Numerical Simulation of a Compartment Fire in a Nuclear Power Plant Containment Building

2002 ◽  
Author(s):  
Jason Floyd
Keyword(s):  
Fire Protection ◽  
Nuclear Power ◽  
Power Plants ◽  
Current Trend ◽  
Test Facility ◽  
Zone Model ◽  
Fire Dynamics ◽  
Computational Tools ◽  
Model Code ◽  
Advantages And Disadvantages

The current trend towards the increased use of risk assessment in the regulation of nuclear power plants will inevitably result in changes in the analysis of fire protection systems and the methods of analysis. Before fire protection can be regulated on a risk basis, a consensus must be reached on a number of issues. One key issue is what types of computational tools will be allowable for analyzing fire events, and what types of scenarios those tools will be approved for use. Reaching this consensus will require an understanding of the types of computational tools available and their inherent advantages and disadvantages. To aid with this understanding, three different methods of fire simulation are applied to an oil pool fire test in the HDR (Heiss Dampf Reaktor) containment test facility [1]. These methods are a hand calculation, the zone model code CFAST (Consolidated Model of Fire Growth and Smoke Transport) [2], and the computational fluid dynamics code FDS (Fire Dynamics Simulator) [3]. Each is applied to a steady-state portion of the test using, to the extent possible, the same set of input parameters. The results of the computation are compared to the test data. The comparisons show that each method is potentially suitable for use depending on the information required from the simulation. Each method will potentially have a role to play in risk based regulation depending on the scenario.

The reactor research and test facility

2020 ◽  
pp. 69-79
Author(s):  
Andrey S. KIRILLOV ◽  
Aleksandr P. PYSHKO ◽  
Andrey A. ROMANENKO ◽  
Valery I. YARYGIN
Keyword(s):  
Measurement System ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Facility ◽  
Physical Parameters ◽  
Pumping System ◽  
Current State ◽  
History Of ◽  
Special Measurement ◽  
Space Nuclear Power

The paper describes an overview of the history of development and the current state of JSC “SSC RF-IPPE” reactor research and test facility designed for assembly, research and full-scale life energy tests of space nuclear power plants with a thermionic reactor. The leading specialists involved in development and operation of this facility are represented. The most significant technological interfaces and upgrade operations carried out in the recent years are discussed. The authors consider the use of an oil-free pumping system as part of this facility during degassing and life testing. Proposed are up-to-date engineering solutions for development of the automated special measurement system designed to record NPP performance, including volt-ampere characteristics together with thermophysical and nuclear physical parameters of a ground prototype of the space nuclear power plant. Key words: reactor research and test facility, thermionic reactor, life energy tests, oil-free pumping system, automated special measurement system, volt-ampere characteristics.

Recent Patents on Valve Mechanism Device

2020 ◽  
Vol 13 (3) ◽  
pp. 230-241
Author(s):  
Ye Dai ◽  
Hui-Bing Zhang ◽  
Yun-Shan Qi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Flow Direction ◽  
Research Progress ◽  
Valve Mechanism ◽  
Advantages And Disadvantages ◽  
Valve Structure ◽  
Valve Leakage ◽  
Safety And Reliability

Background: Valves are an important part of nuclear power plants and are the control equipment used in nuclear power plants. It can change the cross-section of the passage and the flow direction of the medium and has the functions of diversion, cutoff, overflow, and the like. Due to the earthquake, the valve leaks, which will cause a major nuclear accident, endangering people's lives and safety. Objective: The purpose of this study is to synthesize the existing valve devices, summarize and analyze the advantages and disadvantages of various devices from many literatures and patents, and solve some problems of existing valves. Methods: This article summarizes various patents of nuclear-grade valve devices and recent research progress. From the valve structure device, transmission device, a detection device, and finally to the valve test, the advantages and disadvantages of the valve are comprehensively analyzed. Results: By summarizing the characteristics of a large number of valve devices, and analyzing some problems existing in the valves, the outlook for the research and design of nuclear power valves was made, and the planning of the national nuclear power strategic goals and energy security were planned. Conclusion: Valve damage can cause serious safety accidents. The most common is valve leakage. Therefore, the safety and reliability of valves must be taken seriously. By improving the transmission of the valve, the problems of complicated valve structure and high cost are solved.

scholarly journals Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Power Plants ◽  
Volume Fraction ◽  
Characteristic Length ◽  
Initial Conditions ◽  
Test Facility ◽  
Work Package ◽  
Acceptable Agreement

An issue of the stratified atmospheres in the containments of nuclear power plants is still unresolved; different experiments are performed in the test facilities like TOSQAN and MISTRA. MASPn experiments belong to the spray benchmark, initiated in the containment atmosphere mixing work package of the SARNET network. The benchmark consisted of MASP0, MASP1 and MASP2 experiments. Only the measured depressurisation rates during MASPn were available for the comparison with calculations. When the analysis was performed, the boundary conditions were not clearly defined therefore most of the attention was concentrated on MASP0 simulation in order to develop the nodalisation scheme and define the initial and boundary conditions. After achieving acceptable agreement with measured depressurisation rate, simulations of MASP1 and MASP2 experiments were performed to check the influence of sprays. The paper presents developed nodalisation scheme of MISTRA for the COCOSYS code and the results of analyses. In the performed analyses, several parameters were considered: initial conditions, loss coefficient of the junctions, initial gradients of temperature and steam volume fraction, and characteristic length of structures. Parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determine the long-term depressurisation rate.

scholarly journals A modern data measurement system to study and test thermionic heat to electricity converters

2018 ◽  
Vol 4 (3) ◽  
pp. 179-183
Author(s):  
Andrey Kirillov ◽  
Valeriy Yarygin
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Test System ◽  
Temperature Fields ◽  
Test Facility ◽  
Measurement Systems ◽  
Data Measurement ◽  
Functional Capabilities ◽  
Thermionic Nuclear Power ◽  
A Current

Studies and tests are conducted to determine the performance of thermionic nuclear power plants (TNPP) a stage in which is pre-irradiation testing of laboratory thermionic converters (TIC) with flat and cylindrically shaped electrodes using test facilities fitted with automated data measurement systems (DMS). The TIC volt-ampere characteristics (VAC) are measured in the DMS jointly with the measured test section and experimental test facility temperature fields. The structure and the characteristics of a DMS based on products from ICP DAS Co., Ltd are presented. A developed VAC measurement program providing the operator with a convenient graphic interface and enabling adjustment of the measurement parameters has been considered. The VAC recording errors in the process of measurements have been determined using TIC simulators. The error in the VAC diffusion portion on a simulator (with a current of less than 3 A) is not more than 1%. Thanks to the use of modern components, the developed DMS offers extended functional capabilities for measuring the thermocouple signals in an experimental electrophysical test facility. The DMS structure provides for the convenience of scaling (through a larger number of measuring channels) and makes it possible to add modules from other manufacturers. The experience of operating this DMS will be used to develop the DMS for an in-pile test system designed for similar functions.

Development of Advanced Fire Zone Model Applicable to Fire PRA for Nuclear Power Plant

2020 ◽  
Author(s):  
Junghoon Ji ◽  
Koji Shirai ◽  
Koji Tasaka ◽  
Toshiko Udagawa
Keyword(s):  
Oxygen Concentration ◽  
Nuclear Power ◽  
Power Plants ◽  
Fire Risk ◽  
Zone Model ◽  
Radiant Heat Transfer ◽  
Test Results ◽  
Compartment Fire ◽  
Jet Flame ◽  
Fire Source

Abstract In implementing the fire PRA for nuclear power plants, a highly predictive fire model is required for more realistic fire scenarios and fire risk assessment. The fire simulation zone model BRI2002 developed in Japan has been continuously improved to allow analysis considering the characteristics of a compartment fire. In this study, a heat feedback phenomenon was introduced in BRI2002, in which combustion of a fire source can be accelerated by radiant heat transfer inside the compartment during a compartment fire. Not only the thermal radiation from the flame and smoke layers, but also radiation from the hot ceiling surface and the ceiling jet flame were considered when the flame impinges with the ceiling. In addition, in the zone model, the existing model for predicting the oxygen concentration in a compartment was improved so that the oxygen concentration could be predicted considering the vertical location of a fire source (height from the floor). The prediction results were verified by full-scale compartment fire test results. As a result of the calculation in which the fire source is installed at 2 m above the floor, the prediction results for the burning rate and zone temperature were well consistent with the test results.

AFCEN RCC-F: A New Standard for the Fire Protection Design of New Built Light Water Nuclear Power Plants

2018 ◽  
Author(s):  
Bernard Gautier ◽  
Mickael Cesbron ◽  
Richard Tulinski
Keyword(s):  
Fire Protection ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fire Hazard ◽  
Pressurized Water ◽  
Light Water ◽  
Protection Scheme ◽  
Safety Case ◽  
Under Construction

Fire hazard is an important issue for the safety of nuclear power plants: the main internal hazard in terms of frequency, and probably one the most significant with regards to the design costs. AFCEN is publishing in 2018 a new code for fire protection of new built PWR nuclear plants, so-called RCC-F. This code is an evolution of the former ETC-F code which has been applied to different EPR plants under construction (Flamanville 3 (FA3, France), Hinkley Point C (HPC, United Kingdom), Taïshan (TSN, China)). The RCC-F code presents significant enhancement and evolutions resulting from eight years of work by the AFCEN dedicated sub-committee, involving a panel of contributors from the nuclear field. It is now opened to any type of PWR (Pressurized Water Reactor) type of nuclear power plants and not any longer limited to EPR (European Pressurized Reactor) plants. It can potentially be adapted to other light water concepts. Its objective is to help engineers design the fire prevention and protection scheme, systems and equipment with regards to the safety case and the defense in depth taking into account the French and European experience in the field. It deals also with the national regulations, with two appendices dedicated to French and British regulations respectively. The presentation gives an overview of the code specifications and focuses on the significant improvements.

Setup of the Supercritical CO2 Test Facility “SCARLETT” for Basic Experimental Investigations of a Compact Heat Exchanger for an Innovative Decay Heat Removal System

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Wolfgang Flaig ◽  
Rainer Mertz ◽  
Joerg Starflinger
Keyword(s):  
Heat Exchanger ◽  
Supercritical Co2 ◽  
Nuclear Power ◽  
Power Plants ◽  
Heat Removal ◽  
Test Facility ◽  
Experimental Investigations ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

Supercritical fluids show great potential as future coolants for nuclear reactors, thermal power, and solar power plants. Compared to the subcritical condition, supercritical fluids show advantages in heat transfer due to thermodynamic properties near the critical point. A specific field of interest is an innovative decay heat removal system for nuclear power plants, which is based on a turbine-compressor system with supercritical CO2 as the working fluid. In case of a severe accident, this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger (HE) needs to be as compact and efficient as possible. Therefore, a diffusion-bonded plate heat exchanger (DBHE) with mini channels was developed and manufactured. This DBHE was tested to gain data of the transferable heat power and the pressure loss. A multipurpose facility has been built at Institut für Kernenergetik und Energiesysteme (IKE) for various experimental investigations on supercritical CO2, which is in operation now. It consists of a closed loop where the CO2 is compressed to supercritical state and delivered to a test section in which the experiments are run. The test facility is designed to carry out experimental investigations with CO2 mass flows up to 0.111 kg/s, pressures up to 12 MPa, and temperatures up to 150 °C. This paper describes the development and setup of the facility as well as the first experimental investigation.

Experimental Investigation of Steam Bubble Condensation in Flowing Subcooled Water With Two Different Injection Nozzle Geometries

2013 ◽  
Author(s):  
Suleiman Al Issa ◽  
Patricia B. Weisensee
Keyword(s):  
Test Section ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Bubble Diameter ◽  
Large Diameter ◽  
Low Pressure ◽  
Test Facility ◽  
Experimental Investigations ◽  
Subcooled Water

A multiphase flow test facility was built at the Department of Nuclear Engineering at the Technical University Munich. The main goal of this facility is to investigate the condensation of steam bubbles injected into a vertical large diameter pipe (104 mm) with flowing subcooled water (6–15 K) at low pressure conditions (1.1–1.45 bar). Current experimental investigations will contribute to a better understanding of subcooled boiling at low pressures, accidental conditions in nuclear power plants and low-pressure research reactors and correlations for the validation of CFD codes. The test section is a 1 m long transparent pipe that is surrounded by an 18×18 cm rectangular “aquarium” filled with distilled water for refraction correction. High-speed camera (HSC) recording was used to gather data about condensing bubbles including bubble diameter, shape and rising velocity. Steam was injected via two different vertical injection nozzles with an inner diameter of 4 and 6 mm, respectively, directly into the center of the test section. The present experiments were carried out at three different steam superficial velocities, water superficial velocities and water temperatures leading to bubble diameters up to 50 mm and bubble relative velocities around 1 m/s. The measurements enabled the calculation of bubble Reynolds and Nusselt numbers and comparison with correlations given in literature. Even though significant differences could be observed between the two injection nozzles with respect to the bubble’s diameter and velocity, the Nusselt and Reynolds numbers are in the same range of values. The bigger bubbles of the 6 mm with respect to the 4 mm nozzle are almost neutralized by the lower rising velocities.

scholarly journals The Effect of Tungstate and Ethanolamines Added in Tap Water on Corrosion Inhibition of Ductile Cast Iron Pipe for Nuclear Power Plants

Metals ◽  
2020 ◽  
Vol 10 (12) ◽  
pp. 1597
Author(s):  
Butaek Lim ◽  
Kitae Kim ◽  
Hyunyoung Chang ◽  
Heungbae Park ◽  
Youngsik Kim
Keyword(s):  
Cast Iron ◽  
Fire Protection ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Tap Water ◽  
Ductile Cast Iron ◽  
Passivation Film ◽  
Fire Extinguishing ◽  
Immersion Corrosion

Cast iron is primarily used in buried piping to transport water in the fire protection system of nuclear power plants; ductile cast iron is generally used for domestic nuclear power plants. In general, the fluid used as fire-extinguishing water in such fire protection systems is tap water, and corrosion inhibitors are not currently added. In this study, the synergistic effect of an adsorption barrier (monoethanolamine) and oxidized film in an environment with a corrosion inhibitor (tungstate) is examined, and the corresponding passivation properties are presented. An immersion corrosion test and electrochemical test in tap water to which only tungstate was added showed suppression of corrosion compared to molybdate at the same concentration. The polarization resistance value of a passivation film in tap water mixed with monoethanolamine and tungstate showed better results than that of the molybdate control. A surface analysis in mixed addition tap water also demonstrated that oxygen ions were sufficiently distributed, including at some spheroidized graphite sites, when tungstate was added compared to molybdate. In addition, the amount of tungsten ions adsorbed on the surface was larger than that of molybdenum ions, and it was confirmed that tungsten ions were evenly distributed over the entire surface.

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