TRACE and RELAP5 Codes for Beyond Design Accident Condition Simulation in the SPES3 Facility

2012 ◽  
Author(s):  
Roberta Ferri ◽  
Fulvio Mascari ◽  
Paride Meloni ◽  
Giuseppe Vella
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Heat Removal ◽  
Nuclear Fission ◽  
Dynamic Coupling ◽  
Hydraulic Conditions ◽  
Loss Of Coolant ◽  
Accident Condition

Code validation on qualified experimental data is a fundamental issue in the design and safety analyses of nuclear power plants. The SPES3 facility is being built at the SIET laboratories for an integral type SMR simulation, in the frame of an R&D program on nuclear fission, funded by the Italian Ministry of Economic Development and led by ENEA. The facility, based on the IRIS reactor design, reproduces the primary, secondary and containment systems with 1:100 volume scale, full elevation and prototypical fluid and thermal-hydraulic conditions. It is suitable to test the plant response to design and beyond design accidents in order to verify the effectiveness of the primary and containment system dynamic coupling to cope with loss of coolant accidents. Full and complete nodalizations of SPES3 were developed for TRACE and RELAP5 codes in order to investigate the code response to the simulation of the same accidental transient. The DVI line DEG break was simulated in beyond design conditions, assuming the failure of all emergency heat removal systems and relying on PCC intervention for containment depressurization and decay heat removal. The comparison of the code simulation results, other than providing information on the system behavior, allowed to investigate specific phenomena evidenced by the codes, according to the related modeling approach of components with one and three-dimensional volumes. The TRACE and RELAP5 codes will be applied for further transient analyses and will be validated on SPES3 experimental data, once the facility will be available.

scholarly journals On the Root Causes of the Fukushima Daiichi Disaster from the Perspective of High Complexity and Tight Coupling in Large-Scale Systems

Symmetry ◽  
2021 ◽  
Vol 13 (3) ◽  
pp. 414
Author(s):  
Atsuo Murata ◽  
Waldemar Karwowski
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
High Complexity ◽  
Tight Coupling ◽  
Large Scale Systems ◽  
Safety Issues ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Fukushima Daiichi

This study explores the root causes of the Fukushima Daiichi disaster and discusses how the complexity and tight coupling in large-scale systems should be reduced under emergencies such as station blackout (SBO) to prevent future disasters. First, on the basis of a summary of the published literature on the Fukushima Daiichi disaster, we found that the direct causes (i.e., malfunctions and problems) included overlooking the loss of coolant and the nuclear reactor’s failure to cool down. Second, we verified that two characteristics proposed in “normal accident” theory—high complexity and tight coupling—underlay each of the direct causes. These two characteristics were found to have made emergency management more challenging. We discuss how such disasters in large-scale systems with high complexity and tight coupling could be prevented through an organizational and managerial approach that can remove asymmetry of authority and information and foster a climate of openly discussing critical safety issues in nuclear power plants.

scholarly journals Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh

2018 ◽  
Vol 4 (4) ◽  
pp. 251-256 ◽  
Author(s):  
Sergey Shcheklein ◽  
Ismail Hossain ◽  
Mohammad Akbar ◽  
Vladimir Velkin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Good Time ◽  
Diagnostics System ◽  
Seismic Impacts

Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools.

scholarly journals Simulation of Steady States of Helium Loop at Long Time Scenarios

2020 ◽  
Vol 328 ◽  
pp. 01009
Author(s):  
František Világi ◽  
Branislav Knížat ◽  
Róbert Olšiak ◽  
Marek Mlkvik ◽  
Peter Mlynár ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Iteration Process ◽  
Heat Accumulation ◽  
Proposed Model ◽  
Long Time ◽  
Convection Cooling

The natural circulation helium loop is an experimental facility designed for the research of the possibility of utilizing natural convection cooling for the case of decay heat removal from a fast nuclear reactor. This concept would bring an improved automated safety system for future nuclear power plants operating a gas-cooled reactor. The article presents a new possibility of direct use of energy conservation laws in a 1D simulation of natural circulation loops. The calculation is performed by a triple iteration process, nested into each other. The results of the calculations showed good agreement with the measurements at steady state. A calculation with the proposed model at unsteady state is not yet possible, especially because of the exclusion of heat accumulation into the material.

A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV

2017 ◽  
Author(s):  
Eltayeb Yousif ◽  
Zhang Zhijian ◽  
Tian Zhao-fei ◽  
A. M. Mustafa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Systems ◽  
Nuclear Power ◽  
Power Plants ◽  
Computational Simulation ◽  
Primary System ◽  
Transient Time ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

To ensure effective operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, Loss of Coolant Accident (LOCA) is one of the main accidents. RELAP-MV Visualized Modularization software technology is recognized as one of the best estimated transient simulation programs of light water reactors, and also has the options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. In this study, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SB-LOCA accident in AP1000 nuclear power plant (NPP) is carried out by Relap5-MV thermo-hydraulics code. While focusing on LOCA analysis in this study, effort was also made to test the effectiveness of the RELAP5-MV software already developed. The accuracy and reliability of RELAP5-MV have been successfully confirmed by simulating LOCA. The calculation was performed up to a transient time of 4,500.0s. RELAP5-MV is able to simulate a nuclear power system accurately and reliably using this modular modeling method. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially for the beginners.

scholarly journals Modelling of Passive Heat Removal Systems: A Review with Reference to the Framatome BWR Reactor KERENA: Part II

Energies ◽  
2019 ◽  
Vol 13 (1) ◽  
pp. 109 ◽  
Author(s):  
René Manthey ◽  
Frances Viereckl ◽  
Amirhosein Moonesi Shabestary ◽  
Yu Zhang ◽  
Wei Ding ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gravitational Force ◽  
Heat Removal ◽  
Dimensional System ◽  
Thermal Gradients ◽  
Two Phase ◽  
Passive Safety Systems ◽  
External Power ◽  
Safety Systems

Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modeling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I dealt with numerical and experimental methods for modeling of condensation inside the emergency condenser and on the containment cooling condenser. This second part deals with boiling and two-phase flow instabilities.

Wave Propagation Analysis of Piping Structures

2009 ◽  
Author(s):  
Akemi Nishida
Keyword(s):  
Wave Propagation ◽  
Dynamic Behavior ◽  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Beam Theory ◽  
Spectral Element ◽  
Analysis Tool ◽  
Propagation Analysis ◽  
Piping Systems

It is becoming important to carry out detailed modeling procedures and analyses to better understand the actual phenomena. Because some accidents caused by high-frequency vibrations of piping have been recently reported, the clarification of the dynamic behavior of the piping structure during operation is imperative in order to avoid such accidents. The aim of our research is to develop detailed analysis tools and to determine the dynamic behavior of piping systems in nuclear power plants, which are complicated assemblages of different parts. In this study, a three-dimensional dynamic frame analysis tool for wave propagation analysis is developed by using the spectral element method (SEM) based on the Timoshenko beam theory. Further, a multi-connected structure is analyzed and compared with the experimental results. Consequently, the applicability of the SEM is shown.

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