scholarly journals CONCEPTUAL DESIGN OF CONTROL ROD DRIVE MOTOR TYPE MAGNETIC JACK FOR NUCLEAR POWER PLANT

2021 ◽  
Vol 15 (1) ◽  
pp. 35
Author(s):  
Fahmi Alfa Muslimu
Keyword(s):  
Conceptual Design ◽  
Nuclear Power ◽  
Reactor Power ◽  
Motion Simulation ◽  
Working Mechanism ◽  
Control Rod ◽  
Three Phase ◽  
Drive Motor ◽  
Limit Switch ◽  
Transmission Gear

The Control Rod Drive Motor (CRDM) controls the reactor power of a G.A Siwabessy and automatically shuts it down during an emergency using a three-phase motor that drives a ball-nut spindle attached to the magnetic SCRAM through the transmission gear. However, there are several weaknesses associated with this design, such as the inability of the ball nut to rotate when a disturbance occurs at the motor limit switch continuously. This causes the threads on the control rod shaft to wear out due to friction and release from the holder. Therefore, this research aims to develop a CRDM with a magnetic jack for a nuclear plant, which moves the extension shaft and the control rod components vertically and linearly. The control rod's motor must pull, insert, hold, or drop it from any point. This conceptual design is the first step in determining prototyping design criteria with a magnetic jack to understand the working mechanism. The control rod gripping motion simulation was also presented using ANSYS Rigid Dynamics to reduce the failure at the design phase before prototyping. The simulation results showed no collision on each component capable of affecting the overall system performance. Therefore, the control rod motor functions properly in carrying out the pulling and lowering movements on 19.1 mm infrequency.

RETRAN Application of Turbine Trip and Load Rejection of Startup Test Analysis for Lungmen ABWR

2009 ◽  
Author(s):  
Chen-Lin Li ◽  
Chiung-Wen Tsai ◽  
Chunkuan Shih ◽  
Jong-Rong Wang ◽  
Su-Chin Chung
Keyword(s):  
Control System ◽  
Water Level ◽  
Control Systems ◽  
Nuclear Power ◽  
Pressure Control ◽  
Reactor Power ◽  
Level Control ◽  
Bypass Flow ◽  
Control Rod ◽  
Recirculation Flow

This study used RETRAN program to analyze the turbine trip and load rejection transients of Taiwan Power Company Lungmen Nuclear Power Plant’s startup test at 100% power and 100% core flow operating condition. This model includes thermal flow control volumes and junctions, control systems, thermal hydraulic models, safety systems, and 1D kinetics model. In Lungmen RETRAN model, four steam lines are simulated as one line. There are four simulated control systems: pressure control system, water level control system, feedwater control system, and speed control system for reactor internal pumps. The turbine trip event, at above 40% power, triggers the fast open of the bypass valves. Upon the turbine trip, the turbine stop valves close. To minimize steam bypassed to the main condenser, recirculation flow is automatically runback and a SCRRI (selected control rod run in) is initiated to reduce the reactor power. The load rejection event causes the fast opening of the bypass valves. Steam bypass will sufficiently control the pressure, because of their 110% bypass capacity. A SCRRI and RIP runback are also initiated to reduce the reactor power. This study also investigated the sensitivity analysis of turbine bypass flow, runback rate of RIPS and SCRRI to observe how they affect fuel surface heat flux, neutron flux and water level, etc. The results show that turbine bypass flow has larger impacts on dome pressure than RIPS runback rate and SCRRI. This study also indicates that test criteria in turbine trip and load rejection transients are met and Lungmen RETRAN model is performing well and applicable for Lungmen startup test predictions and analyses.

scholarly journals Reactor Core Conceptual Design for a Scalable Heating Experimental Reactor, LUTHER

2021 ◽  
Vol 2 (2) ◽  
pp. 207-214
Author(s):  
Thinh Truong ◽  
Heikki Suikkanen ◽  
Juhani Hyvärinen
Keyword(s):  
Low Temperature ◽  
Conceptual Design ◽  
Nuclear Power ◽  
Water Pressure ◽  
Reactor Core ◽  
Electricity Production ◽  
Design Parameters ◽  
Experimental Reactor ◽  
Fuel Assemblies ◽  
Enriched Uranium

In this paper, the conceptual design and a preliminary study of the LUT Heating Experimental Reactor (LUTHER) for 2 MWth power are presented. Additionally, commercially sized designs for 24 MWth and 120 MWth powers are briefly discussed. LUTHER is a scalable light-water pressure-channel reactor designed to operate at low temperature, low pressure, and low core power density. The LUTHER core utilizes low enriched uranium (LEU) to produce low-temperature output, targeting the district heating demand in Finland. Nuclear power needs to contribute to the decarbonizing of the heating and cooling sector, which is a much more significant greenhouse gas emitter than electricity production in the Nordic countries. The main principle in the development of LUTHER is to simplify the core design and safety systems, which, along with using commercially available reactor components, would lead to lower fabrication costs and enhanced safety. LUTHER also features a unique design with movable individual fuel assembly for reactivity control and burnup compensation. Two-dimensional (2D) and three-dimensional (3D) fuel assemblies and reactor cores are modeled with the Serpent Monte Carlo reactor physics code. Different reactor design parameters and safety configurations are explored and assessed. The preliminary results show an optimal basic core design, a good neutronic performance, and the feasibility of controlling reactivity by moving fuel assemblies.

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.

Thermal–Hydraulic Performance Analysis for AP1000 Passive Containment Cooling System

2013 ◽  
Author(s):  
Yu Yu ◽  
Shengfei Wang ◽  
Fenglei Niu
Keyword(s):  
System Reliability ◽  
Nuclear Power ◽  
Failure Modes ◽  
Cooling System ◽  
Natural Circulation ◽  
Hydraulic Performance ◽  
Reactor Power ◽  
Physical Parameters ◽  
Cold Source ◽  
Process Failure

In order to improve the safety of new generation nuclear power plant, passive containment cooling system is innovatively used in AP1000 reactor design. However, since the system operation is based on natural circulation, physical process failure — natural circulation cannot establish or be maintained — becomes one of the important failure modes. Uncertainties in the physical parameters such as heat and cold source temperature and in the structure parameters have important effect on the system reliability. In this paper, thermal–hydraulic model is established for passive containment cooling system in AP1000 and the thermal–hydraulic performance is studied, the effect of factors such as air temperature and reactor power on the system reliability are analyzed.

Steam Generator Replacement Project at TVA’s Sequoyah Unit 1 Nuclear Power Plant

2004 ◽  
Author(s):  
Myron R. Anderson
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Power ◽  
Original Design ◽  
Pressurized Water ◽  
Repair Work ◽  
Building Roof ◽  
Tennessee Valley ◽  
Generator Replacement

Pressurized Water Reactor Power Plants have at times required that large components be replaced (steam generators weighing 750,000 lbs) which have necessitated performing first time modifications to the plant that were unintended during the original design. The steam generator replacement project at Tennessee Valley Authority (TVA’s) Sequoyah Nuclear Power Station necessitated (1) two large temporary openings (21’×45’) in the plant’s Shield Building roof (2’ thick concrete) by hydro-blasting to allow the removal of the old generators and installation of the new, (2) removal and repair of the concrete steam generator enclosure roofs (20’ diameter, 3’ thick) which were removed by wire saw cutting and (3) the seismic qualification of; the design and construction of an extensive ring foundation for; the use of one of the world largest cranes to remove these components through the roof. This removal and replacement process had to be performed in an expeditious manner to minimize the amount of time the plant is shutdown so the plant could return to providing power to the grid. This paper will address some of the many technical and construction considerations required to perform this demolition and repair work safely, efficiently and in a short as possible duration.

New Human-Machine Interface at VR-1 Training Reactor

2010 ◽  
Author(s):  
Martin Kropi´k ◽  
Jan Rataj ◽  
Monika Jurˇicˇkova´
Keyword(s):  
User Interface ◽  
Foreign Students ◽  
Nuclear Power ◽  
English Language ◽  
Reactor Power ◽  
Text Messages ◽  
Reactor Operation ◽  
The Core ◽  
Machine Interface ◽  
And Training

The paper describes a new human-machine (HMI) interface of the VR-1 nuclear training reactor at the Czech Technical University in Prague. The VR-1 reactor is primarily used for training of university students and future nuclear power plant staff. The new HMI was designed to meet functional, ergonomic and aesthetic requirements. It contains a PC with two monitors. The first alphanumerical monitor presents text messages about the reactor operation and status; next, the operator can enter commands to control the reactor operation. The second graphical monitor provides parameters of reactor operation and shows the course of the reactor power and other parameters. Furthermore, it is able to display the core configuration, perform reactivity calculations, etc. The HMI is also equipped with an alarm annunciator. Due to a high number of foreign students and visitors at the reactor, the Czech and English language versions of the user interface are available. The HMI contains also a History server which provides a very detailed storage and future presentation of the reactor operation. The new HMI improves safety and comfort of the reactor utilization, facilitates experiments and training, and provides better support for foreign visitors.

TRACE/PARCS Analysis of Full Isolation Startup Test for Lungmen ABWR

2012 ◽  
Author(s):  
Ai-Ling Ho ◽  
Jong-Rong Wang ◽  
Hao-Tzu Lin ◽  
Chunkuan Shih
Keyword(s):  
Water Level ◽  
Coupling Model ◽  
Reactor Power ◽  
Signal Delay ◽  
Control Rod ◽  
Pump Speed ◽  
Closure Time ◽  
Delay Times ◽  
Open Position ◽  
Delayed Time

TRACE/PARCS coupling model of Lungmen NPP has been used to analyze the full isolation startup test, defined as a simultaneous full closure of all MSIVs (Main Steamline Isolation Valve). In this analysis, the full closure time of MSIVs was varied by 2 sec, 3sec (base), and 4 sec. As MSIV was closed to 90% full open position, a reactor power scram signal was initiated. For each of MSIV full closure time, the delayed time of scram signal to the start of control rod insertion was varied by base 0.09 sec, 0.04 sec, and 0.14 sec. Because of MSIVs closure, the increasing reactor dome pressure reached the setpoint of relief valves (RVs) and caused RVs to open. The 4 reactor internal pumps (RIPs) not connected to the M/G sets tripped and were runback to minimum pump speed when water level dropped to L3. The 6 RIPs connected to the M/G sets tripped if low water level (L2) or high reactor dome pressure (1140 psia) had been reached. Those important thermal parameters with various MSIV full closure times (2, 3, and 4sec) and various reactor scram signal delay times (0.09, 0.04, and 0.14 sec) are compared and shown in this paper.

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