Volume 81, Issue 6

Main purpose of the study is justifying the use of molybdenum as a structural material of fuel elements for improving the safety of nuclear reactors. Particularity of the used molybdenum is that its isotopic composition corresponds to molybdenum, which is obtained as tailing during operation of the separation cascade for producing a material for medical diagnostics of cancer. The following results were obtained: A method for reducing the thermal constant of fuel elements for light water and fast reactors by using dispersion fuel in cylindrical fuel rods containing, for example, granules of metallic U-Mo-alloy into Mo-matrix was proposed; the necessity of molybdenum enrichment by weakly absorbing isotopes was shown; total use of isotopic molybdenum will be more than 50 %.

Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

Nanofluids have shown to be promising as an alternative for a PWR reactor coolant or as a safety system coolant to cover the core in the event of a loss of coolant accident. The nanoparticles distribution and neutronic parameters are intensively affected by the local boiling of nanofluid coolant. The main goal of this study was the physical-mathematical modeling of the nanoparticles distribution in the nucleate boiling of nanofluids within the viscous sublayer. Nanoparticles concentration, especially near the heat transfer surfaces, plays a significant role in the enhancement of thermal conductivity of nanofluids and prediction of CHF, Hide Out and Return phenomena. By solving the equation of convection-diffusion for the liquid phase near the heating surface and the bulk stream, the effect of heat flux on the distribution of nanoparticles was studied. The steady state mass conservation equations for liquids, vapors and nanoparticles were written for the flow boiling within the viscous sublayer adjacent the fuel cladding surface. The derived differential equations were discretized by the finite difference method and were solved numerically. It was found out that by increasing the surface heat flux, the concentration of nanoparticles increased.

In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTR_PC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

Lead, mercury, tantalum and tungsten were used as target material for calculation of spallation processes in an ADS-TRIGA RC1 reactor. The results show that tungsten has the highest neutron production rate. Therefore it was selected as target material for further calculations. The sensitivity of neutron parameters of the ADS reactor core relative to a change of beam profile and proton energy was determined. The core assembly and parameters of the TRIGA RC1 demonstration facility were used for the calculation model. By changing the proton energy from 115 to 1 400 MeV by using the intra-nuclear cascade model of Bertini (INC-Bertini), the quantity of the relative difference in % for energy gain (G) and spallation neutron yield (Y n/p ), increases to 289.99 % and 5199.15 % respectively. These changes also reduce the amount of relative difference for the proton beam current (I p ) and accelerator power (P acc ), 99.81 % and 81.28 % respectively. In addition, the use of a Gaussian distribution instead of a uniform distribution in the accelerator beam profile increases the quantity of relative difference for energy gain (G), net neutron multiplication (M) and spallation neutron yield (Y n/p ), up to 4.93 %, 4.9 % and 5.55 % respectively.

TRISO-wrapped fuel particle failure is a function of local temperature. The spherical symmetry and hence central temperature peak in PBM-type reactor fuel pebbles can lead to considerably higher failure rates at the center of the pebbles. The so-called “wallpaper fuel” concept has been shown to be a viable solution to this issue. In this paper, the neutronic aspects of this modification were studied using MCNP. According to the obtained results, the inclusion of a central fuel-free zone could lead to a more reactive system with lower actinide production while damping the temperature response of the reactor.

This paper presents a generic Loss of Flow Accident (LOFA) scenario module which is integrated in the LabView-based simulator to imitate a Nuclear Research Reactor (NRR) behavior for different user defined LOFA scenarios. It also provides analyses of a LOFA of a single fuel channel and its impact on operational transactions and on the behavior of the reactor. The generic LOFA scenario module includes graphs needed to clarify the effects of the LOFA under study. Furthermore, the percentage of the loss of mass flow rate, the mode of flow reduction and the start time and transient time of LOFA are user defined to add flexibility to the LOFA scenarios. The objective of integrating such generic LOFA module is to be able to deal with such incidents and avoid their significant effects. It is also useful in the development of expertise in this area and reducing the operator training and simulations costs. The results of the implemented generic LOFA module agree well with that of COBRA-IIIC code and the earlier guidebook for this series of transients.

Dual purpose cask technology is one of the most prominent options for interim storage of spent fuels following their removal from reactors. Criticality safety of the spent fuel assemblies are ensured by design of the basket within these casks. In this study, a set of criticality design calculations of a dual purpose cask for 12 VVER 1000 spent fuel assemblies of Bushehr nuclear power plant were carried out. The basket material of borated stainless steel with 0.5 to 2.5 wt% of boron and Boral (Al-B 4 C) with 1.5 to 40 wt% of boron carbide, were investigated and the minimum required receptacle pitch of the basket was determined. Using the calculated receptacle pitch of the basket, the minimum required diameter of the cavity could be established.

In this research work, the electrical behavior of polycarbonate-carbon nanotube composite, over the radiation absorbed dose under a fixed DC voltage was investigated via finite element method. The predicted electrical percolation threshold value in the composite was validated by experimental results published by other scientists. The absorbed dose value was considered as multiplying of heat capacity and temperature rise of the composite, regarding the calorimetric approach. Results show that this kind of composite can be applied for monitoring and radiation protection utilizations.

The goal of this study is to compare the thermoluminescence properties of nano and micro structure hydroxyapatite. Nano structure hydroxyapatite was synthesized via hydrolysis method, while the micro structure one was from Merck Company. X-ray diffraction and fourier transform infrared spectroscopy were used to determine the crystal structure and chemical composition of the hydroxyapatite samples. Particles sizes of each sample were estimated using Scherer equation and transmission electron microscopy system. Thermoluminescence properties of the samples were investigated under gamma irradiation. The glow curves of micro and nano structure samples show a peak at 150 °C and 200 °C, respectively. Thermoluminescence responses of both the samples were linear in the range of 25 – 1 000 Gy where, nano structure sample show a greater slope and stronger linearity in comparison to the micro sample. The results show that the thermoluminescence response of micro sample faded rapidly in comparison to the nano sample due to the existence of the peak at higher temperature.

During postulated accident sequences in nuclear reactors, hydrogen may get released from the core and form a flammable mixture in the surrounding containment structure. Ignition of such mixtures and the subsequent pressure rise are an imminent threat for safe and sustainable operation of nuclear reactors. Methods for evaluating post ignition characteristics are important for determining the design safety margins in such scenarios. This study presents two thermo-chemical models for determining the post ignition state. The first model is based on internal energy balance while the second model uses the concept of element potentials to minimize the free energy of the system with internal energy imposed as a constraint. Predictions from both the models have been compared against published data over a wide range of mixture compositions. Important differences in the regions close to flammability limits and for stoichiometric mixtures have been identified and explained. The equilibrium model has been validated for varied temperatures and pressures representative of initial conditions that may be present in the containment during accidents. Special emphasis has been given to the understanding of the role of dissociation and its effect on equilibrium pressure, temperature and species concentrations.

In this work, we report a solution to solve the Neutron Point Kinetics Equations applying the Polynomial Approach Method. The main idea is to expand the neutron density and delayed neutron precursors as a power series considering the reactivity as an arbitrary function of the time in a relatively short time interval around an ordinary point. In the first interval one applies the initial conditions and the analytical continuation is used to determine the solutions of the next intervals. A genuine error control is developed based on an analogy with the Rest Theorem. For illustration, we also report simulations for different approaches types (linear, quadratic and cubic). The results obtained by numerical simulations for linear approximation are compared with results in the literature.

One speed, time independent neutron transport equation for a slab geometry with the quadratic anisotropic scattering kernel is considered. The albedo and transmission factor are calculated by the modified F N method. The obtained numerical results are listed for different scattering coefficients.

To calculate the optimum reloading pattern in the second cycle of BUSHEHR VVER-1000 reactor, it was necessary to first calculate the burnup and fuel composition of each fuel assembly at the end of the first operational cycle so that the types of applicable fuel assemblies in the second cycle could be determined. Time-dependent calculations were performed using the WIMSD-5B and CITATION-LDI2 codes coupled with thermo-hydraulic computations during the first cycle. A weighting factor was defined for all 16 types of fuel assemblies in the second cycle based on their types and fuel compositions. In the process of calculating the deterministic reloading pattern, the number of probable arrangements was reduced from 7.55 × 10 22 cases to just 1 856 cases by using a weighting-factor method and the structural constraints of the reactor core. The thermo-neutronic parameters of each of the 1 856 arrangements were then calculated by coupling the WIMSD-5B and CITATION-LDI2 codes with the thermo-hydraulic program. Afterwards, the processes of screening the probable arrangements and searching for the optimal reloading pattern were carried out based on the values of these parameters.