scholarly journals Catalytic activity of Pd-Ni in the oxidation of hydrogen for the safety of nuclear power plant

2016 ◽  
Vol 18 (1) ◽  
pp. 15-18
Author(s):  
Dariusz Łomot ◽  
Zbigniew Karpiński
Keyword(s):  
Bimetallic Catalysts ◽  
Nuclear Power ◽  
Power Plants ◽  
Hydrogen Oxidation ◽  
Precious Metals ◽  
Oxidation Of Hydrogen ◽  
Different Temperatures ◽  
Flowing System ◽  
Over Time ◽  
Excellent Stability

Abstract Pd-Ni/Al2O3 systems were investigated in the reaction of hydrogen oxidation in terms of their possible application as catalysts used in passive autocatalytic recombiners (PARs) used in nuclear power plants. Testing experiments were carried out in a flowing system at different temperatures and humidity of the reaction mixture. The bimetallic catalysts exhibited higher response to the increase of temperature and higher resistance to inhibit water than the monometallic palladium catalyst. They showed excellent stability during a few tens of hours, similarly, like their monometallic counterpart. Our bimetallic catalysts of hydrogen oxidation can be used as cheaper alternatives to catalysts based on the precious metals in the hydrogen oxidation without loss of their activity over time.

The power of empowering team leadership over time: A multi-wave longitudinal study in nuclear power plants

2021 ◽  
Vol 133 ◽  
pp. 105015
Author(s):  
Mario Martínez-Córcoles ◽  
Inés Tomás ◽  
Francisco J. Gracia ◽  
José M. Peiró
Keyword(s):  
Longitudinal Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Team Leadership ◽  
Over Time

A Study of Nanoparticle-Enhanced Heat Transfer

2010 ◽  
Author(s):  
Lawrence M. Jones ◽  
Timothy Sirk ◽  
Eugene Brown
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Molecular Dynamics ◽  
Heat Flux ◽  
Nuclear Power ◽  
Power Plants ◽  
Md Simulations ◽  
Theoretical Description ◽  
Different Temperatures ◽  
Dynamics Simulations

The study of the heat transfer characteristics of nanofluids, i.e. fluids that are suspensions of nanometer size particles, has gained significant attention in the search for new coolants that can effectively service a variety of needs ranging from the increasing heat transfer demands of ever smaller microelectronic devices to mitigating the effects of loss of coolant accidents in nuclear power plants. Experimental data has shown large increases in thermal conductivity and associated increases in the level of critical heat flux in nuclear reactors; however, in some cases the range of the applicability of the experimental results is uncertain and there is a lack of a theory by which this can be resolved. Complicating the theoretical description of heat transfer in nanofluids is the fact that fluids in the vicinity of the nanoparticles are a complex combination of phase transition, interfacial, and transport phenomena. This paper describes a study in which molecular dynamics simulations were used to enhance the understanding of the effect of nanoparticles on heat transfer. The molecular dynamics (MD) simulations presented here model a Lennard-Jones fluid in a channel where the walls are maintained at different temperatures. The heat flux is calculated for a variety of nanoparticle sizes and concentrations. The results are compared to experimental data in order to provide information that will more confidently bound the data and provide information that will guide the development of more comprehensive theories. We also anticipate that this work could contribute to the design of biosensors where suspended molecules are transported through micro- and nano-channels in the presence of heat transfer.

Systematic Evaluation of Creep-Fatigue Life Prediction Methods for Various Alloys

2009 ◽  
Author(s):  
Yukio Takahashi ◽  
Bilal Dogan ◽  
David Gandy
Keyword(s):  
Power Generation ◽  
Nuclear Power ◽  
Power Plants ◽  
Strain Range ◽  
Creep Damage ◽  
Small Strain ◽  
Systematic Evaluation ◽  
Wide Range ◽  
Creep Fatigue ◽  
Different Temperatures

Failure under creep-fatigue interaction is receiving increasing interest due to an increased number of start-up and shut-down in fossil power generation plants as well as development of newer nuclear power plants employing low-pressure coolant. These situations have promoted the development of various approaches for evaluating its significance. However, most of them are fragment and rather limited in terms of materials and test conditions they covered. Therefore applicability of the proposed approaches to different materials or even different temperatures is uncertain in many cases. The present work was conducted in order to evaluate and compare the representative approaches used in the prediction of failure life under creep-fatigue conditions as well as their modifications, by systematically applying them to available test data on a wide range of materials which have been used or are planned to be used in various types of power generation plants. The following observations have been made from this exercise. (i) Time fraction model has a tendency to be unconservative in general, especially at low temperature and small strain range. Because of the large scatter of the total damage, this shortcoming would be difficult to cover by the consideration of creep-fatigue interaction in a fixed manner. (ii) Classical ductility exhaustion model showed a common tendency to be overly conservative in many situations, especially at small strain ranges. (iii) The modified ductility exhaustion model based on the re-definition of creep damage showed improved predictability with a slightly unconservative tendency. (iv) Energy-based ductility exhaustion model developed in this study seems to show the best predictability among the four procedures in an overall sense although some dependency on strain range and materials was observed.

EXAFS Study on Irradiated ODS Ferritic Steel

2007 ◽  
Vol 561-565 ◽  
pp. 1761-1764 ◽  
Author(s):  
Manuel A. Pouchon ◽  
Jia Chao Chen ◽  
Claude Degueldre ◽  
Annick Froideval ◽  
Hermann Emerich ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Oxide Dispersion Strengthened ◽  
Steel Matrix ◽  
Different Temperatures ◽  
Exafs Study ◽  
Ods Steels ◽  
X Ray Absorption ◽  
Very High ◽  
Nuclear Applications

The commercial material PM2000 is investigated as a representative for the material class of Oxide Dispersion Strengthened (ODS) steels. ODS steels are envisaged as substance of structural components in Very High Temperature Reactors (VHTR). The VHTR concept is considered by the Generation IV International Forum, an initiative researching the next generation of nuclear power plants. The chosen ODS is mainly used for non-nuclear applications. In order to justify the applicability of this material within a neutron irradiative environment, the evolution of radiation damage must be investigated. PM2000 samples are therefore exposed to 4He2+ irradiation at different temperatures. The potential structural change is measured as a function of the radiation parameters, using Extended X-Ray Absorption Fine Structure (EXAFS) spectroscopy on the yttrium K-edge and on the iron K-edge. A degradation of the dispersoids or the steel matrix would show the limitation of this material candidate.

scholarly journals Simulation of mixing flows with different temperatures in a T-junction using OpenFOAM code

2021 ◽  
Vol 2088 (1) ◽  
pp. 012030
Author(s):  
Najmeh Jafari Ouregani ◽  
V I Melikhov ◽  
O I Melikhov
Keyword(s):  
Thermal Fatigue ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Stresses ◽  
Temperature Fluctuations ◽  
Experimental Results ◽  
Pipe Material ◽  
Standard Problem ◽  
Different Temperatures ◽  
Rans Models

Abstract This paper aims to evaluate the frequency of velocity and temperature fluctuations in the mixing region using OpenFOAM code. Turbulent mixing of fluids at different temperatures can lead to temperature fluctuations at the pipe material. These fluctuations, or thermal striping, inducing cyclical thermal stresses and resulting thermal fatigue, may cause unexpected failure of pipe material. Therefore, an accurate characterization of temperature fluctuations is important in order to estimate the lifetime of pipe material. Thermal fatigue of the coolant circuits of nuclear power plants is one of the major issues in nuclear safety. To investigate thermal fatigue damage, the OECD/NEA-Vattenfall T-Junction Benchmark was initiated to test the ability of state-of-the-art Computational Fluid Dynamics (CFD) codes to predict the important parameters affecting high-cycle thermal fatigue in mixing tees. In this study, to simulate the standard problem described above, the OpenFOAM code is used, which is an open integrated platform for numerical simulation of problems in continuum mechanics. At the first with Salome-meca code, a computational grid was created, consisting of about 450,000 nodes, and k-eps model and RANS models were used to simulate turbulence. OpenFOAM code results were compared with the available experimental results. The results were found to be in well-agreement with the experimental results in terms of amplitude and frequency of temperature and velocity fluctuations.

Prediction of the Inner Temperatures in a Duct Based on Outer Thermocouples Measurements

2013 ◽  
Author(s):  
Stéphane Gervais ◽  
Alexandre Girard
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Element Model ◽  
Fluid Temperature ◽  
Time Saving ◽  
Gradient Computation ◽  
Different Temperatures ◽  
Fatigue Evaluation ◽  
The Cost ◽  
Nelder Mead Algorithm

Nuclear power plants (NPP) include connections of branches conveying fluids at different temperatures. Some thermohydraulics phenomena such as stratification may affect the inner wall of the ducts and lead to fatigue damage. They are then defined in a conservative way for fatigue evaluation. In order to improve the definitions of these phenomena EDF plans to install thermocouples rings on the outer surfaces of some ducts. The aim of this article is to define a methodology to derive thermal loadings from outer measurements. To estimate the temperature of the inner skin duct through measurements on the outer surface of the ducts, we propose to use optimal control (with quadratic cost functions) to compare the outer temperatures calculated with a finite element model and the measurements. Different methods are investigated: first, a method based on Nelder-Mead algorithm [3] and second, an advanced method based on gradient computation. The advantage of the latter is to explicitly compute the gradient and a Hessian approximation of the cost function with respect to the water temperatures, which is time saving for computation-wise. To enhance the robustness of the methodology, additional conditions on the regularity of the fluid temperature field are added.

scholarly journals Leaching Behavior of Cesium, Strontium, Cobalt, and Europium from Immobilized Cement Matrix

2021 ◽  
Vol 11 (18) ◽  
pp. 8418
Author(s):  
Ja-Young Goo ◽  
Bong-Ju Kim ◽  
Myunggoo Kang ◽  
Jongtae Jeong ◽  
Ho Young Jo ◽  
...  
Keyword(s):  
Portland Cement ◽  
Nuclear Power ◽  
Power Plants ◽  
Ordinary Portland Cement ◽  
Cementitious Materials ◽  
The Other ◽  
Cement Matrix ◽  
Leaching Behavior ◽  
Binding Material ◽  
Over Time

Numerous low-level and intermediate-level radioactive wastes are generated from the decommissioning processes of nuclear power plants; these wastes are immobilized to prevent the release of radionuclides under disposal conditions. In this study, we investigated the leaching behavior of Cs, Sr, Co, and Eu, which are common in immobilized decommissioning wastes. Ordinary Portland cement (OPC) was used as an immobilization agent. During the test, leaching of the nuclides occurred in the order of Cs, Sr, Co, and Eu and decreased over time. The results showed that 41.4% of the total Cs leached over 90 days, although the other elements leached in quantities less than 1.5%. CaCO3 was precipitated by the release of cementitious materials, indicating carbonation of the leachate. The leachability indexes in all cases exceeded the acceptable criteria (>6). The results of the present study suggest that OPC can be effectively used as a binding material to immobilize nuclides (Cs, Sr, Co, and Eu) contained in decommissioning wastes.

Validation of an Estimation Procedure to Predict the Quasi-Static Brittle-to-Ductile Fracture Initiation Transition Temperature (FITT) for Ferritic Piping

2005 ◽  
Author(s):  
Craig Williams ◽  
Bud Brust ◽  
Paul Scott ◽  
Dave Rudland ◽  
Gery Wilkowski
Keyword(s):  
Transition Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Criteria ◽  
Estimation Procedure ◽  
Fracture Initiation ◽  
Compact Tension ◽  
Charpy Specimen ◽  
Different Temperatures ◽  
Shear Area

This work was conducted in order to validate the flaw evaluation criteria for class 2, 3, and balance of plant piping in nuclear power plants. A methodology was developed as part of the Battelle Integrity of Nuclear Piping (BINP) program for predicting fracture initiation transition temperature (FITT) for a surface crack in a pipe from the 85% shear area transition temperature from Charpy specimen data. The predictive relationship is described in a separate paper in this conference [1]. In order to validate this methodology a series of Charpy, dynamic tear test (DTT), compact (tension) (CT), single-edge-notch [tension] (SEN(T)), and full-scale circumferential surface-cracked pipe experiments were conducted at different temperatures. As part of this paper, the results from these tests, along with their potential impact on the validation procedure to predict the quasi-static brittle-to-ductile FITT will be discussed.

An Analytical Model for Thermal Fatigue Crack Initiation and Propagation in Mixing Zones of Piping Systems

2003 ◽  
Author(s):  
Ste´phanie Musi ◽  
Fre´de´ric Beaud
Keyword(s):  
Crack Initiation ◽  
Analytical Model ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Crack Initiation ◽  
Crack Initiation And Propagation ◽  
Thermal Fatigue Crack ◽  
Initiation And Propagation ◽  
Piping Systems ◽  
Different Temperatures

Nuclear power plants piping systems necessarily include connections of branches conveying fluids at different temperatures. Thermal-hydraulic fluctuations arising from the turbulent mixing of the flows possibly affect the inner wall of the pipes and lead to fatigue damage. This paper aims at presenting an analytical model for the mixing zone thermal-mechanical problem and consecutive crack initiation and propagation analyses.

Analysis of the FATHER Experiment With an Engineering Method Devoted to High Cycle Thermal Fatigue

2011 ◽  
Author(s):  
Romain Beaufils ◽  
Stephan Courtin
Keyword(s):  
Fatigue Damage ◽  
Thermal Fatigue ◽  
Turbulent Mixing ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Industry ◽  
Engineering Method ◽  
Different Temperatures ◽  
High Cycle Thermal Fatigue

Nuclear power plants necessarily include connections of branches conveying fluids at different temperatures. Thermo-hydraulic fluctuations arising from the turbulent mixing of the flows may affect the inner wall of the pipes and lead to fatigue damage. The FATHER experiment was carried out in order to better understand this phenomenon of High Cycle Thermal Fatigue (HCTF). The aim of this paper is to present an analysis of the FATHER experiment with a simplified engineering method which is used in French nuclear industry to identify and classify zones presenting a HCTF risk.

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