Experimental Investigations of Control Rod Drive Mechanism

2013 ◽  
Author(s):  
Guangyao Lu ◽  
Zhaohui Lu ◽  
Wenyuan Xiang ◽  
Yonghong Lv ◽  
Wenyou Huang ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Experimental Investigations ◽  
Control Rod ◽  
Drive Mechanism ◽  
Start Up ◽  
Power Regulation ◽  
Set Up ◽  
Control Rods

The control rod drive mechanism (CRDM) is installed on the CRDM socket in reactor pressure vessel (RPV). Directed by Rod Control and Rod Position Indicating System (RGL), CRDM can impel the control rods move up and down in the nuclear reactor core, which implements the functions of reactor start-up, power regulation, power maintaining, normal reactor shutdown and abnormal (accident) shutdown. CRDM was developed by China Nuclear Power Research Institute (CNPRI). Several design improvements were conducted to solve the problems appeared in the operation of nuclear power station. Test bench was also set up and cold tests were carried out to investigate the characteristics of CRDM. The cold tests included lifting experiment, inserting experiment, rod drop experiment. And studies were carried out to analyze the signals of lifting coil, moving coil, stationary coil and the vibration signals. The test results show that the design of CRDM is reasonable and the operation is reliable.

scholarly journals Investigation Of Rod Control System Reliability Of Pwr Reactors

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Syaiful Bakhri
Keyword(s):  
Control System ◽  
System Reliability ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Fault Tree ◽  
Reactor Core ◽  
Pressurized Water ◽  
Control Rod ◽  
Control Rods

<p class="NoSpacing1"><span lang="IN">The Rod Control System is </span>employed<span lang="IN"> to adjust the position of the control rods in the reactor core </span>which corresponds with <span lang="IN">the thermal power generated in the core </span>as well as <span lang="IN">the electric power generated in the turbine. In a Pressurized Water Reactor (PWR) type nuclear power plants, the control-rod drive </span>employs <span lang="IN">magnetic stepping-type mechanism. This </span>type of <span lang="IN">mechanism consists of a pair of circular coils and latch-style jack with the armature. When the </span>electric <span lang="IN">current </span>is <span lang="IN">supplied to the coils sequentially, the control-rods</span>, which <span lang="IN">are held on the drive shaft</span>, can be driven<span lang="IN"> up</span>ward<span lang="IN"> or down</span>ward<span lang="IN"> in increments. </span>This <span lang="IN">sequential current </span>c<span lang="IN">ontrol</span> drive<span lang="IN"> system is called the Control-Rod Drive Mechanism Control System (CRDMCS) or </span>known also as <span lang="IN">the Rod Control System (RCS). The p</span>urpose of this paper is to investigate the RCS reliability <span lang="IN">of APWR </span>using <span lang="IN">the Fault Tree Analysis (FTA)</span> method<span lang="IN"> since </span>the analysis of reliability which considers<span lang="IN"> the FTA</span> for common CRDM <span lang="IN">can </span>not <span lang="IN">be found</span> in <span lang="IN">any </span>public references. <span lang="IN">The FTA method is used to model the system reliability by developing the fault tree diagram of the system. </span>The<span lang="IN"> results show that the failure of the system is very dependent on the failure of most of the individual systems. However, the failure of the system does not affect the safety of the reactor, since the reactor trips immediately if the system fails. The evaluation results also indicate that the Distribution Panel is the most critical component in the system.</span></p>

scholarly journals Measurement and Comparison of Control Rod Worth of BTRR using Inhour Equation and Period reactivity conversion table

2017 ◽  
Vol 41 (1) ◽  
pp. 95-103
Author(s):  
Md Iqbal Hosan ◽  
MAM Soner ◽  
Md Fazlul Huq ◽  
Khorshed Ahmad Kabir
Keyword(s):  
Doubling Time ◽  
Nuclear Reactor ◽  
Reactor Control ◽  
Control Rod ◽  
Fuel Burn ◽  
Start Up ◽  
Conversion Table ◽  
Power Operation ◽  
Academy Of Sciences ◽  
Control Rods

In a nuclear reactor, control rod is a very essential part and plays the elementary role in the reactor control during reactor start up, normal power operation, experimental research and shutdown. To perform all these operations safely, knowledge of differential and integral worth of the control rod is mandatory. In this study, the differential and integral worth curve of all control rods of BAEC TRIGA Research Reactor (BTRR) have been determined by using the positive period method. Reactor period was measured from 1.5 folding time, doubling time, 5 folding time respectively; and in the above three cases reactivity has also been calculated from INHOUR equation and period reactivity conversion table. The total worth of all control rods of BTRR are measured as 14.888 $, 14.672 $, 14.348 $ from INHOUR equation and 13.978 $, 13.672 $, 13.357 $ from period reactivity conversion table for 1.5 folding time, doubling time and 5 folding time respectively. The measured reactivity has also been compared with the previously measured reactivity and due to fuel burn up of the reactor expected lower values were observed.Journal of Bangladesh Academy of Sciences, Vol. 41, No. 1, 95-103, 2017

Developments of the Hydraulic Driving System of the Control Rods for Nuclear Reactors

1999 ◽  
Author(s):  
Yuanqiang Wu
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Reactor Core ◽  
Hydraulic Pressure ◽  
Energy Agency ◽  
Control Rod ◽  
Driving System ◽  
History Of ◽  
The Cost ◽  
Control Rods

Abstract The developments of a new hydraulic driving system of the control rods for nuclear reactors are introduced in this paper. Compared with other driving systems of the control rods, this new hydraulic driving system can be set within the reactor pressure vessel. Under any serious condition, the control rods will not be ejected from the reactor core. Its structure is very simple and the mechanic chain is very short, and thus it is very reliable. It can reduce the height of the nuclear reactor by one-third, and thus dramatically reduce the cost of the reactor. It uses the dynamic hydraulic pressure to control the motion of the control rods. Under extreme conditions, such as the failure of control power supply, the control rods will drop into the reactor core because of their self-weight to shut down the nuclear reaction. Because of these features, International Atomic Energy Agency (IAEA) is very interested in this safe and economical new control rod driving system. A brief history of the developments of the hydraulic driving system is given. Three configurations, the orifice hydraulic step cylinder, the groove-orifice hydraulic step cylinder, and the piston-groove hydraulic step cylinder, are introduced and their working principles are explained. The reliability and safety of the new system are validated by two experimental works: hydraulic step cylinder (HSC) under seismic and rocking conditions. Results from these experiments are presented.

Falling Characteristic Analysis of Passive Drive Mechanism of Absorber Ball Shutdown System for HTR-PM

2013 ◽  
Author(s):  
Feng Chen ◽  
Tianjin Li ◽  
Yuan Liu ◽  
Zhiyong Huang ◽  
Hanliang Bo
Keyword(s):  
Ambient Temperature ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Reactor Safety ◽  
Characteristic Analysis ◽  
Drive Mechanism ◽  
Storage Vessel ◽  
Power Regulation ◽  
Present Design ◽  
Reactor Application

The absorber ball shutdown system is a key system and related to reactivity control in HTR-PM, which has once achieved power regulation and cold shutdown of the nuclear reactor. Passive drive mechanism of absorber ball shutdown system has been designed for HTR-PM reactor safety before. Ball dropping valve can be automatically opened by gravity when loss of offsite power happens, then a mass of balls with BC4 fall into reactor core for function implementation. Therefore drive mechanism reliability is fairly important for absorber ball shutdown system in HTR-PM. Experiments on passive drive mechanism have been carried out in this study. Falling distance and falling time of drive mechanism have been recorded and analyzed with and without absorber ball, as well as at ambient temperature and in the heating conditions. Experimental results demonstrate that designed passive drive mechanism can shift reliably. Ball dropping valve can freely open each time without absorber balls in the storage vessel. Basically present design of passive drive mechanism for absorber ball shutdown system can meet the requirement of actual reactor application.

Calculation on the Distribution of Ex-Core Neutron Flux during Reactor Start-Up

2014 ◽  
Vol 672-674 ◽  
pp. 375-378
Author(s):  
Chun Yu Liu ◽  
Benbicha Mohamed Elghazali
Keyword(s):  
Neutron Flux ◽  
Physical Model ◽  
Pressure Vessel ◽  
Reactor Core ◽  
Mcnp Code ◽  
Control Rod ◽  
Start Up ◽  
Control Rods ◽  
Moving Process ◽  
Detection Effect

The distribution of neutron flux is simulated by MCNP code from reactor start-up to criticality when the control rods are drawn different length from reactor core on ex-core detector area of the WWER reactor of TianWan. Because physical model built is very large, in order to save calculation time, the moving process of control rod is simplified. The results of calculation show that The neutron mainly distributed in the range of 0-400cm outside the pressure vessel. The value of the relative neutron flux ex-core is maximum in the range of 110cm to 180cm, so detection effect is better when the detector is set in this region.

Leakage Flow Investigations on a Through-Wall Crack Related to Thermal Fatigue of Nuclear Power Plant Piping

2017 ◽  
Author(s):  
Stefan Schmid ◽  
Rudi Kulenovic ◽  
Eckart Laurien
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Numerical Models ◽  
Measurement Techniques ◽  
Experimental Investigations ◽  
Leakage Flow ◽  
Test Rig ◽  
Reduced Pressure ◽  
Flow Rates ◽  
Set Up

For the validation of empirical models to calculate leakage flow rates in through-wall cracks of piping, reliable experimental data are essential. In this context, the Leakage Flow (LF) test rig was built up at the IKE for measurements of leakage flow rates with reduced pressure (maximum 1 MPA) and temperature (maximum 170 °C) compared to real plant conditions. The design of the test rig enables experimental investigations of through-wall cracks with different geometries and orientations by means of circular blank sheets with integrated cracks which are installed in the tubular test section of the test rig. In the paper, the experimental LF set-up and used measurement techniques are explained in detail. Furthermore, first leakage flow measurement results for one through-wall crack geometry and different imposed fluid pressures at ambient temperature conditions are presented and discussed. As an additional aspect the experimental data are used for the determination of the flow resistance of the investigated leak channel. Finally, the experimental results are compared with numerical results of WinLeck calculations to prove specifically in WinLeck implemented numerical models.

The analysis of emergency situations related to fires at nuclear power plants

2021 ◽  
Vol 30 (5) ◽  
pp. 66-75
Author(s):  
S. A. Titov ◽  
N. M. Barbin ◽  
A. M. Kobelev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Emergency Situations ◽  
The Usa ◽  
Large Fires ◽  
Short Circuits ◽  
Nuclear Reactor Core

Introduction. The article provides a system and statistical analysis of emergency situations associated with fires at nuclear power plants (NPPs) in various countries of the world for the period from 1955 to 2019. The countries, where fires occurred at nuclear power plants, were identified (the USA, Great Britain, Switzerland, the USSR, Germany, Spain, Japan, Russia, India and France). Facilities, exposed to fires, are identified; causes of fires are indicated. The types of reactors where accidents and incidents, accompanied by large fires, have been determined.The analysis of major emergency situations at nuclear power plants accompanied by large fires. During the period from 1955 to 2019, 27 large fires were registered at nuclear power plants in 10 countries. The largest number of major fires was registered in 1984 (three fires), all of them occurred in the USSR. Most frequently, emergency situations occurred at transformers and cable channels — 40 %, nuclear reactor core — 15 %, reactor turbine — 11 %, reactor vessel — 7 %, steam pipeline systems, cooling towers — 7 %. The main causes of fires were technical malfunctions — 33 %, fires caused by the personnel — 30 %, fires due to short circuits — 18 %, due to natural disasters (natural conditions) — 15 % and unknown reasons — 4 %. A greater number of fires were registered at RBMK — 6, VVER — 5, BWR — 3, and PWR — 3 reactors.Conclusions. Having analyzed accidents, involving large fires at nuclear power plants during the period from 1955 to 2019, we come to the conclusion that the largest number of large fires was registered in the USSR. Nonetheless, to ensure safety at all stages of the life cycle of a nuclear power plant, it is necessary to apply such measures that would prevent the occurrence of severe fires and ensure the protection of personnel and the general public from the effects of a radiation accident.

A novel method for diagnosis and de-noising of control rod drive mechanism within floating nuclear reactor

2022 ◽  
Vol 244 ◽  
pp. 110398
Author(s):  
Liming Zhang ◽  
Qiao Li ◽  
Jingdong Luo ◽  
Minghui Liu ◽  
Yiming Wang ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Control Rod ◽  
Drive Mechanism ◽  
Novel Method

A High Effective Parallel Method for the Coupling Between Neutronics and Thermal-Hydraulic

2016 ◽  
Author(s):  
Xiaomeng Dong ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Lei Li ◽  
Guangliang Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Decisive Role ◽  
Reactor Power ◽  
Coupling Method ◽  
Outlet Temperature

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.

scholarly journals Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

2012 ◽  
Vol 27 (3) ◽  
pp. 229-238
Author(s):  
Ali Sidi ◽  
Zaki Boudali ◽  
Rachid Salhi
Keyword(s):  
Heat Flux ◽  
Critical Velocity ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Thermal Characteristics ◽  
Nuclear Research ◽  
Coolant Fluid ◽  
Critical Velocities ◽  
Set Up ◽  
Physical Quantities

The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate?s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

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