NUCLEAR PHYSICS RESEARCH QUARTERLY REPORT FOR OCTOBER, NOVEMBER, DECEMBER 1958
The experimental program to study the neutroninduced fission in low- energy resonances of nuclides with well defined fast neutron thresholds for fission in Pu/sup 240/, Np/sup 237/, and Am/sup 241/ is summarized. The infinite multiplication factor and thermal utilization factor were measured for U/sub 2/O- moderated UO/sub 2/ lattices. The 19 rod clusters on a nine-inch triangular pitch were either UO/sub 2/ or 16 UO/sub 2/ and 3 Pu-Al rods. Flux traverse data are given. The material bucklings of fuel elements composed of 7-rod clusters of 0.5-inch-diameter natural uranium rods were measured as a function of lattice spacing in graphite. The measurements were made in small exponential assemblies having cross sections of 4 x 4 feet or slightly larger. Radial neutron activation traverses, using Cu, U/sup 235/-Al, and Pu/sup 239/-Al detectors, were made in solid cylindrical fuel elements whose compositions simulated highly exposed natural uranium. An experimental investigation of the spatial dependence of the thermal neutron reaction rate in the vicinity of a temperature discontinuity is in progress. The data taken in moderators with temperatures ranging from 108 to 666 h a description of the experimental procedures, and a statement of the present status of this work are presented. The program of critical mass measurements with 3.063% enriched uranium rods in light water was continued. Critical approach measurements of the 0.600-inch diameter rods are reported. Reactivity parameters were calculated for bare fuel elements of 1.027% U/sup 235/ enrichment in light water lattices. A formal analytical solution of the problem of thermal neutron flux in a nonabsorbing heavy gas medium with a temperature discontinuity was previously obtained in the form of an infinite series. The convergence and numerical evaluation of the analytical solution and the utility of calculated reaction rates are discussed. Two approximation schemes are described. Pertinent formulas and numerical comparisons of the various versions of flux and 1/v reaction rate for a temperature ratio of 2:1 are presented. Data from recent Hanford Test Pile experiments were employed to prediet the effect of various changes in fuel rod cladding on the reactivity of a dry, graphite moderated, infinite lattice, fueled with natural uranium. A general method is developed for attacking problems of variational analysts of multi-dimensional systems through the systematic replacement of a complicated theory involving many independent variables by a simpler theory, involving equations in fewer independent variables. (For preceding period see HW-57861.) (W.D.M.)
- Research Organization:
- General Electric Co. Hanford Atomic Products Operation, Richland, Wash.
- DOE Contract Number:
- W-31-109-ENG-52
- NSA Number:
- NSA-13-012770
- OSTI ID:
- 4240309
- Report Number(s):
- HW-59126
- Resource Relation:
- Other Information: Orig. Receipt Date: 31-DEC-59
- Country of Publication:
- United States
- Language:
- English
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NUCLEAR PHYSICS RESEARCH QUARTERLY REPORT FOR OCTOBER, NOVEMBER, DECEMBER 1957
Related Subjects
ABSORPTION
ACTIVATION
ALUMINUM
AMERICIUM 241
CONFIGURATION
COPPER
CRITICAL MASS
CRITICALITY
CYLINDERS
ENERGY
ENRICHMENT
EQUATIONS
FISSION
FUEL ELEMENT CLUSTERS
FUEL ELEMENTS
GASES
GRAPHITE
ISOTOPES
MASS
MEASURED VALUES
MODERATORS
MULTIPLICATION FACTORS
NEPTUNIUM 237
NEUTRONS
NUCLEAR REACTIONS
NUMERICALS
PLANNING
PLUTONIUM 239
PLUTONIUM 240
RADIATION DETECTORS
REACTIVITY
RESONANCE
RODS
SOLIDS
TEMPERATURE
THERMAL UTILIZATION
URANIUM
URANIUM 235
URANIUM OXIDES
WATER