Computational analysis of neutronic effects of ThO_2 rods loaded in CANDU 6 fuel assemblies

Thorium as a suitable fertile with higher natural resources in comparison with uranium resources has been remarkably considered by different nuclear energy user countries in the last decades. Its prominent features such as suitable possibility for power flattening of a nuclear reactor, applicable breeder blanket to produce^(233)U fissile as well as neutron leakage prevention from a nuclear core has caused its application as power flatter, breeder material or other aimed utilizations be evaluated by the researches. In the present study, neutronics of a modeled CANDU 6loaded with Th O_2 and UO_2fuel rods have been computationally studied. The study aimed at reprocessing of burned Th O_2 seeds at CANDU 6 reactor to recover the total produced uranium, which is to be going under another compound fuel cycle. The obtained results showed all the core reactivity coefficients are sufficiently negative. The modeled core 949 GWd burn-up concluding in 99.99 %depletion of^(235)U initial loads. 18.38 kg of^(233) U was produced in the burnt Th O_2 fuel after 1-year burn-up time. In addition, 31.84 kg of^(239) Pu was produced in the UO_2 spent fuel rods after the burn-up time. After a proposed cooling time, about 50.01 kg of^(233)U will be available in the spent Th O_2 fuel.

Design and Simulation of Solenoid for Buncher of a Linear Accelerator

One of the most important components of a linac is bunchcr. Throughout most of the buncher, the electrons are well forward of the crest and have velocities considerably less than light velocity, thus they arc in a region of radial dcfocusing and a considerable fraction of the beam will be lost unless defocusing action is counteracted by some other applied forces. The simplest way to do this is to sct up a longitudinal magnetic field which intcmcts with the radial motion of the electrons and causes them to follow helical orbits through the space occupied by the field. In this paper, five solenoids were designed to provide necessary magnetic field inside the bunchcr. Magnetic field was analytically calculated and compared with simulation results of Computer Simulation Technology （CST）. Because of resistance in wires, some amount of energy appears in form of heat, so heat power was calculated analytically and cooling system was designed for these solenoids by Analysis System （ANSYS）.

Calibration of Activation Counter by a 2.48 kJ Plasma Focus Device

In many experiments, plastic scintillators are used as a counter instead of Geiger Muller counter, because they are so fast and they are used for pulsed neutron sources. So a silver activation counter has been constructed by plastic scintillator plates along with silver foils for determination the neutron yield of a 2.48 kJ plasma focus device, SBUPF1. This counter was calibrated by source removal method with an Am-Be 5 Ci neutron source which was placed above the anode of plasma focus device. Deuterium gas up to 8 mbar pressure was injected to this device, and the neutron yield produced by pulsed D-D fusion of plasma focus device was measured by the counter. The neutron yield of SBUPF1 in 8 mbar pressure was obtained （3.71± 0.32）×10^7 neutrons per shot. This result has relatively agreed with the neutron yield measured by silver activation Geiger counters.

Dosimetry Methods of Fast Neutron Using the Semiconductor Diodes

Semiconductor detectors based on a silicon pin diode are frequently used in the detection of different nuclear radiations. For the detection and dosimetry of fast neutrons, these silicon detectors are coupled with a fast neutron converter. Incident neutrons interact with the converter and produce charged particles that can deposit their energy in the detectors and produce a signal. In this study, three methods are introduced for fast neutron dosimetry by using the silicon detectors, which are： recoil proton spectroscopy, similarity of detector response function with conversion function, and a discriminator layer. Monte Carlo simulation is used to calculate the response of dosimetry systems based on these methods. In the different doses of an 241Am-Be neutron source, dosimetry responses are evaluated. The error values of measured data for dosimetry by these methods are in the range of 15-25%. We find fairly good agreement in the 241Am-Be neutron sources.

Neutron Radiography System Collimator Design via Monte Carlo Calculation

Neutron radiography uses the unique interaction probabilities of neutrons to create images of materials. This imaging technique is non-destructive. MCNP Monte Carlo Code has been used to design an optimized neutron radiography system that utilizes 241Am-Be neutron source. Many different arrangements have been simulated to obtain a neutron flux with higher amplitude and more uniform distribution in the collimator outlet, next to image plane. In the final arrangement the specifications of neutron filter, Gamma-ray shield and beam collimator has been determined. Simulations has been carried out for a 5Ci 241Am-Be neutron source. In this case 43.8 n/cm2s thermal neutron flux has been achieved at a distance of 35 cm from neutron source.

Combined application of neutron activation analysis using IECF device and neural network for prediction of cement elements

Purpose Gamma spectrum of a cement sample is not straightforward to analyze as a result of peak overlapping produced by Compton effect of gamma rays radiated from activated elements in the neutron activation process and also because of the change into the neutron energy spectrum in the target sample during activation.Methods Artificial neural network(ANN)is an excellent solution for complex and nonlinear systems.Consequently,the use of ANN for quantitative analysis of major cement elements including Ca,Si,Al,and Fe would be very advantageous.Iranian inertial electrostatic confinement fusion device is a fast,monoenergetic and steady neutron generator.It was simulated as a high-energy neutron source for performing neutron activation analysis.In the present study,a library of 29 members of delayed gamma-ray spectra of knowing cement samples were generated via MCNPX version 2.7.Specific photo-peaks related to Ca,Si,Al and Fe obtained from these spectra were used as inputs for ANN(21 of them for training and 8 for testing).Then,using MLP architecture,an ANN model has been presented to model the system to predict the percentages of the elements in cement.The ANN model is optimized to have only a hidden layer with five neurons.Results The comparison between modeling data and results of proposed MLP network shows that there is a good consistency between them.The MAE of training set for Ca,Si and Fe outputs were 0.0351,0.0656 and 0.0660,respectively,and the MAE of testing set for mentioned outputs were 0.0997,0.3046 and 0.8699,respectively.Conclusion Overall,it can be concluded that the defined dispensable errors show that this modeling is a good one as a prediction tool for the mentioned purpose.