Scale 6: Comprehensive nuclear safety analysis code system

Nuclear Technology

Volumn 174, Issue 2, 2011, Pages 126-148

  Bowman, S M ^a  

^a OAK RIDGE NATIONAL LABORATORY   (United States)

Author keywords

Criticality safety; Reactor physics; Sensitivity uncertainty

Indexed keywords

ACCIDENTS; ALARM SYSTEMS; COMPUTER SOFTWARE; CRITICALITY (NUCLEAR FISSION); MONTE CARLO METHODS; NUCLEAR REACTOR ACCIDENTS; RADIATION SHIELDING; REACTOR SHIELDING; SAFETY ENGINEERING; SHIELDING;

CRITICALITY SAFETY; NUCLEAR SAFETY ANALYSIS; OAK RIDGE NATIONAL LABORATORY; REACTOR PHYSICS; SENSITIVITY AND UNCERTAINTY ANALYSIS; SENSITIVITY/UNCERTAINTY; THREE-DIMENSIONAL LATTICES; VARIANCE REDUCTION TECHNIQUES;

UNCERTAINTY ANALYSIS;

EID: 79959237927     PISSN: 00295450     EISSN: None     Source Type: Journal    
DOI: 10.13182/NT10-163     Document Type: Review

References (58)

1. "SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," ORNL/ TM-2005/39, Version 6, Vols. I-III;
2. see also CCC-750, Radiation Safety Information Computational Center, Oak Ridge National Laboratory (Jan. 2009).
3. M. L. WILLIAMS and M. ASGARI, "Computation of Continuous-Energy Neutron Spectra with Discrete Ordinates Transport Theory," Nucl. Sci. Eng., 121, 173 (1995).
4. M. D. DeHART and S. M. BOWMAN, "Reactor Physics Methods and Analysis Capabilities in SCALE," Nucl. Technol., 174, 196(2011).
5. 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, 169 (2011).
6. M. L. WILLIAMS, "Resonance Self-Shielding Methodologies in SCALE," Nucl. Technol., 174, 149 (2011).
7. S. GOLUOGLU and M. L. WILLIAMS, "Modeling Doubly Heterogeneous Systems in SCALE," Trans. Am. Nucl. Soc., 93, 963 (2005).
8. S. GOLUOGLU, "Analysis of a Computational Benchmark for a High-Temperature Reactor Using SCALE," Proc. Int. Topi. Mtg. Reactor Physics (PHYSOR '06), Vancouver, British Columbia, Canada, September 10-14,2006, American Nuclear Society (2006).
9. 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, 214 (2011).
10. M. E. DUNN and N. M. GREENE, "AMPX-2000: A CrossSection Processing System for Generating Nuclear Data for Criticality Safety Applications," Trans. Am. Nucl. Soc., 86,118 (2002).
11. M. E. DUNN and L. C. LEAL, "Calculating Probability Tables for the Unresolved-Resonance Region Using Monte Carlo Methods," Proc. Int. Conf. New Frontiers of Nuclear Technology: Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002), Seoul, Korea, October 7-10, 2002, paper 11C.03, American Nuclear Society (2002).
12. B. T. REARDEN, "Perturbation Theory Eigenvalue Sensitivity Analysis with Monte Carlo Techniques," Nucl. Sci. Eng., 146, 367 (2004).
13. 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 (2004).
14. 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, 236(2011).
15. M. L. WILLIAMS and B. T. REARDEN, "Sensitivity/ Uncertainty Analysis Capabilities and New Covariance Data Libraries in SCALE," Nucl. Data Sheets, 109, 2796 (Dec. 2008).
16. 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 (2003).
17. D. E. MUELLER, B. T. REARDEN, and D. A. REED, "Evaluation of Fission Product Critical Experiments and Associated Biases for Burnup Credit Validation," Proc. IAEA/ CSN Int. Workshop Advances in Applications of Burnup Credit for Spent Fuel Storage, Transport, Reprocessing, and Disposition, Cordoba, Spain, October 27-30, 2009 (2010).
18. 233U," ORNL/TM-2008/196, Oak Ridge National Laboratory (Jan. 2009).
19. G. RADULESCU, D. E. MUELLER, and J. C. WAGNER, "Sensitivity and Uncertainty Analysis of Commercial Reactor Criticais for Burnup Credit," NUREG/CR-6951 (ORNL/TM2006/87), Oak Ridge National Laboratory (Dec. 2007).
20. 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 (2005).
21. M. B. CHADWICK et al., "ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology," Nucl. Data Sheets, 107, 2931 (2006).
22. S. GOLUOGLU, M. E. DUNN, L. M. PETRIE, and T. S. SUMNER, "Development and Validation of the New Continuous Energy Capability in the Criticality Safety Code KENO," Proc. Int. Conf. Physics of Reactors (PHYSOR '08), Interlaken, Switzerland, September 14-19, 2008.
23. M. L. WILLIAMS and G. ILAS, "ENDF/B-VII Nuclear Data Libraries for SCALE-6," Proc. Topi. Mtg. Advances in Nuclear Fuel Management (ANFM IV), Hilton Head Island, South Carolina, April 12-15, 2009, American Nuclear Society (2009).
24. "Evaluated Nuclear Structure Data File-A Computer File of Evaluated Experimental Nuclear Structure Data Maintained by the National Nuclear Data Center," ENSDF decay file as of April 2003, Brookhaven National Laboratory.
25. A. B. PASHCHENKO, H. WIENKE, J. KOPECKY, J.-Ch. SUBLET, and R. A. FORREST, "Neutron Activation CrossSection Data Library for Fusion Applications," IAEA-NDS173, Rev. 1, International Atomic Energy Agency (October 1998).
26. R. A. FORREST, J. KOPECKÝ, and J.-Ch. SUBLET, "The European Activation File: EAF-2003 Cross Section Library," UKAEA FUS-486, UKAEA/EURATOM Fusion Association (Dec. 2002).
27. J.-Ch. SUBLET, A. J. KONING, R. A. FORREST, and J. KOPECKY, "The JEFF-3.0/A Neutron Activation File-EAF2003 into ENDF-6 Format," JEFDOC-982, Commissariat à l'Energie Atomique (Nov. 2003); http://www.nea.fr/html/ dbdata/JEFF/ (current as of August 3, 2010).
28. I. GAULD, "Validation of ORIGEN-S Decay Heat Predictions for LOCA Analysis," Proc. Int. Topi. Mtg. Reactor Physics (PHYSOR'06), Vancouver, British Columbia, Canada, September 10-14, 2006, American Nuclear Society (2006).
29. I. C. GAULD, E. F. SHORES, and R. T. PERRY, "New Neutron Source Algorithms in the ORIGEN-S Code," Proc. Topi. Mtg. Radiation Protection and Shielding Division, Santa Fe, New Mexico, April 14-17, 2002, American Nuclear Society (2002).
30. E. F. SHORES, "SOURCES 4C: A Code for Calculating (a, n), Spontaneous Fission, and Delayed Neutron Sources and Spectra," LA-UR-02-1839, Los Alamos National Laboratory (Apr. 2002).
31. 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), Oak Ridge National Laboratory (Feb. 2010).
32. 2 Fuel)," NUREG/CR-6969 (ORNL/TM-2008/072), Oak Ridge National Laboratory (Feb. 2010).
33. 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), Oak Ridge National Laboratory (Feb. 2010).
34. 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), Oak Ridge National Laboratory (Feb. 2010).
35. 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), Oak Ridge National Laboratory (Oct. 2001).
36. G. RADULESCU and I. C. GAULD, "Enhancements to the Burnup Credit Criticality Safety Analysis Sequence in SCALE," Proc. Topi. Mtg. Nuclear Criticality Safety Division: Realism, Robustness and the Nuclear Renaissance (NCSD 2009), Richland, Washington, September 13-17,2009, American Nuclear Society (2009).
37. "Issue: Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks," Spent Fuel Project Office Interim Staff Guidance-8, Rev. 2, U.S. Nuclear Regulatory Commission (Sep. 27, 2002).
38. M. D. DeHART, "An Advanced Deterministic Method for Spent-Fuel Criticality Safety Analysis," Trans. Am. Nucl. Soc., 78, 170 (1998).
39. M. D. DeHART, "A Deterministic Study of the Deficiency of the Wigner-Seitz Approximation for Pu/MOX Fuel Pins," Proc. Int. Conf. M&C99-Madrid, Mathematics and Computations Meeting, Madrid, Spain, September 27-30,1999, p. 689 (1999).
40. M. D. DeHART, "High-Fidelity Depletion Capabilities of the SCALE Code System Using TRITON," Trans. Am. Nucl. Soc., 97, 598 (2007).
41. J. C. WAGNER and A. HAGHIGHAT, "Automated Variance Reduction of Monte Carlo Shielding Calculations Using the Discrete Ordinates Adjoint Function," Nucl. Sci. Eng., 128, 186(1998).
42. 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," Proc. Int. Conf. Mathematics, Computational Methods & Reactor Physics (M&C 2009), Saratoga Springs, New York, May 3-7, 2009, American Nuclear Society (2009).
43. T. M. EVANS, A. S. STAFFORD, R. N. SLAYBAUGH, and K. T. CLARNO, "Denovo: A New Three-Dimensional Parallel Discrete Ordinates Code in SCALE," Nucl. Technol., 171, 171 (2010).
44. D. E. PEPLOW, T. M. EVANS and J. C. WAGNER, "Simultaneous Optimization of Tallies in Difficult Shielding Problems," Nucl. Technol., 168, 785 (2009).
45. 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), Oak Ridge National Laboratory (Feb. 2009).
46. D. E. PEPLOW and L. M. PETRIE, Jr., "Criticality Accident Alarm System Modeling with SCALE," Proc. 2009 Int. Conf. Advances in Mathematics, Computational Methods, and Reactor Physics (M&C 2009), Saratoga Springs, New York, May 3-7, 2009, American Nuclear Society (2009).
47. S. R. BIERMAN and E. D. CLAYTON, "Critical Experiments to Measure the Neutron Poisoning Effects of Copper and Copper-Cadmium Plates," Nucl. Sci. Eng., 55, 58 (1974).
48. SCALE Web site: http://www.ornl.gov/sci/scale (current as of August 3, 2010).
49. P. B. FOX and D. F. HOLLENBACH, "KENO-VI Validation," ORNL/TM-2004/60, Oak Ridge National Laboratory (May 2005).
50. P. B. FOX and L. M. PETRIE, "Validation and Comparison of KENO V.a and KENO-VI," ORNL/TM-2001/110, Oak Ridge National Laboratory (May 2002).
51. S. M. BOWMAN, W. C. JORDAN, J. F. MINCEY, C. V. PARKS, and L. M. PETRIE, JR., "Experience with the SCALE Criticality Safety Cross-Section Libraries," NUREG/CR6685 (ORNL/TM-1999/322), Oak Ridge National Laboratory (Oct. 2000).
52. J. J. LICHTENWALTER, S. M. BOWMAN, M. D. DeHART, and C. M. HOPPER, "Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages," NUREG/CR-6361 (ORNL/TM-13211) , Oak Ridge National Laboratory (Mar. 1997).
53. 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, Oak Ridge National Laboratory (Nov. 1996).
54. 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 (Mar. 2010).
55. M. D. DeHART and S. M. BOWMAN, "Improved Radiochemical Assay Analyses Using TRITON Depletion Sequences in SCALE," Proc. Int. Conf. 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, p. 99, International Atomic Energy Agency (May 2007).
56. I. C. GAULD, "Strategies for Application of Isotopic Uncertainties in Burnup Credit," NUREG/CR-6811 (ORNL/TM2001/257), Oak Ridge National Laboratory (June 2003).
57. SCALE 6 package, Radiation Safety Information Computational Center, http://www-rsicc.ornl.gov (current as of August 3, 2010);
58. see also Nuclear Energy Agency, http://www. nea.fr http://nea.fr (current as of August 3, 2010).