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The cross-section of the 181Ta(n,2n)180Ta reaction has been measured with respect to the 197Au(n,2n)196Au monitor reaction at the incident neutron energy of 14.78± 0.20 MeV, using neutron activation analysis and off-line γ-ray spectrometric technique. The present measurement has been done at the energy where discrepant measured results are available in the EXFOR data library. The result has been compared with evaluated data libraries JEFF-3.3 and ENDF/BVII. 1. The present result has also been supported by theoretical predictions of nuclear model code TALYS1.8 and TALYS- 1.9. The uncertainty and the correlations among the measured cross-section has been studied using co-variance analysis. [ABSTRACT FROM AUTHOR]

Abstract The neutron capture cross-sections have been measured for the 159Tb(n, γ)160Tb reaction at the spectrum average peak neutron energies of 5.08 ± 0.165, 12.47 ± 0.825, and 16.63 ± 0.95 MeV respectively. The experiment has been carried out using the standard neutron activation technique and off-line γ-ray spectrometry. The present measurement has been done for the energies where very few measured results are available in the data library. The results have been compared with ENDF/B-VII.1 and JENDL-4.0 data libraries. The present results have also been supported by theoretical predictions of nuclear model code TALYS 1.9. Detailed covariance analysis was carried out to find the uncertainty and the correlations among the measured cross-sections. Highlights • Tb being an excellent structural metal has several reactor and nuclear medicines based applications. • Tb isotopes are of keen interest of study since they are widely used in nuclear medicine. • Calculations have been performed within the framework of theoretical code Talys-1.9. [ABSTRACT FROM AUTHOR]

The cross-sections for 86Sr (n, 2n)85Sr reaction are measured at neutron energies 19.44 ± 1.02 MeV and 16.81 ± 0.85 MeV wherein there is scarcity of data. The standard neutron activation analysis technique and offline gamma ray spectroscopy have been employed for measurement and analysis of the data. The results are compared with experimental data available in EXFOR database, JEFF-3.3, JENDL-4.0, TENDL-2017 and ENDF/B-VIII.0 evaluated data. The theoretical prediction was incorporated using nuclear modular codes TALYS 1.8 and EMPIRE 3.2.2. A detailed comparative study of experimental results with the theoretical models and various major evaluations has been presented. [ABSTRACT FROM AUTHOR]

The cross sections of the 186W(n,γ) 187W, 182W(n,p) 182Ta, 154Gd(n,2n) 153Gd, and 160Gd(n,2n) 159Gd reactions were measured at the neutron energies 5.08 ± 0.165, 8.96 ± 0.77, 12.47 ± 0.825, and 16.63 ± 0.95 MeV. Standard neutron activation analysis technique and off-line γ ray spectrometry were used for the measurement and analysis of the data. Measurements were done in the energy range where few or no measured data are available. The results from the present work are compared with the literature data based on the EX FOR compilation. The experimental results are supported by theoretical predictions using nuclear modular codes TALYS 1.8 and EMPIRE 3.2.2. The predictability of different one-dimensional models available in TALYS 1.8 and Levden models in EMPIRE 3.2.2 were tested. A detailed comparison of experimental results with theoretical model calculations is made. [ABSTRACT FROM AUTHOR]

Cross sections for the 64Zn(n,2n)63Zn and 67Zn(n,p)67Cu reactions were measured at neutron energies of 17.10 ± 0.82 and 14.98 ± 0.81 MeV by neutron activation techniques and offline γ-ray spectroscopy method. Production of neutron was done by 7Li (p,n)\documentclass[12pt]{minimal}\usepackage{amsmath}\usepackage{wasysym}\usepackage{amsfonts}\usepackage{amssymb}\usepackage{amsbsy}\usepackage{mathrsfs}\usepackage{upgreek}\setlength{\oddsidemargin}{-69pt}\begin{document}$$(p,n)$$\end{document} reaction. The 27Al (n,α)\documentclass[12pt]{minimal}\usepackage{amsmath}\usepackage{wasysym}\usepackage{amsfonts}\usepackage{amssymb}\usepackage{amsbsy}\usepackage{mathrsfs}\usepackage{upgreek}\setlength{\oddsidemargin}{-69pt}\begin{document}$$(n,\alpha )$$\end{document}24Na reaction was used as a neutron flux monitor. Estimations were performed using TALYS-1.95 and EMPIRE 3.2.2 codes with default values. The measured cross sections were compared with literature data taken from the EXFOR database and the latest evaluated data from the ENDF/B-VIII.0, JEFF-3.3, JENDL-5.0 and TENDL-2021 libraries. [ABSTRACT FROM AUTHOR]

The compound Nb 3 Sn possess superconductivity at suitable temperatures, therefore, it is best suited to be used in the toroidal coils of superconducting magnets which holds the fusion plasma and confine it inside the reactor core. The neutron induced reaction cross-sections are required from threshold to 20 MeV for different isotopes of Tin (Sn). Since limited data is available for the reactions with the Sn isotopes. Therefore, we have optimized the (n, p) and (n, 2 n) reaction cross-sections for all possible Sn isotopes from threshold to 20 MeV with modified input parameters in the nuclear reaction modular codes EMPIRE-3.2.2 and TALYS-1.8. These codes account for the major nuclear reaction mechanisms, including direct, pre-equilibrium, and compound nucleus contributions. The present results from 116 Sn ( n , p ) 116 m In , 117 Sn ( n , p ) 117 In , 117 Sn ( n , p ) 117 m In , 118 Sn ( n , 2 n ) 117 m Sn , 120 Sn ( n , 2 n ) 119 m Sn and 124 Sn ( n , 2 n ) 123 m Sn reactions calculated with nuclear modular codes: TALYS – 1.8, EMPIRE – 3.2.2 were compared with EXFOR data, systematics proposed by several authors and with the existing evaluated nuclear data library ENDF/B-VII.1, as well. The results from the present study can be used for the future development of ITER devices as well as to upgrade the nuclear model codes. [ABSTRACT FROM AUTHOR]

Degradation of material properties under neutron irradiation generates a requirement for studying effects on materials in a fusion environment and optimizing radiation-resistant materials for future applications. In the present work, the durability of stainless steel (SS) alloy used in ITER-like fusion devices is studied. We have predicted the amount of radionuclides produced in the material upon neutron irradiation at various locations is determined using the ACTYS, neutron activation code, for a typical one-dimensional geometry of ITER-like fusion reactor. The ACTYS code is further used to determine the gas production from 55Fe, 59Ni, and other long-lived radionuclides in the material. To further stress the importance of gas production in fusion materials, a comparative study of gas production cross-sections as given in various standard data libraries is examined using TALYS-1.8 and is presented in the paper. • In the present work, our main aim is to highlight the importance of long lived radionuclides produced inside the fusion reactor environment after neutron irradiation from D-T plasma. • The various radionuclides will be formed through different pathways. These radionuclides will produce more problems in the damage. • Neutron induced reaction on radionuclides enhance the gaseous level of hydrogen and helium gases. Transmutation also leads to the production of gas molecules like hydrogen (H) and helium (He) through (n,p) and (n,α) reaction channels. • In the present work, the durability of stainless steel (SS316L(N)-IG) alloy used in ITER-like fusion devices is studied. For one dimensional geometry of ITER like reactor, the amount of radionuclides produced in the material upon neutron irradiation at various locations is determined using ACTYS, neutron activation code. • ACTYS is further used to determine gas production from 55Fe, 59Ni and other long-lived radionuclides in the material. To further stress the importance of gas production in fusion materials, a comparative study of gas production cross-sections as given in various standard data libraries are examined using TALYS-1.8 and is presented in the paper. • The amount of hydrogen, helium produced due to each radionuclides in Ã50–60 is also calculated. [ABSTRACT FROM AUTHOR]

The activation cross-sections of 93Nb(n,2n)92Nbm and 88Sr(n,2n)87Srm reactions were measured using the activation and off-line γ-ray spectrometry technique in the neutron energy range 13.97–20.02 MeV. The present measurements have been done at the neutron energies where there are deficiencies and scarcity in the reaction cross-section data. The neutron flux was determined using the 27Al(n,α)24Na monitor reaction. The γ-ray self-attenuation effect and the low energy background neutron corrections were done in this experiment. The results of the present work were compared with the previously published data and evaluated nuclear data libraries. Theoretically, the cross-section for 93Nb(n,2n)92Nbm and 88Sr(n,2n)87Srm reactions was predicted by TALYS-1.9 nuclear code and compared with the present results. • The cross-section of 93Nb(n,2n)92Nbm and 88Sr(n,2n)87Srm reactions were measured in 13.97 to 20.02 MeV neutron energy range. • The comparisons of present results were done with previously published data, evaluated data and theoretically predicted data by TALYS-1.9 code. • The experimental results are in good agreement with the previously published data and are following trend line of evaluated data. • This work filled the gap in cross-sections data for both the reactions. [ABSTRACT FROM AUTHOR]

The production cross sections for the medical isotope 99Mo were measured using the 100Mo(n,2n) reaction at average neutron energies of 10.95±0.45, 13.97±0.68, 16.99±0.53, and 20.00±0.58 MeV using offline γ-ray spectroscopy. The neutron beam was generated using the natLi(p,n) reaction at the 14UD BARC-TIFR Pelletron facility, Mumbai, India. The uncertainties in the measured data were calculated using covariance analysis. The experimental results were compared with the existing data libraries ENDF/B-VII.1, JENDL-4.0, JEFF-3.2, and CENDL-3.1. The present results were also reproduced and compared with the theoretical nuclear model codes talys-1.9 and alice-2017 using different input descriptions. A modified set of parameters were also used in talys-1.9 to fit the existing data more accurately. The present findings show a good agreement with the theoretical predictions and with the existing experimental results from the different data libraries. In addition to this, a comparative study have also been performed between the literature data of (n,p), (n,2n), and (n,a) reaction cross sections at 14.5 MeV using various systematic formulas. The outcomes of the present work are important for the advancement of medical accelerators, dosimetry, and fast reactor systems. [ABSTRACT FROM AUTHOR]

The cross sections for the 232Th(n,γ) reaction have been measured in the neutron energy range of 11-19 MeV by using activation and offline γ-ray spectrometric techniques. The neutrons of desired energy were obtained by the 7Li(p,n) reaction using a proton beam of 13-21 MeV from the 14UD BARC-TIFR Pelletron facility at Mumbai, India. The 232Th(n,γ) reaction cross sections were calculated using the computer code talys-1.8. The uncertainties in the measurement have been studied using covariance analysis of the experimental data. The results from the present work have been compared with the evaluated data of ENDF/B-VII.1 and JENDL-4.0 as well as the theoretically calculated values based on talys-1.8, and were found to be in good agreement. The outcome of the present work is important for the development of future fast reactors and accelerator driven subcritical systems (ADSs). [ABSTRACT FROM AUTHOR]