RELAP5-3D Code Validation for RBMK-1500 Reactors

2004 ◽  
Author(s):  
Evaldas Bubelis ◽  
Algirdas Kaliatka ◽  
Eugenijus Uspuras
Keyword(s):  
Reasonable Agreement ◽  
3D Model ◽  
Reactor Core ◽  
Measured Data ◽  
Primary Circuit ◽  
Hydraulic Parameters ◽  
Physical Processes ◽  
Code Validation ◽  
Calculation Results ◽  
Plant Data

This paper deals with RELAP5-3D code validation through the modeling of RBMK-1500 specific transients taking place at Ignalina NPP. A successful best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data, which demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors.

RELAP5-3D Code Application for RBMK-1500 Reactor Core Analysis

2002 ◽  
Author(s):  
Evaldas Bubelis ◽  
Algirdas Kaliatka ◽  
Eugenijus Uspuras
Keyword(s):  
3D Model ◽  
Reactor Core ◽  
Core Analysis ◽  
Reactor Operation ◽  
The Core ◽  
Calculation Results ◽  
Plant Data ◽  
Comparison Of The Results ◽  
Real Plant ◽  
Good Agreement

The paper presents an evaluation of RELAP5-3D code suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. A successful best estimate RELAP5-3D model of the Ignalina NPP RBMK-1500 reactor has been developed and validated against real plant data. Certain RELAP5-3D transient calculation results were benchmarked against calculation results obtained using the Russian code STEPAN, specially designed for RBMK reactor analysis. Comparison of the results obtained, using the RELAP5-3D and STEPAN codes, showed quite good mutual coincidence of the calculation results and good agreement with real plant data.

A Data-Constrained 3D Model for Material Compositional Microstructures

2008 ◽  
Vol 32 ◽  
pp. 267-270 ◽  
Author(s):  
Sam Yang ◽  
Scott Furman ◽  
Andrew Tulloh
Keyword(s):  
Mathematical Model ◽  
Material Properties ◽  
Reasonable Agreement ◽  
3D Model ◽  
Compositional Data ◽  
Simulated Data ◽  
Measured Data ◽  
Porosity Distribution ◽  
Numerical Predictions ◽  
Tomography Data

A mathematical model has been developed for predicting material compositional microstructures using measured data as constraints. Examples of measured data include 3-D sets of tomography data, 2-D sets of compositional data on surfaces and sections, and material absorption and interaction properties. The model has been partially implemented as a MS-Windows application. Reasonable agreement has been obtained between the numerical predictions from the software and the simulated data. The predicted microstructures could be used to study various material properties such as porosity distribution, diffusion and corrosion.

scholarly journals RELAP5/MOD3.3 Code Validation with Plant Abnormal Event

2008 ◽  
Vol 2008 ◽  
pp. 1-10 ◽  
Author(s):  
Andrej Prošek ◽  
Borut Mavko
Keyword(s):  
Nuclear Power ◽  
Calculated Data ◽  
Computer Code ◽  
Measured Data ◽  
Power Reduction ◽  
Code Validation ◽  
Periodic Testing ◽  
Plant Data ◽  
Input Model

Measured plant data from various abnormal events are of great importance for code validation. The purpose of the study was to validate the RELAP5/MOD3.3 Patch 03 computer code with the abnormal event which occurred at Krško Nuclear Power Plant (NPP) on April 10, 2005. The event analyzed was a malfunction, which occurred during a power reduction sequence when regular periodic testing of the turbine valves was performed. Unexpected turbine valve closing caused safety injection signal, followed by reactor trip. The RELAP5 input model delivered by Krško NPP was used. In short term, the calculation agrees very well with the plant measured data. In the long term, this is also true when operator actions and special plant systems are modeled. In the opposite, the transient would progress quite differently. Finally, the calculated data may be supplemental to plant measured data when the information is missing or the measurement is questionable.

Kinematics and Dynamics Simulation Research for Roller Coaster Multi-Body System

2011 ◽  
Vol 421 ◽  
pp. 276-280 ◽  
Author(s):  
Ge Ning Xu ◽  
Hu Jun Xin ◽  
Feng Yi Lu ◽  
Ming Liang Yang
Keyword(s):  
Structural Design ◽  
3D Model ◽  
Structure Design ◽  
Dynamics Simulation ◽  
Body System ◽  
Load Spectrum ◽  
Roller Coaster ◽  
Simulation Research ◽  
Calculation Results ◽  
Multi Body

To assess the roller coaster multi-body system security, it is need to extract the running process of kinematics, dynamics, load spectrum and other features, as basis dates of the roller coaster structural design. Based on Solidworks/motion software and in the 3D model, the calculation formula of the carrying car velocity and acceleration is derived, and the five risk points of the roller coaster track section are found by simulation in the running, and the simulation results of roller coaster axle mass center velocity are compared with theoretical calculation results, which error is less than 4.1%, indicating that the calculation and simulation have a good fit and providing the evidence for the roller coaster structure design analysis.

Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Design Solution ◽  
Fast Neutrons ◽  
Steel Shell ◽  
Primary Circuit ◽  
Coated Particles ◽  
The Core ◽  
Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

Analysis of Passive Residual Heat Removal System at Primary Side when Station Blackout Occurs

2014 ◽  
Vol 986-987 ◽  
pp. 231-234
Author(s):  
Jun Teng Liu ◽  
Qi Cai ◽  
Xia Xin Cao
Keyword(s):  
Nuclear Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Saturation Temperature ◽  
Heat Removal ◽  
Primary Circuit ◽  
Station Blackout ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

This paper regarded CNP1000 power plant system as the research object, which is the second-generation half Nuclear Reactor System in our country, and tried to set Westinghouse AP1000 passive residual heat removal system to the primary circuit of CNP1000. Then set up a simulation model based on RELAP5/MOD3.2 program to calculate and analyze the response and operating characteristic of passive residual heat removal system on assumption that Station Blackout occurs. The calculation has the following conclusions: natural circulation was quickly established after accident, which removes core residual heat effectively and keep the core safe. The residual heat can be quickly removed, and during this process the actual temperature was lower than saturation temperature in reactor core.

scholarly journals Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Author(s):  
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Computational Study ◽  
Reactor Core ◽  
Primary Circuit ◽  
Feed Water ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

Thermo-chemical model of the Pierce-Smith copper converter

2007 ◽  
Vol 221 (3) ◽  
pp. 129-138 ◽  
Author(s):  
N Aminizadeh ◽  
S. H. Mansouri
Keyword(s):  
Reaction Rates ◽  
Oxygen Enrichment ◽  
Measured Data ◽  
Chemical Model ◽  
Process Automation ◽  
Melt Temperature ◽  
Copper Converter ◽  
Parametric Studies ◽  
Oxygen Efficiency ◽  
Plant Data

A thermo-chemical model of the Pierce-Smith copper converter has been developed to predict the melt temperature. The primary assumption of the model was that all the matte, the slag, and the gaseous phases in the converter were in thermal and chemical equilibrium. In this model, the matte components considered were copper, iron, and sulphur elements. In the slag-making stage, all the reacted iron was converted to fayalite. On the basis of the model prediction and comparison with the plant data, the reaction rates of iron and oxygen in the slag-making stage and sulphur and oxygen in the blowing copper stage were assumed to be mixing controlled. The converter temperature was calculated by heat balance over time. The model was validated by detailed comparisons with measured industrial data. Good agreements between predicted and measured data were obtained. The results of parametric studies show that the melt temperature in the converter depends on the process parameters such as oxygen efficiency, melt emissivity, and oxygen enrichment. The results of this research can be used for process automation and optimizing.

Study on Heating Process of Dehumidifying Experiment in HTGR

2020 ◽  
Author(s):  
Kaiyue Shen ◽  
Wei Zheng ◽  
Shengchao Ma ◽  
Huaqiang Yin ◽  
Xuedong He ◽  
...  
Keyword(s):  
High Temperature ◽  
Kinetic Energy ◽  
Thermal Energy ◽  
Reactor Core ◽  
Primary Circuit ◽  
Heating Process ◽  
Transfer Model ◽  
Thermal Source ◽  
Flow And Heat Transfer ◽  
Selection Of

Abstract A large number of carbon materials are used in high temperature gas-cooled reactor (HTGR). As a kind of porous material, the carbon material contains a certain amount of moisture and other impurities. In order to reduce the corrosion of internal material in reactor core of HTGR, the initial core or post-accident core must be strictly heated and dehumidified. The current primary circuit heating mainly relies on the rotation of the primary pump to convert the kinetic energy into thermal energy. Obviously, the current scheme was flawed: (1) Due to the insufficient heat generated by rotation of the primary pump, the temperature rising process of the primary circuit is sluggish; (2) The rotation of the primary pump converts the kinetic energy into thermal energy of the helium, at the meantime, the primary circuit dissipates heat outward. For the above reasons, it is difficult to achieve the desired dehumidification temperature in the heating process. While in this paper, an additional thermal source will be added to the steam generator to heat the primary circuit in a new scheme. A proper flow and heat-transfer model of heating the primary circuit in high-temperature reactor was established based on software COMSOL Multiphysics. The numerical analysis of the primary circuit heating process provides rewarding guidance for the selection of the dehumidification scheme in HTGR.

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