Determination of 1-keV to 1-MeV Neutron Flux by Radiative Capture (n, $\gamma $ ) on92,94 Zr Dosimeters
The purpose of this paper is to demonstrate <sup>92,94</sup>Zr application in neutron reactor dosimetry. We aim to detect neutron flux in 1-keV to 1-MeV energy range by irradiating <sup>92,94</sup>Zr samples under boron nitride (BN) filter. The presence of BN is needed in order to cut off thermal neutron contribution. This paper describes Zr irradiation and experimental results that were obtained by <inline-formula> <tex-math notation="LaTeX">$\gamma $ </tex-math></inline-formula> spectrometry and accelerator mass spectrometry. In parallel, Zr irradiation is modeled with Monte Carlo-based code for neutron transport—TRIPOLI-4v10 developed by CEA (Saclay). Experimental values of <sup>95</sup>Zr activity and <sup>93</sup>Zr/<sup>92</sup>Zr ratio are compared to calculation results. Since crosssection values are necessary for the neutron flux determination, we present here the comparison between several nuclear data libraries for radiative capture on <sup>92,94</sup>Zr. Uncertainties for calculation/experience (C/E) ratios are obtained combining both experimental and nuclear data ones. This paper shows the feasibility of 1-keV to 1-MeV neutron detection by using <sup>92,94</sup>Zr (n, <inline-formula> <tex-math notation="LaTeX">$\gamma $ </tex-math></inline-formula>) reaction and introduces a postirradiation analysis for reactor dosimetry based on accelerator mass spectrometry. It also reveals that more precise evaluation of <sup>92,94</sup>Zr (n, <inline-formula> <tex-math notation="LaTeX">$\gamma $ </tex-math></inline-formula>) cross section is required in order to improve C/E values.