REACTIVITY LIFETIME AND BURNUP IN NUCLEAR FUELS (thesis)

Technical Report · Tue Jan 01 04:00:00 EST 1963

OSTI ID:4728370

de Ladonchamps, J R.L.

Analytical methods for the prediction of the reactivity lifetime and burnup of nuclear fuels are developed. The analysis applies to those nuclear fuels whose changes in composition with time are due solely to neutron-absorption processes, so that the composition of any fuel species is a function only of the integrated flux time of its irradiation exposure. Reactivity lifetime can then be expressed as a function of the appropriate average flux time of the fuel at the end of the irradiation. Local and average burnup of the irradiated fuel are calculated without necessarily specifying the magnitude of the irradiation flux or the power program of irradiation. A generalized perturbation method is developed which allows calculation of the above results, and which takes into account the spatial variation in neutron flux within the reactor and changes in this spatial variation during irradiation resulting from changes in fuel composition. Tabulated functions allow hand computations for the batch irradiation of fixed fuel in cylindrical or spherical reactors that have a uniform initial fuel loading. Such functions also apply to radial mixing and graded irradiation. The perturbation method is most easily applied to the one- group diffusion model, but the method is extended to the multigroup model with only slight modification for reactors with energy-independent boundary conditions. An exact analytical solution for the reactivity lifetime and fuel burnup was developed for continuous fueling schemes, if the one-group model applies and if the characteristic excess neutron production of the fuel varies as a quadratic function of the flux time of irradiation exposure. A comparative study of various continuous fueling schemes was made for a fuel with typical properties. The validity of the approximate solutions is determined by comparing results of the second-order perturbation method with the exact solutions of the same equations. Numerical computations on high-speed digital computers were used to obtain exact solutions of those equations that could not be solved analytically by means of elliptic functions. The computational procedures developed allow survey studies comparing the performances of various fuels and various reactor designs. Also, it predicts the magnitude of the errors to be expected when various neutron behavior models (one-group, two-group, continuous slowing down) are used in more elaborate computations on high-speed digital computers. Although attention was focused mainly on the properties related to the variations of the flux in the equivalent homogeneous fuel-moderator cell, corrections for lumped fuel were also investigated. 58 references. (auth)

• Research Organization: California. Univ. of Berkeley. Lawrence Radiation Lab.
• DOE Contract Number: W-7405-ENG-48
• NSA Number: NSA-17-021260
• OSTI ID: 4728370
• Report Number(s): UCRL-10614
• Country of Publication: United States
• Language: English

Related Subjects

ABSORPTION
BIBLIOGRAPHY
BURNUP
COMPUTERS
CYLINDERS
DIFFUSION
EQUATIONS
ERRORS
FUELS
GROUP THEORY
HOMOGENEOUS REACTORS
IRRADIATION
LIFETIME
NEUTRON FLUX
NEUTRONS
NUMERICALS
PLANNING
REACTIVITY
REACTOR TECHNOLOGY
REACTORS
SLOWDOWN
SPHERES
TABLES
TRANSPORT THEORY
VARIATIONS