Multi-resolution analysis for determining control rod drop time

This study describes a multi-resolution analysis (MRA) to determine the onset and end drop times of control rods. The measurement test of the drop times of control rods is normally performed during the start-up test of each reactor cycle since it is a crucial safety function to guarantee the reactor safe shutdown. The MRA with wavelet transform is particularly useful in analyzing the onset transients of rod drop as a means of capturing the unique attributes of such signals in an efficient way. This approach also allows the automated determination of rod drop time which reduces the uncertainty induced by ad hoc heuristic approaches. The test signal is generated by adding the random noise measured from real rod drop tests subtracting the wavelet-filtered noise free signal from the noisy signal leaving the noise. The signal is similar to both high sharp spikes noise and sine wave noise from the real voltage trace generated during the rod drop test. The effectiveness of the method is demonstrated by the MRA process.     

Methodology for a numerical simulation of an insertion or a drop of the rod cluster control assembly in a PWR

In a pressurised water reactor, the rod cluster control assembly is a system which controls the neutronic activity of the core. It consists of long rods, connected by a spider fixture and a cylindrical system for the control drive mechanism. At its withdrawn position, the activity of the core is maximum, and at its completely inserted position, the activity of the core vanishes. In case of emergency, an effective way to shutdown the reactor is to let it drop under its own weight. An other way to verify the efficiency of the rod cluster control assembly is the insertion test. It consists in inserting the rod into its guides and in checking if the reaction friction force is not high enough to block the movement of the rod cluster control assembly.We present in this paper a methodology for a numerical simulation of an insertion or a drop of the rod cluster control assembly into its guides (discontinuous and continuous guides, guide thimble). A numerical model is elaborated in which many loads are taken into account: fluid load, gravity and friction force between the rod and the guide. The numerical results are compared to experimental measurements obtained from a full-scale structure. A good agreement between the calculated and the measured data is observed.The numerical model takes into account the possible deflection of the guide. It shows clearly that the friction force cannot be neglected when the guide is bowed. So one can locate a faulty guiding system by examining the reaction force during the insertion test. Then, the numerical model can help the decider to make his choice among different rod cluster/fuel assembly components.     

Control rod drop hydrodynamic analysis for a pressurized water reactor

During operation of nuclear power reactors, reactivity initiated accidents can take place such as a control rod drop. If this occurs, the reactivity increases significantly and leads to an enhancement in power, fuel temperature and damage of reactor eventually. Exact assessment of these accidents depends on the hydrodynamic information. In this research, it is tried to simulate the unsteady flow field around the control rod for a pressurized water reactor power plant. In order to simulate the flow field around the control rod inside the guide tube, averaged Navier–Stokes equations accompanied by the layering dynamic mesh strategy have been used. The information exchange between the two computational stationary and moving grids, the computational grid around the control rod and the grid next to the guide tube, has been taken place through the interface. It was concluded that the time duration of control rod to reach the bottom of the core depends on the leakage. It was also observed that the velocity and acceleration of the control rod would be reduced by decreasing leakage flow rate and in certain leakages, the acceleration of the control rod approaches zero due to equilibrium conditions. During this research, a correlation based on the achieved data was proposed which would provide useful information on the relation between the leakage and the time for control rod to reach the bottom of the core.     

Pressure drop for parallel flow through a roughened rod cluster

A method is described of calculating the pressure drop for parallel flow through rod clusters with artificial surface roughnesses in order to improve the heat removal. The method allows the extensive experimental data on artificial roughness to be applied to rod clusters. The method is based on universal flow parameters for roughness and on general laws describing the pressure drop in non-circular channels with rough walls. The method is tested on the basis of measured results obtained from a rod cluster with 19 rough rods. The parameters determined are in excellent agreement with data measured in rough tubes and annular gaps. In this way it is possible, given the geometrical shape of the roughness elements, to calculate pressure drops and flow distributions in artificially roughened rod clusters contained in a smooth channel. The simplistic approach using an average friction factor and an overall hydraulic diameter results in rather unrealistic flow distributions and hence temperature distributions across the cluster.     

A simple mathematical model of the core criticality control by burnable poisons

A simple mathematical model is proposed and developed for the core criticality control by burnable poisons (BPs) distributed only throughout the peripheral region of the core while its central region remains free from BPs. The numerical burnup calculations confirm the effectiveness of the considered BP distribution for the criticality control of nuclear reactors.     

Pressure drop and heat transfer in gas-cooled rod bundles

Extensive experimental and analytical investigations of fluid flow and heat transfer in gas-cooled rod bundles have been carried out. Different bundle geometries with partially or fully roughened rod surfaces were tested in a carbon dioxide loop. An advanced and comprehensive measuring control and instrumentation are important design features of this experiment. Comprehensive thermal hydraulic subchannel analysis computer codes have been developed in order to assist fuel element design calculation for gas-cooled reactors. The experiments, codes and their verification procedure are described and the results of comparisons between measured and calculated pressure and temperature distributions are given.     

Wave propagation visualization in an experimental model for a control rod drive mechanism assembly

Nondestructive inspection techniques such as ultrasonic testing, eddy current testing, and visual testing are being developed to detect primary water stress corrosion cracks in control rod drive mechanism (CRDM) assemblies of nuclear power plants. A unit CRDM assembly consists of a reactor upper head including cladding, a penetration nozzle, and J-groove dissimilar metal welds with buttering. In this study, we fabricated a full-scale CRDM assembly mock-up. An ultrasonic propagation imaging (UPI) method using a scanning laser ultrasonic generator is proposed to visualize and simulate ultrasonic wave propagation around the thick and complex CRDM assembly. First, the proposed laser UPI system was validated for a simple aluminium plate by comparing the ultrasonic wave propagation movie (UWPM) obtained using the system with numerical simulation results reported in the literature. Lamb wave mode identification and damage detectability, depending on the ultrasonic frequency, were also included in the UWPM analysis. A CRDM assembly mock-up was fabricated in full-size and its vertical cross section was scanned using the laser UPI system to investigate the propagation characteristics of the longitudinal and Rayleigh waves in the complex structure. The ultrasonic source location and frequency were easily simulated by changing the sensor location and the band pass filtering zone, respectively. The ultrasonic propagation patterns before and after cracks in the weld and nozzle of the CRDM assembly were also analyzed. Since this visualization method is not limited in the flat cross section, it will be useful in developing ultrasound-based structural health monitoring technologies, advanced nondestructive methods, and numerical models. In addition, the proposed laser UPI system could be a useful tool in optimizing the receiver and transmitter locations, the ultrasonic path, and the ultrasonic frequency.     

Uncontrolled withdrawal of a control rod without scram

In the present work the thermal-hydraulics of reactivity-induced transients in low enriched uranium (LEU) core of a typical material test research reactor (MTR) are analyzed using the previous program developed by Khater et al. The analysis was done for uncontrolled withdrawal of a control rod with scram-disabled conditions. Initiating reactivity events with and without the influence of reactivity efficiency curve (“S” curve) were considered. The results of the proposed transients are analyzed and compared with each other. In transient without the “S” curve influence, a high primary peak power of 406.18 MW is attained and a clad melt down takes place after 1.85 s. In the transient with the “S” curve influence, a high super prompt-critical situation is produced (1.762$ at 0.895 s) with a very high primary peak power of 801.05 MW at 0.912 s. Also, a fast clad melt down is resulted in the hot channel at 1.088 s and a stable film boiling is established. This study indicates that, compared to the application of linear reactivity curve, the application of the reactivity efficiency curve results in the prediction of higher peaks in power and temperatures (fuel, clad and coolant) with a fast clad melt down.     

Derivation of coupling coefficient from rod drop measurements in two-point reactors

A method is proposed to determine the reactivity coupling coefficient of a zero-power coupled-core system, based on the control rod drop measurements. We derive a two-point version of the formula for a rod drop experiment. The formula is very simple and the experimental procedures are convenient as well as conventional rod drop measurements.

For experimental determination of the coupling coefficient, experiments were carried out in the UTR-KINKI reactor, a light-water-moderated and graphite-reflected reactor. The validity of the proposed method is demonstrated by close agreement in the coefficients between the present result and the previous ones from reactor-noise measurements.     

Pressure drop and velocity distribution in rod bundles with spacer grids

The paper contains experimental data and analysis of the pressure drop of turbulent flow through rod bundles. For laminar flow the dependence of the pressure drop on the pitch-to-diameter and wall-to-diameter ratios is discussed on the basis of theoretical analysis. In addition, correlations for the calculation of the pressure loss due to spacer grids are presented and compared with experimental data.Detailed measurements of the velocity distribution in a full bundle of 19 rods are compared with predictions for fully developed turbulent flow. Moreover, detailed measurements of the velocity distributions upstream and downstream of spacer grids typical for LMFBRs are discussed together with the mass flow separation and redistribution between the subchannels. The mass flow distribution found experimentally is compared with the predictions by a subchannel code. The status of experimental knowledge is shown.     

A model for predicting pellet-cladding interaction-induced fuel rod failure

A model for predicting pellet-cladding mechanical-interaction-induced fuel rod failure is presented. Cladding failure is predicted by explicitly modelling the formation and propagation of radial cladding cracks by the use of non-linear fracture mechanics concepts in a finite element computational framework. The failure model is intended for implementation in finite element fuel performance codes in which local pellet-clad interaction is modelled. Crack initiation is supposed to take place at pre-existing cladding flaws, the size of which is estimated by simple probabilistic concepts, and the subsequent crack propagation is assumed to be due to either iodine-induced stress corrosion cracking or ductile fracture. The novelty of the outlined approach is that the development of cladding cracks which may ultimately lead to fuel rod failure can be treated as a dynamic and time-dependent process. The influence of complex or cyclic loading, ramp rates and material creep on the failure mechanism can thereby be investigated. The presented failure model has been incorporated in the ABB Atom transient fuel performance code. Numerical results from some applications of the code are used to illustrate the usefulness of the model.     

Investigation of axial power gradients near a control rod tip

Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations.     

Estimation of irradiated control rod worth

When depleted control rods are planned to be used in new core configurations, their worth has to be accurately predicted in order to deduce key design and safety parameters such as the available shutdown margin. In this work a methodology is suggested for the derivation of the distributed absorbing capacity of a depleted rod, useful in the case that the level of detail that is known about the irradiation history of the control rod does not allow an accurate calculation of the absorber’s burnup. The suggested methodology is based on measurements of the rod’s worth carried out in the former core configuration and on corresponding calculations based on the original (before first irradiation) absorber concentration. The methodology is formulated for the general case of the multi-group theory; it is successfully tested for the one-group approximation, for a depleted control rod of the Greek Research Reactor, containing five neutron absorbers. The computations reproduce satisfactorily the irradiated rod worth measurements, practically eliminating the discrepancy of the total rod worth, compared to the computations based on the nominal absorber densities.     

A simplified mathematical dynamic model of the HTR-10 high temperature gas-cooled reactor with control system design purposes

A simplified mathematical dynamic model of the HTR-10 high temperature gas-cooled reactor is developed based upon the fundamental conservation of fluid mass, energy and momentum. The model is formulated for coupling reactor neutron kinetics with reactivity feedback and reactor thermal-hydraulics. The reactor is nodalized to employ the lumped parameter modeling methodology, which is mathematically described by differential algebraic equations (DAEs). The developed model is implemented on a personal computer using the MATLAB/Simulink tool. A lot of numerical simulation experiments are investigated and discussed. The transient results show that the model can properly predict the reactor dynamics and can serve as the basis for the model-based control system design.     

秦山核电一期的棒控棒位系统数字化改造

秦山核电一期棒控棒位系统经数字化改造后,采用PLC(programmable logic controller)(现逻辑控制功能,建立人机界面对系统状态进行在线监测和授权参数修改,控制部分采用冗余配置,提高了系统的灵活性和可靠性;对模拟电路进行修改,采用故障双保持的电路设计;控制棒控制系统和棒位指示系统间通过冗余配置的...     

Analysis of flow distribution a PWR fuel rod bundle model containng A 90% blockage

An experimental investigation, covering a Reynolds number range from 2 × 10³ to 3.5 × 10⁴, was conducted to study the velocity and turbulence intensity distributions due to the presence of a blockage in an unheated 7 × 7 rod bundle. The blockage configuration, consisting of a 4 × 4 rod array, created a maximum flow area reduction of 90% in the central nine subchannels. The blockage sleeve length was 38.3 × rod diameter and the 90% blockage zone length extended for 16.4 × rod diameter. The results showed that upstream of the blockage, the flow was not influenced by the blockage until it reached the location where the inlet taper section of the swelling started. At the downstream end, the flow disturbance was extensive and persisted over a distance of about 83 rod diameters. Compared to the downstream velocity profiles, the turbulence intensity measurements however showed a faster recovery from the blockage influence. At the higher Reynolds number, velocity profiles calculated using the COBRA subchannel computer code compared consistently with the experimental data. The general flow behaviour of the various subchannels was reasonably well predicted. However, at low Reynolds number, due mainly to the frictional form loss calculation scheme in COBRA and uncertainty in the flow transition, the flow diversion due to the blockage to the surrounding unblocked subchannels was overestimated. The influence of the degree of recovery from the rod swelling on the flow was also studied using COBRA.     

A mathematical model for crevice corrosion under porous deposits

A model has been developed to investigate the corrosion of steels in a thin, narrow crevice formed between the metal surface and an oxygen-permeable, porous deposit. A thin electrolyte layer exists within the deposit, due to geochemical fluids dripping onto a deposit-covered surface or due to the adsorption of moisture by a hygroscopic deposit. Mass transfer by diffusion and ion migration is considered in both the electrolyte films inside and outside of the crevice. The main reactions considered are the anodic dissolution of the alloy substrate, hydrolysis of the alloying element cations, dissociation of water, and the cathodic reduction of oxygen, hydrogen ion, and water. Special attention has been given to the role of parameters connected with the porous layer (porosity, tortuosity, and the layer thickness) on the rate of crevice corrosion. It is shown that the cavity acts as an ‘electrochemical amplifier’ from the point of view of the concentration of aggressive anions that leads to increasing of corrosion rate and to a higher probability of pit nucleation within the crevice.     

Combination coding control rod position-indicating system

A new control rod position-indicating system for water reactors—the combination coding, control rod position-indicating system—is introduced in this paper. The sensor of the system consists of several grouped transformers and one coding bar which is composed of m lengths of magnetic core and q lengths of non-magnetic portion. As the control-rod is withdrawn, the magnetic cores of the coding bar pass the transformers and change their output which is related to the rod position. Furthermore, we describe the coding principle of the indicating system and its experiments. The conclusion shows that this system has many advantages, such as smart coding, simple structure and excellent reliability. Therefore, it is a good control rod position-indicating system for water reactors.