CFD Simulation of Debris Transport in a PWR Reactor Sump Following a Loss of Coolant Event

2005 ◽  
Author(s):  
Stuart A. Cain ◽  
Fariba Gartland ◽  
Andrew E. Johansson
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Cfd Simulation ◽  
Energy Levels ◽  
High Energy ◽  
Material Transport ◽  
Pressurized Water ◽  
Laboratory Flume ◽  
Debris Transport

High Energy Line Breaks (HELBs) inside nuclear reactor containment are recognized as challenges to Pressurized Water Reactor (PWR) nuclear power plants arising from the collateral damage due to insulation, fireproofing, coatings, and other miscellaneous materials (such as tags, stickers, signs, etc) which are shredded and transported during the event. These materials, as well as latent debris (dirt and dust) will be washed towards the containment floor and the recirculation sump screens by flow from both the HELB and the containment spray headers. This debris, if washed towards the recirculation pumps, could potentially impede the performance of the ECCS system. To evaluate transport of material towards the sump and the potential for degradation in performance of the ECCS system, Computational Fluid Dynamics (CFD) has been used to predict the flow patterns and energy levels in the containment pool during the recirculation phase of the event. Further, a unique methodology has been applied to correlate the CFD results with material-specific laboratory flume data and predict the volume of material transported to the sump screens. The predicted volume of debris transported to the sump screens is then used to determine if there is sufficient suction head for the pumps to operate without the potential for cavitation. In this paper, the CFD-based methodology used to predict material transport to the sump screens is discussed and the results of a prototype containment analysis are presented. Of particular interest is the analytical method for introducing the HELB flow into the containment pool and the quasi-steady treatment of the water surface to simulate the gradual filling of the pool. Coupling of the velocity and kinetic energy fields from the CFD simulations with material-specific incipient tumbling velocities (as predicted during a series of laboratory flume tests) are presented and used to demonstrate overall material transport to the sump screens.

The Autoignition of Nuclear Reactor Power Plant Explosions

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactions ◽  
High Energy ◽  
Reactor Power ◽  
Common Cause ◽  
Fluid Transients ◽  
Accident Conditions

Abstract Explosive research proves that there is a common cause for most explosions in nuclear reactor power plants during normal operations and accident conditions. The autoignition of flammable hydrogen is a common cause for nuclear power plant explosions, where complex corrosion processes, nuclear reactions, and thermal-fluid transients autoignite explosions. Research evaluated increasingly complicated accidents. First, piping explosions occurred at Hamaoka and Brunsbuttel. Fluid transients compressed oxygen and flammable hydrogen to heat these gases to autoignition, where resultant explosions shredded steel pipes. This identical mechanism was responsible for pipe and pump damages to U.S. reactor systems since the 1950s, where water hammer alone has been assumed to cause damages. Small explosions inside the piping actually cause damages during nuclear reactor startups and flow rate changes. Second, explosions are caused by thermal-fluid transients during nuclear reactor restarts, following accidental nuclear reactor meltdowns. Disastrous explosions destroyed nuclear reactor buildings (RBs) at Fukushima Daiichi. Previously considered to be a fire, a 319 kilogram hydrogen explosion occurred at Three Mile Island (TMI). The explosion cause following each of these loss-of-coolant accidents was identical, i.e., after meltdowns, pump operations heated gases, which in turn acted as the heat source to autoignite sequential hydrogen explosions in reactor systems to ignite RBs. Third, the Chernobyl explosion followed a reactor meltdown that was complicated by a high energy nuclear criticality. The hydrogen ignition and explosion causes are more complicated as well, where two sequential hydrogen explosions were ignited by high-temperature reactor fuel.

scholarly journals Anthropogenic Radiocarbon: Past, Present, and Future

Radiocarbon ◽  
1986 ◽  
Vol 28 (2A) ◽  
pp. 668-672 ◽  
Author(s):  
Pavel Povinec ◽  
Martin Chudý ◽  
Alexander Šivo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Pressurized Water Reactors ◽  
Anthropogenic Radionuclides ◽  
Pressurized Water ◽  
Heavy Water Reactor ◽  
Weapon Testing ◽  
Water Reactors

14C is one of the most important anthropogenic radionuclides released to the environment by human activities. Weapon testing raised the 14C concentration in the atmosphere and biosphere to +100% above the natural level. This excess of atmospheric C at present decreases with a half-life of ca 7 years. Recently, a new source of artificially produced 14C in nuclear reactors has become important. Since 1967, the Bratislava 14C laboratory has been measuring 14C in atmospheric 14CO2 and in a variety of biospheric samples in densely populated areas and in areas close to nuclear power plants. We have been able to identify a heavy-water reactor and the pressurized water reactors as sources of anthropogenic 14C. 14C concentrations show typical seasonal variations. These data are supported by measurements of 3H and 85Kr in the same locations. Results of calculations of future levels of anthropogenic 14C in the environment due to increasing nuclear reactor installations are presented.

Conceptual Formation of Advanced HIS Design Method for Digital I&C + HMIT

2017 ◽  
Author(s):  
Hidekazu Yoshikawa ◽  
Zhanguo Ma ◽  
Amjad Nawaz ◽  
Ming Yang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Design Method ◽  
Severe Accident ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Human Interface ◽  
Conceptual Frame ◽  
Loss Of Coolant Accident

A new conceptual frame of how to design and validate a digital HIS (human interface system) on an innovative numerical simulation basis is proposed for the support of plant operators’ supervisory control of various types of automated complex NPPs (nuclear power plants). The proposed conceptual framework utilizes the object-oriented AI softwares for plant DiD (defense-in depth) risk monitor with the combination of nuclear reactor accident simulation by an advanced nuclear safety analysis code RELAP5/MOD4 and severe accident analysis code MAAP. The developed conceptual frame proposed in this paper will be applied for an example practice for the SBLOCA (small break loss of coolant accident) case of passive safety PWR (pressurized water reactor) AP1000.

A Review on Specific Features of Small and Medium Sized Nuclear Power Plants

2010 ◽  
Author(s):  
Salah Ud-din Khan ◽  
Minjun Peng ◽  
Muhammad Zubair ◽  
Shaowu Wang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Fixed Bed ◽  
Nuclear Industry ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Energy Supply System

Due to global warming and high oil prices nuclear power is the most feasible solution for generating electricity. For the fledging nuclear power industry small and medium sized nuclear reactors (SMR’s) are instrumental for the development and demonstration of nuclear reactor technology. Due to the enhanced and outstanding safety features, these reactors have been considered globally. In this paper, first we have summarized the reactor design by considering some of the large nuclear reactor including advanced and theoretical nuclear reactor. Secondly, comparison between large nuclear reactors and SMR’s have been discussed under the criteria led by International Atomic Energy Agency (IAEA). Thirdly, a brief review about the design and safety aspects of some of SMR’s have been carried out. We have considered the specifications and parametric analysis of the reactors like: ABV which is the floating type integral Pressurized water reactor; Long life, Safe, Simple Small Portable Proliferation Resistance Reactor (LSPR) concept; Multi-Application Small Light Water Reactor (MASLWR) concept; Fixed Bed Nuclear Reactor (FBNR); Marine Reactor (MR-X) & Deep Sea Reactor (DR-X); Space Reactor (SP-100); Passive Safe Small Reactor for Distributed energy supply system (PSRD); System integrated Modular Advanced Reactor (SMART); Super, Safe, Small and Simple Reactor (4S); International Reactor Innovative and Secure (IRIS); Nu-Scale Reactor; Next generation nuclear power plant (NGNP); Small, Secure Transportable Autonomous Reactor (SSTAR); Power Reactor Inherently Safe Module (PRISM) and Hyperion Reactor concept. Finally we have point out some challenges that must be resolved in order to play an effective role in Nuclear industry.

Nuclear Power Plant Fires and Explosions: Part I — Plant Designs and Hydrogen Ignition

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Hydrogen Generation ◽  
Reactor Power ◽  
Pressurized Water ◽  
Hydrogen Ignition ◽  
History Of ◽  
Water Reactors

Requiring further investigation, hydrogen explosions and fires have occurred in several operating nuclear reactor power plants. Major accidents that were affected by hydrogen fires and explosions included Chernobyl, Three Mile Island, and Fukushima Daiichi. Smaller piping explosions have occurred at Hamaoka and Brunsbüttel Nuclear Power Plants. This paper is the first paper in a series of publications to discuss this issue. In particular, the different types of reactors that have a history of fires and explosions are discussed here, along with a discussion of hydrogen generation in commercial reactors, which provides the fuel for fires and explosions in nuclear power plants. Overall, this paper is a review of pertinent information on reactor designs that is of particular importance to this multi-part discussion of hydrogen fires and explosions. Without a review of reactor designs and hydrogen generation, the ensuing technical discussions are inadequately backgrounded. Consequently, the basic designs of pressurized water reactors (PWR’s), boiling water reactors (BWR’s), and pressure-tube graphite reactors (RBMK) are discussed in adequate detail. Of particular interest, the Three Mile Island design for a PWR is presented in some detail.

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 Activation Sources of Reactor Internals During Decommissioning

2017 ◽  
Author(s):  
Jianchun Han ◽  
Yan Zhou ◽  
Hui Li ◽  
Qiliang Mei
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Neutron Flux ◽  
Nuclear Power ◽  
Power Plants ◽  
Neutron Transport ◽  
Three Dimensional ◽  
High Energy ◽  
Pressurized Water ◽  
Dimensional Distribution

As China’s first nuclear power plant connected to the grid, the first Qinshan nuclear power plant is approaching the decommissioning period. Other nuclear power plants also turn into the preparation phase of decommissioning in succession. In order to facilitate decommissioning, source survey is conducted during the pre-decommissioning phase, which can provide radioactive inventory, contamination distribution, species and quantities of nuclides. The internals of the reactor work under the most severe radiation environment. During the reactor operation, the materials of internals are irradiated by high-energy neutrons. So activated nuclides are generated due to the neutron capture reaction, which are the main radioactive waste to be treated during decommissioning. In this paper, the neutron irradiation and the generated activation source of the internals for pressurized water reactors (PWR) are studied and analyzed. Firstly, core modeling was carried out, and the neutron transport calculation is performed to obtain three-dimensional distribution of the neutron flux. Secondly, according to the three-dimensional distribution of the material composition and the neutron flux rate of the reactor, the activation calculation is carried out to obtain the activation source.

scholarly journals A Global Review of PWR Nuclear Power Plants

2020 ◽  
Vol 10 (13) ◽  
pp. 4434
Author(s):  
Pablo Fernández-Arias ◽  
Diego Vergara ◽  
José A. Orosa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Real Option ◽  
Nuclear Reactor ◽  
Reactor Design ◽  
Life Extension ◽  
Pressurized Water ◽  
The North ◽  
The World

Nuclear energy is presented as a real option in the face of the current problem of climate change and the need to reduce CO2 emissions. The nuclear reactor design with the greatest global impact throughout history and which has the most ambitious development plans is the Pressurized Water Reactor (PWR). Thus, a global review of such a reactor design is presented in this paper, utilizing the analysis of (i) technical aspects of the different variants of the PWR design implemented over the past eight years, (ii) the level of implementation of PWR nuclear power plants in the world, and (iii) a life extension scenario and future trends in PWR design based on current research and development (R&D) activity. To develop the second analysis, a statistical study of the implementation of the different PWR variants has been carried out. Such a statistical analysis is based on the operating factor, which represents the relative frequency of reactors operating around the world. The results reflect the hegemony of the western variants in the 300 reactors currently operating, highlighting the North American and French versions. Furthermore, a simulation of a possible scenario of increasing the useful life of operational PWRs up to 60 years has been proposed, seeing that in 2050 the generation capacity of nuclear PWRs power plants will decrease by 50%, and the number of operating reactors by 70%.

scholarly journals Estimating Safety Valve Stochastic Failure-to-Close Probabilities for the Purpose of Nuclear Reactor Severe Accident Analysis

2017 ◽  
Author(s):  
S. Tina Ghosh ◽  
Alfred Hathaway ◽  
Hossein Esmaili ◽  
Kyle W. Ross ◽  
Douglas M. Osborn ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Safety Valve ◽  
Important Determinant ◽  
Severe Accident ◽  
Pressurized Water ◽  
Actual Operating ◽  
The U.S ◽  
Failure Probabilities

Recent consequences analyses of potential station blackout (SBO) accidents at nuclear power plants have shown that an important uncertainty in accident progression and radionuclide release is the probability that a safety valve (SV) will fail-to-close after it has opened to relieve pressure [1]. The U.S. Nuclear Regulatory Commission’s (NRC’s) State-of-the-Art Reactor Consequence Analyses (SOARCA) and associated uncertainty analyses for SBOs at a pressurized-water reactor (PWR) indicated that SV behavior is an important determinant of whether an induced steam-generator tube rupture (an undesirable bypass event) may develop [2], and an important determinant of whether a PWR with an ice condenser containment may experience an early containment failure [3]. Given the importance of SV failure-to-close probabilities in these accidents, available information was reviewed to help develop better estimates of the probability for a SV’s failure-to-close on demand. The SVs of interest in the SOARCA PWR analyses are the PWR code SVs, designated SVVs in a study of SVs published in 2007 (NUREG/CR-7037) [4]. There are two sets of failure probabilities reported in NUREG/CR-7037: failure probabilities based on behavior after reactor scrams i.e., after actual operating events, and failure probabilities based on tests. Information is included for both the secondary-side, main steam system (MSS) valves, as well as reactor coolant system (RCS) valves. The NUREG/CR-7037 failure probabilities based on actual operating events differ markedly from the failure probabilities based on tests. Further inquiries on valve testing and review of testing requirements show that the focus of testing is to demonstrate that the valves will open to relieve pressure during design-basis accidents to prevent overpressure events. The reseating or closing capability is not tested under severe accident conditions, in other words, the valve’s repeated full-stroking and passing steam. As such, the testing data was not considered applicable for severe accident modeling purposes. Furthermore, the assumption was made that MSS data was representative of RCS valve failures too during severe accident scenarios, as it is judged that they are similar enough in weighing the difference between the valves against the lack of operational data on the RCS SVs (only four data points, and one of two failures having a cause of failure now-defunct in the majority of operating PWRs in the U.S.). Lastly, recovered valve function, e.g., a previously stuck-open valve closing when pressure reduces, was not considered as a successful valve operation based on a review of licensee event reports. Paper published with permission.

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.

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