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Experimental and theoretical analysis of neutron induced reaction cross section for 107,109Ag and 115In isotopes with covariance analysis.

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Title: Experimental and theoretical analysis of neutron induced reaction cross section for 107,109Ag and 115In isotopes with covariance analysis.
Authors: Sharma, Shivani1 (AUTHOR), Vandana1 (AUTHOR), Singh, N. L.2 (AUTHOR), Mehta, Mayur3 (AUTHOR), Singh, R. K.4 (AUTHOR), Makwana, R. J.4 (AUTHOR), Chauhan, R. D.4 (AUTHOR), Abhangi, Mitul3,5 (AUTHOR), Kumar, Ratnesh3 (AUTHOR), Sharma, Himanshu3 (AUTHOR), Vala, Sudhirsinh3,5 (AUTHOR), Katovsky, K.6 (AUTHOR), Bangotra, Pargin1 (AUTHOR) parginkumar@gmail.com
Source: European Physical Journal A -- Hadrons & Nuclei. May2026, Vol. 62 Issue 5, p1-17. 17p.
Subjects: Nuclear cross sections, Neutron irradiation, Nuclear activation analysis, Nuclear reactors, Covariance matrices, Nuclides
Geographic Terms: India
Abstract: Precise and up-to-date cross-sections with uncertainty propagation data of materials used in control rods are essential for the smooth functioning of nuclear reactors, since fast neutrons interact with control rods that regulate chain reactions by absorbing excess neutrons. The present study aims to measure the neutron-induced reaction cross-section values for the isotopes of silver (Ag) and indium (In). Deuterium–tritium (D–T) fusion neutrons with an energy of 14.96 ± 0.22 MeV were produced and used to irradiate natural samples of Ag and In targets, inducing measurable activation for the reactions 109Ag(n,2n)108Agg, 109Ag(n,p)109Pdm, 107Ag(n,2n)106Agm, 115In(n,p)115Cdg, and 115In(n,α)112Ag at the Neutron and Ion Irradiation Facility, Institute for Plasma Research (NIIF-IPR), Gujarat, India. The radioactive samples were subsequently taken for offline γ-ray counting in a high purity germanium detector (HPGe), with a fine resolution of 2.1 keV at 1.33 MeV γ-ray energy of 60Co connected with GENIE software. Two well-known standard reactions of Aluminium, 27Al(n,p)27 Mg and 27Al(n,α)24Na were employed for neutron flux measurements as per the irradiation period. Appropriate correction factors were applied in cross-section assessment, and uncertainties from all input parameters were rigorously propagated to the final values through covariance analysis. The experimental results were also validated by predictions from the nuclear code TALYS-2.0. Furthermore, prior reported studies from EXFOR and evaluated libraries from ENDF were compared with the presently measured results. [ABSTRACT FROM AUTHOR]
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Database: Engineering Source
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Abstract:Precise and up-to-date cross-sections with uncertainty propagation data of materials used in control rods are essential for the smooth functioning of nuclear reactors, since fast neutrons interact with control rods that regulate chain reactions by absorbing excess neutrons. The present study aims to measure the neutron-induced reaction cross-section values for the isotopes of silver (Ag) and indium (In). Deuterium–tritium (D–T) fusion neutrons with an energy of 14.96 ± 0.22 MeV were produced and used to irradiate natural samples of Ag and In targets, inducing measurable activation for the reactions 109Ag(n,2n)108Agg, 109Ag(n,p)109Pdm, 107Ag(n,2n)106Agm, 115In(n,p)115Cdg, and 115In(n,α)112Ag at the Neutron and Ion Irradiation Facility, Institute for Plasma Research (NIIF-IPR), Gujarat, India. The radioactive samples were subsequently taken for offline γ-ray counting in a high purity germanium detector (HPGe), with a fine resolution of 2.1 keV at 1.33 MeV γ-ray energy of 60Co connected with GENIE software. Two well-known standard reactions of Aluminium, 27Al(n,p)27 Mg and 27Al(n,α)24Na were employed for neutron flux measurements as per the irradiation period. Appropriate correction factors were applied in cross-section assessment, and uncertainties from all input parameters were rigorously propagated to the final values through covariance analysis. The experimental results were also validated by predictions from the nuclear code TALYS-2.0. Furthermore, prior reported studies from EXFOR and evaluated libraries from ENDF were compared with the presently measured results. [ABSTRACT FROM AUTHOR]
ISSN:14346001
DOI:10.1140/epja/s10050-026-01841-4