[en] Counter current buoyancy driven flow (CCBDF) is a type of flow in which two streams flows in opposite direction due to driving force caused by density gradient between two fluids. To understand the nature of counter current flows, related instabilities, local perturbation at interface of two counter streams, experiment were carried out. Experiments were performed in a transparent tilted pipe having constant temperature source tanks at both ends. Studies were carried out by temperature measurement in the test section at various locations. Flow visualizations were carried out using solution of potassium permanganate as dye and flow velocity was measured using Particle Image Velocimetry (PIV). Experiments have been conducted for different inclination of pipe and temperature of hot fluid inside the lower tank of test setup. This paper gives the details of experimental setup, experimental techniques used and the results obtained. (author)

[en] In present work, we measured ⁷⁸Se(n, p)⁷⁸As cross section at 13.52±0.67 and 19.86±0.59 MeV neutron energies and estimates the theoretically from using TALYS-1.8 and EMPIRE-3.2.2 code

[en] The use of copper as a first wall material has been considered in reactor designs that have high thermal loads on the first wall or require a shield of high electrically conductive material surrounding the plasma to help stabilize its location. In the present work, the cross section of the ⁶⁵Cu(𝑛, 𝑝)⁶⁵Ni reaction was measured above 13 MeV incident neutron energies relative to the ²⁷Al(𝑛, 𝛼)²⁴Na reference reaction by activation and offline γ-ray spectrometric technique. The present results were compared with the literature data taken from the EXFOR database and evaluated data of the ENDF/B-VIII.0,JEFF-3.3, JENDL4.0/HE, CENDL-3.2, TENDL-2019 and FENDL-3.2 libraries. The statistical reaction TALYS (ver. 1.9) code were used for the cross section calculation using the different optical potential, level density and pre-equilibrium models

[en] The present paper contemplates the cross section data for isotopes of Ag for (n,2n) reaction. Nuclear reactor technology is the field where nuclear reaction cross-section possess prime importance. Silver plays very significant role in nuclear reactor, establishing itself in the form of control rods as it is capable enough to absorb neutrons without undergoing fission. It is well known fact that control rods is one of the key component of nuclear reactor to control the fission rate of uranium and plutonium for the sustain operation of reactor. As the composition of control rods must be designed according to reactor's neutron spectrum, knowledge of interaction probability of composition material with energy range of neutron is quite important. Apart from this, theoretical predictions for nuclear reactions may help us to understand the nuclear reaction mechanisms and in general nucleon-nucleon interactions. Therefore neutron cross sections are playing a significant role in nuclear model developments. Here we have calculated neutron induced reaction cross-section for ¹⁰⁷Ag (n,2n)¹⁰⁶Ag at two energy points by Neutron Activation Analysis (NAA) and offline-γ ray spectrometry technique. These cross sections were calculated by nuclear model codes TALYS, EMPIRE and ALICE as well as compared with already available data on the EXFOR data libraries. The experimental details and theoretical calculations are discussed here

[en] Nuclear reactor in its operating stage through fission or fusion reaction, produces different radiation (α, β, γ, n, p, etc.). Among all, neutron and gamma are considered as a prime radiation corresponds to its energy and penetration. The kind and quantity of material which is used for shielding depend upon the factors like a type of radiation, the dose rate, intensity of radiations. Including these, the other factors such as cost, weight, fabrication and most important its availability in the country are one has to consider. On the above mention points, the studies of the absorption of radiation in materials have become an important issue and thus it is desirable to have the knowledge about the effective materials for neutron and gamma. For neutron shielding low atomic elements (Z) are used, mainly which have more hydrogen contain. In the present paper, comparison of previous theoretical data with experimental data for gamma-ray (⁶⁰Co) source is provided. The final results match with 2-5 % uncertainty

[en] Study of the neutron induced reactions are of immense interest in reactor applications. Nuclear reactor consists of structural materials, control rods, fuel and shielding materials, therefore, when neutrons originated from fusion or fission reaction interact with these materials, they change mechanical and physical properties of materials. It is necessary to have cross section data for these materials at all possible neutron energies. Lots of measured experimental nuclear data is available in the EXchange FORmat (EXFOR) library. In present work, we have studied ¹⁰³Rh(n,2n)¹⁰²Rh reaction at neutron energy of 20.12±12.03. The cross section was measured using neutron activation analysis (NAA) technique. The experimentally measured neutron cross-section data is also compared with evaluated nuclear data libraries from ENDF/BVII, JEFF 3.2. The measured cross section is important to enhance the nuclear data, as well as for nuclear reactor applications

[en] The cross sections for the reaction ¹⁶⁰Gd(n,2n)¹⁵⁹Gd have been measured experimentally using standard neutron activation analysis (NAA) technique at average neutron energies 10.72±0.95, 14.72±0.86 and 18.72±0.93 MeV. Neutrons were generated by ⁷Li(p,n) reaction and samples were irradiated for production of sufficient activity of the desired isotopes. The offline gamma-ray spectrometric technique has been employed for the measurement of the activity of the samples using HPGe detector. Calculation of flux and cross sections have been done with NAA technique with standard literature data of EXFOR and ENDF database. The experimentally measured cross sections were also calculated and compared using nuclear modular codes: TALYS-1.6 and EMPIRE-3.2.2. The result will be presented during the conference. (author)

[en] In the present work cross sections for the reaction ¹⁶⁰Gd(n,2n)¹⁵⁹Gd have been measured experimentally using standard neutron activation analysis (NAA) technique. The cross sections were also calculated and compared using nuclear modular codes: TALYS-1.8 and EMPIRE-3.2.2

[en] In fusion reactor different reaction channels i.e. (n,g), (n,p), (n,2n) were opened when 14 MeV neutrons interact with first-wall and different structural materials. The knowledge of neutron induced reaction cross-section of these materials is important for the study and operation of reactors. Strontium (Sr) is used as super conducting magnet material due to its superconducting properties. In present case, the reaction cross-section of ⁸⁸Sr(n,2n)^87mSr has been measured at an average neutron energies 13.97±0.68 and 16.99±0.53 MeV respectively. The results were also compared with the previously published data in EXFOR data library as well as evaluated data and theoretically predicated data by TALYS-1.95 code. In present experiment the ⁸⁸Sr(n,2n)^87mSr reaction cross-section was calculated with ²⁷Al(n,α)²⁴Na reference monitor reaction at an average neutron energies 13.97±0.68 MeV and 16.99±0.53 MeV. The comparison of present results were done with available literature data, evaluated data as well as theoretically predicated data by TALYS-1.95 as shown in the article. The results show that the cross-section data at 13.97±0.68 MeV was in good agreement with the previously measured data and data from JENDL/AD-2017 library. At 16.99±0.53 MeV, the measured cross-section of ⁸⁸Sr(n,2n)⁸⁷mSr reaction is lower than the data

[en] Several isotopes of silver are formed as fission products in reactor. Silver is also used as a control rod in fission nuclear reactors. Hence the neutron induced reaction cross sections data of Ag-isotopes are important for nuclear applications. In the present work, the ¹⁰⁷Ag(n,2n)^106mAg reaction cross sections within the neutron energies of 10.50±0.68, 13.52±0.67, 16.86±0.58 and 19.86±0.59 MeV were measured by using the activation and off-line γ-ray spectrometric technique. The ⁷Li(p, n)⁷Be reaction was used to produce the high energy neutrons with the proton beam from the 14UD BARC-Tata Institute of Fundamental Research Pelletron facilities at Mumbai, India. The γ-ray activity of the irradiated samples was measured by using a pre-calibrated HPGe detector. From the photo-peak activities of the γ-lines of ^106mAg, the cross-sections of ¹⁰⁷Ag(n,2n) reaction were estimated by using usual decay equation. Neutron flux was determined by using the monitor reactions ²⁷AI(n,α)²⁴Na and ¹¹⁵In(n,n)^115mIn. The uncertainties in the measured cross sections data were calculated using the covariance analysis. The measured data were compared with the available data in EXFOR and ENDF data library. Cross section for the ¹⁰⁷Ag(n,2n)^106mAg reaction was also theoretically calculated using computer codes TALYS-1.9 and EMPIRE-3.2.3 with different level density models. The experimental data are compared with the theoretical values based on both the codes. However, our data show good agreement with some of the previous experimental data and with the calculated values from TALYS-1.9. (author)