SOME EFFECTS OF Pu-240, Pu-241, AND Pu-242 ON THE Pu-239 CRITICAL MASS OF A FAST REACTOR

Thesis/Dissertation · Mon Jan 01 04:00:00 EST 1962

OSTI ID:4717475

Mertes, Jr, F P

The effects of variation in the isotopic composition of Pu used as a fast reactor fuel material are considered. When the Pu fuel material is produced in a thermal reactor, the primary factor affecting composition is the flux time of irradiation , 10/sup 21/ cm/sup -2/. Flux times of = 1, = 5, and = 10 were studied. It was found that for a flux time of = 10, the concentration of Pu-239 which yielded a critical mass was reduced to a concentration of 7.42% of the fuel from a concentration of 14.1% for the pure Pu- 239 case. In all cases studied the spherical core consisted of 35% fuel, 15% Fe, and 50% Na by volume and had a radius of 31 centimeters. The core was surrounded by a blanket of 70% U, 10% Fe, and 20% Na by volume which extended to a radius of 75 centimeters. The critical masses of the fast reactor were calculated by a one- dimensional, two-region, 16-neutron-energygroup computer code written in IBM FORTRANSIT. This code also was used to tabulate the neutron flux in 16 energy groups at each of 75 radius points. No significant change in the flux spectra resulted from the changes in Pu isotopic composition. The peak flux was in the energy range 0.3 Mev < E < 0.5 Mev. The program is compatible with the FORTRAN coding system and could be used to compute the critical mass of any spherical fast reaclor with one or two regions. The choice of fuel, coolant, and structural materials in the reactor is in no way restricted by the program instructions. For the particular reactor studied, a one-neutron-energy-group diffusion theory calculation predicted approximately the same changes in critical mass as were obtained from the computer solution. The Pu fuel was assumed to be produced in a thermal Pu recycle reactor with an average enrichment of 1%. The set of differential equations detailing the buildup of Pu-239, Pu-240, Pu-241, and Pu-242 in the thermal reactor was solved by use of an electronic analog computer. (Dissertation Abstr.)

• Research Organization: Originating Research Org. not identified
• NSA Number: NSA-17-022826
• OSTI ID: 4717475
• Country of Publication: Country unknown/Code not available
• Language: English

Related Subjects

COMPUTERS
CRITICALITY
DIFFERENTIAL EQUATIONS
DIFFUSION
FAST NEUTRONS
FORTRAN
GROUP THEORY
IRON
IRRADIATION
ISOTOPE EFFECTS
MASS
NEUTRON FLUX
PLUTONIUM 239
PLUTONIUM 240
PLUTONIUM 241
PLUTONIUM 242
PLUTONIUM ISOTOPES
PRODUCTION
PROGRAMMING
REACTOR CORE
REACTOR TECHNOLOGY
REACTORS
SODIUM
SPECTRA
SPHERES
TABLES
THERMAL NEUTRONS