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Technical Paper

The Determination of Ratios of Effective Cross Sections from Burnup Data from the Saxton Plutonium Program

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Pages 229-238 | Published online: 12 May 2017

References

  • W. R. SMALLEY, “Saxton Plutonium Program Semiannual Progress Report for the Period Ending December 31, 1969,” WCAP-3385-22, Westinghouse Electric Corporation (1970).
  • R. P. MATSEN, “A Technique for the Determination of Ratios of Effective Cross Sections from Reactor Fuel Burnup Data,” Nucl. Sci. Eng., 32, 262 (1968).
  • R. P. MATSEN, “DBUFIT-I, A Least Squares Analysis Code for Nuclear Burnup Data,” BNWL-1396, Pacific Northwest Laboratory (1970).
  • R. J. NODVIK, “Saxton Core II Fuel Performance Evaluation Part II: Irradiated Saxton Plutonium Fuel,” WCAP-3885-56, Westinghouse Electric Corporation (1970).
  • M. E. MEEK and B. F. RDDER, “Summary of Fission Product Yields for U235, U238, Pu239 and Pu241 at Thermal, Fission Spectrum and 14 MeV Neutron Energies,” APED-5398-A, General Electric Company (1968).
  • F. L. LISMAN et al., “Burnup Determination of Nuclear Fuels,” IN-1277, Idaho Nuclear Corporation (1969).
  • R. P. MATSEN, “An Analysis of Yankee-Rowe Burnup Data,” BNWL-1122, Pacific Northwest Laboratory (1969).
  • G. C. HANNA, C. H. WESTCOTT, H. D. LEMMEL, B. R. LEONARD, Jr., J. S. STORY, and P. M. ATTREE, “Revision of Values for the 2200 m/s Neutron Constants for Four Fissile Nuclides,” Atomic Energy Review, Vol. VII, No. 4, International Atomic Energy Agency, Vienna (1969).
  • C. G. PONCELET, “LASER-A Depletion Program for Lattice Calculations Based on MUFT and THERMOS,” WCAP-6073, Westinghouse Electric Corporation (1966).
  • H. BOHL, Jr., E. M. GELBARD, andG. M. RYAN, “MUFT-4 Fast Neutron Spectrum Code for the IBM-704,” WAPD-TM-72, Westinghouse Electric Corporation (1957).
  • H. C. HONECK, “THERMOS-A Thermalization Transport Theory Code for Reactor Lattice Calculations,” BNL-5826, Brookhaven National Laboratory (1961).
  • D. F. NEWMAN and C. R. GORDON, “Measurement and Analytical Correlation of Neutronic Parameters for a Water Moderated Lattice Containing UO2-PuO2 Particulate Fuels,” BNWL-1603, Pacific Northwest Laboratory (1971).
  • R. P. MATSEN, G. J. BUSSELMAN, R. H. HOLEMAN, and R. C. LIIKALA, “An Analysis of Uranium Fuel Irradiated in Yankee Reactor,” Trans. Am. Nucl. Soc., 12, 31 (1969).
  • L. C. SCHMID, D. E. CHRISTENSEN, B. H. DUANE, R. C. LIIKALA, and R. P. MATSEN, Proc. Joint Intern. Conf. Physics Problems in Thermal Reactor Design, London, England, p. 241 (1967). ( Unfortunately the last two pages of this report were not reproduced in the proceedings. The missing pages may be obtained from the authors, Battelle, Pacific Northwest Laboratories.)

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