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 Performance Analyses and Evaluation of Co2 and N2 as Coolants in a Recuperated Brayton Gas Turbine Cycle for a Generation IV Nuclear Reactor Power Plant

2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Emmanuel O. Osigwe ◽  
Arnold Gad-Briggs ◽  
Theoklis Nikolaidis ◽  
Pericles Pilidis ◽  
Suresh Sampath
Keyword(s):  
Power Plant ◽  
Gas Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Power ◽  
Critical Conditions ◽  
Working Fluid ◽  
Closed Cycle ◽  
Gas Turbine Cycle

Abstract As demands for clean and sustainable energy renew interests in nuclear power to meet future energy demands, generation IV nuclear reactors are seen as having the potential to provide the improvements required for nuclear power generation. However, for their benefits to be fully realized, it is important to explore the performance of the reactors when coupled to different configurations of closed-cycle gas turbine power conversion systems. The configurations provide variation in performance due to different working fluids over a range of operating pressures and temperatures. The objective of this paper is to undertake analyses at the design and off-design conditions in combination with a recuperated closed-cycle gas turbine and comparing the influence of carbon dioxide and nitrogen as the working fluid in the cycle. The analysis is demonstrated using an in-house tool, which was developed by the authors. The results show that the choice of working fluid controls the range of cycle operating pressures, temperatures, and overall performance of the power plant due to the thermodynamic and heat properties of the fluids. The performance results favored the nitrogen working fluid over CO2 due to the behavior CO2 below its critical conditions. The analyses intend to aid the development of cycles for generation IV nuclear power plants (NPPs) specifically gas-cooled fast reactors (GFRs) and very high-temperature reactors (VHTRs).

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

A High Effective Parallel Method for the Coupling Between Neutronics and Thermal-Hydraulic

2016 ◽  
Author(s):  
Xiaomeng Dong ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Lei Li ◽  
Guangliang Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Decisive Role ◽  
Reactor Power ◽  
Coupling Method ◽  
Outlet Temperature

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.

scholarly journals The Prospects for Nuclear Reactor-MHD Commercial Power Production

1966 ◽  
Vol 88 (4) ◽  
pp. 345-354 ◽  
Author(s):  
L. A. Booth
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Ad Hoc ◽  
Nuclear Reactor ◽  
Power Production ◽  
Reactor Power ◽  
Foreseeable Future ◽  
Staff Members ◽  
Component Development ◽  
Scientific Laboratory

Cycle analysis and preliminary component designs are made for prospective nuclear reactor-MHD commercial power plants, and the problems associated with component development are discussed. The concept of commercial MHD power production is compared with other nuclear reactor power plants, and recommendations are made as to the feasibility of the MHD concept. The results of the study, conducted by an ad hoc committee of staff members at the Los Alamos Scientific laboratory, indicate that nuclear reactor-MHD power production is not feasible for the foreseeable future. The development of major components for MHD plants would require extensive programs, and this cannot be justified in view of the presently undefined criteria. Moreover, no significant gains in power economy can be predicted that will allow the nuclear-MHD plant to compete with more conventional nuclear power plants.

Performance Analysis of Generation IV Nuclear Reactor Power Plant Using CO2 and N2: Case Study of a Recuperated Brayton Gas Turbine Cycle

2018 ◽  
Author(s):  
Emmanuel O. Osigwe ◽  
Arnold Gad-Briggs ◽  
Theoklis Nikolaidis ◽  
Pericles Pilidis ◽  
Suresh Sampath
Keyword(s):  
Power Plant ◽  
Performance Analysis ◽  
Gas Turbine ◽  
Energy Demand ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Reactor Power ◽  
Critical Conditions ◽  
Working Fluid ◽  
Closed Cycle

With renewed interest in nuclear power to meet the world’s future energy demand, the Generation IV nuclear reactors are the next step in the deployment of nuclear power generation. However, for the potentials of these nuclear reactor designs to be fully realized, its suitability, when coupled with different configurations of closed-cycle gas turbine power conversion systems, have to be explored and performance compared for various possible working fluids over a range of operating pressures and temperatures. The purpose of this paper is to carry out performance analysis at the design and off-design conditions for a Generation IV nuclear-powered reactor in combination with a recuperated closed-cycle gas turbine and comparing the influence of carbon dioxide and nitrogen as working fluid in the cycle. This analysis is demonstrated in GTACYSS; a performance and preliminary design code developed by the authors for closed-cycle gas turbine simulations. The results obtained shows that the choice of working fluid controls the range of cycle operating pressures, temperatures and overall performance of the power plant due to the thermodynamic and heat properties of the fluids. The performance results favored the nitrogen working fluid over CO2 due to the behavior CO2 below its critical conditions.

Assessing the Design Margins for an Ultimate Heat Sink Sizing

2013 ◽  
Author(s):  
Dong Zheng ◽  
Julie M. Jarvis ◽  
Allen T. Vieira
Keyword(s):  
Power Plant ◽  
Heat Load ◽  
Nuclear Power ◽  
Power Plants ◽  
Cooling Tower ◽  
Design Parameters ◽  
Evaporative Water ◽  
Basin Water ◽  
Accident Conditions ◽  
New Generation

The Ultimate Heat Sink (UHS) is a large body of water supply that can be used to cool vital nuclear power plant systems during normal operation and for accident conditions. Due to more stringent environmental and water permit requirements, many new nuclear design proposals have selected the relatively smaller sized mechanical-draft cooling tower with a basin for their UHS. UHS sizing analysis is a critical licensing task for some new generation nuclear power plants Combined Operating License Applications (COLA). In this paper, a potential UHS is sized for a representative new generation nuclear power plant considering worst case design inputs and modeling assumptions. Over 30 years of historical site meteorological data are processed using an automated technique to identify limiting conditions based on resulting worst UHS design parameters, such as the maximum basin evaporative water loss and the highest basin temperature. The impacts of the cooling tower entrance recirculation effect to these design parameters are also investigated. This paper models the transient plant heat loads in detail for various design basis accident conditions. The large-break LOCA heat load is determined to be bounding for the basin evaporative water loss, while a small-break LOCA heat load may result in the highest basin water temperature. This paper also illustrates that the bounding basin water temperature can result when the peak wet bulb temperature is coincident with the peak UHS heat load. The results of this paper are of interest for new generation nuclear power plants as the paper determines impacts of limiting conditions in assessing the design margins for UHS sizing.

Implementation Status of Performance Demonstration for Nuclear Power Plant Piping Weld Ultrasonic Examination in Korea

2008 ◽  
Vol 580-582 ◽  
pp. 135-138
Author(s):  
Yong Sik Kim ◽  
Hee Jong Lee ◽  
Se Kyoung Kim
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Power ◽  
Ultrasonic Examination ◽  
Asme Code ◽  
Heavy Water Reactor ◽  
Implementation Status ◽  
Performance Demonstration

The ASME B&PV code section XI adopted the performance demonstration requirements (Appendix VIII) for the ultrasonic examination of nuclear power plant piping weld in the 1989 winter addenda for the first time. Korea Electric Power Research Institute(KEPRI) and Korea Hydro and Nuclear Power Company(KHNP) finished developing Korean Performance Demonstration(KPD) system for the ultrasonic examination to apply pressurized light-water and pressurized heavy-water reactor power plants piping weld in accordance with ASME code section XI, appendix VIII on January 2004. KEPRI has been accomplishing the performance demonstration for nuclear power plant piping weld ultrasonic examination in Korea from April 2004. This paper describes the implementation status of the performance demonstration for nuclear power plant piping weld ultrasonic examination in Korea.

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.

Nuclear Power Plant Fires and Explosions: Part III — Hamaoka Piping Explosion

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Noble Metals ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Water Hammer ◽  
Preventive Action ◽  
Hydrogen Ignition ◽  
Accident Conditions

An explosion that burst a steel pipe like a paper fire cracker at the Hamaoka Nuclear Power Station, Unit-1 is investigated in this paper, which is one of a series of papers investigating fires and explosions in nuclear power plants. The accumulation of flammable hydrogen and oxygen due to radiolysis has long been recognized as a potential problem in nuclear reactors, where radiolysis is the process that decomposes water into hydrogen and oxygen by radiation exposure in the reactor core. Hydrogen ignition and explosion has long been considered the cause of this Hamaoka piping explosion, but the cause of ignition was considered to be a minor fluid transient, or water hammer, that ignited flammable gasses in the piping, which was made possible by the presence of catalytic noble metals inside the piping. The theory presented here is that a much larger pressure surge occurred due to water hammer during operations. In fact, calculations presented here serve as proof of principle that this explosion mechanism may be present in many operating nuclear power plants. Chubu Electric, the operator of the Hamaoka plant, took appropriate actions to prevent this type of explosion in their plants in the future. In fact, this accident indicates one potential preventive action from explosions for other operating plants. Ensure that a system high point is available, where mixed hydrogen and oxygen may be removed during routine operations and during off-normal accident conditions, such as nuclear reactor meltdowns and loss of coolant accidents.

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.

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