Volume 79, Issue 3

In the present feasibility study the system code ATHLET, which originates from nuclear engineering, is applied to a parabolic trough test facility. A model of the DISS (DIrect Solar Steam) test facility at Plataforma Solar de Almería in Spain is assembled and the results of the simulations are compared to measured data and the simulation results of the Modelica library “DissDyn”. A profound comparison between ATHLET Mod 3.0 Cycle A and the “DissDyn” library reveals the capabilities of these codes. The calculated mass and energy balance in the ATHLET simulations are in good agreement with the results of the measurements and confirm the applicability for thermodynamic simulations of DSG processes in principle. Supplementary, the capabilities of the 6-equation model with transient momentum balances in ATHLET are used to study the slip between liquid and gas phases and to investigate pressure wave oscillations after a sudden valve closure.

The objective of the paper is to measure the velocity profiles of water based nanofluids for flow through channels in order to understand whether the nanofluids behave Newtonian. For this purpose, experiments were carried for flow through a rectangular channel in laminar regime. Four different nanofluids were used, i. e. Al 2 O 3 , CuO, TiO 2 and SiO 2 with base fluid as water. Experiments were conducted at low concentration of these particles. The velocity profiles were measured using Particle Image Velocimetry. The results indicate that the velocity profiles are similar for all the fluids indicating the flows to be Newtonian.

A common characteristic of many industrial two-phase natural circulation systems is the presence of a large number of parallel boiling channels. Sensitivity of the steady state behavior of such a two-phase natural circulation system to different system parameters has many implications vis-à-vis performance of the system as per the design intent under various operating conditions. Experimental studies were carried out, to study the characteristics of a low pressure two-phase natural circulation system having ten transparent parallel channels with their individual heat sources. In view of its particular significance in nuclear industry, a special system condition with zero power in one of the parallel channels was also studied. The flow in the unheated channel and that in the heated channels were observed to be strongly interdependent in their sensitivity to downcomer resistance. A simple theory based on electrical analogy has been developed to predict the hydraulic behavior in such a system. Agreement between the theoretical predictions and the experimental observations has been found to be good. This article describes the experimentally observed phenomenon, the electrical analogy and the comparison between the experimental data and the analogical predictions.

In this study, the effect of (n,γ) reactions in spent fuel-grade (SFG)-PuO 2 , UO 2 , NpO 2 and UCO contents was investigated in a designed hybrid reactor system. In this system, the molten salt-heavy metal mixtures 93 – 85 % Li 20 Sn 80 + 5 % SFG-PuO 2 and 2 – 10 % UO 2 , 93 – 85 % Li 20 Sn 80 + 5 % SFG-PuO 2 and 2 – 10 % NpO 2 , 93 – 85 % Li 20 Sn 80 + 5 % SFG-PuO 2 and 2 – 10 % UCO were used as fluids. Beryllium (Be) is used as neutron multiplier by (n,2n) reactions. Thence, a Be zone with a width of 3 cm was used in order to contribute on fissile fuel breeding between the liquid first wall and blanket. (n,γ) reactions were calculated in the liquid first wall, blanket and shield zones, which SFG-PuO 2 , UO 2 , NpO 2 and UCO contents. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code and nuclear data library ENDF/B-VII.0.

In part two of the present paper, neutronic properties of the pool-type research reactor core are used to assess the similitude laws derived for out-of-pile modelling of the fuel element. The benchmark reactor used for this purpose is an IAEA 5 MW thermal pool-type research reactor currently in operation. The neutronic properties analysis are based on typical 2 200 m/sec and neutrons having 0.025 eV energy. The non-leakage capability of the system is estimated in terms of diffusion length. Also the slowing down power and the moderating ratio of the modelled methanol coolant are calculated in terms of lethargy of the diffusing medium. It is shown that the Iron which is substituted for Aluminium cladding is a relatively low absorber of neutrons but has a high neutron leakage. Methanol which replaced ordinary water as coolant is not a suitable coolant due to high neutrons absorbing substance. It is concluded that although Iron as a cladding material and methanol as a coolant meet the modelling out-of-pile criteria but are not satisfying neutronic properties. Therefore, use of them as a model clad and coolant are not suggested for research reactors.

In this paper, we introduce a novel 60 Co-ray-irradiation-based one-step synthesis method of Pt-reduced graphene oxide composites (Pt-RGO) in acid aqueous solution. The compositional distribution of the particles in the samples was characterized by transmission electron microscopy. The structure and composition of the nanocomposite has been determined with a scanning electron microscope (SEM) equipped with an energy dispersion X-ray (EDS) analyzer. Surface enhanced Raman scattering (SERS) of graphene deposited by the Pt nanoparticles were investigated with the 514.5 nm excitation. It was found that small-sized and highly-dispersed Pt nanoparticles could easily grow on the RGO surface under acidic conditions. In addition, the obtained homogeneous dispersions exhibit long-term stability, which will facilitate the production of homogeneous composites.

This work presents the Hamming method generalized corrector for numerically resolving the differential equation of delayed neutron precursor concentration from the point kinetics equations for reactivity calculation, without using the nuclear power history or the Laplace transform. A study was carried out of several correctors with their respective modifiers with different time step calculations, to offer stability and greater precision. Better results are obtained for some correctors than with other existing methods. Reactivity can be calculated with precision of the order h 5 , where h is the time step.

The core cooling of upward flow MTR pool type Research Reactor (RR) at the later stage of pump coast down is experimentally handled to clarify the effect of some operating parameters on RR core cooling. Therefore, a test rig is designed and built to simulate the core cooling loop at this stage. The core is simulated as two vertical channels, electrically heated, and extended between upper and lower plenums. Two elevated tanks filled with water are connected to the two plenums. The first one constitutes a left branch, connected to the lower plenum, and is electrically heated to simulate the core return pipe. The second one constitutes the right branch, connected to the upper plenum, and is cooled by refrigerant circuit to simulate the reactor pool. Channel coolant and wall temperatures at different power and branch temperatures are measured, registered and analyzed. The results show that at this stage of core cooling two cooling loops are established; an internal circulation loop between the channels dominated by the difference in channel's power and an external circulation loop between the branches dominated by the temperature difference between branches. Also, there is a double inversion in core flow, upward-downward-upward flow. This double inversion increases largely the channel's wall temperature. Complementary safety analysis to evaluate this phenomenon must be performed.

Reactor kinetic parameters are very important for safety analysis in power reactors. This paper focused on calculations of kinetic parameters such as the effective delayed neutron fraction (β eff ), the neutron generation time (Λ) and the prompt neutron lifetime l p for PWR. In this paper the burn-up calculations were performed to calculate kinetic parameters as a function of burn-up for a PWR when fueled with standard UO 2 fuel and MOX (Mixed-Oxide) fuel in order to study the impact of burn-up on these parameters. The calculations were performed using the Monte Carlo code MCNPX. The core chosen for the simulation is based on a 4-loop Westinghouse PWR power plant. The calculations were performed from beginning of the life to the end of life 210 days and compared with safety values taken from the design control document of an AP1000 PWR. The results are in agreement for UO 2 but not for MOX until end of cycle (EOC).

The diffusion lengths for one-speed neutrons spreading in a uniform homogeneous slab with backward, forward and linear anisotropic scattering are calculated by using first order approximation of the Chebyshev polynomials of first kind, i. e. T 1 method. The method is applied to neutron transport equation by expanding first the neutron angular flux in terms of the Chebyshev polynomials of first kind, and then diffusion lengths for one-speed neutrons are determined for various values of the collision and anisotropy parameters. Numerical results obtained from the present method are tabulated in tables together with the ones already presented in literature.

This paper describes a new system that has the ability to reduce background components to apply for environmental studies. This system uses gamma spectrometry with semi-empirical self–absorption correction and practical method for routine measurements of the mass activity radionuclides at low energy such as 210 Pb (46.5 keV), 234 Th (63.3 keV) and 226 Ra (186.2 keV). The reliability and precision of proficiency test must pass for final scores all the analytical determinations of received “acceptable” for all radionuclides. Our work shows an experiment developed for the calculation of self-absorption correction in that case that the sample's chemical composition is unknown.

In this paper the Hamiltonian theory of particle motion has been applied for developing the motion equations of electrons in linear accelerator of Iris-loaded waveguides. Using J. C. Slater asumption for determining electric field in Oz direction, the energy increase of electron in the guide wave pipe following the linacs resonance cavity with circulated electromagnetic distribution and repeat-cycle of given number of resonance cavities has been developed. The energy gain of electron following the electron way in Oz axle direction of the accelerator with the different injection phase and phase shift of RF has been obtained. The results indicate that the energy increase of electron depends on the injection phase of RF and cell-to-cell phase shift.

Age-dependent effective doses for external exposure to photons emitted by radionuclides uniformly distributed in air are reported. The calculations were performed for 160 radionuclides, which are important for safety assessment of nuclear facilities. The energies and intensities of photons emitted from radionuclides were taken from the decay data DECDC used for dose calculations. The results are tabulated in the form of effective dose per unit concentration and time (Sv per Bq s m −3 ) for 6 age groups: newborn, 1, 5, 10 and 15 years-old and adult. The effective doses for the adult are also compared to values given in the literature.