Activation cross section for the (n,2n) and (n,p) reactions on 103Rh, 48Ti and 52Cr from reaction threshold up to 25 MeV energy region.

Activation and off-line γ-ray spectrometric methods were used to measure the ground and isomeric state (n,2n) reaction cross section for 103Rh at two different neutron energies. The standard 27Al (n,α)24Na reference reaction was used to normalise neutron flux. The proton beam from the 14UD BARC-TIFR Pelletron facility in Mumbai, India, was utilised to create high-energy quasi-monoenergetic neutrons via the 7Li (p,n) reaction. Statistical model calculations including the level density, pre-equilibrium and optical potential model were performed using the TALYS (ver. 1.95) and EMPIRE (ver. 3.2.3) reaction codes. In addition, because of considerable discrepancies in measured data, the literature (n,p) reaction cross section of 52Cr and 48Ti targets were examined theoretically in the present work. The measured cross sections are discussed and compared with the latest evaluated data of the FENDL-3.2b, CENDL-3.2, TENDL-2019, JENDL-5.0, and ENDF/B-VIII.0 libraries, and experimental data based on the EXFOR compilation. The theoretical investigation of the (n,2n) reaction cross section was performed for the ground and isomeric state for the first time from reaction threshold to 25 MeV energies. The experimental data corresponding to the ground, isomeric state and isomeric ratio were reproduced consistently by the theoretical calculations. The present experimental results are good with certain literature data and theoretical values.

Estimation of production cross-sections, transmutation and gas generation from radionuclides (A ∼50-60) in fusion environment.

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.

Cross-section of (n,2n) reaction for niobium and strontium isotopes between 13.97 to 20.02 MeV neutron energies.

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.

Measurement of neutron induced 86Sr(n, 2n)85Sr reaction cross sections at different neutron energies.

The cross-sections for 86Sr (n, 2n)85Sr reaction are measured at neutron energies 19.44 ±â€¯1.02 MeV and 16.81 ±â€¯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.

Neutron capture cross-sections for 159Tb isotope in the energy range of 5 to 17 MeV.

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 ±â€¯0.165, 12.47 ±â€¯0.825, and 16.63 ±â€¯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.

Investigation of (n, p), (n, 2n) reaction cross sections for Sn isotopes for fusion reactor applications.

The compound Nb3Sn 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 20MeV 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, 2n) reaction cross-sections for all possible Sn isotopes from threshold to 20MeV 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 116Sn(n,p)116mIn, , 117Sn(n,p)117mIn, 118Sn(n,2n)117mSn,120Sn(n,2n)119mSn and 124Sn(n,2n)123mSn 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.

Spectrum average cross section measurement of 183W (n, p)183Ta and 184W (n, p)184Ta reaction cross section in 252Cf(sf) neutron field.

Neutron induced nuclear reactions are of prime importance for both fusion and fission nuclear reactor technology. Present work describes the first time measurement of spectrum average cross section of nuclear reactions 183W(n,p)183Ta and 184W(n,p)184Ta using 252Cf spontaneous fission neutron source. Standard neutron activation analysis (NAA) technique was used. The neutron spectra were calculated using Monte Carlo N Particle Code (MCNP). The effects of self-shielding and back scattering were taken into account by optimizing the detector modeling. These effects along with efficiency of detector were corrected for volume sample in the actual source-detector geometry. The measured data were compared with the previously measured data available in Exchange Format (EXFOR) data base and evaluated data using EMPIRE - 3.2.2.