Modeling and Simulation of Subcooled Coolant Loss Through Circumferential Pipe Leakage

2020 ◽  
Author(s):  
Sandip Ghosh ◽  
Samir Saha
Keyword(s):  
Flow Behavior ◽  
High Energy ◽  
Computational Techniques ◽  
Narrow Slit ◽  
Thermal Systems ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Experimental Findings ◽  
Coolant Loss ◽  
Leak Before Break

Abstract The present work demonstrates the leak flow behavior of subcooled water at high pressure and high temperature through a narrow slit analogous to a pipe crack. The modeling and simulations are based on the loss of coolant accident in the primary loop piping of pressurized water-cooled reactors where a subcooled liquid is subjected to a rapid depressurization. Prediction of critical leak flow pattern is crucial in the design methodology of costly high energy pipelines in the perspective of leak before break consideration. Computational techniques have been used to replace costly experiments required for simulating leak flow conditions. For a variety of entry and exit conditions, the interphase mass transfer was studied with a change of boundary conditions. Presence of thermodynamic nonequilibrium has been detected on several occasions due to high transit velocities. A comparison with experimental findings indicates the validity of the flashing model for safety analysis of similar high energy thermal systems.

Rupture Hardware Minimization in Pressurized Water Reactor Piping

1989 ◽  
Vol 111 (1) ◽  
pp. 64-71 ◽  
Author(s):  
S. K. Mukherjee ◽  
J. J. Szy Slow Ski ◽  
V. Chexal ◽  
D. M. Norris ◽  
N. A. Goldstein ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Ferritic Steel ◽  
High Energy ◽  
Pressurized Water Reactor ◽  
Pipe Failure ◽  
Pressurized Water ◽  
Screening Criteria ◽  
Water Reactor ◽  
Leak Before Break ◽  
Generic Sense

For much of the high-energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but also improves the overall safety and integrity of the plant since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station—Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in. (152-mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel line as small as 3-in. (76-mm) diameter (outside containment) can qualify for pipe rupture hardware elimination.

Analysis of Protein-Protein Interaction Networks through Computational Approaches

2020 ◽  
Vol 27 (4) ◽  
pp. 265-278 ◽  
Author(s):  
Ying Han ◽  
Liang Cheng ◽  
Weiju Sun
Keyword(s):  
Protein Interaction ◽  
Biological Networks ◽  
Interaction Networks ◽  
Computational Techniques ◽  
Cellular Functions ◽  
Protein Protein Interaction ◽  
Comprehensive Information ◽  
Protein Interaction Prediction ◽  
Or Gene ◽  
Experimental Findings

The interactions among proteins and genes are extremely important for cellular functions. Molecular interactions at protein or gene levels can be used to construct interaction networks in which the interacting species are categorized based on direct interactions or functional similarities. Compared with the limited experimental techniques, various computational tools make it possible to analyze, filter, and combine the interaction data to get comprehensive information about the biological pathways. By the efficient way of integrating experimental findings in discovering PPIs and computational techniques for prediction, the researchers have been able to gain many valuable data on PPIs, including some advanced databases. Moreover, many useful tools and visualization programs enable the researchers to establish, annotate, and analyze biological networks. We here review and list the computational methods, databases, and tools for protein−protein interaction prediction.

Analysis of loss of coolant accident without ECCS and DHRS in an integral pressurized water reactor using RELAP/SCDAPSIM

2021 ◽  
Vol 134 ◽  
pp. 103648
Author(s):  
Katarzyna Skolik ◽  
Chris Allison ◽  
Judith Hohorst ◽  
Mateusz Malicki ◽  
Marina Perez-Ferragut ◽  
...  
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

scholarly journals Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison
Keyword(s):  
Driving Forces ◽  
Natural Circulation ◽  
Safety System ◽  
Operating Conditions ◽  
Severe Accident ◽  
Passive Safety ◽  
System Component ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Accident Conditions

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

scholarly journals Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach

Nukleonika ◽  
2015 ◽  
Vol 60 (2) ◽  
pp. 339-345 ◽  
Author(s):  
Tomasz Bury
Keyword(s):  
Hydrogen Generation ◽  
Cooling System ◽  
Component Mixture ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Issues ◽  
Loss Of Coolant Accident ◽  
Lumped Parameter ◽  
Core Cooling ◽  
Parameter Approach

Abstract The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal-hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational efficiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.

LOCA Simulation: Analysis of Rarefaction Waves Propagating Through Geometric Singularities

2011 ◽  
Author(s):  
Fabien Crouzet ◽  
Vincent Faucher ◽  
Pascal Galon ◽  
Philippe Piteau ◽  
Patrick Izquierdo
Keyword(s):  
Simulation Analysis ◽  
Rarefaction Waves ◽  
Test Rig ◽  
Transient Wave ◽  
Pressurized Water ◽  
Dissipation Term ◽  
Loss Of Coolant Accident ◽  
Nuclear Plants ◽  
Proposed Model ◽  
Single Pipe

The propagation of a transient wave through an orifice is investigated for applications to Loss Of Coolant Accident in nuclear plants. An analytical model is proposed for the response of an orifice plate and implemented in the EUROPLEXUS fast transient dynamics software. It includes an acoustic inertial effect in addition to a quasi-steady dissipation term. The model is experimentally validated on a test rig consisting in a single pipe filled with pressurized water. The test rig is designed to generate a rapid depressurization of the pipe, by means of a bursting disk. The proposed model gives results which compare favourably with experimental data.

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