Assessment of TRITON and PARCS for full-core MOX fuel calculations

Transactions of the American Nuclear Society

Volumn 92, Issue , 2005, Pages 763-766

  Dehart, Mark D ^a   Saccheri, Jacopo ^b   Diamond, David ^b  

^a OAK RIDGE NATIONAL LABORATORY   (United States)

^b BROOKHAVEN NATIONAL LABORATORY   (United States)

Author keywords

[No Author keywords available]

Indexed keywords

EID: 33244471320     PISSN: 0003018X     EISSN: None     Source Type: Conference Proceeding    
DOI: None     Document Type: Conference Paper

References (5)

1. M. D. DeHART, "An Advanced Deterministic Method for Spent-Fuel Criticality Safety Analysis," Proc. ANS 1998 Annual Meeting and Embedded Topical Meeting, June 7-11, 1998, Nashville, TN.
2. Trans. Am. Nucl. Soc., 78, 170-172 (1998).
3. T. J. DOWNAR, et al., "PARCS: Purdue Advanced Reactor Core Simulator," Proc. Int. Conf. on the New Frontiers of Nuclear Technology: Reactor Physics, Safety and High-Performance Computing, PHYSOR 2002, Seoul, Korea, October 7-10, 2002.
4. M. D. DeHart, Z. ZHONG, T. J. DOWNAR, "TRITON: An Advanced Lattice Code for MOX Fuel Calculations," 14-01.pdf in Proc. of American Nuclear Society, Advances in Nuclear Fuel Management III, October 5-8, 2003, Hilton Head Island, South Carolina.
5. O. W. HERMANN,R. M. WESTFALL, "ORIGEN-S: SCALE System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms," Vol. 2, Sect. F7, of SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, NUREG/CR-0200, Rev. 7 (ORNL/NUREG/CSD-2/R7) (2004).