NEUTRON-FLUX MEASUREMENTS IN A FLAT PLATE FUEL ELEMENT

The equipment and experiments performed to measure the thermal-neutron- flux distribution in a fuel assembly of an experimental loop mock-up of a gas:cooled reactor at the Battelle Research Reactor (BRR) are described. The loop was located adjacent to the core of the BRR and contained one fuel assermbly composed of seven flat fuel plates each containing approximately 29.5 g of U/sup 235/. The plates consisted of a core 0.050 in. thick of UO/sub 2/ dispersed in Type 347 stainless steel and clad on each side with 0.005 in. of Type 347 stainless steel. The measurements showed that with the present design of the loop system an average thermal-neutron flux of 4.09 x 1O/sup 12/ neutrons/(cm/sup 2/)(sec) or a power generation of 45 kw in the assembly can be conveniently obtained. The ratio of the peak thermal-neutron flux to average thermal flux in the entire element was found to be 1.87. At any horizontal cross section, thermal-flux depression from the edge of the element to the center of less than a factor of two was observed for the final loopcore arrangement. (auth)