Volume 79, Issue 5

To cope with increasing spent fuel disposals and limited domestic spent fuel storages in Korea, the Korea Atomic Energy Research Institute has developed an advanced sodium-cooled fast reactor for TRU transmutation with an electricity output of 600 MWe (called the KALIMER-600 TRU burner). The design philosophy of the KALIMER-600 TRU burner concept is highly focused on inherent safety mechanisms, i.e., passive responses to abnormal and emergency conditions, and thereby minimizes the need for engineered safety systems. Accordingly, the main concern is on the sodium coolant void reactivity, a very important safety parameter of the KALIMER-600 TRU burner. This study was therefore performed to analyze the sodium void reactivity of the KALIMER-600 TRU burner to the finest resolution possible, e.g., contributions from any isotope in each core region. Such detailed analysis could be valuable and applicable to further optimization of the core passive safety characteristics against severe coolant voiding accident conditions.

The Indian Pressurized Heavy Water Reactor (PHWR) core consists of a number of horizontal channels containing nuclear fuel bundles. Parallel coolant channels are connected to Inlet and Outlet header through feeder pipes. Coolant from Reactor Inlet Header is distributed to the coolant channels and after removing heat combines at Reactor Outlet Header. Due to space constraints the feeder pipes are joined to the channel with one or two elbows close to the end fittings of the coolant channels. The carbon steel feeder pipes carry high temperature fluid at higher velocity and are liable to undergo Flow Accelerated Corrosion (FAC). In the recent inspection it has been found that feeders having double elbow are more susceptible to FAC on the intrados of second elbow. But it was found that in some of the elbows maximum thinning due to FAC was observed on the intrados of the first elbow. Hence to resolve this, effect of first bend orientation with respect of upstream direction has been studied. Two different approaches are used for predicting the FAC rate from calculated value of wall shear stress by CFD. One method is based on evaluating of wear rate using Colburn analogy and the other using an empirical equation between wear rate and shear stress. In Colburn analogy, mass transfer coefficient is evaluated by knowing shear stress and equilibrium concentration. For a case study, wall shear stress obtained from k -∊ turbulence model was compared with k -ω SST turbulence model and no appreciable change in the wall shear stress has been found. Hence for subsequent analysis k -∊ turbulence model was chosen because large mesh size near to the surface (first layer thickness) is permitted due to higher y + value.

The instability occurring of the OTSG (Once-Through Steam Generator) in naval craft nuclear power plants is presented by the multivariable frequency domain theory. As concerning coupling interactions of the OTSG tubing, it is more accurate for analyzing the instability of OTSG compared to the common single variable method. A mathematical model for the system is derived from the fundamental equations by using the perturbation, Laplace-transform and the nodalization techniques. The stable boundary and parameters which influence the stability of the system are evaluated through computer simulation.

In the paper experimental modelling and investigation of the MSFR concept will be presented. MSFR is a homogeneous, single region liquid fuelled fast reactor concept. In case of molten salt reactors the core neutron flux and fission distribution is determined by the flow field through distribution and transport of fissile material and delayed neutron precursors. Since the MSFR core is a single region homogeneous volume without internal structures, it is a difficult task to ensure stable flow field, which is strongly coupled to the volumetric heat generation. These considerations suggest that experimental modelling would greatly help to understand the flow phenomena in such geometry. A scaled and segmented experimental mock-up of MSFR was designed and built in order to carry out particle image velocimetry measurements. Basic flow behaviour inside the core region can be investigated and the measurement data can also provide resource for the validation of computational fluid dynamics models. Measurement results of steady state conditions will be presented and discussed.

Recently thorium-based fuel matrixes are taken into consideration for nuclear waste incineration because of thorium proliferation resistance feature moreover its breeding or convertor ability in both thermal and fast reactors. In this work, neutronic influences of adding Am to (Th- 235 U)O 2 on effective delayed neutron fraction, reactivity coefficients and burn up of a fed CANDU core has been studied using MCNPX 2.6.0 computational code. Different atom fractions of Am have been introduced in the fuel matrix to evaluate its effects on neutronic parameters of the modeled core. The computational data show that adding 2% atom fraction of Am to thorium-based fuel matrix won't noticeably change reactivity coefficients in comparison with the fuel matrix containing 1% atom fraction of Am. The use of 2% atom fraction of Am resulted in a higher delayed neutron fraction. According to the obtained data, 32.85 GWd burn up of the higher Americium-containing fuel matrix resulted in 55.2%, 26.5%, 41.9% and 2.14% depletion of 241 Am, 243 Am, 235 U and 232 Th respectively. 132.8 kg of 233 U fissile element is produced after the burn up time and the nuclear core multiplication factor increases in rate of 2390 pcm. The less americium-containing fuel matrix resulted in higher depletion of 241/ 243 Am, 235 U and 232 Th while the nuclear core effective multiplication factor increases in rate of 5630 pcm after the burn up time with 9.8 kg additional 233 U production.

9–12% Cr ferritic steel structures at atomic scale were modeled by LAMMPS (Large-scale Atomic/Molecular Massively Parallel Simulator) Molecular Dynamics package with high accuracy. Embedded-Atom Model (EAM) potential parameters were applied for Fe–Fe, Fe–Cr and Cr–Cr atomic interactions. Nuclear reactor temperature conditions were used in the simulations. Heat flux, kinetic energy, potential energy, total energy, pressures, and atomic displacements of Fe–Cr steels including 9%, 10%, 11%, and 12% Cr were given.

In this work the authors solved the steady state neutron diffusion equation for a multi-layer slab assuming the multi-group energy model. The method to solve the equation system is based on an expansion in Taylor Series resulting in an analytical expression. The results obtained can be used as initial condition for neutron space kinetics problems. The neutron scalar flux was expanded in a power series, and the coefficients were found by using the ordinary differential equation and the boundary and interface conditions. The effective multiplication factor k was evaluated using the power method. We divided the domain into several slabs to guarantee the convergence with a low truncation order. We present the formalism together with some numerical simulations.

In this paper we implement the local analytical solution technique to the problem of heat transfer in axisymmetric annulus geometry with internal heating element. This method has shown to be very accurate in estimating the temperature field for axisymmetric problems even for coarse mesh. It is shown that this method reduces to the analytical solution for unidirectional heat transfer in the radial direction in homogeneous media. The technique is based on finding an analytical expression for the temperature field in the radial direction within each grid cell. This means that the temperature field in each cell is allowed to change in a nonlinear fashion along the radial direction. We compare this technique with the traditional finite volume technique and show that; with only few cells in the radial direction, this technique arrives at the mesh-independent solution quite accurately whereas it required denser mesh to arrive closer to this solution using traditional techniques. This method is proposed to the 1D codes that are currently being used to simulate thermalhydraulic characteristics of reactor systems. Furthermore, we also implement the experimental temperature field algorithm in which the governing equations are approximated for each cell as it would without extra manipulation to the governing equations. This technique is very simple and separates the physics from the solving part.

Synthetic nano-structure hydroxyapatite samples were doped by carbon in furnace via exposing to CO 2 circulating gas in different time of 30, 120, and 360 min under 100°C and 400°C of temperatures. FTIR, TEM and XRD analyses results confirmed the doping of carbonate group to the hydroxyapatite structure. The samples were irradiated in absorbed doses of 2, 5, 10, 20, 50, and 80 kGy using a 60 Co γ-ray source and subjected to EPR measurement, subsequently. The EPR signal intensities were constructed as a function of radiation dose, were compared with the results of raw sample and were studied from dosimetric point of view. The obtained results show a considerable increment in EPR signal intensities of the samples which were treated at 400°C. At the doses less than 20 kGy the carbon doped samples show a better EPR respond. This is much better for doses below 5 kGy in comparison to the raw samples and the response is quite linear.

In this paper, an investigation on the dependence of the effective multiplication factor, k eff , on moderator temperature for various thicknesses of the upper beryllium reflector in reactor conditions with different fuel burnups for the Miniature Neutron Source Reactor is carried out. Based on the linear dependence of k eff on moderator temperature, an approach to calculate the moderator temperature coefficient of reactivity, α T , at different temperatures and its average value, α , in a range of temperatures directly through the moderator temperature is developed. Calculations are performed to evaluate the effect of change in the upper reflector thickness on the moderator temperature coefficient of reactivity for the fresh core and reactor conditions with different fuel burnups. Calculated results indicate that α T increases with the increased beryllium thickness, but decreases with the increasing fuel burnup. Analysis of calculated results provides an additional insight into the relation of the upper reflector thickness, the neutron energy spectrum in the reactor core, and the moderator temperature coefficient of reactivity.