THERMAL AND RESONANCE ABSORPTION CROSS SECTIONS OF THE U$sup 233$, U$sup 235$, AND Pu$sup 239$ FISSION PRODUCTS

Technical Report · Tue Mar 01 04:00:00 EST 1960

OSTI ID:4177514

Nephew, E A

ission product aggregate of U/sup 233/, U/sup 235/, and Pu/sup 239/ was calculated as a function of irradiation and the degree of flux thermaliziation. Attention was given to the resonance integral of the fission products in the range of 0 to 100 ev. The saturation characteristics of the individual fission products which determine the composition of the fission product aggregate are calculated for fluxes with different epithermal contents. The parameter P used to specify the degree of flux thermalization represents the fraction of fissions due to the neutrons in the l/E portion of the flux. By summing over the individual fission produce, a resulting thermal cross section and an energy distributed resonance integral were obtained for the fission product aggregate of the three fuels as a function of P and F where F is the fractional burnup of the fuel. For F = 0 and T = 1000 deg K, the effective thermal cross sections are calculated to be 37, 43. and 54 barns per fission for 239/, respectively. At the same conditions, the excess resonance integrals below 100 ev are calculated to be 134, 172, and 266 barns per fission for U/sup 233/, U/sup 235/, and ping the individual fission products which have similar saturation characteristics, it was possible to designate four pseudoelements which when taken together yield the same total thermal cross section and resonance integral as that of the fission product aggregate over the range of P and F values considered. Multigroup cross sections were computed and are listed for each pseudoelement for use in a 32- group code. Thermal group cross sections are listed for each pseudoelement for neutron temperatures between 69 and 2704 deg F. (auth)

• Research Organization: Oak Ridge National Lab., Tenn.
• DOE Contract Number: W-7405-ENG-26
• NSA Number: NSA-14-008946
• OSTI ID: 4177514
• Report Number(s): ORNL-2869
• Country of Publication: United States
• Language: English

Related Subjects

ABSORPTION
CROSS SECTIONS
DISTRIBUTION
ENERGY
ENERGY RANGE
EPITHERMAL NEUTRONS
FISSION
FISSION PRODUCTS
FUELS
IRRADIATION
NEUTRON FLUX
NUMERICALS
PHYSICS
QUANTITY RATIO
RESONANCE
RESONANCE INTEGRALS
SLOWDOWN
TABLES
TEMPERATURE
THERMAL NEUTRONS
THERMALIZATION
URANIUM 233
URANIUM 235