Speaker
Description
In spent nuclear fuel, $^{241}$Am is an important contributor to the decay heat and radiotoxicity for cooling times between several decades and a few centuries. Inherent $\gamma$-ray emission from $^{241}$Am aggravate time-of-flight (TOF) measurements of $^{241}$Am capture cross section. Alternatively, $^{241}$Am thermal neutron capture cross section and resonance integral may be estimated by analysis of activation or pile oscillation measurements. Potentially, the former may be more accurate than the one obtained from TOF measurements. It may thus serve for normalisation of the energy dependent capture yields measured by TOF.
Studies within the framework of the OECD/NEA WPEC SG-41 have shown that there are significant biases in the derivation of the thermal $^{241}$Am capture cross section from activation measurements with cadmium transmission filters using the conventional Westcott method. It was shown that this problem can be overcome either by a higher order correction to the Westcott method or by Monte Carlo calculated correction factors.
Neutron activation analysis of $^{241}$Am is comparatively complex. First, $^{241}$Am cross section contains two resonances below and around the cadmium transmission filter cut-off energy ($\sim$ 0.55 eV). Second, the activation product is produced in both ground and metastable state, and the latter has a much longer half-life (141 a for $^{242m}$Am vs. $\sim$ 16 h for $^{242g}$Am). And finally, the decay scheme of activation products is relatively complicated. $\gamma$-ray spectrometry of activation products is difficult due to low $\gamma$-ray energies and increased $\gamma$-ray background from $^{241}$Am fission products and $^{241}$Am itself. Therefore, the only realistic option is $\alpha$-particle spectrometry. $\alpha$-particles originating from $^{242}$Cm, which is a decay product of $^{242g,m}$Am, are measured relative to $\alpha$-particles originating from $^{241}$Am. Due to the much longer half-life of $^{242m}$Am and a relatively small branching fraction for its production by neutron capture ($\sim$ 0.09), its contribution to the $^{242}$Cm activity is negligible compared to the contribution from $^{242g}$Am for a few years after irradiation.
In contrast, pile oscillations are performed by introducing a sample containing $^{241}$Am into the reactor during operation, and the resulting decrease in system reactivity is assumed proportional to the neutron capture rate in $^{241}$Am. Compared to activation measurements, the required amount of $^{241}$Am in the samples is several orders of magnitudes higher.
Two $^{241}$Am samples, one with and one without cadmium cover, were irradiated Central Channel (CC) of the TRIGA reactor at Jožef Stefan Institute (JSI). The neutron fluence was monitored by reactions $^{59}$Co(n,$\gamma$), $^{197}$Au(n,$\gamma$) and $^{58}$Ni(n,p). In addition, oscillation measurements of two $^{241}$Am samples were performed in the MINERVE reactor of CEA Cadarache in the frame of the OSMOSE program.
The first step is to determine the specific reaction rates starting from the measured detector count rates. In the case of JSI measurements, the reaction rates were deduced from $\alpha$-particle spectrometry ($^{241}$Am) and $\gamma$-ray spectrometry (other materials). On the other hand, a C/E comparison was performed for the oscillation measurements at CEA Cadarache.
Preliminary trends on the $^{241}$Am neutron capture cross section have been obtained from the simultaneous analysis of the OSMOSE and JSI results by using a simple approach. The optimization process results in obtaining final C/E values close to unity. In the case of the JSI results, the Cd areal density was deduced from the measurements with the monitor sample containing gold. After optimisation, the discrepancies in the $^{241}$Am cross section are low. Using the corrected cross sections, the resulting thermal capture cross section and resonance integral are estimated as $\sigma(E_{th}) = 745$ b and $I_0 = 1560$ b, respectively. This confirms that the $^{241}$Am neutron capture resonance integral in JEFF-3.3 ($I_0 = 1826$ b) is overestimated.
[1] G. Žerovnik et al., ''Systematic effects on cross section data derived from reaction rates in reactor spectra and a re-analysis of $^{241}$Am reactor activation measurements,'' Nucl. Instr. Meth. A 877 (2018) 300-313.
[2] G. Žerovnik et al., ''Method for Analysis of Neutron Activation Measurements of Am-241 with Uncertainty Propagation,'' Proc. Int. Conf. NENE 2021, Sept. 6-9, 2021, Bled, Slovenia, pp. 1006.1-1006.8.
