR-6102 (ORNL/TM-12460), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., September 1994.

C. V. Parks, "The SCALE Criticality Safety Analysis Sequences: Status and Future Directions," Trans. Am. Nucl. Soc. 68(A), 237-238 (1993).

D. F. Hollenbach, L. M. Petrie, and N. F. Landers, "KENO-VI: A Monte Carlo Criticality Program with Generalized Quadratic Geometry," Proc. ANS 1993 Topical Meeting on Physics and Methods in Criticality Safety, Nashville, Tennessee, September 19-23, 1993.

S. M. Bowman, "OFFSCALE: PC Input Processor for SCALE-4 Criticality Sequences," Proceedings of Seminar on SCALE-4 and Related Modular Systems for the Evaluation of Nuclear Fuel Facilities and Package Design Featuring Criticality, Shielding, and Heat Transfer Capabilities, OECD NEA Data Bank, Saclay, France, September 17-19, 1991.

L. M. Petrie, N. F. Landers, N. M. Greene, and C. V. Parks, "New SCALE-4 Features Related to Cross-Section Processing," Presented at Seminar on SCALE-4 and Related Modular Systems for the Evaluation of Nuclear Fuel Facilities and Package Design Featuring Criticality, Shielding, and Heat Transfer Capabilities, OECD NEA Data Bank, Saclay, France, September 17-19, 1991.

W. C. Jordan, C. V. Parks, and L. M. Petrie, "An Improved Dancoff Correction Factor for the SCALE Code System," Trans. Am. Nucl. Soc. 56, 324-325 (1988).

C. V. Parks, L. M. Petrie, N. F. Landers, and J. A. Bucholz, "Computational Methods for Criticality Safety Analysis within the SCALE System," Workshop on the SCALE-3 Modular System, Saclay, France, June 24-27, 1986, Newsletter of the NEA Data Bank No. 33, pp. 31-44, October 1986.

SCALE Depletion/Source Terms/Decay Heat

I. C. Gauld and M. W. Francis, "Investigation of Passive Gamma Spectroscopy to Verify Spent Nuclear Fuel Content," in Proceedings of INMM 51st Annual Meeting, Baltimore, MD, July 11-15, 2010.

M. D. DeHart, I. C. Gauld, and K. 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," Interlaken, Switzerland, September 14-19, 2008.

G. Ilas, B. D. Murphy, and I. C. Gauld, "Overview of ORIGEN-ARP and its Applications to VVER RBMK," Trans. Am. Nucl. Soc. 97, 601-603 (2007).

G. Ilas, B. Murphy, and I. 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, Washington, D.C., November 11-15, 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).

G. Ilas, B. D. Murphy, and I. 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, St. Petersburg, Russia, May 28-June 1, 2007.

B. D. Murphy, ORIGEN-ARP Cross-Section Libraries for the RBMK-1000 System, ORNL/TM-2006/139, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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).

I. 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, Vancouver, British Columbia, Canada, September 10-14, 2006.

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, Nashville, Tennessee, July 16-20, 2006.

G. Ilas and I. 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).

G. Ilas and I. 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, Oak Ridge National Laboratory, Oak Ridge, Tenn., May 2005.

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, Chattanooga, Tennessee, April 17-21, 2005, 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, and EAF-99 Data, ORNL/TM-2003/118, Oak Ridge National Laboratory, Oak Ridge, Tenn., May 2004.

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, Chicago, Illinois, April 25-29, 2004, 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, Oak Ridge National Laboratory, Oak Ridge, Tenn. 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, Phoenix, Arizona, July 13-17, 2003.

I. C. Gauld, MOX Cross-Section Libraries for ORIGEN-ARP, ORNL/TM-2003/2, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., 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, Santa Fe, New Mexico, April 14-18, 2002.

O. W. Hermann, San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses, ORNL/TM-1999/108/R1, Oak Ridge National Laboratory, Oak Ridge, Tenn., March 2000.

B. D. Murphy, J. Kravchenko, A. Lazarenko, A. Pavlovitchev, V. Sidorenko, and A. Chetverikov, Simulation of Low-Enriched Uranium (LEU) Burnup in Russian VVER Reactors with the HELIOS Code Package, ORNL/TM-1999/168, Oak Ridge National Laboratory, Oak Ridge, Tenn., March 2000.

O. W. Hermann, Benchmark of SCALE (SAS2H) Isotopic Predictions of Depletion Analyses for San Onofre PWR MOX Fuel, ORNL/TM-1999/326, Oak Ridge National Laboratory, Oak Ridge, Tenn., February 2000.

O. W. Hermann, San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses, ORNL/TM-1999/108, Oak Ridge National Laboratory, Oak Ridge, Tenn., September 1999.

M. D. DeHart, Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term Disposal Criticality Safety, ORNL/TM-1999/99, Oak Ridge National Laboratory, Oak Ridge, Tenn., August 1999.

B. D. Murphy, Prediction of the Isotopic Composition of UO2 Fuel from a BWR Analysis of the DU1 Sample from the Dodewaard Reactor, ORNL/TM-13687, Oak Ridge National Laboratory, Oak Ridge, Tenn., October 1998.

B. D. Murphy, Characteristics of Spent Fuel From Plutonium Disposition Reactors, Vol. 4: Westinghouse Pressurized-Water-Reactor Fuel Cycle Without Integral Absorber, ORNL/TM-13170/V4, Oak Ridge National Laboratory, Oak Ridge, Tenn., April 1998.

B. D. Murphy, Verification of the LWRARC Code for Light-Water-Reactor Afterheat Rate Calculations, NUREG/CR-6536 (ORNL/TM-13396), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., February 1998.

O. W. Hermann, P. R. Daniel, and J. C. Ryman, ORIGEN-S Decay Data Library and Half-Life Uncertainties, ORNL/TM-13624, Oak Ridge National Laboratory, Oak Ridge, Tenn., September 1998.

O. W. Hermann and M. D. DeHart, Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel, ORNL/TM-13315, Oak Ridge National Laboratory, Oak Ridge, Tenn., September 1998.

L. C. Leal, O. W. Hermann, S. M. Bowman, and C. V. Parks, ARP: Automatic Rapid Process for the Generation of Problem-Dependent SAS2H/ORIGEN-S Cross-Section Libraries, ORNL/TM-13584, Oak Ridge National Laboratory, Oak Ridge, Tenn., April 1998.

O. W. Hermann and C. V. Parks, "SAS2H: A Coupled One-Dimensional Depletion and Shielding Analysis Module," Vol. I, Sect. S2 of SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluations, NUREG/CR-0200, Rev. 5 (ORNL/NUREG/CSD-2R5), March 1997. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-545.

M. D. DeHart and O. W. Hermann, An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel, ORNL/TM-13317, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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, Oak Ridge National Laboratory, Oak Ridge, Tenn., March 1995.

O. W. Hermann, SAS2H Input for Computing Core Activities of 4.5, 5.0, and 5.5 Weight % 235U Fuel for Sequoyah Nuclear Plant, ORNL/M-3739, Oak Ridge National Laboratory, Oak Ridge, Tenn., August 1994.

L. C. Leal, O. W. Hermann, and C. V. Parks, "Automatic Rapid Processing SCALE/SAS2H-Produced Parameter-Dependent Cross Sections for ORIGEN-S," Trans. Am. Nucl. Soc. 70, 356-358 (1994).

S. M. Bowman, OFFSCALE: A PC Input Processor for the SCALE Code System, The ORIGNATE Processor for ORIGEN-S, NUREG/CR-6182, Vol. 2 (ORNL/TM-12263/V2), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., November 1994.

O. W. Hermann, J. P. Renier, and C. V. Parks, Technical Support for a Proposed Decay Heat Guide Using SAS2/ORIGEN-S Data, NUREG/CR-5625 (ORNL-6698), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., September 1994.

O. W. Hermann and R. Salmon, "Borosilicate Glass (,n) Sources Used with ORIGEN-Type Calculations," Proc. 1992 International High-Level Radioactive Waste Management Conference, Las Vegas, Nevada, April 12-16, 1992.

C. V. Parks, "Overview of ORIGEN2 and ORIGEN-S: Capabilities and Limitations," Vol. 1, pp. 57-63, in Proc. of Third International Conference on High Level Radioactive Waste Management, Las Vegas, Nevada, April 12-16, 1992.

S. M. Bowman, "ORIGNATE-PC Input Processor for ORIGEN-S," 1991 International High-Level Radioactive Waste Management Conference, Las Vegas, Nevada, April 12-16, 1992.

S. M. Bowman, "ORIGNATE: PC Input Processor for ORIGEN-S," Proceedings of Seminar on SCALE-4 and Related Modular Systems for the Evaluation of Nuclear Fuel Facilities and Package Design Featuring Criticality, Shielding, and Heat Transfer Capabilities, OECD NEA Data Bank, Saclay, France, September 17-19, 1991.

O. W. Hermann and C. V. Parks, "SAS2H: The SCALE-4 Analysis Sequence for LWR Fuel Depletion," Presented at Seminar on SCALE-4 and Related Modular Systems for the Evaluation of Nuclear Fuel Facilities and Package Design Featuring Criticality, Shielding, and Heat Transfer Capabilities, OECD NEA Data Bank, Saclay, France, September 17-19, 1991.

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, Pittsburgh, Pennsylvania, April 28-May 1, 1991.

B. Duchemin and C. Nordborg, DECAY HEAT CALCULATION An International Nuclear Code Comparison, Nuclear Energy Agency Report NEACRP-319"L," NEANDC-275"U," 1989.

C. V. Parks, O. W. Hermann, and J. C. Ryman, "Evaluation of Spent Fuel Isotopics, Radiation Spectra, and Decay Heat Using the SCALE Computational System," pp. 96-124 Workshop on the SCALE-3 Modular System, 24-27 June 1986, Saclay, France. Newsletter of the NEA Data Bank No. 33, October 1986.

C. V. Parks, J. S. Tang, O. W. Hermann, J. A. Bucholz, and M. B. Emmett, "Shielding Analysis Methods Available in the SCALE Computational System," pp. 45-66 Workshop on the SCALE-3 Modular System, 24-27 June 1986, Saclay, France. Newsletter of the NEA Data Bank NO. 33, October 1986.

SCALE Shielding

D. E. Peplow, "Monte Carlo Shielding Analysis Capabilities with MAVRIC," Nucl. Technol. 174(2), 289-313, May 2011.

M. L. Williams, "Resonance Self-Shielding Methodologies in SCALE 6," Nucl. Technol. 174(2), 149-168, May 2011.

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.

J. C. Wagner, E. D. Blakeman, and D. 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), Saratoga Springs, New York, May 3-7, 2009.

D. E. Peplow and L. 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), Saratoga Springs, New York, May 3-7, 2009.

D. E. Peplow, E. D. Blakeman, and J. C. Wagner, "Advanced Variance Reduction Strategies for Optimizing Mesh Tallies in MAVRIC," 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," Trans. Am. Nucl. Soc. 97, 630-633 (2007).

M. L. Williams and S. Goluoglu, "Sensitivity Analysis for Coupled Neutron-Gamma Calculations,"Trans. Am. Nucl. Soc. 96, 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, "TORTSQ--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, Nashville, Tennessee, April 19-23, 1998.

S. M. Bowman and D. L. Barnett, "SINEX: SCALE Shielding Analysis GUI for X-Windows," Vol. 2, pp. 331-337 in Proceedings 1998 ANS Radiation Protection and Shielding Division Topical Conference on Technologies for the New Century, Nashville, Tennessee, April 19-23, 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), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., December 1996.

S. M. Bowman, "SCALE-PC Shielding Analysis Sequences," ANS 1996 Radiation Protection and Shielding Division Topical Meeting, Cape Cod, Massachusetts, April 21-25, 1996.

B. L. Broadhead and C. V. Parks, "Overview of SCALE System Shielding Capabilities," Trans. Am. Nucl. Soc. 73, 360-61 (1995).

J. S. Tang and B. L. Broadhead, "Applications of SAS4 Monte Carlo Shielding Sequence to Deep-Penetration Problems," Trans. Am. Nucl. Soc. 73, 361-63 (1995).

J. S. Tang and B. L. Broadhead, "Development and Application of the Automated Monte Carlo Biasing Procedure in SAS4," Proc. Seminar on Advanced Monte Carlo Computer Programs for Radiation Transport, Centre d'Etudes, Saclay, France, April 27-29, 1993.

B. L. Broadhead, J. S. Tang, and C. V. Parks, "SAS1 and SAS4, Two New Shielding Analysis Sequences for Spent Fuel Casks," pp. 210-221 in Proc. Seminar on SCALE-4 and Related Modular Systems for the Evaluation of Nuclear Facilities and Package Design Featuring Criticality, Shielding and Transfer Capabilities, Saclay, France, September 17-19, 1991.

B. L. Broadhead and C. V. Parks, "SCALE-4 Shipping Cask Shielding Applications," pp. 33-46 in Proc. Seminar on SCALE-4 and Related Modular Systems for the Evaluation of Nuclear Facilities and Package Design Featuring Criticality, Shielding and Transfer Capabilities, Saclay, France, September 17-19, 1991.

B. L. Broadhead, "An Overview of the QADS Code," pp. 163-174 in Proc. Seminar on SCALE-4 and Related Modular Systems for the Evaluation of Nuclear Facilities and Package Design Featuring Criticality, Shielding and Transfer Capabilities, Saclay, France, September 17-19, 1991.

SCALE Sensitivity/Uncertainty

J. A. Roberts, B. T. Rearden, and P. H. Wilson, "Determination and Application of Partial Biases in Criticality Safety Validation," Nucl. Sci. Eng. 173, 43–57 (2013).

B. T. Rearden, M. L. Williams, M. A. Jessee, D. E. Mueller, and D. A. Wiarda, "Sensitivity and Uncertainty Analysis Capabilities and Data in SCALE," Nucl. Technol. 174(2), 236-288, May 2011.

M. A. Jessee, M. L. Williams, and M. 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), Saratoga Springs, New York, May 3-7, 2009.

B. T. Rearden and D. E. Mueller, "Recent Use of Covariance Data for Criticality Safety Assessment," Nuclear Data Sheets 109, 2739-2744 (2008).

B. T. Rearden and D. E. Mueller, "TSUNAMI Methods for Validation, Gap Analysis and Experiment Design Verification," presented at the Validation for Nuclear Systems Analysis Workshop, Idaho Falls, Idaho, July 24, 2008.

M. L. Williams and B. T. Rearden, "SCALE-6 Sensitivity/Uncertainty Methods and Covariance Data," Nuclear Data Sheets, 109(12), 2796-2800 (December 2008), Special Issue on Workshop on Neutron Cross Section Covariances, Port Jefferson, USA, June 24-28, 2008.

B. T. Rearden, "TSUNAMI Sensitivity and Uncertainty Analysis Capabilities in SCALE 5.1," 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), Pocatello, Idaho, July 30-August 2, 2007.

B. T. Rearden and M. L. Williams, "Overview of the SCALE TSUNAMI Sensitivity and Uncertainty Analysis Tools," Trans. Am. Nucl. Soc. 96, 535-537 (2007).

M. L. Williams and S. Goluoglu, "Sensitivity Analysis for Coupled Neutron-Gamma Calculations," Trans. Am. Nucl. Soc. 96, 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," London, United Kingdom, August 29-September 2, 2005; IAEA-TECDOC-CD-1547, Session 2.2, pg. 183-195 (May 2007).

B. T. Rearden and M. 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, St. Petersburg, Russia, May 28-June 1, 2007.

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, St. Petersburg, Russia, May 28-June 1, 2007.

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), Monterey, California, April 15-19, 2007, on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2007).

M. L. Williams, "Sensitivity and Uncertainty Analysis for Eigenvalue-Difference Responses," Nucl. Sci. Eng., 155(1), 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, Vancouver, British Columbia, Canada, September 10-14, 2006.

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, Vancouver, British Columbia, Canada, September 10-14, 2006.

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, Knoxville, Tennessee, September 19-22, 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).

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, Chattanooga, Tennessee, April 17-21, 2005, 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, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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)", Tokai-mura, Japan, October 20-24, 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).

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).

M. L. Williams, B. L.Broadhead,and C. V.Parks, "Eigenvalue Sensitivity Theory for Resonance-Shielded Cross Sections," Nucl. Sci. Eng. 138, 177 (2001).

SCALE Reactor Physics

G. Ilas, D. Ilas, R. P. Kelly, and E. E. Sunny, Validation of SCALE for High Temperature Gas-Cooled Reactor Analysis, NUREG/CR-7107 (ORNL/TM-2011/161), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., July 2012.

D. Ilas, Development of a SCALE Model for High Flux Isotope Reactor Cycle 400, ORNL/TM-2011/367, Oak Ridge National Laboratory, Oak Ridge, Tenn., February 2012.

M. D. DeHart and S. M. Bowman, "Reactor Physics Methods and Analysis Capabilities in SCALE," Nucl. Technol. 174(2), 196-213, May 2011.

M. L. Williams and G. 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).

M. D. DeHart and M. L. Pritchard, "Validation of SCALE and the TRITON Depletion Sequences for Gas Reactor Analysis," Trans. Am. Nuc. Soc., 99, 683-685 (November 2008).

M. D. DeHart, I. C. Gauld, and K. 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," Interlaken, Switzerland, September 14-19, 2008.

M. 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, Washington, D.C., November 11-15, 2007, Trans. Am. Nucl. Soc. 97, 598-600 (2007).

G. Ilas, B. D. Murphy, and I. 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, St. Petersburg, Russia, May 28-June 1, 2007.

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), Monterey, California, April 15-19, 2007, 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), Monterey, California, April 15-19, 2007, 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), Monterey, California, April 15-19, 2007, on CD-ROM, American Nuclear Society, LaGrange Park, Illinois (2007).

M. L. Williams, "Sensitivity and Uncertainty Analysis for Eigenvalue-Difference Responses," Nucl. Sci. Eng. 155(1), 18-36 (January 2007).

B. D. Murphy, ORIGEN-ARP Cross-Section Libraries for the RBMK-1000 System, ORNL/TM-2006/139, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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, Vancouver, British Columbia, Canada, September 10-14, 2006.

K. T. Clarno and J. 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, Vancouver, British Columbia, Canada, September 10-14, 2006.

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, Vancouver, British Columbia, Canada, September 10-14, 2006.

M. 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, Vancouver, British Columbia, Canada, September 10-14, 2006.

M. 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, Vancouver, British Columbia, Canada, September 10-14, 2006.

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, Vancouver, British Columbia, Canada, September 10-14, 2006.

Z. Zhong, T. J. Downar, M. D. DeHart, and M. L. Williams, "Parallelization of the SCALE Continuous-Energy Resonance Processing Module GEMINEWTRN," Trans. Am. Nucl. Soc., 94, 473-475 (2006).

Z. Zhong and M. D. DeHart, "Coarse-Mesh Finite-Difference Acceleration in the NEWT Generalized-Geometry Lattice Physics Package," Trans. Am. Nucl. Soc., 94, 471-472 (2006).

M. D. DeHart , J. Sacchari and D. 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, Oak Ridge National Laboratory, Oak Ridge, Tenn., 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, Hilton Head Island, South Carolina, October 5-8, 2003.

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, Gatlinburg, Tennessee, April 6-11, 2003.

Burnup Credit

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.

J. A. Roberts and D. 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), prepared for the U. S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., 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), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., 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," London, United Kingdom, August 29-September 2, 2005; 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," London, United Kingdom, August 29-September 2, 2005; 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, 14(2), 32-41, March/April 2007.

S. N. Williams and D. 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, Knoxville, Tennessee, September 19-22, 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.

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), Berlin, Germany, September 20-24, 2004.

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, Honolulu, Hawaii, March 21-25, 2004.

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), Tokai-mura, Japan, October 20-24, 2003.

I. C. Gauld, Strategies for Application of Isotopic Uncertainties in Burnup Credit, NUREG/CR-6811 (ORNL/TM-2001/257), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., 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," Las Vegas, Nevada, March 30-April 2, 2003.

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.

M. D. DeHart, Sensitivity and Parametric Evaluations of Significant Aspects of Burnup Credit for PWR Spent Fuel Packages, ORNL/TM-12973, Lockheed Martin Energy Research Systems, Inc., Oak Ridge National Laboratory, May 1996.

M. D. DeHart and S. M. Bowman, Analysis of Fresh Fuel Critical Experiments Appropriate for Burnup Credit Validation, ORNL/TM-12959, Lockheed Martin Energy Research Systems, Inc., Oak Ridge National Laboratory, October 1995.

B. L. Broadhead, M. D. DeHart, J. C. Ryman, J. S. Tang, and C. V. Parks, Investigation of Nuclide Importance to Functional Requirements Related to Transport and Long-Term Storage of LWR Spent Fuel, ORNL/TM-12742, Lockheed Martin Energy Systems, Inc., Oak Ridge National Laboratory, June 1995.

S. M. Bowman, M. D. DeHart, and C. V. Parks, "Validation of SCALE-4 for Burnup Credit Applications," Nucl. Tech. 110, 53 (April 1995).

M. D. DeHart, SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 1-Summary, ORNL/TM-12294/V1, Lockheed Martin Energy Research Systems, Inc., Oak Ridge National Laboratory, March 1995.

S. M. Bowman, O. W. Hermann, and M. C. Brady, SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3, ORNL/TM-12294/V2, Lockheed Martin Energy Research Systems, Inc., Oak Ridge National Laboratory, March 1995.

S. M. Bowman and O. W. Hermann, SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 3-Surry Unit 1 Cycle 2, ORNL/TM-12294/V3, Lockheed Martin Energy Research Systems, Inc., Oak Ridge National Laboratory, March 1995.

M. D. DeHart, SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5, ORNL/TM-12294/V4, Lockheed Martin Energy Research Systems, Inc., Oak Ridge National Laboratory, March 1995.

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 Systems, Inc., Oak Ridge National Laboratory, March 1995.

S. M. Bowman and H. Taniuchi, "Burnup Credit Validation of SCALE-4 Using Mixed-Oxide Critical Experiments," Trans. Am. Nucl. Soc. 68(A), 241 (1993).

S. M. Bowman, O. W. Hermann, and M. C. Brady, "Burnup Credit Validation of SCALE-4 Using Light Water Reactor Criticals," Trans. Am. Nucl. Soc. 68(A), 243, (1993).

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, Antwerp, Belgium, April 5-7, 2005.

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, Antwerp, Belgium, April 5-7, 2005.

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), Seoul, Korea, October 7-10, 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.

D. E. Mueller, "Evaluation of the HTC Critical Experiment Data for Spent Nuclear Fuel," Trans. Am. Nucl. Soc. 98, 219-222 (2008).

K. Raskach and C. 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, St. Petersburg, Russia, May 28-June 1, 2007.

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, Berlin, Germany, September 20-24, 2004.

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.

B. L. Broadhead, C. M. Hopper, and C. V. Parks, Sensitivity and Uncertainty Analyses Applied to Criticality Safety Validation, Volume 2: Illustrative Applications and Initial Guidance, NUREG/CR-6655, Vol. 2 (ORNL/TM-13692/V2), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, November 1999.

R. L. Childs, SEN1: A One-Dimensional Cross-Section Sensitivity and Uncertainty Module for Criticality Safety Analysis, NUREG/CR-5719 (ORNL/TM-13738), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, July 1999.

M. E. Dunn and P. B. Fox, Criticality Safety Scoping Study for the Transport of Weapons Grade Mixed Oxide Fuel Using the MO-1 Shipping Package, ORNL/TM-13741, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, May 1999.

C. V. Parks, C. M. Hopper, and J. L. Lichtenwalter, Assessment and Recommendations for Fissile-Material Packaging Exemptions and General Licenses Within 10 CFR Part 71, NUREG/CR-5342 (ORNL/TM-13607), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, July 1998.

C. M. Hopper and B. L. Broadhead, An Updated Nuclear Criticality Slide Rule. Volume 2: Functional Slide Rule, NUREG/CR-6504, Vol. 2 (ORNL/TM-13322/V2), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, April 1998.

C. V. Parks, W. C. Jordan, L. M. Petrie, and R. Q. Wright, "Use of Metal/Uranium Mixtures to Explore Data Uncertainties," Trans. Am. Nucl. Soc. 73, 217-18 (1995).

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 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), Funchal, Portugal, May 9-14, 2004.

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), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., September 2003.

H. 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, Gatlinburg, Tenn., April 6-11, 2003.

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).

C. V. Parks and B. L. Broadhead, "Review of Criticality Safety and Shielding Analysis Issues for Transportation Packages," PATRAM'95, Las Vegas, Nevada. December 3-8, 1995.

S. M. Bowman and B. L. Broadhead, "Dose Rates in the ABWR Upper Drywell for a Dropped Fuel Bundle Accident," Trans. Am. Nucl. Soc. 73, 364-66 (1995).

M. B. Emmett, "Status of the MORSE Multigroup Monte Carlo Radiation Transport Code," Proc. Seminar on Advanced Monte Carlo Computer Programs for Radiation Transport, Centre d'Etudes, Saclay, France, April 27-29, 1993.

General Reactor Physics

T. Greifenkamp, K. Clarno, and J. Gehin, "Effect of Fuel Temperature on Eigenvalue Calculations," in Proc. of the 2008 American Nuclear Society National Student Conference "Expanding the Nuclear Family", Texas A&M University, College Station, Texas, February 28-March 1, 2008.

K. 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.

K. 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.

K. 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, Vancouver, British Columbia, Canada, September 10-14, 2006.

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, Vancouver, British Columbia, Canada, September 10-14, 2006.

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

A. M. Fleckenstein and B. T. Rearden, "Extensible SCALE Intelligent Text Editor-ExSITE," Trans Am. Nucl. Soc., 98, 223-226 (2008).