Atr WG-MOX fuel pellet burnup measurement by Monte Carlo - Mass spectrometric method

Proceedings of the PHYSOR 2002 - International Conference on the New Frontiers of Nuclear Technology : Reactor Physics, Safety and High-Performance Computing - The ANS 2002 RPD Topical Meeting

Volumn , Issue , 2002, Pages

  Chang, G S ^a  

^a IDAHO NATIONAL LABORATORY   (United States)

Author keywords

[No Author keywords available]

Indexed keywords

FISSION PRODUCTS; ISOTOPES; LABORATORIES; LIGHT WATER REACTORS; MASS SPECTROMETERS; MASS SPECTROMETRY; MONTE CARLO METHODS; NUCLEAR FUELS; NUCLEAR WEAPONS; OXIDE MINERALS; PLASMA DIAGNOSTICS; PLUTONIUM; URANIUM; URANIUM DIOXIDE;

ADVANCED TEST REACTORS; IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORIES; ISOTOPIC CONCENTRATIONS; LINEAR HEAT GENERATION RATE; MIXED URANIUM-PLUTONIUM OXIDES; POST IRRADIATION EXAMINATIONS; SPECTROMETRIC METHODS; WEAPONS-GRADE PLUTONIUM;

REACTOR SHIELDING;

EID: 79959511165     PISSN: None     EISSN: None     Source Type: Conference Proceeding    
DOI: None     Document Type: Conference Paper

References (12)

1. Standard Test Method for Atom Percent Fission in Uranium and Plutonium Fuel (Neodymium-148 Method)”, ASTM Standards, E-321-96.
2. G. S. Chang And J. M. Ryskamp, "Depletion Analysis of Mixed Oxide Fuel Pins in Light Water Reactors and the Advanced Test Reactor," Nucl. Technol., Vol. 129, No. 3, pp. 326-337 (2000).
3. R. N. Morris, C. A. Baldwin, B. S. Cowell, S. A. Hodge, et. al. “MOX Average Power Early PIE: 8 GWd/MT Final Report,” Oak Ridge National Laboratory, ORNL/MD/LTR-172, (November 1999).
4. R. N. Morris, C. A. Baldwin, S. A. Hodge, L. J. Ott, C. M. Malone, N. H. Packan, “MOX Average Power Intermediate PIE: 21 GWd/MT Final Report,” Oak Ridge National Laboratory, ORNL/MD/LTR-199, (December 2000).
5. R.N. Morris, C. A. Baldwin, S. A. Hodge, N. H. Packan, “MOX Average Power 30 GWd/MT PIE: Final Report,” Oak Ridge National Laboratory, ORNL/MD/LTR-212, (November 2001).
6. Standard Test Method for Atom Percent Fission in Uranium and Plutonium Fuel (Mass Spectrometric Method), ASTM Standards, E-244-80, (Re-approved 1995).
7. J. Briesmeister (Editor), “MCNP-A General Monte Carlo N-Particle Transport Code”, Version 4B, LA-12625-M, Rev. 2, Los Alamos National Laboratory (1993).
8. A. G. Croff, “ORIGEN2: A Versatile Computer Code for Calculating the Nuclide Compositions and Characteristics of Nuclear Materials,” Nuclear Technology, Vol. 62, pp. 335-352 (1983).
9. G. S. Chang, R. C. Pedersen, “Weapons-Grade MOX Fuel Burnup Validation in ATR,” Trans. Am. Nucl. Soc., Vol 84, pp. 239, 2001.
10. G. S. Chang, JW Rogers, and J. M. Ryskamp, "Neutron Flux and Spectrum Variation in a MOX Fuel Experiment," Trans. Am. Nucl. Soc., Vol. 81, 284 (1999).
11. G. S. Chang and J. M. Ryskamp, "Monte-Carlo Boundary Source Approach in MOX Fuel Test Capsule Design," Trans. Am. Nucl. Soc., Vol. 80, pp. 268-269 (1999).
12. S. A. Hodge, L. J. Ott, F. P. Griffin, C. R. Luttrell, “Implications of the PIE for the 30 GWd/MTWithdrawal MOX Capsules,” Oak Ridge National Laboratory, ORNL/MD/LTR-212, Vol. 2, Rev. 1, (February 2002).