Prediction of spent mox and leu fuel composition and comparison with measurements

Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium

Volumn , Issue , 2000, Pages

  Murphy, B D ^a   Primm, R T ^a  

^a OAK RIDGE NATIONAL LABORATORY   (United States)

Author keywords

[No Author keywords available]

Indexed keywords

ACTINIDES; BOILING WATER REACTORS; CESIUM; ESTIMATION; EUROPIUM; FISSION PRODUCTS; FORECASTING; ISOTOPES; NEODYMIUM; NUCLEAR FUELS; OXIDE MINERALS; SAMARIUM; URANIUM DIOXIDE;

ACCURATE ESTIMATION; BOILING WATER REACTOR (BWR); COMPARISON WITH MEASUREMENTS; IRRADIATED SAMPLES; NEODYMIUM CONCENTRATIONS; PRODUCT CONCENTRATION; RADIOCHEMICAL ANALYSIS; SIMULATION CODE;

PRESSURIZED WATER REACTORS;

EID: 44849133596     PISSN: None     EISSN: None     Source Type: Conference Proceeding    
DOI: None     Document Type: Conference Paper

References (6)

1. B. D. Murphy, Characteristics of Spent Fuel from Plutonium Disposition Reactors, Vol. 3: A Westinghouse Pressurized-Water Reactor Design, ORNL/TM-13170, Oak Ridge Natl. Lab., July 1997.
2. M. D. DeHart, Sensitivity and Parametric Evaluations of Significant Aspects of Burnup Credit for PWR Spent Fuel Packages, ORNL/TM-12973, Oak Ridge Natl. Lab., May 1996.
3. S. E. Fisher and F. C. Difilippo, Neutronics Benchmark for the Quad Cities-1 (Cycle 2) Mixed-Oxide Assembly Irradiation, ORNL/TM-13567, Oak Ridge Natl. Lab., April 1998.
4. J. J. Casal, R. J. J. Stamm’ler, E. A. Villarino, and A. A. Ferri,HELIOS: Geometric Capabilities of a New Fuel-Assembly Program, International Topical Meeting on Advances in Mathematics, Computations, and Reactor Physics, Pittsburgh, Pa., April 28-May 2, 1991.
5. SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, NUREG/CR-0200, Rev. 5 (ORNL/NUREG/CSD-2/R5), Vols. I, II, and III, U.S. Nuclear Regulatory Commission, March 1997. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-545.
6. ASTM, Annual Book of Standards, Vol. 12.01, E321, Standard Test Method for Atom Percent Fission in Uranium and Plutonium Fuel (Neodymium-148 Method), American Society for Testing Materials, West Conshohocken, Pa. 19428 (1996).