Operating Experience and Further Development of Advanced Water Chemistry of Nuclear Reactor Systems ・ The 1st Semin. on Water Chemistry and Corrosion of Nuclear Power Plants in Japan and Korea

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.

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The analysis of emergency situations related to fires at nuclear power plants

ПОЖАРОВЗРЫВОБЕЗОПАСНОСТЬ ◽

10.22227/0869-7493.2021.30.05.66-75 ◽

2021 ◽

Vol 30 (5) ◽

pp. 66-75

Author(s):

S. A. Titov ◽

N. M. Barbin ◽

A. M. Kobelev

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Nuclear Reactor ◽

Reactor Core ◽

Emergency Situations ◽

The Usa ◽

Large Fires ◽

Short Circuits ◽

Nuclear Reactor Core

Introduction. The article provides a system and statistical analysis of emergency situations associated with fires at nuclear power plants (NPPs) in various countries of the world for the period from 1955 to 2019. The countries, where fires occurred at nuclear power plants, were identified (the USA, Great Britain, Switzerland, the USSR, Germany, Spain, Japan, Russia, India and France). Facilities, exposed to fires, are identified; causes of fires are indicated. The types of reactors where accidents and incidents, accompanied by large fires, have been determined.The analysis of major emergency situations at nuclear power plants accompanied by large fires. During the period from 1955 to 2019, 27 large fires were registered at nuclear power plants in 10 countries. The largest number of major fires was registered in 1984 (three fires), all of them occurred in the USSR. Most frequently, emergency situations occurred at transformers and cable channels — 40 %, nuclear reactor core — 15 %, reactor turbine — 11 %, reactor vessel — 7 %, steam pipeline systems, cooling towers — 7 %. The main causes of fires were technical malfunctions — 33 %, fires caused by the personnel — 30 %, fires due to short circuits — 18 %, due to natural disasters (natural conditions) — 15 % and unknown reasons — 4 %. A greater number of fires were registered at RBMK — 6, VVER — 5, BWR — 3, and PWR — 3 reactors.Conclusions. Having analyzed accidents, involving large fires at nuclear power plants during the period from 1955 to 2019, we come to the conclusion that the largest number of large fires was registered in the USSR. Nonetheless, to ensure safety at all stages of the life cycle of a nuclear power plant, it is necessary to apply such measures that would prevent the occurrence of severe fires and ensure the protection of personnel and the general public from the effects of a radiation accident.

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Digital I&C Operating Experience in the US

Volume 3: Thermal Hydraulics; Instrumentation and Controls ◽

10.1115/icone16-48862 ◽

2008 ◽

Author(s):

Bruce Geddes ◽

Ray Torok

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Failure Modes ◽

Operating Experience ◽

Data Set ◽

Common Cause ◽

Hardware Failure ◽

Common Cause Failure ◽

Defensive Measures

The Electric Power Research Institute (EPRI) is conducting research in cooperation with the Nuclear Energy Institute (NEI) regarding Operating Experience of digital Instrumentation and Control (I&C) systems in US nuclear power plants. The primary objective of this work is to extract insights from US nuclear power plant Operating Experience (OE) reports that can be applied to improve Diversity and Defense in Depth (D3) evaluations and methods for protecting nuclear plants against I&C related Common Cause Failures (CCF) that could disable safety functions and thereby degrade plant safety. Between 1987 and 2007, over 500 OE events involving digital equipment in US nuclear power plants were reported through various channels. OE reports for 324 of these events were found in databases maintained by the Nuclear Regulatory Commission (NRC) and the Institute of Nuclear Power Operations (INPO). A database was prepared for capturing the characteristics of each of the 324 events in terms of when, where, how, and why the event occurred, what steps were taken to correct the deficiency that caused the event, and what defensive measures could have been employed to prevent recurrence of these events. The database also captures the plant system type, its safety classification, and whether or not the event involved a common cause failure. This work has revealed the following results and insights: - 82 of the 324 “digital” events did not actually involve a digital failure. Of these 82 non-digital events, 34 might have been prevented by making full use of digital system fault tolerance features. - 242 of the 324 events did involve failures in digital systems. The leading contributors to the 242 digital failures were hardware failure modes. Software change appears as a corrective action twice as often as it appears as an event root cause. This suggests that software features are being added to avoid recurrence of hardware failures, and that adequately designed software is a strong defensive measure against hardware failure modes, preventing them from propagating into system failures and ultimately plant events. 54 of the 242 digital failures involved a Common Cause Failure (CCF). - 13 of the 54 CCF events affected safety (1E) systems, and only 2 of those were due to Inadequate Software Design. This finding suggests that software related CCFs on 1E systems are no more prevalent than other CCF mechanisms for which adherence to various regulations and standards is considered to provide adequate protection against CCF. This research provides an extensive data set that is being used to investigate many different questions related to failure modes, causes, corrective actions, and other event attributes that can be compared and contrasted to reveal useful insights. Specific considerations in this study included comparison of 1E vs. non-1E systems, active vs. potential CCFs, and possible defensive measures to prevent these events. This paper documents the dominant attributes of the evaluated events and the associated insights that can be used to improve methods for protecting against digital I&C related CCFs, applying a test of reasonable assurance.

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Structural Modeling and Dynamic Analysis of a Nuclear Reactor Building

10.5772/intechopen.94956 ◽

2020 ◽

Author(s):

Evrim Oyguc ◽

Abdul Hayır ◽

Resat Oyguc

Keyword(s):

Nonlinear Dynamic ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Nuclear Reactor ◽

Energy Sources ◽

Finite Element Program ◽

Design Spectrum ◽

Reactor Building ◽

Set Up

Increasing energy demand urge the developing countries to consider different types of energy sources. Owing the fact that the energy production capacity of renewable energy sources is lower than a nuclear power plant, developed countries like US, France, Japan, Russia and China lead to construct nuclear power plants. These countries compensate 80% of their energy need from nuclear power plants. Further, they periodically conduct tests in order to assess the safety of the existing nuclear power plants by applying impact type loads to the structures. In this study, a sample third-generation nuclear reactor building has been selected to assess its seismic behavior and to observe the crack propagations of the prestressed outer containment. First, a 3D model has been set up using ABAQUS finite element program. Afterwards, modal analysis is conducted to determine the mode shapes. Nonlinear dynamic time history analyses are then followed using an artificial strong ground motion which is compatible with the mean design spectrum of the previously selected ground motions that are scaled to Eurocode 8 Soil type B design spectrum. Results of the conducted nonlinear dynamic analyses are considered in terms of stress distributions and crack propagations.

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Approaches to Water Chemistry Subjects in Nuclear Power Plants―Dreams of Water Chemistry Division under Age One

Journal of the Atomic Energy Society of Japan ◽

10.3327/jaesjb.50.8_506 ◽

2008 ◽

Vol 50 (8) ◽

pp. 506-510

Author(s):

Shunsuke UCHIDA ◽

Yosuke KATSUMURA ◽

Motomasa FUSE ◽

Takahiro SHIOKAWA ◽

Hideki TAKIGUCHI

Keyword(s):

Water Chemistry ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Chemistry Division

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Editorial. Fire protection in nuclear power plants: current status and further development

Kerntechnik ◽

10.1515/kern-2000-652-305 ◽

2000 ◽

Vol 65 (2-3) ◽

pp. 76-76

Author(s):

H.-P. Berg

Keyword(s):

Fire Protection ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Current Status ◽

Further Development

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Steady-State Thermal-Hydraulic Analysis of a Particle-Bedded Boiling Water Reactor

Volume 4: Codes, Standards, Licensing and Regulatory Issues; Student Paper Competition ◽

10.1115/icone17-75050 ◽

2009 ◽

Author(s):

Xiaoyu Cai ◽

Suizheng Qiu ◽

Guanghui Su ◽

Changyou Zhao

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Superheated Steam ◽

Nuclear Reactor ◽

Hydraulic Analysis ◽

Light Water ◽

Safety Systems ◽

Fuel Elements ◽

Thermal Hydraulic Analysis

The current Light Water Reactors both BWR and PWR have extensive nuclear reactor safety systems, which provide safe and economical operation of Nuclear Power Plants. During about forty years of operation history the safety systems of Nuclear Power Plants have been upgraded in an evolutionary manner. The cost of safety systems, including large containments, is really high due to a capital cost and a long construction period. These conditions together with a low efficiency of steam cycle for LWR create problems to build new power plants in the USA and in the Europe. An advanced Boiling Water Reactor concept with micro-fuel elements (MFE) and superheated steam promises a radical enhancement of safety and improvement of economy of Nuclear Power Plants. In this paper, a new type of nuclear reactor is presented that consists of a steel-walled tube filled with millions of TRISO-coated fuel particles (Micro-Fuel Elements, MFE) directly cooled by a light-water coolant-moderator. Water is used as coolant that flows from bottom to top through the tube, thereby fluidizing the particle bed, and the moderator water flows in the reverse direction out of the tube. The fuel consists of spheres of about 2.5 mm diameter of UO2 with several coatings of different carbonaceous materials. The external coating of steam cycle the particles is silicon carbide (SiC), manufactured with chemical vapor deposit (CVD) technology. Steady-State Thermal-Hydraulic Analysis aims at providing heat transport capability which can match with the heat generated by the core, so as to provide a set of thermal hydraulic parameters of the primary loop. So the temperature distribution and the pressure losses along the direction of flow are calculated for equilibrium core in this paper. The calculation not only includes the liquid region, but the two phase region and the superheated steam region. The temperature distribution includes both the temperature parameters of micro-fuel elements and the coolant. The results show that the maximum fuel temperature is much lower than the limitation and the flow distribution can meet the cooling requirement in the reactor core.

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