Published: Jan 1994
| ||Format||Pages||Price|| |
|PDF Version (120K)||7||$25||  ADD TO CART|
|Complete Source PDF (15M)||7||$131||  ADD TO CART|
The 233U fission spectrum-averaged cross sections for twelve threshold reactions were measured relative to the average cross section of 0.688±0.040 mb for the 27Al(n,α)24Na reaction. The reference value was obtained by calculation using the energy dependent cross section in the Japanese Evaluated Nuclear Data Library (JENDL) Dosimetry File and the Watt-type fission spectrum in ENDF/B-VI. General agreement was seen between the measured and the calculated fission-spectrum averaged cross sections. However, there exist discrepancies of more than 10 % between the measured and the calculated average cross sections for the 24Mg(n,p)24Na, 47Ti(n,p)47Sc, and 64Zn(n,p)64Cu reactions. The tendencies in the calculated-to-measured ratios are similar to those for 235U fission spectrum-averaged cross sections we previously measured.
The measured average cross sections were also applied for the spectrum adjustment of the 233U fission neutrons using the Neutron Unfolding Package Code (NEUPAC). The adjusted spectrum is close to the Watt-type fission spectrum of 233U within the uncertainties of the obtained spectrum, although there exist some fluctuations in the ratio spectrum of the adjusted to the Watt-type.
233, U fission spectrum, 233, U fission foil, activation foil, spectrum-averaged cross section, threshold reaction, dosimetry cross section, JENDL Dosimetry File, activation measurement, spectrum adjustment, 27, Al(n,α), 24, Na reaction
Associate Professor, Research Reactor Institute, Kyoto University, Osaka,
Paper ID: STP15168S