Mitul Abhangi

Investigation of X-ray spectral response of D-T fusion produced neutron irradiated PIPS detectors for plasma X-ray diagnostics

This paper describes the fusion-produced neutron irradiation induced changes in the X-ray spectral response of commercially available Passivated Implanted Planar Silicon (PIPS) detectors using the accelerator based D-T generator. After 14.1 MeV neutron irradiation up to a fluence of 3.6× 1010 n/cm2, the energy resolution (i.e. FWHM) of the detectors at room temperature is found to degrade by about 3.8 times that of the pre-irradiated value. From the X-ray spectral characteristics, it has been observed that the room temperature spectral response of PIPS detectors is too poor even at low neutron fluences. Irradiation is also carried out with Am-Be neutron source for studying the effect of scattered neutrons from the reactor walls on the detector performance. Comparative studies of the damage caused by 14.1 MeV neutrons and Am-Be source produced neutrons at the same neutron fluence are carried out by analyzing the irradiated detector characteristics. The degradation in the energy resolution of the detectors is attributed to the radiation induced changes in the detector leakage current. No considerable changes in the full depletion voltage and the effective doping concentration up to the neutron fluence of 3.6× 1010 n/cm2, are observed from the measured C-V characteristics. Partial recovery of the neutron irradiated detector characteristics is discussed.

Numerical Simulation of 14.1 MeV Neutron Irradiation Effects on Electrical Characteristics of PIPS Detector for Plasma X-Ray Tomography

Numerical simulation studies of D-T fusion produced neutron irradiation-induced changes in the electrical characteristics of commercially available Passivated Implanted Planar Silicon (PIPS) detectors (with ρ = 7.4 kΩcm) are carried out using a commercial device simulator for plasma X-ray tomography in nuclear fusion reactors. The proposed radiation damage model is validated by comparing with the experimental data of 14.1 MeV neutron irradiated PIPS detector for fluence up to 3.6 ×1010 n/cm2. The possible changes in the irradiated detector characteristics for higher neutron fluence levels are predicted. Furthermore, the simulated results for higher neutron fluence are validated by comparing with the reported experimental results. The probable deterioration in the X-ray energy response of PIPS detectors is analyzed from the changes in the electrical characteristics due to neutron irradiation. The simulated investigations are also repeated for detectors with different resistivity substrates (4 kΩ cm, and 300 Ωcm) and thin detector structures (100 μm). From the simulation studies, the possible use of the silicon detectors for the plasma X-ray tomography diagnostics is discussed.

Mechanical performance degradation of glass fiber insulation material after neutron irradiation

Epoxy based glass fiber reinforced plastic composites G-10 CR grade are suitable material for electrical insulation breaks of the superconducting fusion device. At IPR, under the national fusion program, characterization of G-10 CR material with the neutron irradiation dose is studied. This study will help us towards the indigenous development project of electrical insulation breaks for future superconducting fusion magnets under the radiation environment. The insulation material has been irradiated by 14 MeV neutrons D-T generator in the Fusion Neutronics Laboratory, I.P.R. The mechanical and electrical performance was investigated after irradiation (with a fluence of 8.63x1014 n/m2) at room and cryogenic temperature condition. The results show that the mechanical degradations (by ∼ 9 to 30%) after neutron irradiation in tensile strength and impact strength at 300 K and 77 K. In this paper, we present the experimental details of the neutron irradiation and the data have been compared using MCNP code for flux calculation on glass fiber insulation composite along with Stycast 2850 GT epoxy. Investigation of mechanical, electrical, SEM analysis for internal surface morphology, glass transition temperature measurement of insulation material and its test results at 300 K and 77 K before and after neutron irradiation will be presented.

Development of indigenous insulation material for superconducting magnets and study of its characteristics under influence of intense neutron irradiation

Epoxy based glass fiber reinforced composites are the main insulation system for the superconducting magnets of fusion machines. 14MeV neutrons are generated during the DT fusion process, however the energy spectra and flux gets modified to a great extent when they reach the superconducting magnets. Mechanical properties of the GFRP insulation material is reported to degrade up to 30%. As a part of R & D activity, a joint collaboration with IGCAR, Kalpakkam has been established. The indigenous insulation material is subjected to fast neutron fluence of 1014 - 1019 n/m2 (E>0.1 MeV) in FBTR and KAMINI Reactor, India. TRIM software has been used to simulate similar kind of damage produced by neutrons by ion irradiation with 5 MeV Al ions and 3 MeV protons. Fluence of the ions was adjusted to get the same dpa. We present the test experiment of neutron irradiation of the composite material (E-glass, S-glass fiber boron free and DGEBA epoxy). The test results of tensile, inter laminar shear and electrical breakdown strength as per ASTM standards, assessment of micro-structure surface degradation before and after irradiation will be presented. MCNP simulations are carried out for neutron flux, dose and damages produced in the insulation material.

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.

Effect of D-T Fusion-Produced Neutron Irradiation on X-Ray Spectral Characteristics of PIPS Detectors

D-T generator produced neutron irradiation induced changes in electrical and X-ray spectral characteristics of PIPS detectors are investigated. Changes in detector resolution and leakage current are analyzed. Partial recovery of detector characteristics is discussed.

Experimental Studies on the Self-Shielding Effect in Fissile Fuel Breeding Measurement in Thorium Oxide Pellets Irradiated with 14 MeV Neutrons

The 14 MeV neutrons produced in the D-T fusion reactions have the potential of breeding Uranium-233 fissile fuel from fertile material Thorium-232. In order to estimate the amount of U-233 produced, experiments are carried out by irradiating thorium dioxide pellets with neutrons produced from a 14 MeV neutron generator. The objective of the present work is to measure the reaction rates of 232Th + 1n → 233Th → 233Pa → 233U in different pellet thicknesses to study the self-shielding effects and adopt a procedure for correction. An appropriate assembly consisting of high-density polyethylene is designed and fabricated to slow down the high-energy neutrons, in which Thorium pellets are irradiated. The amount of fissile fuel (233U) produced is estimated by measuring the 312 keV gammas emitted by Protactinium-233 (half-life of 27 days). A calibrated High Purity Germanium (HPGe) detector is used to measure the gamma ray spectrum. The amount of 233U produced by Th232 (n, γ) is calculated using MCNP code. The self-shielding effect is evaluated by calculating the reaction rates for different foil thickness. MCNP calculation results are compared with the experimental values and appropriate correction factors are estimated for self-shielding of neutrons and absorption of gamma rays.