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.

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.

Full Implicit Integrate Solution to the Coupled Neutronics/Thermal-Hydraulics Problems

2013 ◽  
Author(s):  
Han Zhang ◽  
Fu Li
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Thermal Hydraulics ◽  
Field Equations ◽  
Krylov Method ◽  
Tightly Coupled ◽  
Physical Insight ◽  
Second Order Convergence ◽  
The Individual ◽  
Integrate Solution

The traditional solution of the coupled neutronics/ thermal-hydraulics problems has typically been performed by solving the individual field separately and then transferring information between each other. In this paper, full implicit integrate solution to the coupled neutronics/ thermal-hydraulic problem is investigated. There are two advantages compared with the traditional method, which are high temporal accuracy and stability. The five equations of single-phase flow, the solid heat conduction and the neutronics are employed as a simplified model of a nuclear reactor core. All these field equations are solved together in a tightly coupled, nonlinear fashion. Firstly, Newton-based method is employed to solve nonlinear systems due to its local second-order convergence rate. And then the Krylov iterative method is used to solve the linear systems which are from the Newton linearization. The two procedures above are the so-called Newton-Krylov method. Furthermore, in order to improve the performance of the Krylov method, physics-based preconditioner is employed, which is constructed by the physical insight. Finally, several Newton-Krylov solution approaches are carried out to compare the performance of the coupled neutronics / thermal-hydraulic equations.

Identification of a nuclear reactor core (VVER) using recurrent neural networks

2002 ◽  
Vol 29 (10) ◽  
pp. 1225-1240 ◽  
Author(s):  
Mehrdad Boroushaki ◽  
Mohammad B. Ghofrani ◽  
Caro Lucas
Keyword(s):  
Neural Networks ◽  
Recurrent Neural Networks ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Nuclear Reactor Core

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.

The Measurement of Light Element Segregation Using EDS and EELS

1998 ◽  
Vol 4 (S2) ◽  
pp. 772-773
Author(s):  
J.T. Busby ◽  
E.A. Kenik ◽  
G.S. Was
Keyword(s):  
Grain Boundaries ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Light Element ◽  
Free Surfaces ◽  
Spatial Redistribution ◽  
Core Components ◽  
Radiation Induced ◽  
Potential Interactions ◽  
Impurity Elements

Radiation-induced segregation (RIS) is the spatial redistribution of elements at defect sinks such as grain boundaries and free surfaces during irradiation. This phenomenon has been studied in a wide variety of alloys and has been linked to irradiation-assisted stress corrosion cracking (IASCC) of nuclear reactor core components. However, several recent studies have shown that Cr and Mo can be enriched to significant levels at grain boundaries prior to irradiation as a result of heat treatment. Segregation of this type may delay the onset of radiation-induced Cr depletion at the grain boundary, thus reducing IASCC susceptibility. Unfortunately, existing models of segregation phenomena do not correctly describe the physical processes and therefore are grossly inaccurate in predicting pre-existing segregation and subsequent redistribution during irradiation. Disagreement between existing models and measurement has been linked to potential interactions between the major alloying elements and lighter impurity elements such as S, P, and B.

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