Effectiveness Of Control Rods Research Articles

Investigation of the combined effect of control rods and forced rotation on a cylinder

A novel structure combining the application of control rods and forced rotation on a cylinder is proposed based on the cylindrical vibration suppression, and the combined structure is numerically simulated at a low Reynolds number of 200, an attack angle of 0°–105°, and a rotation rate of 0−1. The vortex-induced vibration responses, fluid forces, and cylindrical wake evolution are analyzed, and the VIV suppression is compared and discussed. The results show that the merging of the vortex layers on the cylinder and control rods promotes cylindrical vortex shedding, causing a high amplitude cylinder response. The cylinder vibrates at a low amplitude for no vortex layer merging. Rotation causes increased directional sensitivity of the control rod to cylindrical amplitude suppression. A 98%-cylinder amplitude suppression can be achieved by combining the control rod and rotation, while only 60% can be achieved by the control rods or rotation alone, indicating that the combined structure is highly effective for amplitude suppression.

Advanced Multi-Modeling of PWR Dynamics and Deep Learning based Computational Tool in SIMULINK and LabVIEW

The reactivity monitoring, prediction, and investigation is the most important parameter to ensure the safety and reliable operation of a nuclear power plant. This parameter is gained further importance in Pressurized Water Reactor (PWR) due to more sophisticated reactivity insertion mechanisms and innovative reactor core fuel loading scheme. Based on deterministic internal and external dynamics and neutronics analysis of Advanced PWR, all the reactivity feedback effects such as Doppler effect, moderator effect, control rod effect, liquid boron effect and reactor poisons effect are investigated, modeled and stochastically optimized using deep artificial intelligence. Advance Pressurized Water Reactor (APWR) of 600 MWe rating (AP-600) is used as a reference reactor model and based on the dynamics of AP-600, an Advanced Pressurized Water Reactor Dynamics and Intelligent Stochastic Optimization based Deterministic Neutronics Simulation (APD-ISO-DNS) Code is developed in the hybrid SIMULINK andLabVIEW environments. AP-600 reactor model is fine-tuned and adjusted for 300 MWe PWR (P-300) and 1070 MWe Advanced Chinese PWR (ACP-1000) using neutronics parameters and operational dynamic data of operating PWR nuclear power plants in Pakistan. Various load reduction transient experiments are conducted and dynamic transient simulations of P-300, AP-600 and ACP-1000 are evaluated in SIMULINK and in LabVIEW environments and found as per design basis.

Efficient Neutron Irradiation Procedure for Uniform Dopant Distributions in Silicon Doping

Abstract Neutron transmutation doping (NTD) is one of the several techniques used for n-type doping with good homogeneity. NTD performances were investigated in this work within a project funded by the Academy of Scientific Research and Technology (ASRT), Egypt, for the “production of doped Silicon Wafers for Solar Panels and Basic Electronics,” 2015–2018. The Egypt second research reactor (ETRR-2) has two irradiation positions for silicon doping using NTD. The two positions can accommodate silicon ingots with large diameters up to six inches. One crucial parameter that affects the quality of the product is the distribution of the dopant in the irradiated ingot. The irradiation facility provides the option of rotating the sample during irradiation for flattening the radial distribution of the dopant. Since the neutron flux is not flat along the irradiation position due to leakage and the effect of control rods, the generated dopant will be not flat along the axial direction. To overcome this problem, this paper proposed an irradiation procedure to smooth the dopant distribution along the axial direction. Such procedure can be applied for NTD facilities that has not an acceptable axial neutron flux distribution in silicon doping process. The procedure is based on loading the irradiation position by two ingots staked end to end and, at the middle of the irradiation duration, the two ingots swap their places. This strategy was verified by MCNP5 calculations and experimentally.

Demonstration of power distribution estimation using ex-core detectors by reactor experiment at UTR-KINKI

ABSTRACT A novel core power distribution reconstruction method based on ex-core detectors is demonstrated at the Kindai University Reactor UTR-KINKI. Technology to monitor the interior of the core is critical for assuring nuclear safety. Typical core monitoring systems utilize in-core detectors. However, the harsh environment inside the core degrades detector maintainability and reliability. On the other hand, monitoring using ex-core detectors is difficult due to the lack of information of neutrons. A previous study, therefore, proposed a new power distribution reconstruction method using power correlation between fuel regions, PHOEBE. However, PHOEBE has not been verified experimentally. Thus, the present study demonstrates PHOEBE at UTR-KINKI. The experiment measures the neutron counts using four detectors outside the reflector, and the power distribution was found to be well reconstructed. Furthermore, the present study estimates the power distribution using only two detectors for confirming the effect of power correlation. When the control rod had a large impact on detectors, estimated power distribution was emphasized. By contrast, small control rod effect cases suggested the advantage of PHOEBE. However, a more complex experimental geometry, such as a multi-layered fuel region, is required to more clearly evaluate the effect of considering the power correlation between fuel regions.

The effect of control rods on the reactivity and flux distribution of BWR 4 bundle using MCNPX Code

This paper has three main objectives related to the neutronic and burnup analysis of the BWR (Boiling Water Reactor) Four-Lattice. The first objective is to provide partial validation of the MCNPX code for this lattice by comparing its results with Scale-5.1 results. Validation of the MCNPX to calculate effective multiplication factor and reactivity rod worth for the F-Lattice is provided. This is carried out in case of instantly removing the control blade and replacing it with a graphite moderator. Moreover, spatial neutron flux distributions using F-mesh card over the bundle and the control blade are investigated at inserting and withdrawing the B4C. The second objective is to perform parametric design studies of the F-Lattice. Areas of particular interest are the effect of increased or decreased blade width on the neutron flux throughout the bundle. It is found that the presence of carbon in the control blade at withdrawing the B4C makes the reactor supercritical, (K-eff = 1.22206). On the other hand, the use of B4C blade presents (K-eff = 0.93521). Consequently, the reactivity of 10% B4C thinner case is higher that of 10% B4C thicker. The simulation also showed that the B4C blade had an effective role in decreasing the thermal flux at the periphery of the bundle. This is contrast to the effect of carbon that moderates fast to thermal neutrons. The third part of this work aims at studying the burnup calculations using MCNPX code for 30 days burn with 1 day time step then for 20 months burn with 2 week time steps for the lattice. At the end of the work, it is very important to determine the most proper bundle model that achieves a prolonged fuel burn and flatting thermal flux distribution. For reaching this goal, three cases (B4C, 10% thinner of B4C and 10% thicker) are simulated by MCNPX code till 70 GWd/ton. It is found that the B4C and 10% thicker are the appropriate models that can satisfy the safety considerations of the Compact Modular Boiling Water Reactor.

Модель реактивнісної аварії РБМК-1000 енергоблока № 4 Чорнобильської АЕС

The article presents the results of modeling of the reactivity accident, which resulted in the destruction of reactor RBMK-1000 of the 4th power unit of the Chornobyl NPP on April 26, 1986. The RBMK-1000 reactivity accident model was developed on the basis of the kinetics of the nuclear reactor, taking into account the change in the reactivity of the reactor. Reactivity changes as a result of both external influence (movement of control rods; change in the reactor inlet coolant temperature (density)) and due to the action of reactivity feedback by the parameters of the reactor core (change in the fuel temperature, coolant temperature, concentration of 135Хе, graphite stack temperature, etc.). A similar approach was applied by the authors of the article for the study of transient processes with the operation of accelerated unit unloading mode on VVER-1000, and the validity of such model is confirmed. The study of the reactivity accident on RBMK-1000 was carried out for various combinations of values of the effectiveness of control rods; reactivity coefficients of the coolant temperature and fuel temperature; changes in the temperature of the coolant at the inlet to the reactor. In most of the studied RBMK-1000 reactor accident scenarios, the critical values of fuel enthalpy, at which the process of fuel destruction begins, are reached first. An important result of the research is the conclusion that it is not necessary to reach supercriticality on instantaneous neutrons, supercriticality on delayed neutrons is also sufficient to initiate fuel destruction.

Numerical Simulation of Drag Reduction on a Square Rod Detached with Two Control Rods at Various Gap Spacing via Lattice Boltzmann Method

Numerical simulations are performed to examine the effect of size of control rods (d1) and spacing ratio (g) on flow around a square rod with upstream and downstream control rods aligned in-line using the lattice Boltzmann method (LBM). The Reynolds number (Re) is fixed at Re = 160, while the spacing between the main rod and control rods is taken in the range 1 ≤ g ≤ 5 and the size of the control rod is varied between 4 and 20. Seven different types of flow mods are observed in this study at different values of g and d1. Variation in force statistics, like mean drag coefficient (Cdmean), Strouhal number (St), root mean square values of drag (Cdrms) and lift coefficients (Clrms), and percentage reduction in mean drag coefficient is discussed in detail. It was examined that vortex shedding completely suppressed at (g, d1) = (1, 12), (2, 12), and (2, 16) where steady flow mode exists. Moreover, it was found that at large gap spacing, where g = 5, the effect of control rods on the main rod vanishes. Due to this strong vortex shedding produced and as a result, maximum value of Cdmean is found at (g, d1) = (5, 8). The negative values of mean drag force are also observed at some gap spacing and size of control rods are due to the effect of thrust. Furthermore, the maximum percentage reduction in Cdmean is 121%, found at (g, d1) = (2, 20).

An investigation into the fatigue damage of a long flexible cylinder with multiple control rods in crossflow

In the field of offshore engineering, the control rod is a common passive device to suppress the vortex-induced vibration (VIV) of cylindrical structures. Some key parameters of control rods (diameter ratio, spacing distance, coverage rate and attack angle) have a great influence on the VIV fatigue life of the cylinder. This paper aims to investigate the fatigue damage of the main flexible cylinder with multiple control rods caused by VIV based on rainflow counting and a standard S–N curve, with the final purpose of studying the effectiveness of multiple control rods, in different configurations around the cylinder in improving the structural fatigue life. The results show that control rods can suppress the VIV response of the cylindrical structure and reduce its fatigue damage effectively. The reduction effect of the control rods is very sensitive to the spatial arrangement. The reduction effect of the 3 control rods is the best at the attack angle of 40°, while that of the 4 control rods is the best at the attack angle of 30°. The fatigue damage of the flexible cylinder can be reduced to a lower level with 4 control rods. In addition, “mode transition”, a specific feature of VIV in uniform flow, is found in the experiment and has a strong influence on fatigue life.

NEUTRONIC AND THERMAL HYDRAULICS ANALYSIS OF CONTROL ROD EFFECT ON THE OPERATION SAFETY OF TRIGA 2000 REACTOR

NEUTRONIC AND THERMAL HYDRAULICS ANALYSIS OF CONTROL ROD EFFECT ON THE OPERATION SAFETY OF TRIGA 2000 REACTOR. Analysis of neutronic and thermal-hydraulics parameters of whole operation cycle is very important for the safety of reactor operation. During the reactor operation cycle, the position of the control rods will change due to reactivity changes. The purpose of this study is to determine the effect of control rods position on neutronic and thermal-hydraulics parameters in relation to the safety of reactor operation of the TRIGA 2000 reactor using silicide fuel of MTR plate type. Those parameters are power peaking factor, reactivity coefficients, and steady-state thermohydraulic parameters. Neutronic calculations are performed using a combination of WIMSD/5 and Batan-3DIFF codes and for thermal-hydraulics the calculations are done using WIMSD/5 and MTRDYN codes. The calculation results show that the reactivity coefficient values are negative for all control rod positions both at CZP and HFP conditions. The MTC value decreases when the control rod is inserted into the active core while the FTC value increases. The total ppf results and temperature in steady-state rise when the control rods are inserted of into the active core whereby the maximum value occurs at the position of the control rods of 20 cm from the bottom of the active core. The calculation results of ppf, reactivity coefficient, and thermal-hydraulics parameters lay below safety limits, indicating that the TRIGA 2000 reactor can safely use U3Si2-Al silicide fuel as a substitute fuel for cylindrical type fuel.Keywords: neutronic, thermal-hydraulic parameter, control rod effect, TRIGA 2000, silicide fuel.

Fluid Flow Past a Circular Cylinder with Tandem Rod and Staggered Rod at Low Reynolds Number

The flow past a primary cylinder with one tandem control rod and one staggered control rod is simulated in this paper through solving the Navier-Stokes equations. Two examples are simulated to validate the model, and the results matched well with those of previous researches. The Reynolds number based on the diameter of the primary cylinder is 500. The diameter ratio between the control rod and the primary cylinder (d/D) is 0.25. It was found that the effect of the combination of one upstream tandem control rod and one staggered control rod on the hydrodynamics of the primary cylinder is a linear superposition of the effect of a corresponding single control rod, and the effect of the upstream tandem control rod is dominant at larger spacing ratios such as G/D = 2. For the combination of a downstream tandem control rod and a staggered control rod, the effect of the control rods is different from that of the corresponding single control rod in the region of 0.2<G/D<0.5 & 30°<α<120° and 0.9<G/D<1.4 & 30°<α<50°, where the additional effect is obvious. In this case, the effect of the downstream tandem control rod is dominant at small spacing ratios (such as G/D = 0.1). At moderate spacing ratios such as G/D = 0.4, the effects of the tandem control rod and the staggered control rod are comparable in both cases.

Benchmarking verification of control rod effects on CMRR using MCNP codes throughout 3D core modeling and rod-drop experiment

In this research, MCNP code and ORIGEN code are used to calculate the control rod reactivity worth effects by simulating the 3D core model of CMRR reactor. The integral and differential behaviors of reactivity worth effects are measured by rod-drop experiments and digital inverse kinetic method with each other. The calculated and measured results are well accorded. The integral reactivity worth of one safety rod is about 4%k/k. Even in an accident when one safety rod gets stuck, the CMRR shutdown margin is still greater than 10%k/k, and CMRR is totally safe. So the physical design of CMRR is highly reliable and the operation could be safe.

Dependence of Control-Rod Efficiency on the Initial Conditions of Rod Motion

The effectiveness of control rods is measured in a nonstationary regime at minimum reactor power. The measurement algorithm entails a buildup of reactor power caused by lifting of the measured or any other rod followed by its release, which is due to the need to maintain a detector signal above the background level up to the end of the experiment. It is known that some fission neutrons (the prompt neutrons) are created directly during nuclear fission while others are formed from the decay of the precursors of delayed neutrons, which possess a low-energy spectrum or lower average energy. For fast reactors, this can be reflected on the efficiency of the control rods during measurements and during the development of the processes occurring during an accident. This work is devoted to determining the effect of the conditions preceding the movement of control rods on their effectiveness and the discrepancy between the efficiency of the rods realized in practice and the efficiency calculated on the basis of stationary states.

Experimental investigation of the effect of control rod guide tubes on the breakup of a molten metal jet in the lower plenum of a boiling water reactor under isothermal conditions

It is important to clarify the molten material jet breakup process to estimate corium behavior in the lower plenum of a boiling water reactor (BWR). To identify the effect of control rod guide tubes (CRGTs) on the jet breakup behavior, a molten material (U-alloy) breakup experiment considering CRGTs in a BWR lower plenum was conducted under isothermal conditions. The experiment results show that jet breakup fraction for the case with CRGTs (pitch/diameter ratio of 1.37) was only approximately 20 % of that for the case without CRGTs. A coarser pitch/diameter ratio of 2.47 was also tested, but this configuration only slightly reduced the amount of jet breakup. The experiments also indicate that the CRGTs had no significant effect on the fragmentation of droplet diameter. Furthermore, the velocity distribution was measured around the jet with particle image velocimetry. The velocities of the water surrounding the jet for the cases with CRGTs were relatively larger than those in the case without CRGTs.

Assessment of Contributions by Void Reactivity Coefficient and End Effect of Control Rods into ChNPP-4 Accident Progression

The paper analyzes physical peculiarities of RBMK leading to catastrophic progression of the accident at ChNPP-4 on 26 April 1986. The modeled transient, which preceded the accident, and modeling of its first phase showed that introduction of absorbing rods into the core played a decisive role in the accident progression.

Benchmarking verification of the control rod effects on the MTR core parameters using the MTR-PC and MCNP codes throughout 3D core modeling and rod-drop experiment

Verification of nuclear data and codes is highly recommended for reactor safety analyses. In this research, MTR-PC package and MCNP 4C are used as deterministic and Monte Carlo simulation code, respectively. They are taken into account safety parameters as a function of control rods. Control absorbers classified in two different types. The first called Shim Safety Rods (SSR) made of an Ag–In–Cd composition, and the second one is the Fine Regulating Rod (FRR) made of Stainless Steel. One startup test of a 5 MW Material Testing Reactor (MTR) is simulated throughout the 3-D core modeling. It checks the overall simulating performance. Reactivity worth effects of control rods are calculated and benchmarked against rod-drop experimental results. Both types of codes are taken into account the integral and differential behavior of reactivity worth effects. The startup state has been simulated very carefully, and also total reactivity worth effect of control rods differs from the measured value less than 1.1% (in pcm). Results are in good agreement and highly reliable to calculate the total reactivity worth effect of control rods, and to simulate the startup state of MTRs as Best Estimate (BE) tools.

Quantifying the effects of depletion parameters on the PWR spent fuel reactivity based on nuclide sensitivity coefficients

This study employs nuclide sensitivity coefficients to investigate the spent fuel reactivity of pressurized water reactors (PWRs) in terms of differences in predicted inventories under various depletion conditions. Considering a representative PWR spent fuel, the study began by selecting the top 20 nuclides of large macroscopic neutron absorption cross sections, which are expected to have a substantial effect on reactivity. The sensitivity coefficients (Δk∞/ΔNi) of each nuclide were first evaluated at various burnups and cooling times using MCNPX. These coefficients were used to facilitate a detailed analysis of the spent fuel reactivity and to identify which nuclides cause the effect in a quantitative manner. The approach was demonstrated by analyzing the effects on spent fuel reactivity of several depletion parameters, including the initial enrichment, specific power, operating history, boron concentration, fuel temperature, moderator temperature, control rod effect, and axial burnup profile. The results were generally in good agreement with those obtained from direct MCNPX criticality calculations. Two major advantages of this approach are: (1) the changes in spent fuel reactivity with various depletion parameters can be closely related to composition changes; and (2) it provides a quantitative estimate of the contribution of each nuclide to the reactivity effect based on the predicted inventory differences.

Assessment of the control rods shadow effect in the VENUS-F core

Abstract The partitioning and transmutation (P&amp;T) of spent nuclear fuel is an important field of present development of nuclear energy technologies. One of the possible ways to carry out the P&amp;T process is to use the accelerator driven systems (ADS). This technology has been developed within the EURATOM Framework Programmes for several years now. Current research in this field is carried out within the scope of 7th FP project FREYA. Important parts of the project are experiments performed in the GUINEVERE facility devoted to characterising the subcritical core kinetics and development of reactivity monitoring techniques. The present paper considers the effects of control rods use on the core reactivity. In order to carry out the evaluation of the experimental results, it is important to have detailed core characteristics at hand and to take into consideration the differences in the effect of control rods acting separately or together (the so-called shadow effect) on both the reactivity value and the measured neutron flux. Also any core asymmetry should be revealed. This goal was achieved by both MCNP simulations and the experimental results. However, in the case of experimental results, the need for calculating respective correction factors was unavoidable.

Parameter analysis of the neutron point kinetics equations with feedback temperature effects

The paper presents a solution technique for modeling point nuclear kinetic equations based on one group of delayed neutrons and temperature feedback. Explicit and implicit solutions for point kinetic equations are used for this purpose. Analytical treatment is conducted. Mathematical models describing neutron density and reactor reactivity are deduced. Also, Matlab Simulink and VisSim environments are used to achieve the implicit solution. Using of graphical user interface allows a quick experimentation with alternative values of performance parameters such as initial reactivity, initial neutron flux density, temperature coefficient of reactivity and reciprocal of thermal capacity of reactor. Moreover, implicit solutions of dynamic equations governing point nuclear kinetic equations provide exact handling of the device performance. Proposed mathematical models and block diagram simulation results are validated against published work and full agreements are obtained. Several performance parameters are tuned to enhance the performance of these point nuclear kinetic equations through the presented methodology. The resultant performance characteristics and comparison among investigated models are presented in this work. The obtained results confirm that the implicit solution showed less accurate representation of the studied point nuclear kinetic equations compared to mathematical models. Furthermore, the effect of control rod on the neutron flux density was discussed.

PWR-UO2 nuclear fuel criticality study: control rod effects on infinite neutron multiplication factor and spent fuel composition

Deterministic and stochastic nuclear codes are software packages used to perform reactor physics calculations, especially in PWRs, the most common type of nuclear reactor currently in operation. The NEA Expert Group on Burn-up Credit Criticality Safety has published a Benchmark with results obtained from simulations of PWR-UO2 nuclear fuel. The same simulations were performed at DEN/UFMG with SCALE 6.0, a modular nuclear system code developed by Oak Ridge National Laboratory using two different neutron energy libraries (238 and 44 groups). The results obtained using a three-dimensional model with the T6-DEPL sequence of the TRITON module in SCALE 6.0 for spent fuel inventory and infinite neutron multiplication factor calculations show very good agreement with those published by the OECD. The main goal of this work is to validate the methodology at DEN/UFMG for future use in simulations related to Angra I, II and III Nuclear Power Plants.

Efficient code system of core characteristics for small cylindrical thermal reactor

A new code system for the overall neutronic calculation of a thermal reactor by a simple and effective way is presented. The code covers microscopic library compilation, macroscopic constant generation, cell calculations by multi-group treatment for neutron transport equation and core calculations over three zones for fuel and one zone for moderator. The Dancoff correction factor required in the interpolation of the self-shielding factors of resonance nuclides is automatically calculated by the installed collision probability routines. The burn-up calculation and Garrison and Ross model of fission product have been included. Also the effect of control rod on the reactivity of the reactor with special treatment for the control rod based on the homogenization technique has been included. Making a comparison with SRAC95 code system has checked the adopted code.