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For the accuracy improvement of nuclear data of minor actinides and long-lived fission products in the project of “Research and development for Accuracy Improvement of neutron nuclear data on Minor Actinides”, the isotopic compositions of two Am samples (241Am sample and 243Am sample) were analyzed by thermal ionization mass spectrometry. Only the...
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... results of the three times analysis of 243 Am sample were summarized in Table 4. The abundances of 243 Am and 241 Am were evaluated as 97.7% and 2.3% as shown in Table 4. ...
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Citations
... To make detailed isotopic and impurity analyses, liquid sample from the same batch of the samples was also supplied by the manufacturers. Isotopic and impurity abundances were determined by thermal ionization mass spectrometry (TIMS) and alpha-ray spectrometry [32]. The results are described in Table 4. ...
Neutron total and capture cross sections of ²⁴³Am have been measured in Accurate Neutron Nucleus Reaction measurement Instrument at Materials and Life Science Experimental Facility of Japan Proton Accelerator Research Complex with a neutron time-of-flight (TOF) method. The neutron capture cross section in the energy region from 10 meV to 100 eV was determined using an array of Ge detectors. Three samples with different activities were used for measurements of the capture cross section. The neutron total cross section in the energy region from 4 meV to 100 eV was measured using Li-glass detectors. Derived cross-section value at neutron energy of 0.0253 eV is 87.7±5.4 b for the capture cross section and 101±11 b for the total cross section.
... Decay heat of the Am sample was measured accurately with a precision of 200 nW by using the TAM-IV calorimeter manufactured by TA Instruments, and thus the activity of the Am sample was obtained with an uncertainty of 0.05% [8]. The isotopic composition of the Am sample was analyzed by thermal ionization mass spectrometry and alpha-ray spectroscopy [39]. The determined isotopic purity of 241 Am was more than 99.9996%, and the isotopic contamination due to other isotopes was less than 0.0004%. ...
Neutron total and capture cross sections of ²⁴¹Am have been measured with a new data acquisition system and a new neutron transmission measurement system installed in Accurate Neutron Nucleus Reaction measurement Instrument at Materials and Life Science Experimental Facility of Japan Proton Accelerator Research Complex. The neutron total cross sections of ²⁴¹Am were determined by using a neutron time-of-flight (TOF) method in the neutron energy region from 4 meV to 2 eV. The thermal total cross section of ²⁴¹Am was derived with an uncertainty of 2.9%. A pulse-height weighting technique was applied to determine neutron capture yields of ²⁴¹Am. The neutron capture cross sections were determined by the TOF method in the neutron energy region from the thermal to 100 eV, and the thermal capture cross section was obtained with an uncertainty of 4.1%. The evaluation data of JENDL-4.0 and JEFF-3.2 were compared with the present results.
The thermal-neutron capture cross section (σ0) and resonance integral (I0) for⁹³Nb were measured by an activation method; a half-life of ⁹⁴Nb was derived by mass analysis. Niobium-93 samples were irradiated with the use of the hydraulic conveyer installed in the research reactor in Institute for Integral Radiation and Nuclear Science, Kyoto University. Gold-aluminum and cobalt-aluminum alloy wires were used to monitor thermal-neutron fluxes and Westcott’s indexes at the irradiation position. A 25-μm-thick gadolinium foil was used to separate reactions ascribed to thermal neutrons. Its thickness provided a cut-off energy of 0.133 eV. In order to attenuate the radioactivity of ¹⁸²Ta produced by impurities, the Nb samples were cooled for nearly 2 years. The induced radio activity in the monitors and Nb samples was measured by gamma-ray spectroscopy. In the analysis based on Westcott’s convention, the σ0 and I0 values were calculated as 1.11 ± 0.04 barns and 10.5 ± 0.6 barns, respectively. After the γ-ray measurements, mass analysis was applied to the Nb samples to obtain the reaction rate. By combining data obtained by both γ-ray spectroscopy and mass analysis, a half-life of ⁹⁴Nb was determined as (2.00 ± 0.15) ×10⁴ years.
Neutron total and capture cross–section measurements of ¹⁵⁵Gd and ¹⁵⁷Gd were performed with Li-glass detectors and a NaI(Tl) spectrometer in the ANNRI at the MLF of the J-PARC. The neutron total cross sections were determined in the energy region from 5 to 100 meV, and those at the thermal neutron energy were obtained to be 59.4 ± 1.7 and 251.9 ± 4.6 kilobarn for ¹⁵⁵Gd and ¹⁵⁷Gd, respectively. The neutron capture cross sections were determined in the energy region from 3.5 to 100 meV with an innovative method, and those at the thermal energy were obtained as 59.0 ± 2.5 and 247.4 ± 3.9 kilobarn for ¹⁵⁵Gd and ¹⁵⁷Gd, respectively. The present total and capture cross sections agree well within the standard deviations. The results for ¹⁵⁵Gd were found to be consistent with the values in JENDL-4.0 and the experimental data given by Møller et al., Mastromarco et al. and Leinweber et al. Those for ¹⁵⁷Gd agreed with the evaluated data in JENDL-4.0 and the experimental data by Møller et al. and Mastromarco et al. However, they disagree (11% larger) with the experimental result by Leinweber et al.
In terms of nuclear transmutation studies of minor actinides in nuclear wastes, the present work selected ²³⁷Np among them and aimed to measure the thermal-neutron capture cross-section of ²³⁷Np using a well-thermalized neutron field by a neutron activation method because there have been discrepancies among reported cross-section data. A ²³⁷Np standard solution was used for irradiation samples. The thermal-neutron flux at an irradiation position was measured with flux monitors: ⁴⁵Sc, ⁵⁹Co, ⁹⁸Mo, ¹⁸¹Ta, and ¹⁹⁷Au. The ²³⁷Np sample was irradiated together with the flux monitors for 30 minutes in the graphite thermal column equipped in the Kyoto University Research Reactor. The similar irradiation was repeated once more to confirm the reproducibility of the results. After irradiation, the ²³⁷Np samples were quantified using 312-keV gamma-ray emitted from ²³³Pa in radiation equilibrium with ²³⁷Np. The reaction rates of ²³⁷Np were obtained from the peak net counts of gamma-rays emitted from generated ²³⁸Np, and then the thermal-neutron capture cross-section of ²³⁷Np was found to be 173.8 ± 4.7 barn by averaging the results obtained by the two irradiations. The present result was in agreement with the reported data given by a time-of-flight method within a limit of uncertainty.
