Reactor physics methods and analysis capabilities in SCALE

Nuclear Technology

Volumn 174, Issue 2, 2011, Pages 196-213

  De Hart, Mark D ^a   Bowman, Stephen M ^a,b  

^a OAK RIDGE NATIONAL LABORATORY   (United States)

^b IDAHO NATIONAL LABORATORY   (United States)

Author keywords

Lattice physics; Reactor analysis; TRITON

Indexed keywords

LIGHT WATER REACTORS;

ANALYSIS CAPABILITIES; COMPLEX DESIGN PROJECTS; LATTICE PHYSIC; REACTOR ANALYSIS; REACTOR CONCEPTS; REACTOR PHYSICS; RIGOROUS APPROACH; TRITON;

NUCLEAR FUELS;

EID: 79959208362     PISSN: 00295450     EISSN: None     Source Type: Journal    
DOI: 10.13182/NT174-196     Document Type: Review

References (36)

1. M. D. DeHART, Z. ZHONG, and T. J. DOWNAR, "TRITON: An Advanced Lattice Code for MOX Fuel Calculations," Proc. Advances in Nuclear Fuel Management III (ANFM 2003), Hilton Head Island, South Carolina, October 5-8, 2003, 14-01.pdf, American Nuclear Society (2003) (CD-ROM).
2. "SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," ORNL/ TM-2005/39, Version 6, Vols. I-III, CCC-725, Radiation Safety Information Computational Center, Oak Ridge National Laboratory (Feb. 2009).
3. M. D. DeHART, "An Advanced Deterministic Method for Spent-Fuel Criticality Safety Analysis," Trans. Am. Nucl. Soc., 78, 170 (1998).
4. M. D. DeHART, "Advancements in Generalized-Geometry Discrete Ordinates Transport for Lattice Physics Calculations," Proc. Topi Mtg. Reactor Physics: Advances in Nuclear Analysis and Simulation (PHYSOR 2006), Vancouver, British Columbia, Canada, September 10-14, 2006, American Nuclear Society (2006).
5. 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).
6. 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).
7. S. M. BOWMAN, M. D. DeHART, and L. M. PETRIE, Jr., "Integrated KENO Monte Carlo Transport for 3-D Depletion with SCALE," Proc. The Monte Carlo Method: Versatility Unbounded in a Dynamic Computing World, Chattanooga, Tennessee, April 17-21, 2005, American Nuclear Society (2005) (CD-ROM).
8. M. L. WILLIAMS and M. ASGARI, "Computation of Continuous-Energy Neutron Spectra with Discrete Ordinates Transport Theory," Nucl. Sci. Eng., 121, 173 (1995).
9. M. L. WILLIAMS, S. GOLUOGLU, and L. M. PETRIE, "Recent Enhancements to the SCALE 5 Resonance SelfShielding Methodology," Trans. Am.Nucl Soc.,92, 751 (2005).
10. M. D. DeHART, R. E. PEVEY, and T. A. PARISH, "An Extended Step Characteristic Method for Solving the Transport Equation in General Geometries," Nucl. Sci. Eng., 118, 79 (1994).
11. 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 (2003).
12. G. ILAS and I. C. GAULD, "SCALE Analysis of CLAB Decay Heat Measurements for LWR Spent Fuel Assemblies," Ann. Nucl. Energy, 35, 37 (2008).
13. S. M. BOWMAN and L. C. LEAL, "ORIGEN-ARP, A Fast and Easy-to-Use Source Term Generation Tool," Proc. Ninth Int. Conf. Radiation Shielding (ICRS-9), Tsukuba, Japan, October 17-22, 1999.
14. T. J. DOWNAR et al., "PARCS: Purdue Advanced Reactor Core Simulator," Proc. Int. Conf. New Frontiers of Nuclear Technology: Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002), Seoul, Korea, October 7-10, 2002.
15. M. D. DeHART, J. SACCHERI, and D. DIAMOND, "Assessment of TRITON and PARCS for Full-Core MOX Fuel Calculations," Trans. Am. Nucl. Soc., 92, 763 (2005).
16. P. J. TURINSKY et al., "NESTLE: A Few-Group Neutron Diffusion Equation Solver Utilizing the Nodal Expansion Method (NEM) for Eigenvalue, Adjoint, and Fixed-Source Steady-State and Transient Problems," EGG-NRE-11406, Idaho National Engineering Laboratory (June 1994).
17. G. I. MALDONADO, J. GALLOWAY, H. HERNANDEZ, K. T. CLARNO, E. L. POPOV, and M. A. JESSEE, "Integration of the NESTLE Core Simulator with SCALE," Trans. Am. Nucl. Soc., 100, 619 (2009).
18. G. ILAS, R. T. PRIMM III, and J. C. GEHIN, "Designing a New Fuel for HFIR - Performance Parameters for LEU Core Configurations," Proc. Advances in Nuclear Fuel Management IV (ANFM 2009), Hilton Head Island, South Carolina, April 12-15, 2009, usr-gilas-l-paper.pdf, American Nuclear Society (2009) (CD-ROM).
19. T. A. WAREING, J. M. McGHEE, J. E. MOREL, and S. D. PAUTZ, "Discontinuous Finite Element SN Methods on ThreeDimensional Unstructured Grids," Nucl. Sci. Eng., 138, 256 (2001).
20. D. S. LUCAS et al., "Comparison of the 3-D Deterministic Neutron Transport Code Attilāto Measured Data, MCNP and MCNPX for the Advanced Test Reactor," INL/CON-05-00662, Idaho National Laboratory (Sep. 2005).
21. M. D. DeHART and S. M. BOWMAN, "Improved Radiochemical Assay Analyses Using TRITON Depletion Sequences in SCALE," Proc. Technical Mtg. 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, pp. 99-108, International Atomic Energy Agency (May 2007).
22. 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," Proc. Topi Mtg. Reactor Physics: Advances in Nuclear Analysis and Simulation (PHYSOR 2006), Vancouver, British Columbia, Canada, September 10-14, 2006, American Nuclear Society (2006) (CD-ROM).
23. M. A. JESSEE, M. L. WILLIAMS, and M. D. DeHART, "Development of Generalized Perturbation Theory Capability Within the SCALE Code Package," Proc. Int. Conf. Mathematics, Computational Methods and Reactor Physics (M& C 2009), Saratoga Springs, New York, May 3-7, 2009, American Nuclear Society (2009).
24. D. BOULANGER, M. LIPPENS, L. MERTENS, J. BASSELIER, and B. LANCE, "High Burnup PWR and BWR MOX Fuel Performance: A Review of Belgonucleaire Recent Experimental Programs," Proc. Int. Mtg. LWR Fuel Performance, Orlando, Florida, September 19-22, 2004, American Nuclear Society (2004) (CD-ROM).
25. G. ILAS and I. C. GAULD, "Analysis of Isotopic Assay Data from the MALIBU Program," Proc. Int. Conf. Physics of Reactors (PHYSOR 2008), Interlaken, Switzerland, September 1,2008(CD-ROM).
26. B. C. NA, "Benchmark on the VENUS-2 MOX Core Measurements," NEA/NSC/DOC(2000)7, Organisation for Economic Co-operation and Development/Nuclear Energy Agency (2000).
27. M. D. DeHART, Z. ZHONG, and T. J. DOWNAR, "TRITON: An Advanced Lattice Code for MOX Fuel Calculations," Proc. Advances in Nuclear Fuel Management III, Hilton Head Island, South Carolina, October 5-8, 2003, American Nuclear Society (2003) (CD-ROM).
28. M. D. DeHART et al., "Application of TRITON-TRACE/ PARCS to the Analysis of Peach Bottom Unit 2, Cycle 1," ORNL/NRC/LTR-06/07, Oak Ridge National Laboratory (Dec. 2006).
29. "Evaluation of High-Temperature Gas-Cooled Reactor Performance: Benchmark Analysis Related to Initial Testing of the HTTR and HTR-IO," IAEA-TECDOC-1382, International Atomic Energy Agency (2003).
30. J. GALLOWAY, H. HERNANDEZ, and G. I. MALDONADO, "BWR Modeling Capability and SCALE/TRITON Lattice-to-Core Integration of the Nestle Nodal Simulator," Proc. Advances in Reactor Physics to Power the Nuclear Renaissance (PHYSOR 2010), Pittsburgh, Pennsylvania, May 9-14, 2010, American Nuclear Society (2010) (CD-ROM).
31. J. D. GALLOWAY, "Boiling Water Reactor Simulation with Generalized Isotopic Inventory Tracking for Actinide Management," Ph.D. Thesis, University of Tennessee Department of Nuclear Engineering (July 2010).
32. S. GOLUOGLU and M. L. WILLIAMS, "Modeling Doubly Heterogeneous Systems in SCALE," Trans. Am. Nucl. Soc., 93, 963 (2005).
33. G. HOSKING and T. D. NEWTON, "Benchmark Specification for an HTR Fuelled with Reactor Grade Plutonium (or Reactor Grade Pu/Th and U/Th)," Proposal, NEA/NSC/ DOC(2003)22, Nuclear Energy Agency (Mar. 2005).
34. M. D. DeHART and A. P. ULSES, "Benchmark Specification for HTGR Fuel Element Depletion," NEA/NSC/ DOC(2009) 13, Nuclear Energy Agency (Feb. 2009).
35. J. LEPPÄNEN, "Development of a New Monte Carlo Reactor Physics Code," D.Sc. Thesis, Helsinki University of Technology, VTT Publications 640 (2007).
36. R. E. MacFARLANE and D. W. MUIR, "NJOY 97.0: Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data," RSICC Code Package PSR-368, Oak Ridge National Laboratory (May 1998).