Sharif Digital Repository

An embedded time interval data acquisition system (DAS) is developed for zero power reactor (ZPR) noise experiments. The system is capable of measuring the correlation or probability distribution of a random process. The design is totally implemented on a single Field Programmable Gate Array (FPGA). The architecture is tested on different FPGA platforms with different speed grades and hardware resources. Generic experimental values for time resolution and inter-event dead time of the system are 2.22 ns and 6.67 ns respectively. The DAS can record around 48-bit × 790 kS/s utilizing its built-in fast memory. The system can measure very long time intervals due to its 48-bit timing structure... 

In this paper an approach based on system identification is used for fault detection in a nuclear reactor. A continuous-time Extended Kalman Filter (EKF) is presented, which allows the parameters of the nonlinear system to be estimated. Also a fault tolerant control system is designed for the nuclear reactor during power changes operation. The proposed controller is an adaptive critic-based neuro-fuzzy controller. Performance of the controller in terms of transient response and robustness against failures is very good and considerable  

The present research work aims to employ plasma focus in order to deposit molybdenum (Mo) on the 304 stainless steel substrate. The processing parameters were shot numbers as well as the distance of substrate from the anode tip. Stereo, atom force microscopy (AFM) and field emission-scanning electron microscopy (FE-SEM) equipped with energy dispersive X-ray spectroscopy (EDS) were used to study the deposited coatings. Microhardness measurements were also performed on the coatings. Results indicated that the plasma focus can be successfully applied to deposit Mo anode on the stainless steel substrate. The coatings contained discrete pores with sizes varying by processing parameters. The... 

In the present study, the development of a neutron noise simulator, DYN-ACNEM, using the Average Current Nodal Expansion Method (ACNEM) in 2-G, 2-D hexagonal geometries is reported. In first stage, the static neutron calculation is performed. The neutron/adjoint flux distribution and corresponding eigen-values are calculated using the algorithm developed based on power iteration method by considering the coarse meshes. The results of the static calculation are validated against the well-known IAEA-2D benchmark problem. In the second stage, the dynamic calculation is performed in the frequency domain in which the dimension of the variable space of the noise equations is lower than the time... 

Power reactor noise analysis is one of the most powerful tools in online monitoring and diagnostics of nuclear power reactors. Unfortunately, since such an analysis belongs to the non-linear “parametric excitation” realm, its theoretical aspects and relations have been mostly carried out after linearization. In this paper a general framework, i.e. the Ladder Expansion Method, is developed to convert such equations to a series of coupled linear equations, up to any desired accuracy. This method is then applied to the single mode random fluctuations of the absorption cross sections in a power reactor which is modelled by the multigroup diffusion equation with multiple delayed neutron groups. A... 

In this study, the SD-HACNEM (Sharif Dynamic - High order Average Current Nodal Expansion Method) neutron noise simulator in two energy groups using a second-order flux expansion method for two-dimensional rectangular X Y-geometry has been developed. In the first step, the calculations were performed for the steady state and results of ACNEM (Average Current Nodal Expansion Method) and HACNEM (High order Average Current Nodal Expansion Method) were examined and compared. To solve the problem, the power iteration algorithm has been used to calculate the distribution of neutron flux and neutron multiplication factor by considering the coarse-mesh (each fuel assembly one node). To validate the... 

Supercritical water reactor is one of the generation IV reactors which is basically a creative mixture of conventional PWRs and supercritical pressure steam boilers. Application of nanoparticles provides an effective way of improving heat transfer characteristics of conventional coolants; thus, utilization of a nanofluid coolant in the conceptual design of this reactors is quite reasonable and inevitable. Reactor coolant at supercritical pressure dose not experience any phase change and is heated up to 500 °C in three pass core design. In this paper, thermal–hydraulic analysis of applying a water base Al2O3 nanofluid with different nanoparticle mass fractions were investigated using a porous... 

The Lateral deformation of Fuel Assembly (FA) under the operational conditions of the reactor cores is called FA bowing. This phenomenon is caused by factors such as thermal and hydraulic loads on FAs in the reactor core. It can lead to disturbances in the movement of control rods inside of FAs, cross-contact of FAs in refueling, and also changes in power distribution. Changing the distance between the fuels along the assemblies due to bowing, leads to non-uniform distribution of water (coolant) around the FAs and results in neutronic perturbation. In this research, by developing a calculation model for the bowed FAs of the VVER-1000 reactor, based on the distribution of water around FA at... 

In normal operation conditions of nuclear power plants, the distribution of primary coolant between fuel channels would be considered almost uniform. When different number of Reactor Circulation Pumps (RCPs) are switched off, known as an abnormal condition, this uniform distribution is disturbed and different conditions occur for each channel depending on its position in the core. In this research, the normal and abnormal condition (with one or two tripped RCPs) for a VVER-1000/446 is investigated. For evaluation of the core neutronic calculations and thermal power distribution, USNRC's PARCS system code is employed. Then a thermal-hydraulics module was developed for performing the T/H... 

The time-dependent diffusion of neutrons in a spheroid as a function of the focal distance has been studied. The solution is based on an orthogonal basis and an extrapolation distanced related boundary condition for the spheroidal geometry. It has been shown that spheres and disks are two limiting cases for the spheroids, for which there is a smooth transition for the systems properties between these two limits. Furthermore, it is demonstrated that a slight deformation from a sphere does not affect the fundamental mode properties, to the first order. The calculations for both multiplying and non-multiplying media have been undertaken, showing good agreement with direct Monte Carlo... 

Small and medium break LOCA accidents at low pressure and under low velocity conditions have been studied in the TTL-2 Thermo-hydraulic Test Loop, experimentally. TTL-2 is a thermal hydraulic test facility which is designed and constructed in NSTRI to study thermal hydraulic parameters under normal operational and accident conditions of nuclear research reactors. A nodalization has been developed for the TTL-2 and experimental results have been compared with RELAP5/MOD3.2 results. The considered accidents are a 25% and 50% cold leg break without emergency core cooling systems. Results show good agreement between experiments and RELAP5/MOD3.2 results. This research provides experimental data... 

A two-channel time analyser data acquisition system is developed for analysis of stochastic processes of random time interval pulses. The system is implemented on a typical low cost FPGA device. Two stochastic processes of nuclear interactions can be recorded by the system independently without any inter-channel dead time behaviour. The experimental results without any hardware based data reduction are transferred to the computer to perform arbitrary post analysis of the data using powerful software engineering tools to estimate the statistical properties of the processes. The performance of the system is verified experimentally. The maximum time digitization period and the minimum channel... 

A correlation measuring tool for an endogenous pulsed neutron source experiment is developed in this work. Paroxysmal pulses generated by a bursts of neutron chains are detected by a 10-kbit embedded shift register with a time resolution of 100 ns. The system is implemented on a single reprogrammable device making it a compact, cost-effective instrument, easily adaptable for any case study. The system was verified experimentally in the Esfahan heavy-water zero power reactor (EHWZPR). The results obtained by the measuring tool are validated by the Feynman-α experiment, and a good agreement is seen within the boundaries of statistical uncertainties. The theory of the methods is briefly... 

The present research aims at depositing molybdenum on stainless steel substrate using dense plasma focus (DPF) device. The varying parameter was the angular position of substrate surface from the anode tip. The deposited layers were characterized using various analyses such as X-ray diffraction (XRD), scanning electron microscopy (SEM) equipped with energy dispersive X-ray spectroscopy (EDS), atomic force microscopy (AFM), and microhardness measurements. Microstructural observations revealed that the deposited coatings had rough surfaces containing holes as well as a stripe-featured structure covering the whole surface including the interior regions of the holes and the regions among them.... 

This paper proposes a new method for solving the time-dependent neutron transport equation based on nodal discretization using the MCNPX code. Most valid nodal codes are based on the diffusion theory with differences in approximating the leakage term until now. However, the Monte Carlo (MC) method is able to estimate transport parameters without approximations usual in diffusion method. Therefore, improving the nodal approach via the MC techniques can substantially reduce the errors caused by diffusion approximations. In the proposed method, the reactor core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates and leakage ratio are... 

A highly reliable tool for transient simulation is vital in the safety analysis of a nuclear reactor. Despite this fact most tools still use diffusion theory and point-kinetics that involve many approximation such as discretization in space, energy, angle and time. However, Monte Carlo method inherently overcomes these restrictions and provides an appropriate foundation to accurately calculate the parameters of a reactor. In this paper fundamental parameters like multiplication factor (K eff ) and mean generation time (t G ) are calculated using Monte Carlo method and then employed in transient analysis for computing the neutron population, proportional to K eff , during a generation time... 

Transient analysis which is vital in safety analysis requires a reliable calculation method. Most valid tools use diffusion theory with many approximations by now. However, the Monte Carlo method inherently overcomes these approximations and accurately calculates the parameters of a reactor. In this paper, a new time-dependent transport approach is described to simulate the nuclear reactor dynamic correctly using the MCNPX code. In this approach the fundamental parameters of a nuclear reactor like multiplication factor (K eff ) and mean generation time (t G ) are calculated using MCNPX code. They are then employed in the formulas to compute neutron population, proportional to K eff , during... 

In this paper a probabilistic methodology based on core nodalization is proposed to estimate the core power in the presence of xenon oscillation. A time-dependent Monte Carlo neutron transport code named MCSP-NOD is developed for dynamic analysis in arbitrary 3D geometries to simulate xenon oscillations as well as sub-critical condition with feedbacks. The new code is based on the approach adopted in MCNP-NOD which was previously introduced as a tool for core transient analysis using the MCNPX platform. As before, the core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates, leakage ratio are estimated using the MC techniques.... 

In this paper, a simpler approach compared to the existing approaches is developed to analyze nuclear reactor dynamics based on the explicit Monte Carlo method. A new population control method is also introduced to prevent neutron population growth and consequent computer memory shortages, which also increases simulation speed. The scheme is applied for time-dependent particle tracking in three-dimensional arbitrary geometries in the presence of feedbacks through a code named MCSP-Explicit. Changes in material density, as well as geometry dimensions, are also considered during simulation. MCSP-Explicit can be run with either continuous or multigroup data libraries, and it is further boosted... 

In this paper, a simpler approach compared to the existing approaches is developed to analyze nuclear reactor dynamics based on the explicit Monte Carlo method. A new population control method is also introduced to prevent neutron population growth and consequent computer memory shortages, which also increases simulation speed. The scheme is applied for time-dependent particle tracking in three-dimensional arbitrary geometries in the presence of feedbacks through a code named MCSP-Explicit. Changes in material density, as well as geometry dimensions, are also considered during simulation. MCSP-Explicit can be run with either continuous or multigroup data libraries, and it is further boosted...