Research on Modular Modeling Method of Reactor Coolant System Based on THEATRe

2014 ◽  
Author(s):  
Bo Shi ◽  
Zhao-Fei Tian
Keyword(s):  
Real Time ◽  
Data Exchange ◽  
Reactor Core ◽  
Simulation Software ◽  
Modeling Method ◽  
Time Step ◽  
Coolant Pump ◽  
Modular Modeling ◽  
Reactor Coolant ◽  
Coolant System

At present, research on the reactor coolant system is less yet, though modular modeling method has been widely used in the second-loop system of reactor. This paper takes the reactor coolant system of Qinshan-1 nuclear power plant as the object of study, analyses and researches on modular modeling method of reactor coolant system based on THEATRe, which is a large Thermal-Hydraulic real time simulation software developed by GSE Company and adopts NMNP (Nodal Momentum Nodal Pressure) solving method. This research establishes the modular model of the reactor coolant system equipments (including reactor core, main coolant pump, pressurizer, steam generator) using the THEATRe code. Due to each module is wrote into through different input cards, they can be solved by using their own matrix of velocity-pressure to guarantee the independence of the numerical calculation for different modular modules. THEATRe code does not have its own TDV like relap-5, meanwhile it also needs to ensure the pressurizer module can play a role in the multi-pressure node system. So this paper modifies solving method of the THEATRe source code to get suitable pressure boundary and flux boundary for RCS equipment modular module, and selects reasonable time step and data exchange frequency to achieve the data exchange of boundary pressure, flux and enthalpy among the equipment modules, which lays the foundation of establishing the real-time modular simulation model of the reactor coolant system in the future.

A Modular Code for Real Time Dynamic Simulation of Gas Turbines in Simulink

2002 ◽  
Vol 128 (3) ◽  
pp. 506-517 ◽  
Author(s):  
S. M. Camporeale ◽  
B. Fortunato ◽  
M. Mastrovito
Keyword(s):  
Mathematical Model ◽  
Real Time ◽  
Gas Turbine ◽  
Dynamic Behavior ◽  
Gas Turbines ◽  
Simulation Software ◽  
Computational Time ◽  
High Fidelity ◽  
Time Step ◽  
The Mathematical Model

A high-fidelity real-time simulation code based on a lumped, nonlinear representation of gas turbine components is presented. The code is a general-purpose simulation software environment useful for setting up and testing control equipments. The mathematical model and the numerical procedure are specially developed in order to efficiently solve the set of algebraic and ordinary differential equations that describe the dynamic behavior of gas turbine engines. For high-fidelity purposes, the mathematical model takes into account the actual composition of the working gases and the variation of the specific heats with the temperature, including a stage-by-stage model of the air-cooled expansion. The paper presents the model and the adopted solver procedure. The code, developed in Matlab-Simulink using an object-oriented approach, is flexible and can be easily adapted to any kind of plant configuration. Simulation tests of the transients after load rejection have been carried out for a single-shaft heavy-duty gas turbine and a double-shaft aero-derivative industrial engine. Time plots of the main variables that describe the gas turbine dynamic behavior are shown and the results regarding the computational time per time step are discussed.

Flow Transient in Primary Coolant System During Reactor Coolant Pump Start-Up Period

2013 ◽  
Author(s):  
Hong Gao ◽  
Feng Gao ◽  
Xianchao Zhao ◽  
Jie Chen ◽  
Xuewu Cao
Keyword(s):  
Primary Coolant ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Start Up ◽  
Coolant System

scholarly journals Decontamination of a Contaminated RCP Shaft Using the SP-HyBRID Process

Processes ◽  
2021 ◽  
Vol 9 (10) ◽  
pp. 1725
Author(s):  
Hee-Chul Eun ◽  
Na-On Chang ◽  
Wang-Kyu Choi ◽  
Sang-Yoon Park ◽  
Seon-Byeong Kim ◽  
...  
Keyword(s):  
Decontamination Factor ◽  
Hybrid Process ◽  
Waste Generation ◽  
Nuclear Facility ◽  
Experimental Conditions ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Pump Shaft ◽  
Shaft Surface ◽  
Coolant System

It is very important to minimize the waste generation for decontamination of the reactor coolant system in a nuclear facility. As an alternative to commercial decontamination technologies, an inorganic acid chemical decontamination (SP-HyBRID) process can be effectively applied to the decontamination because it can significantly reduce the waste generation. In this study, the decontamination of a contaminated reactor coolant pump shaft from a nuclear facility was conducted using the SP-HyBRID process. First, equipment for a mock-up test of the decontamination was prepared. Detailed experimental conditions for the decontamination were determined through the mock-up test. Under the detailed conditions, the contaminated shaft was successfully decontaminated. The dose rate on the shaft surface was greatly reduced from 1400 to 0.9 μSv/h, and the decontamination factor showed a very high value (>1500).

scholarly journals ANALISIS DESAIN PROSES SISTEM PENDINGIN PADA REAKTOR RISET INOVATIF 50 MW

2015 ◽  
Vol 17 (1) ◽  
pp. 19 ◽  
Author(s):  
Sukmanto Dibyo ◽  
Endiah Puji Hastuti ◽  
Ign. Djoko Irianto
Keyword(s):  
Mass Flow Rate ◽  
Mass Flow ◽  
Flow Rate ◽  
Cooling System ◽  
Reactor Core ◽  
Material Testing ◽  
Thermal Design ◽  
Coolant Pump ◽  
Energy Balances ◽  
Coolant System

Reaktor Riset Inovatif (RRI) merupakan jenis MTR (Material Testing Reactor) yang dipersiapkan ke depan sebagai desain reaktor baru. Daya RRI telah ditetapkan dari perhitungan neutronik dan termohidrolika teras yaitu 50 MW termal. Reaktor bertekanan 8 kgf/cm2 dan laju aliran massa pendingin primer 900 kg/s. Tantangan yang penting dalam menindak lanjuti desain reaktor ini adalah analisis desain pada sistem pendingin. Makalah ini bertujuan untuk menganalisis desain proses sistem pendingin utama reaktor RRI daya 50 MW (RRI-50) dengan menggunakan program Chemcad 6.1.4. Dalam analisis ini dilakukan perhitungan neraca massa dan energi (mass/energy balances) pada sistem pendingin primer dan sekunder sebagai pendingin utama. Masing-masing sistem pendingin tersebut terdiri dari 2 jalur beroperasi secara paralel dan 1 jalur redundansi. Disamping itu untuk desain termal unit komponen telah dianalisis dengan program RELAP5, frenchcreek dan Metoda Analitik. Hasil analisis yang diperoleh adalah desain diagram sistem pendingin yang mencakup data parameter entalpi, temperatur, tekanan dan laju aliran massa pendingin untuk masing-masing jalur. Adapun hasil desain unit komponen utama pada RRI-50 adalah tangki tunda dengan volume 51,5 m3, 2 unit pompa sentrifugal dan 1 unit pompa cadangan pada pendingin primer daya 141 kW/pompa dan pendingin sekunder daya 206 kW/pompa, 2 unit penukar panas tipe shell-tube dengan koefisien termal overall 1377 W/m2.oC dan 4 unit menara pendingin yang mampu melepaskan panas ke udara dengan desain temperatur approach 5,0 oC dan temperatur range 9,0 oC. Desain sistem pendingin reaktor RRI-50 ini telah menetapkan parameter operasi sistem pendingin yaitu temperatur, tekanan dan laju aliran massa pendingin dengan mempertimbangkan tuntutan aspek keselamatan teras reaktor sehingga desain temperatur maksimum pendingin masuk ke teras 44,5 oC. Kata kunci : RRI 50 MW, desain sistem pendingin, program Chemcad 6.1.4   Innovative Research Reactor RRI is a type of MTR (Material Testing Reactor), which is being prepared in the future as a design of new reactor. The power of RRI has been determined based on the core thermalhydraulic and neutronic calculation, which is 50 MWt. The reactor pressure is 8 kgf/cm 2 and coolant mass flow rate is 900 kg/s. The important challenge in the follow up of this reactor design is the design analysis of cooling system. The purpose of this study is to analyze the design of RRI reactor main coolant system at the power of 50 MWt (RRI-50) using ChemCAD 6.1.4. In this analysis the mass and energy balances at the primary and secondary cooling system are calculated as main coolant. Each of the cooling system consists of two lines operating in parallel and redundancy lines. Besides that, the thermal design of the component units have been analyzed using RELAP5, FrenchCreek and Analytical Methods. The analyses result obtained is a design of cooling system diagram which includes parameter of enthalpy, temperature, pressure and coolant mass flow rate of each line. Meanwhile, design result of main component unit are delay tank of 51.5 m3 volume, 2 unit centrifugal pumps and 1 unit stand-by pump for the primary coolant pump each of 141 kW power and secondary coolant pump each of 206 kW power, 2 unit of shell-tube heat exchanger with overall thermal coefficient of 1377 W/m2.oC and 4 unit cooling tower that capable to release the heat to the air at approach temperature of 5,0 oC and range temperature of 9,0 oC. design of reactor coolant system RRI-50 has decided the operating parameters of cooling system are temperature, pressure and mass flow rate by considering into the demands of the safety aspects of the reactor core therefore design of maximum coolant temperature to the reactor core is 44,5 oC. Keywords : RRI 50MW,  design of cooling system, program Chemcad 6.1.4.

Analysis of reactor coolant pump transient performance in primary coolant system during start-up period

2013 ◽  
Vol 54 ◽  
pp. 202-208 ◽  
Author(s):  
Hong Gao ◽  
Feng Gao ◽  
Xianchao Zhao ◽  
Jie Chen ◽  
Xuewu Cao
Keyword(s):  
Transient Performance ◽  
Primary Coolant ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Start Up ◽  
Coolant System

Margin Assessment of AP1000 Loss of Flow Transient

2006 ◽  
Author(s):  
Edward L. Carlin ◽  
Peter A. Hilton ◽  
Yixing Sung
Keyword(s):  
Three Dimensional ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Improved Method ◽  
Margin Assessment ◽  
Reactor Coolant ◽  
Power Profile ◽  
Canned Motor ◽  
Reactivity Insertion ◽  
Coolant System

The Reactor Coolant System (RCS) of the AP1000 plant consists of two circulating loops. Each loop contains two canned motor Reactor Coolant (RC) pumps that have a rotating inertia to provide RCS flow coastdown if power to the pumps is lost. Westinghouse analysis of the complete loss of flow (CLOF) accident in support of the AP1000 design certification was based on the USNRC-approved traditional methodology applied to operating plants. The RCS response during the transient was predicted using the LOFTRAN code based on a reactivity insertion curve highly skewed to the bottom of the reactor core, but the calculation of Departure from Nucleate Boiling Ratio (DNBR) was performed assuming a top-skewed axial power profile. A more realistic margin assessment can be made by using an improved method similar to Westinghouse RAVE methodology recently approved by the USNRC. The improved method uses the three-dimensional kinetic nodal code SPNOVA coupled with the reactor core thermal-hydraulic code VIPRE-W for predicting the reactor core response during the CLOF transient. The improved method significantly improves margin predictions by generating core power distributions consistent with the trip reactivity changes for the DNBR calculation. The margin assessment showed that the improved method resulted in a 19% DNBR increase as compared to the traditional method for the AP1000 CLOF transient.

scholarly journals Hardware in the Loop Real-Time Simulation for the Associated Discrete Circuit Modeling Optimization Method of Power Converters

Energies ◽  
2018 ◽  
Vol 11 (11) ◽  
pp. 3237 ◽  
Author(s):  
Xizheng Guo ◽  
Jiaqi Yuan ◽  
Yiguo Tang ◽  
Xiaojie You
Keyword(s):  
Power Electronics ◽  
Real Time ◽  
Optimization Method ◽  
Modeling Method ◽  
Resource Optimization ◽  
Switching Frequency ◽  
Hardware In The Loop ◽  
Time Step ◽  
Real Time Simulation ◽  
Time Simulation

Due to the complicated circuit topology and high switching frequency, field-programmable gate arrays (FPGA) can stand up to the challenges for the hardware in the loop (HIL) real-time simulation of power electronics converters. The Associated Discrete Circuit (ADC) modeling method, which has a fixed admittance matrix, greatly reduces the computation cost for FPGA. However, the oscillations introduced by the switch-equivalent model reduces the simulation accuracy. In this paper, firstly, a novel algorithm is proposed to determine the optimal discrete-time switch admittance parameter, Gs, which is obtained by minimizing the switching loss. Secondly, the FPGA resource optimization method, in which the simulation time step, bit-length, and model precision are taken into consideration, is presented when the power electronics converter is implemented in FPGA. Finally, the above method is validated on the topology of a three-phase inverter with LC filters. The HIL simulation and practicality experiments verify the effect of FPGA resource optimization and the validity of the ADC modeling method, respectively.

Nonlinear Dynamic Analysis of RCS LOCA Based on Secondary Development of ANSYS

2011 ◽  
Vol 105-107 ◽  
pp. 334-338
Author(s):  
Huan Huan Qi ◽  
Zhong Xiu Zeng
Keyword(s):  
Nonlinear Dynamic ◽  
Nuclear Reactor ◽  
System Modeling ◽  
Nonlinear Dynamic Analysis ◽  
Secondary Development ◽  
Modular System ◽  
Coolant Pump ◽  
Modular Modeling ◽  
Reactor Coolant ◽  
Loss Of Coolant Accident

Research of loss of coolant accident (LOCA) postulated on the nuclear reactor coolant system (RCS) was investigated with ANSYS program. Secondary development of ANSYS was performed to form the customized module for implementing effective and efficient RCS LOCA nonlinear analysis. A standard analysis procedure was established. It has following functions, such as parameters and modular system modeling, automatically set break, static pilot analysis, LOCA nonlinear dynamic calculation and automatic reports generation. Main pipes, Steam generator and reactor coolant pump models were molded by parameterized modular modeling method. Those models considered the nonlinear factors, such as material nonlinearities, gap and so on, constructed component model libraries of RCS. Comparing the results calculated by ANSYS and program-specific, it is showed that the results are generally consistent.

Analysis of the anticipated transient without scram (ATWS) initiated by emergency power mode through the full scope simulator

Kerntechnik ◽  
2022 ◽  
Vol 0 (0) ◽  
Author(s):  
Alexandre de Souza Soares ◽  
Antonio C. M. Alvim
Keyword(s):  
Reactor Core ◽  
Heat Removal ◽  
Reactor Power ◽  
Operating Personnel ◽  
Reactor Coolant ◽  
Operational Variables ◽  
Emergency Power ◽  
Power Mode ◽  
Reactor Shutdown ◽  
Coolant System

Abstract The integrity of the reactor coolant system is severely challenged as a result of an Emergency Power Mode – ATWS event. The purpose of this paper is to simulate the Anticipated Transient without Scram (ATWS) using the full scope simulator of Angra 2 Nuclear Power Plant with the Emergency Power Case as a precursor event. The results are discussed and will be used to examine the integrity of the reactor coolant system. In addition, the results were compared with the data presented in Final Safety Analysis Report (FSAR – Angra 2) in order to guarantee the validation of the methodology and from there analyze other precursor events of ATWS which presented only plausibility studies in FSAR – Angra 2. In this way, the aim is to provide and develop the knowledge and skill necessaries for control room operating personnel to ensure safe and reliable plant operation and stimulate information in the nuclear area through the academic training of new engineers. In the presented paper the most severe scenario is analyzed in which the Reactor Coolant System reaches its highest level of coolant pressure. This scenario is initiated by the turbine trip jointly with the loss of electric power systems (Emergency Power Mode). In addition, the failure of the reactor shutdown system occurs, i.e., control rods fail to drop into the reactor core. The reactor power is safely reduced through the inherent reactivity feedback of the moderator and fuel, together with an automatic boron injection. Several operational variables were analyzed and their profiles over time are shown in order to provide data and benchmarking references. At the end of the event, it was noted that Reactor shutdown is assured, as is the maintenance of subcriticality. Residual heat removal is ensured.

The IRIS Spool-Type Reactor Coolant Pump

2002 ◽  
Author(s):  
J. M. Kujawski ◽  
D. M. Kitch ◽  
L. E. Conway
Keyword(s):  
Reactor Core ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Steam Generators ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
The Us ◽  
Loss Of Coolant ◽  
The Individual ◽  
Integral Reactor

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called “spool” pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper.

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