The program KRUDE, which solves the slowing down and adjoint equations for a mixture of resonance and nonresonance isotopes in an infinite homogeneous medium, was used to evaluate some resonance capture approximations (variational and successive) that use linear combinations of narrow and wide resonance fluxes. Results obtained for resonance capture in the 291-ev resonance of Zr/sup 91/ and the 192-ev resonance of U/sup 238/ are compared, and three methods for including Doppler effects in the variational method are considered.

The variational method is applied to the monoenergetic time dependent transport equation to obtain a simple relation for the asymptotic decay constant in small pulsed assemblies. The results indicate that flat trial functions may be a reasonable representation of the flux distributions in the thin slab limit. This approach is superior to many of the usual transport approximations.

A presentation of the values for the Pu²³⁹ fission cross section f factor extended into a higher energy range than previously reported in BNL-250. The values were obtained by integrating the standard expression defining the f factor on the Hanford IBM-CPC equipment. The un-fit experimental data of Leonard, as corrected for resolution and doppler effect, was used above 0.03 ev being attached to an analytic fit applicable below this point. Results are shown on the accompanying graph.

Monthly letter report on the research on the development of a beta-ray particle size analyzer (BPSA).

The purpose of the test was to determine the radiation levels in a Reactor Plant Boiler Compartment with the respective loop shut down and the remainder of the plant operating at power. The highest radiation levels recorded in the IC Boiler Compartment were located in the vicinity of the reactor coolant loop pipe penetrations of the Compartment. Except for the above mentioned areas, personnel could continuously occupy the IC Boiler Compartment without exceeding the maximum dose rate of 300 mr per week. The areas in the vicinity of the reactor coolant pipe penetrations, however, could be occupied only 1 to 2 hours per person per week because of the high radiation levers (200-250 mr/hr) present at these locations.

The ORNL analog computer was used to examine one aspect of a safety system for the EGCR. The thermal behavior of the fuel as a result of increases in reactivity at various power levels and gas flows was examined. Safety actions initiated by wither a flux-minus-flow signal or a reactor outlet gas temperature signal were included in simulation.