The thermal neutron capture cross sections and the neutron capture resonance integrals of 241Am leading to the production of the isomer 242mAm and the ground-state 242gAm were measured radiochemically by the Cd-ratio technique with neutron flux monitors of Co/Al and Au/Al alloy. Highly-purified 241Am targets were irradiated in an aluminum capsule by using JMTR. The neutron fluxes and their epithermal neutron fractions were determined by measuring y-rays of 6oCo and ' "Au. The yields of 242mAm and 242gAm were decided by analyzing growth and decay curves of the a-ray activity ratios 242Cm/241Am. The resultant thermal neutron capture cross sections are 85.7 & 6.3 b and 768 zk 58 b for 242mAm and 242gAm, and the resonance integrals 114rt7 b and 1,694f146 b, respectively. The differences between the present results and the evaluated values by Mughabghab are 38-59%. The isomeric ratios, g/(m+g), of 0.90f0.09 for thermal neutrons and 0.94zk0.11 for epithermal neutrons are, however, almost consistent with evaluated values.
The thermal neutron capture cross sections and the neutron capture resonance integrals of 241Am leading to the production of the isomer 242mAm and the ground-state 242gAm were measured radiochemically by the Cd-ratio technique with neutron flux monitors of Co/Al and Au/Al alloy. Highly-purified 241Am targets were irradiated in an aluminum capsule by using JMTR. The neutron fluxes and their epithermal neutron fractions were determined by measuring grays of 60Co and 198Au. The yields of 242mAm and 242gAm were decided by analyzing growth and decay curves of the a-ray activity ratios 242Cm/241Am. The resultant thermal neutron capture cross sections are 85.7+-6.3b and 768+-58b for 242mAm and 242gAm, and the resonance integrals 114+-7b and 1,694+-146b, respectively. The differences between the present results and the evaluated values by Mughabghab are 38-59%. The isomeric ratios, g/(m+g), of 0.90+-0.09 for thermal neutrons and 0.94+-0.11 for epithermal neutrons are, however, almost consistent with evaluated values.
Destructive analyses for five spent fuel samples taken from a Gd bearing fuel assembly were done. The measured amounts of actinides of zs4-zssU, zs7Np, 23*-z4zP~, z41~242m~z43Am and , and fission products of Is4Cs and 1 5 4 E~ were used for evaluating the accuracy of calculation made by CASMO-MICBURN and ORIGEN-2 codes. T h e effect of Gd on the neutron spectrum was taken into account in the CASMO-MICBURN calculation.The amounts of 235U, 239Pu and 241Pu calculated by CASMO-MICBURN agreed well with the observed values within about 3%. On the other hand, the amounts obtained from ORIGEN-2 calculation showed lower values than those observed, especially by -12% in average in 236ufor Gdz03-U02 fuel. 'The main cause of this large difference may be attributed to the effect of Gd on the neutron spectrum. The amounts of the other actinides by both calculation codes revealed no significant difference in nearly 10% except for 24zmAm, in which a large fluctuation among the samples was observed. About 10% difference between the measured values and the calculated values was also observed for Is4Cs, but the calculated values for IS4Eu showed a significant difference from measured values.
242,244Cm
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