scholarly journals Tests under irradiation of optical fibers and cables devoted to corium monitoring in case of severe accident in a Nuclear Power Plant

2020 ◽  
Vol 225 ◽  
pp. 08006
Author(s):  
G. Cheymol ◽  
L. Maurin ◽  
L. Remy ◽  
V. Arounassalame ◽  
H. Maskrot ◽  
...  
Keyword(s):  
Optical Fiber ◽  
Optical Fibers ◽  
Nuclear Power ◽  
Carbon Coating ◽  
Hydrogen Diffusion ◽  
Coating Layer ◽  
Distributed Sensing ◽  
Severe Accident ◽  
Normal Operation ◽  
The Status

The DISCOMS project, which stands for “DIstributed Sensing for COrium Monitoring and Safety”, considers the potential of distributed sensing technologies, based on remote instrumentations and Optical Fiber Sensing cables embedded into the concrete floor under the reactor vessel, to monitor the status of this third barrier of confinement. This paper focuses on the selection and testing of singlemode (SM) optical fibers with limited RIA (Radiation Induced Attenuation) to be compliant with remote distributed instruments optical budgets, the ionizing radiation doses to sustain, and their reduction provided by the concrete basemat shielding. The tests aimed at exposing these fibers and the corresponding sensitive optical cables, to the irradiation doses expected during the normal operation of the reactor (up to 60 years for the European Pressurized Reactor), followed by a severe accident. Several gamma and mixed (neutron-gamma) irradiations were performed at CEA Saclay facilities: POSÉÏDON irradiator and ISIS reactor, up to a gamma cumulated dose of about 2 MGy and fast neutron fluence (E > 1 MeV) of 6 x 1015 n/cm2. The first gamma test permitted to assess the RIA at various optical wavelengths, and to select three radiation tolerant singlemode fibers (RIA < 5 dB/100 m, at 1550 nm operating wavelength). The second one was performed on voluminous strands of sensitive cables encapsulating the selected optical fibers, up to approximately the same accumulated dose, at two temperatures: 30°C and 80°C. A significant increase of the RIA, without any saturation tendency, appeared for fibers inserted into cables, correlated with the increase of the hydroxyl attenuation peak at 1380 nm. Molecular hydrogen generated by the radiolysis of compounds of the cable is at the origin of this phenomenon. A third gamma irradiation run permitted to measure the radiolytic hydrogen production yield of some compounds of a dedicated temperature cable sample. The efficiency of a carbon coating layer over the silica cladding, acting as a barrier against hydrogen diffusion, was also successfully confirmed. Finally, the efficiency of this carbon coating layer has also been tested under neutron irradiation, then qualified as a protection barrier against hydrogen diffusion in the optical fiber cores.

Numerical Simulation of Hydrogen Dispersion Inside a Compartment Using HYDRAGON Code

2016 ◽  
Author(s):  
M. Saeed ◽  
Yu Jiyang ◽  
B. X. Hou ◽  
Aniseh A. A. Abdalla ◽  
Zhang Chunhui
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Concentration Distribution ◽  
Nuclear Power ◽  
Turbulence Models ◽  
Hydrogen Combustion ◽  
Severe Accident ◽  
Normal Operation ◽  
Accurate Estimation ◽  
Nuclear Power Plant Accident

During severe accident in the nuclear power plant, a considerable amount of hydrogen can be generated by an active reaction of the fuel-cladding with steam within the pressure vessel which may be released into the containment of nuclear power plant. Hydrogen combustion may occur where there is sufficient oxygen, and the hydrogen release rates exceed 10% of the containment. During hydrogen combustion, detonation force and short term pressure may be produced. The production of these gas species can be detrimental to the structural integrity of the safety systems of the reactor and the containment. In 1979, the Three Mile Island (1979) accident occurred. This accident compelled experts and researchers to focus on the study of distribution of hydrogen inside the containment of nuclear power plant. However after the Fukushima Dai-ichi nuclear power plant accident (2011), the modeling of the gas behavior became important topic for scientists. For the stable and normal operation of the containment, it is essential to understand the behavior of hydrogen inside the containment of nuclear power plant in order to mitigate the occurrence of these types of accidents in the future. For this purpose, it is important to identify how burnable hydrogen clouds are produced in the containment of nuclear power plant. The combustion of hydrogen may occur in different modes based on geometrical complexity and gas composition. Reliable turbulence models must be used in order to obtain an accurate estimation of the concentration distribution as a function of time and other physical phenomena of the gas mixture. In this study, a small scale hydrogen-dispersion case is selected as a benchmark to address turbulence models. The computations are performed using HYDRAGON code developed by Department of Engineering Physics, Tsinghua University, China. HYDRAGON code is a three dimensional thermal-hydraulics analysis code. The purpose of this code is to predict the behavior of hydrogen gas and multiple gas species inside the containment of nuclear power plant during severe accident. This code mainly adopts CFD models and structural correlations used for wall flow resistance instead of using boundary layer at a wall. HYDROGAN code analyzes many processes such as hydrogen diffusion condensation, combustion, gas stratification, evaporation, mixing process. The main purpose of this research is to study the influence of turbulence models to the concentration distribution and to demonstrate the code thermal-hydraulic simulation capability during nuclear power plant accident. The calculated results of various turbulence models have different prediction values in different compartments. The results of k–ε turbulence model are in reasonable agreement as compared to the benchmark experimental data.

scholarly journals ANALYTICAL SIMULATION OF A PWR USING MELCOR

2021 ◽  
Vol 8 (3A) ◽  
Author(s):  
Maritza Rodríguez Gual ◽  
Nathalia N. Araújo ◽  
Marcos C. Maturana
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
Severe Accident ◽  
Normal Operation ◽  
Detailed Knowledge ◽  
Pressurized Water ◽  
Severe Accidents ◽  
Reference Plant

After the two most significant nuclear accidents in history – the Chernobyl Reactor Four explosion in Ukraine(1986) and the Fukushima Daiichi accident in Japan (2011) –, the Final Safety Analysis Report (FSAR) included a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), covering accidents with core melting. FSAR is the most important document for licensing of siting, construction, commissioning and operation of a nuclear power plant. In the USA, the elaboration of the FSAR chapter 19 is according to the review and acceptance criteria described in the NUREG-0800 and U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200. The same approach is being adopted in Brazil by National Nuclear Energy Commission (CNEN). Therefore, the FSAR elaboration requires a detailed knowledge of severe accident phenomena and an analysis of the design vulnerabilities to the severe accidents, as provided in a PSA – e.g., the identification of the initiating events involving significant Core Damage Frequency (CDF) are made in the PSA Level 1. As part of the design and certification activities of a plant of reference, the Laboratory of Risk Analysis, Evaluating and Management (LabRisco), located in the University of São Paulo (USP), Brazil, has been preparing a group of specialists to model the progression of severe accidents in Pressurized Water Reactors (PWR), to support the CNEN regulatory expectation – since Brazilian Nuclear Power Plants (NPP), i.e., Angra 1, 2 and 3, have PWR type, the efforts of the CNEN are concentrated on accidents at this type of reactor. The initial investigation objectives were on completing the detailed input data for a PWR cooling system model using the U.S. NRC MELCOR 2.2 code, and on the study of the reference plant equipment behavior – by comparing this model results and the reference plant normal operation main parameters, as modeled with RELAP5/MOD2 code.

Development and Verification of Severe Accident Simulation Capability for a NPP Simulator

2013 ◽  
Author(s):  
Zhifei Yang ◽  
Xiaofei Xie ◽  
Xing Chen ◽  
Shishun Zhang ◽  
Yehong Liao ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Simulation System ◽  
Severe Accident ◽  
Normal Operation ◽  
Feed Water ◽  
Accident Simulation ◽  
Accident Conditions ◽  
Relap5 Code

It is reflected in the severe accident in Fukushima Daiichi that the emergency capacity of nuclear power plant needs to be enhanced. A nuclear plant simulator that can model the severe accident is the most effective means to promote this capacity. Until now, there is not a simulator which can model the severe accident in China. In order to enhance the emergency capacity in China, we focus on developing a full scope simulator that can model the severe accident and verify it in this study. The development of severe accident simulation system mainly includes three steps. Firstly, the integral severe accident code MELCOR is transplanted to the simulation platform. Secondly, the interface program must be developed to switch calculating code from RELAP5 code to MELCOR code automatically when meeting the severe accident conditions because the RELAP5 code can only simulate the nuclear power plant normal operation state and design basis accident but the severe accident. So RELAP5 code will be stopped when severe accident conditions happen and the current nuclear power plant state parameters of it should be transported to MELCOR code, and MELCOR code will run. Finally, the CPR1000 nuclear power plant MELCOR model is developed to analyze the nuclear power plant behavior in severe accident. After the three steps, the severe accident simulation system is tested by a scenario that is initiated by the station black out with reactor cooling pump seal leakage, HHSI, LHSI and auxiliary feed water system do not work. The simulation result is verified by qualitative analysis and comparison with the results in severe accident analysis report of the same NPP. More severe accident scenarios initiated by LBLOCA, MBLOCA, SBLOCA, SBO, ATWS, SGTR, MSLB will be tested in the future. The results show that the severe accident simulation system can model the severe accident correctly; it meets the demand of emergency capacity promotion.

Irradiation Tests of Optical Fibers and Cables Devoted to Corium Monitoring in Case of a Severe Accident in a Nuclear Power Plant

2020 ◽  
Vol 67 (4) ◽  
pp. 669-678 ◽  
Author(s):  
G. Cheymol ◽  
L. Maurin ◽  
L. Remy ◽  
V. Arounassalame ◽  
H. Maskrot ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Optical Fibers ◽  
Nuclear Power ◽  
Severe Accident

Effect on Hydrogen Risk of Initial Gas Injection Time During AP1000 Post-Inerting

2016 ◽  
Author(s):  
XueFeng Lyu ◽  
YanLin Chen ◽  
XiaoBo Li ◽  
ShengFei Wang ◽  
Yu Yu ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Hydrogen Diffusion ◽  
Gas Injection ◽  
Inert Gas ◽  
Severe Accident ◽  
Injection Time ◽  
Steam Generation ◽  
Flame Acceleration ◽  
Average Temperature ◽  
Total Elimination

To calculate the hydrogen risk at severe accident of small break of cold leg in 1# steam generation compartment with ADS4 invalid in AP1000 nuclear power plant, and apply the results to level2 probabilistic safety analysis, we study the effect of initial gas injection time on reducing hydrogen risk during AP1000 post-inerting, the initial gas injection times are 300 second, 500second, 700second, respectively. First, analyzing the total elimination of hydrogen by recombiners. Then, analyzing the average hydrogen mole fraction, flame acceleration factor in 1# steam generation compartment and in upper space of containment. Finally, analyzing the pressure and average temperature in containment. The results show that, the premature injection of inert gas can slow down hydrogen diffusion rate from 1# steam generation compartment to the upper space of containment, which causes hydrogen risk rising in 1# steam generation compartment. Post-inerting can ease hydrogen risk in upper space of containment, but can’t ease hydrogen risk in 1# steam generation compartment effectively. The pressure of containment is only relevant to the total mass of inert gas, and the pressure of containment is always less than limit pressure, so the containment breakage is nearly impossible.

Development of a Water Injection System Fusible Plug for Severe Accidents at Nuclear Power Plants

2020 ◽  
Author(s):  
Sei Hirano ◽  
Daisuke Hirasawa ◽  
Yoshihisa Kiyotoki ◽  
Keisuke Sakemura ◽  
Keiji Sasaki ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Reliability ◽  
Water Injection ◽  
Severe Accident ◽  
Injection System ◽  
Normal Operation ◽  
Autonomous Operation ◽  
Performance Requirements

Abstract Background: When terminal stage of Severe Accident (SA) with no coolant injection at a nuclear power plant, the equipment that has cooled and solidified through water injection to a molten core that has ex-vessel and fallen outside of the pressure vessel will then be required to operate autonomously by heat detection, without external signals or power (e.g. electricity, air). The fusible plug operation is triggered by fusible alloy which receives heat from molten core and will melt. Because the fusible plug is also the boundary of Suppression Pool (S/P), high reliability is required for sealing performance. It is for that reason that Hitachi GE Nuclear Energy Ltd. (Hitachi-GE) has developed a fusible plug to serve as a device necessary to operate this system. Features of the Fusible Plug: The autonomous operation of the fusible plug is triggered by the melting of a fusible alloy, which is part of the fusible plug. However, the fusible alloy has a remarkably low mechanical strength and therefore is not suitable as a strength member. As such, it is necessary to ensure reliable plug sealing without applying a load to the fusible alloy so as to prevent the fusible plug from malfunctioning during normal operation. Therefore, to reduce the load to be applied to the fusible alloy, Hitachi-GE has developed a fusible plug structure that operates autonomously by detecting the ambient temperature without using the fusible alloy as a strength member. We have performed a verification test using this fusible plug and confirmed that it satisfies the predetermined performance requirements. Future Actions: Hitachi-GE is holding discussions on using the fusible plug at nuclear power plants in Japan. In the future, we plan to expand to the overseas.

scholarly journals Engineering nanoparticle features to tune Rayleigh scattering in nanoparticles-doped optical fibers

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Victor Fuertes ◽  
Nicolas Grégoire ◽  
Philippe Labranche ◽  
Stéphane Gagnon ◽  
Ruohui Wang ◽  
...  
Keyword(s):  
Optical Fiber ◽  
Long Range ◽  
Optical Fibers ◽  
Rayleigh Scattering ◽  
Distributed Sensing ◽  
Optical Losses ◽  
Sensing Applications ◽  
Fiber Fabrication ◽  
Silica Fibers ◽  
Potential Applications

AbstractRayleigh scattering enhanced nanoparticles-doped optical fibers are highly promising for distributed sensing applications, however, the high optical losses induced by that scattering enhancement restrict considerably their sensing distance to few meters. Fabrication of long-range distributed optical fiber sensors based on this technology remains a major challenge in optical fiber community. In this work, it is reported the fabrication of low-loss Ca-based nanoparticles doped silica fibers with tunable Rayleigh scattering for long-range distributed sensing. This is enabled by tailoring nanoparticle features such as particle distribution size, morphology and density in the core of optical fibers through preform and fiber fabrication process. Consequently, fibers with tunable enhanced backscattering in the range 25.9–44.9 dB, with respect to a SMF-28 fiber, are attained along with the lowest two-way optical losses, 0.1–8.7 dB/m, reported so far for Rayleigh scattering enhanced nanoparticles-doped optical fibers. Therefore, the suitability of Ca-based nanoparticles-doped optical fibers for distributed sensing over longer distances, from 5 m to more than 200 m, becomes possible. This study opens a new path for future works in the field of distributed sensing, since these findings may be applied to other nanoparticles-doped optical fibers, allowing the tailoring of nanoparticle properties, which broadens future potential applications of this technology.

In-Vessel Inspection Probing Technique Using Optical Fibers Under High Radiation Dose

2014 ◽  
Author(s):  
Chikara Ito ◽  
Hiroyuki Naito ◽  
Hironori Ohba ◽  
Morihisa Saeki ◽  
Keisuke Ito ◽  
...  
Keyword(s):  
Optical Fiber ◽  
Optical Fibers ◽  
Nuclear Power ◽  
Near Infrared ◽  
Gamma Ray ◽  
Pressure Vessels ◽  
High Radiation ◽  
Remote Imaging ◽  
Pure Silica ◽  
Radiation Resistant

A high-radiation resistant optical fiber has been developed in order to investigate the interiors of the reactor pressure vessels and the primary containment vessels of the Fukushima Daiichi Nuclear Power Station. The radiation resistance of an optical fiber was improved by increasing the amount of hydroxyl up to 1000 ppm in pure silica fiber. We have tried to apply the optical fiber for remote imaging technique by means of fiberscope. The improved image fiber consists of common cladding and a large number of fiber cores made from pure silica that contains 1000 ppm hydroxyl. The transmissive rate of an infrared image was not affected after the irradiation of 1 MGy. The radiation resistant optical fiber is available for remote ultimate analysis by laser induced breakdown spectroscopy (LIBS) in order to identify whether a material is fuel debris or not. We have developed the fiber-coupled LIBS system to detect plasma emission efficiently in near-infrared region. In addition, we have performed a gamma ray dose rate measurement using an optical fiber of which scintillator is attached to the tip. As a result, the concept of applicability of a probing system using the high-radiation resistant optical fibers has been confirmed.

scholarly journals Durability of Functionalized Carbon Structures with Optical Fiber Sensors in a Highly Alkaline Concrete Environment

2019 ◽  
Vol 9 (12) ◽  
pp. 2476
Author(s):  
Kort Bremer ◽  
Lourdes S. M. Alwis ◽  
Yulong Zheng ◽  
Frank Weigand ◽  
Michael Kuhne ◽  
...  
Keyword(s):  
Optical Fiber ◽  
Optical Fibers ◽  
Carbon Coating ◽  
Carbon Filament ◽  
Fiber Sensors ◽  
Bending Tests ◽  
Carbon Structures ◽  
Straight Line ◽  
Naoh Solution ◽  
Bonding Length

The paper presents an investigation into the durability of functionalized carbon structures (FCS) in a highly alkaline concrete environment. First, the suitability of optical fibers with different coatings—i.e., acrylate, polyimide, or carbon—for the FCS was investigated by subjecting fibers with different coatings to micro/macro bending and a 5% sodium hydroxide (NaOH) (pH 14) solution. Then, the complete FCS was also subjected to a 5% NaOH solution. Finally, the effects of spatial variation of the fiber embedded in the FCS and the bonding strength between the fiber and FCS was evaluated using different configurations —i.e., fiber integrated into FCS in a straight line and/or with offsets. All three coatings passed the micro/macro bending tests and show degradation after alkaline exposure, with the carbon coating showing least degradation. The FCS showed relative stability after exposure to 5% NaOH. The optimum bonding length between the optical fiber and the carbon filament was found to be ≥150 mm for adequate sensitivity.

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