Measurements of Adhesion Force Between a Single Silver Particle Interacting With Hastelloy X and Inconel 617

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
Naphtali M. Mokgalapa ◽  
Tushar K. Ghosh ◽  
Robert V. Tompson ◽  
Sudarshan K. Loyalka
Keyword(s):  
Surface Roughness ◽  
High Temperature ◽  
Silver Particle ◽  
Adhesion Force ◽  
Reactor Core ◽  
Normal Operation ◽  
Hastelloy X ◽  
Inconel 617 ◽  
High Temperature Reactors ◽  
Predicted Values

Abstract Silver is generated in the reactor core during normal operation of very high temperature reactors (VHTRs). It may be transported as aerosols in the reactor circuit and plates out at different points in the reactor including the intermitted heat exchanger. The resuspension of particles in high temperature reactors is closely associated with source term analysis including the environment safety assessment. Adhesion force is important in determining resuspension rate used in resuspension calculations. In this work, the atomic force microscope was used to measure the adhesive force between a spherical silver particle and VHTRs structural materials of Hastelloy X and Inconel 617. These forces were also predicted through the Johnson-Kendall-Roberts (JKR) theoretical model. The theoretically calculated values (when the particle size of 15.1 μm diameter was used) are higher than the measured results by a factor of a 1000. However, when surface roughness was taken into account, an improved comparison between the theoretically predicted values and the measured values was observed. In fact, the comparisons between the theoretically predicted values and the measured results show a deviation that ranges from 0.155 to 0.422 in the case of Hastelloy X and from 0.243 to 0.503 in the case of Inconel 617. The data generated provide insight into the significant influence that surface roughness has on the adhesion force. This adhesion force data may be important in understanding the adhesion of silver particles to Hastelloy X and Inconel 617, and may contribute to particle resuspension calculations.

scholarly journals Computational Fluid Dynamics Applied to Study Coolant Loss Regimes in Very High Temperature Reactors

2021 ◽  
Vol 9 (2A) ◽  
Author(s):  
Uebert Gonçalves Moreira ◽  
Dany Sanchez Dominguez ◽  
Leorlen Yunier Rojas Mazaira ◽  
Carlos Alberto Brayner de Oliveira Lira ◽  
Carlos Rafael García Henández
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
High Temperature ◽  
Reactor Core ◽  
Normal Operation ◽  
Pebble Bed ◽  
Continuous Increase ◽  
High Temperature Reactors ◽  
Fuel Elements ◽  
Very High

The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this pers-pective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its   inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). We obtain the temperature profile distribution in the core for regimes where the coolant flow rate is smaller than recommended in a normal operation. In general, the temperature distributions calculated are consistent with phenomenological behavior. Even without considering the reactivity changes to reduce the reactor power or other safety mechanisms, the maximum temperatures do not exceed the recommended limits for TRISO fuel elements.

Chromium Activity Measurements in Nickel Based Alloys for Very High Temperature Reactors: Inconel 617, Haynes 230, and Model Alloys

2009 ◽  
Vol 131 (6) ◽  
Author(s):  
Stéphane Gossé ◽  
Thierry Alpettaz ◽  
Sylvie Chatain ◽  
Christine Guéneau
Keyword(s):  
High Temperature ◽  
Heat Exchangers ◽  
Mass Spectrometric ◽  
Operating Conditions ◽  
Pure Substance ◽  
Haynes 230 ◽  
Inconel 617 ◽  
High Temperature Reactor ◽  
High Temperature Reactors ◽  
Very High

The alloys Haynes 230 and Inconel 617 are potential candidates for the intermediate heat exchangers (IHXs) of (very) high temperature reactors ((V)-HTRs). The behavior under corrosion of these alloys by the (V)-HTR coolant (impure helium) is an important selection criterion because it defines the service life of these components. At high temperature, the Haynes 230 is likely to develop a chromium oxide on the surface. This layer protects from the exchanges with the surrounding medium and thus confers certain passivity on metal. At very high temperature, the initial microstructure made up of austenitic grains and coarse intra- and intergranular M6C carbide grains rich in W will evolve. The M6C carbides remain and some M23C6 richer in Cr appear. Then, carbon can reduce the protective oxide layer. The alloy loses its protective coating and can corrode quickly. Experimental investigations were performed on these nickel based alloys under an impure helium flow (Rouillard, F., 2007, “Mécanismes de formation et de destruction de la couche d’oxyde sur un alliage chrominoformeur en milieu HTR,” Ph.D. thesis, Ecole des Mines de Saint-Etienne, France). To predict the surface reactivity of chromium under impure helium, it is necessary to determine its chemical activity in a temperature range close to the operating conditions of the heat exchangers (T≈1273 K). For that, high temperature mass spectrometry measurements coupled to multiple effusion Knudsen cells are carried out on several samples: Haynes 230, Inconel 617, and model alloys 1178, 1181, and 1201. This coupling makes it possible for the thermodynamic equilibrium to be obtained between the vapor phase and the condensed phase of the sample. The measurement of the chromium ionic intensity (I) of the molecular beam resulting from a cell containing an alloy provides the values of partial pressure according to the temperature. This value is compared with that of the pure substance (Cr) at the same temperature. These calculations provide thermodynamic data characteristic of the chromium behavior in these alloys. These activity results call into question those previously measured by Hilpert and Ali-Khan (1978, “Mass Spectrometric Studies of Alloys Proposed for High-Temperature Reactor Systems: I. Alloy IN-643,” J. Nucl. Mater., 78, pp. 265–271; 1979, “Mass Spectrometric Studies of Alloys Proposed for High-Temperature Reactor Systems: II. Inconel Alloy 617 and Nimomic Alloy PE 13,” J. Nucl. Mater., 80, pp. 126–131), largely used in the literature.

Design values of Inconel 617 for high temperature reactors operating at temperatures above 800°C

1984 ◽  
Vol 83 (3) ◽  
pp. 397-402 ◽  
Author(s):  
H.J. Penkalla ◽  
H.H. Over ◽  
F. Schubert ◽  
H. Nickel
Keyword(s):  
High Temperature ◽  
Design Values ◽  
Inconel 617 ◽  
High Temperature Reactors

scholarly journals The Stress and Reliability Analysis of HTR’s Graphite Component

2014 ◽  
Vol 2014 ◽  
pp. 1-8 ◽  
Author(s):  
Xiang Fang ◽  
Haitao Wang ◽  
Suyuan Yu
Keyword(s):  
High Temperature ◽  
Reliability Analysis ◽  
Reactor Core ◽  
Mechanical Analysis ◽  
Normal Operation ◽  
Reliability Models ◽  
Analysis Process ◽  
Close Relationship ◽  
Calculation Results ◽  
Stress Accumulation

The high temperature gas cooled reactor (HTR) is developing rapidly toward a modular, compact, and integral direction. As the main structure material, graphite plays a very important role in HTR engineering, and the reliability of graphite component has a close relationship with the integrity of reactor core. The graphite components are subjected to high temperature and fast neutron irradiation simultaneously during normal operation of the reactor. With the stress accumulation induced by high temperature and irradiation, the failure risk of graphite components increases constantly. Therefore it is necessary to study and simulate the mechanical behavior of graphite component under in-core working conditions and forecast the internal stress accumulation history and the variation of reliability. The work of this paper focuses on the mechanical analysis of pebble-bed type HTR's graphite brick. The analysis process is comprised of two procedures, stress analysis and reliability analysis. Three different creep models and two different reliability models are reviewed and taken into account in simulation. The stress and failure probability calculation results are obtained and discussed. The results gained with various models are highly consistent, and the discrepancies are acceptable.

scholarly journals Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up

2019 ◽  
Vol 5 (4) ◽  
pp. 289-295 ◽  
Author(s):  
Olga I. Bulakh ◽  
Oleg K. Kostylev ◽  
Vladimir N. Nesterov ◽  
Eldar K. Cherdizov
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Alternative Energy ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Calculated Data ◽  
Fuel Burnup ◽  
The Core ◽  
High Temperature Reactors ◽  
High Temperature Gas

High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources. The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability. Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors. Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated. The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan.

Design of the Essential Material Test Equipment for the Pebble Bed Effective Thermal Conductivity Measurement Experiment

2013 ◽  
Author(s):  
Cheng Ren ◽  
Xing-Tuan Yang ◽  
Cong-Xin Li ◽  
Zhi-Yong Liu ◽  
Sheng-Yao Jiang
Keyword(s):  
Thermal Conductivity ◽  
High Temperature ◽  
Effective Thermal Conductivity ◽  
Reactor Core ◽  
Test Equipment ◽  
Normal Operation ◽  
Pebble Bed ◽  
Material Test ◽  
Temperature Detector ◽  
High Temperature Gas

High Temperature Gas-cooled Reactor (HTGR) is a typical representation of Generation IV nuclear power system for its advantages like inherent safety, high efficiency, widely application as high-temperature heat source. The first two 250-MWt high temperature reactor pebble bed modules (HTR-PM) have be installing at the Shidaowan plant in Shandong Province, China, which have the cylindrical core structure with thousands of spherical fuel elements randomly packed inside. The values of the effective thermal conductivity of the pebble bed core under different temperatures are essential parameters for the design of HTGR, which are needed to analyze the maximum fuel temperature, temperature distribution and residual heat releasing ability in reactor core. For this purpose, Tsinghua University in China has proposed a full-scale heat transfer experiment to conduct comprehensive thermal transfer tests in packed pebble bed and to determine the effective thermal conductivity through the pebble bed under vacuum condition and helium environment with temperature up to 1600°C. An essential material test equipment is built in advance to provide reliable materials and technical support for the design of the final experimental device aimed at measuring the effective thermal conductivity of pebble bed type reactor core of the high temperature gas-cooled reactor. The design of the essential material test equipment is introduced in detail, including the heat element, the insulation structure, the temperature detector, cooling water system, vacuum system, hydraulic lifting system, data acquisition system and so on. Several key technologies in design are described in detail. Test temperature in the equipment was elevated up to 1600°C, which covers the whole temperature range of the normal operation and accident condition of HTGR and could fully meet the test requirements of materials used in the reactor. The construction and commissioning of the test equipment shows that the test equipment has met the design requirements and verified the feasibility of the related materials and structures.

Analysis of Bypass Effect on a 250 MW High Temperature Gas-Cooled Reactor

2014 ◽  
Author(s):  
Yanhua Zheng ◽  
Fubing Chen ◽  
Lei Shi
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Thermal Inertia ◽  
Hydraulic Calculation ◽  
Normal Operation ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Cooling Flow ◽  
Safety Features ◽  
High Temperature Gas

Pebble bed modular high temperature gas-cooled reactors (HTR), due to their characteristics of low power density, slender structure, large thermal inertia of fuel elements and reactor component materials (graphite), have good inherent safety features. However, the reflectors consisting of large piles of graphite blocks will form huge numbers of certain bypass gaps in the radial, axial and circumferential directions, thus affecting the effective cooling flow into the reactor core, which is one of the concerned issues of HTRs. According to the preliminary design of the Chinese high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the thermal-hydraulic calculation model is established in this paper. Based on this model, considering different bypass flow, that is to say, different core cooling flow, fuel element temperature, outlet helium temperature and the core pressure drop in the normal operation, as well as the maximal fuel temperature during the depressurized loss of forced cooling (DLOFC) accident are analyzed. This study on bypass effects on the steady-state and transient phases can further demonstrate the HTR safety features.

scholarly journals Assessing the Functional Properties of TiZr Nanotubular Structures for Biomedical Applications, through Nano-Scratch Tests and Adhesion Force Maps

Molecules ◽  
2021 ◽  
Vol 26 (4) ◽  
pp. 900
Author(s):  
Maria Vardaki ◽  
Aida Pantazi ◽  
Ioana Demetrescu ◽  
Marius Enachescu
Keyword(s):  
Surface Roughness ◽  
Functional Properties ◽  
Bacterial Adhesion ◽  
Biomedical Applications ◽  
Adhesion Force ◽  
Roughness Parameters ◽  
Force Microscopy ◽  
Atomic Force ◽  
The Mean ◽  
Scratch Tests

In this work we present the results of a functional properties assessment via Atomic Force Microscopy (AFM)-based surface morphology, surface roughness, nano-scratch tests and adhesion force maps of TiZr-based nanotubular structures. The nanostructures have been electrochemically prepared in a glycerin + 15 vol.% H2O + 0.2 M NH4F electrolyte. The AFM topography images confirmed the successful preparation of the nanotubular coatings. The Root Mean Square (RMS) and average (Ra) roughness parameters increased after anodizing, while the mean adhesion force value decreased. The prepared nanocoatings exhibited a smaller mean scratch hardness value compared to the un-coated TiZr. However, the mean hardness (H) values of the coatings highlight their potential in having reliable mechanical resistances, which along with the significant increase of the surface roughness parameters, which could help in improving the osseointegration, and also with the important decrease of the mean adhesion force, which could lead to a reduction in bacterial adhesion, are providing the nanostructures with a great potential to be used as a better alternative for Ti implants in dentistry.

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