Browse TRAIL Inventories

Fast-flux measurements in the ORR core

The experimental design for BeO irradiation experiments ORNL 41-8 and ORNL 41-9

Program STATEST :an application of the method of statistical estimation to the calculation of neutron flux in anisotropically scattering media by Monte Carlo

Calculation of fission-source thermal-neutron distribution in water by the transfusion method

Calculations of thermal-neutron flux distributions in concrete-walled ducts using an albedo model with Monte Carlo techniques

Stress-strain measuring system for in-reactor use in high neutron flux

Determination of thermal-neutron flux distributions in the Bulk Shielding Reactor II by copper wire activation techniques

Neutron flux spectra in the experimental facilities of the Oak Ridge Research Reactor

Reactor yield calculations for 81 radioisotopes produced by (n,[gamma]) reactions at fluxes of 10⁷ to 10¹⁶ n/cm©ʹ²℗·sec for irradiation times of 30 minutes to one year


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