scholarly journals Simulation System for PeLUIt 150 MW Nuclear Reactor by using vPower

2021 ◽  
Vol 2048 (1) ◽  
pp. 012034
Author(s):  
M Yunus ◽  
A A Budiman ◽  
S Zhe ◽  
Kiswanta ◽  
W Chunlin ◽  
...  
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Nuclear Power ◽  
Thermal Power ◽  
Rankine Cycle ◽  
Operating Conditions ◽  
Transient Condition ◽  
Design Data ◽  
Water Steam ◽  
Main Steam

Abstract In developing the PeLUIt 150 MW nuclear power plant based on the High Temperature Gas-cooled Reactor (HTGR) technology, with the helium-coolant and output thermal power of 150 MW, the PeLUIt simulator is also developed for training the operators and educating other technical personnel. Referred to the balance of plant (BOP) design of the PeLUIt, the simulator utilized the vPower simulation platform to simulate the secondary loop for power generation with a water-steam Rankine cycle. The paper focuses on developing the secondary loop’s main components: steam generator, steam turbine, condenser, deaerator, and feedwater pump. The reactor module in the primary loop is simplified as a heat source with 150 MW output. The steam generator that connects the primary and secondary loops is modeled with the heat exchanger module by transferring heat from helium to water/steam. Meanwhile, pressure and flow parameters can also be simulated for both helium and water/steam flows in steady-state and transient operating conditions. The steady-state simulation results are almost the same as the design data. The differences in the main steam temperature, feedwater pressure, and feedwater temperature, are 0.03%, 0.53%, and 0.02%, respectively. Meanwhile, the transient condition carried out in the loss of coolant accident showed a decrease in flowrate of 43.31 kg/s and an increase in temperature of feedwater and main-steam of 52.32 and 15.38 °C, respectively. In addition, there was a pressure drop of around 10.37 (feedwater) and 10.16 MPa (main-steam).

Maanshan PWR Loss of Flow Transients Analysis With TRACE

2009 ◽  
Author(s):  
Jong-Rong Wang ◽  
Hao-Tzu Lin ◽  
Wei-Chen Wang ◽  
Yi-Hsiang Cheng ◽  
Chunkuan Shih
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Nuclear Power ◽  
Thermal Power ◽  
Complete Loss ◽  
Transient Condition ◽  
Partial Loss ◽  
Core Flow ◽  
System Parameters ◽  
Trace Model

TRACE model of Maanshan Nuclear Power Plant (three-loop PWR) was used to analyze Loss of Flow transient as defined in FSAR Chapter 15. The results were compared with those from RETRAN02 and LOFTRAN/THINC licensing analysis of Westinghouse Inc. Three different initiation events were involved in this analysis: Partial Loss of Flow (PLOF), Complete Loss of Flow-Under Voltage (CLOF-UV) and Complete Loss of Flow-Under Frequency (CLOF-UF). This paper compared important thermal hydraulic parameters at steady state, such as the pressure of pressurizer, cold-leg temperature, and the pressure of steam generator, etc.. It also compared system parameters under transient conditions, such as core thermal power, core flow rate, and pressure of pressurizer, etc.. It is concluded that the steady state results of TRACE calculations are in general good agreements with those from RETRAN02 and have a largest error of 3.03% in the steam generator flow. For transient condition, TRACE results are also comparable with those from LOFTRAN and RETRAN02. In summary, our studies show that Maanshan TRACE model is correct and accurate enough for future safety analysis applications.

The Research of Marine Nuclear Power Two Loop Simulation Software Based on the Thermal System Analysis

2014 ◽  
Vol 983 ◽  
pp. 288-291
Author(s):  
Guo Lei Zhang ◽  
Xiang Dong Jin ◽  
Zhan Zhao ◽  
Zhi Jun Shi
Keyword(s):  
Nuclear Power ◽  
System Analysis ◽  
Simulation Analysis ◽  
Simulation Software ◽  
Operating Conditions ◽  
Two Phase ◽  
Design Data ◽  
Software Optimization ◽  
Power Simulation ◽  
Basic Functions

To study of Nuclear power simulation software's basic functions and mathematical model based on thermal analysis. Describes the two-phase flow model of GSE software superiority, as well as the software optimization program .Use of software tools for normal operating conditions of the simulation calculation and analysis of the results. Comparison with design data shows that,the software use in marine nuclear power two loop system simulation analysis field, the accuracy of it is higher.

Application of Supercritical Fluids in Thermal- and Nuclear-Power Engineering

2021 ◽  
pp. 601-658
Author(s):  
Igor L. Pioro
Keyword(s):  
Heat Transfer ◽  
Carbon Dioxide ◽  
Supercritical Fluids ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Rankine Cycle ◽  
Working Fluid ◽  
Thermal Power Plants ◽  
Power Cycles

Supercritical Fluids (SCFs) have unique thermophyscial properties and heat-transfer characteristics, which make them very attractive for use in power industry. In this chapter, specifics of thermophysical properties and heat transfer of SCFs such as water, carbon dioxide, and helium are considered and discussed. Also, particularities of heat transfer at Supercritical Pressures (SCPs) are presented, and the most accurate heat-transfer correlations are listed. Supercritical Water (SCW) is widely used as the working fluid in the SCP Rankine “steam”-turbine cycle in fossil-fuel thermal power plants. This increase in thermal efficiency is possible by application of high-temperature reactors and power cycles. Currently, six concepts of Generation-IV reactors are being developed, with coolant outlet temperatures of 500°C~1000°C. SCFs will be used as coolants (helium in GFRs and VHTRs, and SCW in SCWRs) and/or working fluids in power cycles (helium, mixture of nitrogen (80%) and helium (20%), nitrogen and carbon dioxide in Brayton gas-turbine cycles, and SCW/“steam” in Rankine cycle).

Main Steam Line Break Calculations Using a Coupled RELAP5/PARCS Model for the Ringhals-3 Pressurized Water Reactor

2008 ◽  
Author(s):  
Mathias Sta˚lek ◽  
Jo´zsef Ba´na´ti ◽  
Christophe Demazie`re
Keyword(s):  
Steady State ◽  
Fuel Assembly ◽  
Thermal Power ◽  
Steam Line ◽  
Pressurized Water ◽  
The Core ◽  
Different Types ◽  
Main Steam Line ◽  
Main Steam ◽  
Control Rods

A Main Steam Line Break (MSLB) is an important transient for Pressurized Water Reactors (PWR) due to the strong positive reactivity introduced by the over-cooling of the core. Since this effect is stronger when the Moderator Temperature Coefficient (MTC) has a large amplitude, a conservative result will be obtained for a high burnup of the fuel due to the more negative MTC late in the cycle. The calculations have been performed at a cycle burnup of 12.9742 GWd/tHM. The Swedish Ringhals-3 PWR is a three loop Westinghouse design, currently with a thermal power of 3000 MW. The PARCS model has 157 fuel assemblies of 8 different types. Four different types of reflector are used. The cross sections, and kinetic data were obtained from CASMO-4 calculations, using a cross section interface developed at the department. There are 24 axial nodes, and 2×2 radial nodes for each assembly. The transient option for calculating the effect of poisoning was used. The PARCS model has been validated against steady-state measurements from Ringhals-3 of the Relative Power Fraction (RPF) and of the core criticality. The RELAP5 model has 157 channels for the core which means that there is a one to one correspondence between the thermal hydraulics model and the neutronics model. There is eight axial nodes. Originally, the intention was to have 24 axial nodes but this proved not to work because of some limitation in RELAP5. There is currently no mixing between the different channels in the core. The feedwater, and turbines are modelled as boundary conditions. The stand-alone RELAP5 model has been validated against steady state measurements from Ringhals-3. A number of different cases were considered. In the first case, both the isolation of the feedwater for the broken loop, and all the control rods were assumed to work properly. For the second case one of the control rods was assumed to be stuck. The stuck rod was located in the fuel assembly with the highest power. This rod has also one of the highest rod worths. In the final case, the feedwater control valve for the broken loop was fully open. None of the cases led to any recriticality. The increase in power for each fuel assembly was also investigated. With the control rod located in the assembly with the highest power, the maximum power increase before scram turned out to be about 25% compared to the initial power.

Investigation of Blast Wave Effects on Containment Wall and Steam Generator

2018 ◽  
Author(s):  
Tae Jin Kim ◽  
Yoon-Suk Chang
Keyword(s):  
Blast Wave ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Criteria ◽  
High Energy ◽  
Element Analysis ◽  
Secondary Damage ◽  
Main Steam Line ◽  
Main Steam

When a sudden rupture occurs in high energy lines such as MSL (Main Steam Line) and safety injection line of nuclear power plants, ejection of inner fluid with high temperature and pressure causes blast wave, and may lead to secondary damage of adjacent major components and/or structures. The objective of this study is to assess integrity of containment wall and steam generator due to the blast wave under a postulated high energy line break condition at the MSL piping. In this context, a preliminary analysis was conducted to examine the blast wave simulation using coupled Eulerian-Lagrangian technique. Subsequently, a finite element analysis was carried out to assess integrity of the structures. As typical results, strain and stress values were calculated at the containment wall and steam generator, which did not exceed their failure criteria.

A Methodological Approach to Support the Design of Induction Hobs

2016 ◽  
Author(s):  
Daniele Landi ◽  
Paolo Cicconi ◽  
Michele Germani ◽  
Anna Costanza Russo
Keyword(s):  
Methodological Approach ◽  
Virtual Prototyping ◽  
Thermal Power ◽  
Industrial Applications ◽  
Thermal Simulation ◽  
Operating Conditions ◽  
Electromagnetic Simulation ◽  
Design Data ◽  
Operational Phase ◽  
Whole Life Cycle

Nowadays in many industrial applications, i.e. electrical household appliances, it is necessary to have a robust and safe control for some variables involved in the analysis of the performances of different products. In addition, the recent eco-design directives require products increasingly eco-friendly and eco-efficient, preserving high-performance but a low power consumption. For these reasons, the physical prototypes of products require many expensive and complex tests in term of time, resources and qualified personnel involved. To overcome these limitations, the proposed approach is focused on the use of virtual prototyping tools, which support and reduce the expensive physical experiments. The main objective of this paper is the development, implementation and testing of an innovative methodology, which could be an improvement for the sustainable design of induction hobs. Induction heating applied to the domestic cooking has significantly evolved since the first cooking hobs appeared. Different issues such as maximum power available for heating a pot, dimensional compactness of the hobs, or inverter electronics efficiency have achieved a great development. The proposed methodology provides the development of a multi-physic model which is able to estimate the efficiency of the induction hobs starting from the design data of the project. In particular, the multi-physic model is composed by an electromagnetic simulation and a thermal simulation. The electromagnetic simulation, starting from electrical values such as voltage, current and frequency, is able to simulate the eddy current induced in the bottom of the pot, and resistance leads to the Joulean heating of the material. The thermal simulation is able to measure the energy consumption during the operational phase and the temperature reached by the materials. Therefore, the thermal power obtained by the Joulean heating is, at the same time, the output of the electromagnetic simulation and the input of the thermal one. The proposed model can be applied to design product and simulate the performance considering different operating conditions such as different types of cookers, different coils and different materials. Through the use of virtual prototyping tools is possible to control the heat flux on the whole system (stove, pot, water), and to evaluate the energy efficiency during the operational phase. The proposed tool makes the product-engineer more aware about decision-making strategies in order to achieve an energy saving, calculated over the whole life cycle.

Wear Scar Progression of Impact-Fretting at Elevated Temperature for Steam Generator Tubes in Nuclear Power Plants

2006 ◽  
Vol 326-328 ◽  
pp. 1251-1254 ◽  
Author(s):  
Chi Yong Park ◽  
Jeong Keun Lee
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Wear Scar ◽  
Fretting Wear ◽  
Test Material ◽  
The Impact ◽  
Steam Generator Tubes

Fretting wear generated by flow induced vibration is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Understanding of tube wear characteristics is very important to keep the integrity of the steam generator tubes to secure the safety of the nuclear power plants. Experimental examination has been performed for the purpose of investigating the impact fretting. Test material is alloy 690 tube and 409 stainless steel tube supports. From the results of experiments, wear scar progression is investigated in the case of impact-fretting wear test of steam generator tubes under plant operating conditions such as pressure of 15MPa, high temperature of 290C and low dissolved oxygen. Hammer imprint that is actual damaged wear pattern, has been observed on the worn surface. From investigation of wear scar pattern, wear mechanism was initially the delamination wear due to cracking the hard oxide film and finally transferred to the stable impact-fretting pattern.

Study on Load Capacity of Main Steam Superpipe Nozzle Used in Nuclear Power Plant

2017 ◽  
Author(s):  
Zhou Gengyu ◽  
Liang Shuhua ◽  
Sun Lin ◽  
Lv Feng
Keyword(s):  
Finite Element ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Load Capacity ◽  
Operating Conditions ◽  
Integrity Assessment ◽  
Inner Radius ◽  
Main Steam

The main steam super pipe used in nuclear power plant is an important safety class2 component. There are several nozzles located on it and linked with main steam safety valves. In the past two decades, the hot extrusion forming technology has been widely used to manufacture the super pipe nozzles. Comparing with traditional insert weldolet, the wall thickness of the extruded nozzle is relative small, and the nozzle inner radius is hard to control precisely in the fabrication process. Due to high temperature working condition and complicated loading conditions, the load capacity of the super pipe extruded nozzle has become an issue of concern for manufacturers and users. This paper presents a structural integrity assessment of a super pipe extruded nozzle. The nozzle stresses due to internal pressure and external loads for different operating conditions are obtained by the three-dimensional finite element analysis. The extruded nozzle is evaluated against the RCCM code Subsection C3200 Service Levels O, B and D stress limits for design, upset and faulted conditions. A parametric sensitivity analysis of the extruded nozzle inner radius size is also carried out. In addition, in order to reduce the calculation effort, an efficient calculation method is developed by using the commercial finite element program ANSYS.

The Load Rejection Transient Analysis of Chinshan NPP (BWR/4) Using TRACE/PARCS

2014 ◽  
Author(s):  
Jhih-Jhong Huang ◽  
Hsiung-Chih Chen ◽  
Jong-Rong Wang ◽  
Lih-Yih Liao ◽  
Chunkuan Shih ◽  
...  
Keyword(s):  
Control System ◽  
Nuclear Power ◽  
Thermal Power ◽  
Reactor Core ◽  
Second Step ◽  
System Response ◽  
Recirculation Flow ◽  
Fuel Assemblies ◽  
Flow Control System ◽  
Main Steam

Chinshan Nuclear Power Plant (NPP) is the first boiling water reactor (BWR) NPP in Taiwan. It has two units of BWR/4 reactor made by GE Company and each rated thermal power was 1775 MW without power uprate (now its rated thermal power is 1805 MW after power uprate). This research focuses on the development of the Chinshan NPP TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS (Purdue Advanced Reactor Core Simulator) model. The model is done in two steps: The first step is the development of a TRACES/PARCS model of Chinshan NPP which includes the vessel, fuel assemblies, the main steam lines and important control systems (such as the feedwater control system, recirculation flow control system, etc.). Key parameters (such as power, feedwater flow rate, reactor dome temperature, etc.) were identified to refine the model further in the frame of a steady state analysis. The second step is development of TRACE/PARCS model for the load rejection transient. In order to check the system response of the Chinshan NPP TRACE/PARCS model, this study uses the load rejection transient results of startup tests to benchmark the analysis results of Chinshan NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results were consistent with the startup test data. It indicates that there is a respectable accuracy in the Chinshan NPP TRACE/PARCS model for the load rejection transient.

Crack Initiation Factor by Impact Fretting Wear at INCONEL Alloy for Steam Generator Tube in Nuclear Power Plants

2007 ◽  
Vol 26-28 ◽  
pp. 1269-1272
Author(s):  
Chi Yong Park ◽  
Jeong Kun Kim ◽  
Tae Ryong Kim ◽  
Sun Young Cho ◽  
Hyun Ik Jeon
Keyword(s):  
Steam Generator ◽  
Wear Mechanism ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Fretting Wear ◽  
Steam Generator Tube ◽  
Inconel Alloy ◽  
The Impact

Inconel alloy such as alloy 600 and alloy 690 is widely used as the steam generator tube materials in the nuclear power plants. The impact fretting wear tests were performed to investigate wear mechanism between tube alloy and 409 stainless steel tube support plates in the simulated steam generator operating conditions, pressure of 15MPa, high temperature water of 290°C and low dissolved oxygen(<10 ppb). From investigation of wear test specimens by the SEM and EDS analysis, hammer imprint, which is known to be an actual damaged wear pattern, has been observed on the worn surface, and fretting wear mechanism was investigated. Wear progression of impact-fretting wear also has been examined. It was observed that titanium rich phase contributes to the formation of voids and cracks in sub-layer of fretting wear damage by impact fretting wear.

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