Rajnikant Makwana

Estimation of (n, p) and (n, α) Cross Section of Radionuclide 60Co for Fusion Technology Applications

Abstract For fusion application, there is a high demand for nuclear data for long-lived radionuclides produced in a neutron environment. Cobolt-60 (t1/2 = 5.3 years) is one of the radionuclides produced in a large amount inside the fusion reactor via different pathways. In this context, the excitation function of 60Co(n, p) and 60Co(n, α) reaction from threshold to 20 MeV has been calculated using TALYS-1.6 in the framework of the Hauser Feshbach statistical model along with preequilibrium effects. Outgoing (proton and alpha) particle energy spectra (dσ/dEp, dσ/dEα) and double-differential cross section (d2σ/dE dΩ) has also been estimated at 14 MeV incident neutron energy. Optimized input parameters used during the model calculation were determined by fitting the (n, p) and (n, α) cross sections to the experimental data for the adjacent stable nuclide 59Co. The activation analysis has also been carried out for 1 kg of stainless steel (SS316) using FISPACT-2007.

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 ± 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.

Measurement of 232Th and 238U neutron capture cross-sections in the energy range 5 to 17 MeV

The neutron capture cross sections of 232Th and 238U at the average neutron energies of 5.08 ± 0.17, 8.96 ± 0.77, 12.47 ± 0.83, and 16.63 ± 0.95 MeV have been measured by using the activation technique and off-line γ-ray spectroscopy. The 232Th and 238U were irradiated with neutrons produced from the 7Li(p, n) reaction using the proton energies of 7, 11, 15 and 18.8 MeV from the 14UD BARC-TIFR Pelletron facility in Mumbai, India. Detailed covariance analysis was also performed to evaluate the uncertainties in the measured cross-sections. The excitation function of the 232Th(n, γ) and 238U(n, γ) reactions were calculated using the theoretical model code TALYS-1.9. The experimental and theoretical results from the present work were compared with the ENDF/B-VII-1 and JENDL-4.0 nuclear data libraries and were found to be in good agreement.

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.

Measurement of (n, p) cross section for some structural materials at 14.2 MeV

The (n, p) reaction cross section for some structural materials such as 75As, 66Zn, 64Zn, 55Mn, 51V and 58Ni was measured at 14.2±0.2 MeV using activation and off line gamma ray spectroscopic technique. For the purpose of safe and economical design of reactors, the neutron cross section data for structural materials are required with high precession and accuracy. The neutron cross section data for important structural materials are collected and evaluated systematically and data files are prepared for the reactor design. However, considerably large discrepancies exist among different evaluated nuclear data files. Hence there is a need to study these reactions with better accuracy. The results were compared with existing data available in EXFOR data base. The measured cross sections were also estimated theoretically using nuclear modular codes: TALYS-1.6 and EMPIRE-3.2.2.