Modeling of Steam Explosion in Partially Flooded PWR Reactor Cavity

2005 ◽  
Author(s):  
Bosˇtjan Koncˇar ◽  
Matjazˇ Leskovar ◽  
Leon Cizelj
Keyword(s):  
Steam Explosion ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Fission Products ◽  
General Purpose ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fit For Purpose ◽  
Computational Fluid Dynamics Cfd ◽  
Reactor Accidents

When the hot molten core comes into contact with the water in the reactor cavity a steam explosion can occur. The steam explosion might be triggered during some scenarios of severe nuclear reactor accidents, when extremely hot molten nuclear fuel interacts with the coolant water. A highly energetic steam explosion in a nuclear power plant could cause the containment failure and the release of radioactive fission products to the environment. The purpose of the performed analysis is to provide a first estimation of the expected pressure loadings on the typical PWR cavity structures during a steam explosion. To achieve this, the fit-for-purpose steam explosion model is proposed, followed by a Computational Fluid Dynamics (CFD) analysis. In the present work two steam explosion scenarios in the partially flooded Pressurized Water Reactor (PWR) cavity were simulated with the general purpose code CFX-5 [1] to estimate pressure loadings on cavity walls.

scholarly journals Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

2021 ◽  
Vol 13 (14) ◽  
pp. 7964
Author(s):  
Alain Flores y Flores ◽  
Danilo Ferretto ◽  
Tereza Marková ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Accident Simulation ◽  
Experimental Facilities ◽  
The Difference

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

Multiphysics Modeling of Pressurized Water Reactor Fuel Performance

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Wang Zhu ◽  
Zhang Chungyu ◽  
Yuan Cenxi
Keyword(s):  
Three Dimensional ◽  
Fission Products ◽  
Axial Direction ◽  
General Purpose ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Rods ◽  
Reactivity Insertion

Nuclear fuel rods operate under complex radioactive, thermal, and mechanical conditions. Nowadays, fuel rod codes usually make great simplifications on analyzing the multiphysics behavior of fuel rods. The present study develops a three-dimensional (3D) module within the framework of a general-purpose finite element solver, i.e., abaqus, for modeling the major physics of the fuel rods. A typical fuel rod, subjected to stable operations and transient conditions, is modeled. The results show that the burnup levels have an important influence on the thermomechanical behavior of fuel rods. The swelling of fission products causes a dramatically increasing strain of pellets. The variation of the stress and the radial displacement of the cladding along the axial direction can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress in the outer regime of the pellet and may cause further fragmentation to the pellets. Fission products migration effects and differential thermal expansion become more severe if there are flaws or imperfections on the pellet.

scholarly journals Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

2011 ◽  
Vol 32 (4) ◽  
pp. 67-79
Author(s):  
Tomasz Bury
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Computer Code ◽  
Hydrogen Combustion ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Lumped Parameter ◽  
Loss Of Coolant

Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

Design Process for an Advanced Reactor Coolant Pump for a 1400 MW Nuclear Power Plant

2005 ◽  
Author(s):  
Claus Knierim ◽  
Sven Baumgarten ◽  
Jochen Fritz ◽  
Michael T. Coon
Keyword(s):  
Fluid Dynamics ◽  
Hydraulic System ◽  
Nuclear Power ◽  
Pressurized Water Reactor ◽  
Power Station ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Computational Fluid Dynamics Cfd ◽  
Specific Speed

As part of the planning activities for a new 1400 MW nuclear power station with a pressurized water reactor, a new hydraulic system had to be designed for the reactor coolant pumps (RCPs). Starting from the design principles and the main dimensions of an existing pump a new diffuser and impeller had to be designed for the specified requirements which provided a specific speed of almost ns = 100 rpm (≈ 5100 in US units). The authors describe how impeller and diffuser of the hydraulic system were gradually optimized with the aid of computational fluid dynamics (CFD). The system had to meet demanding requirements, thus it was decided to build a model pump (on a scale ≈ 1:2) to demonstrate that the new pump would satisfy the specified duty parameters. Based on the spectra of tests performed on the model pump the authors discuss the resulting pump characteristics (Q, H, η, NPSH).

SuperCritical Water-Cooled Nuclear Reactor With Intermediate Heat Exchangers

2010 ◽  
Author(s):  
H. Thind ◽  
I. Pioro ◽  
G. Harvel
Keyword(s):  
Heat Exchanger ◽  
Supercritical Water ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Primary Loop ◽  
Reactor Building

At present, there are a number of Generation-IV nuclear reactor concepts under development worldwide, and the SuperCritical Water-cooled nuclear Reactor (SCWR) type is one of them. The main objective of developing SCWRs is to: 1) Increase the thermal efficiency of current Nuclear Power Plants (NPPs) from 30–35% to approximately 45–50%, and 2) Decrease capital and operational costs. SCW NPPs will have much higher operating parameters compared to current NPPs (i.e., pressures of about 25 MPa and outlet temperatures up to 625°C). This paper presents a SCWR single-reheat indirect cycle concept with intermediate heat exchangers. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, heat exchangers separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in the reactor building. The nuclear activities stay within the reactor building, and there is a reduced possibility for radioactive contamination of equipment in the turbine building. As SCW NPPs will have much higher operating thermal hydraulic parameters this paper analyzes the technical challenges and higher costs typically associated with heat exchangers. The double-pipe heat exchanger is analyzed in depth to determine the heat-transfer surface area, number of units and physical dimensions of the heat exchanger. This study will help to determine whether the advantages of the indirect cycle justify implementation of heat exchangers at a SCW NPP.

The Failure Analysis and Processing of Digital Reactor Protection System

2013 ◽  
Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Communication Failure ◽  
Pressurized Water ◽  
Power Failure ◽  
Self Test ◽  
Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

Depletion Calculation for a Nodal Reactor Physics Code

2010 ◽  
Author(s):  
Antonio Carlos Marques Alvim ◽  
Fernando Carvalho da Silva ◽  
Aquilino Senra Martinez
Keyword(s):  
Diffusion Equation ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Expansion Method ◽  
Design System ◽  
Pressurized Water Reactor ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Nodal Method ◽  
Orthogonal Polynomial Expansion

This paper deals with an alternative numerical method for calculating depletion and production chains of the main isotopes found in a pressurized water reactor. It is based on the use of the exponentiation procedure coupled to orthogonal polynomial expansion to compute the transition matrix associated with the solution of the differential equations describing isotope concentrations in the nuclear reactor. Actually, the method was implemented in an automated nuclear reactor core design system that uses a quick and accurate 3D nodal method, the Nodal Expansion Method (NEM), aiming at solving the diffusion equation describing the spatial neutron distribution in the reactor. This computational system, besides solving the diffusion equation, also solves the depletion equations governing the gradual changes in material compositions of the core due to fuel depletion. The depletion calculation is the most time-consuming aspect of the nuclear reactor design code, and has to be done in a very precise way in order to obtain a correct evaluation of the economic performance of the nuclear reactor. In this sense, the proposed method was applied to estimate the critical boron concentration at the end of the cycle. Results were compared to measured values and confirm the effectiveness of the method for practical purposes.

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