scholarly journals Diamond Coating Reduces Nuclear Fuel Rod Corrosion at Accidental Temperatures: The Role of Surface Electrochemistry and Semiconductivity

Materials ◽  
2021 ◽  
Vol 14 (21) ◽  
pp. 6315
Author(s):  
Lucie Celbová ◽  
Petr Ashcheulov ◽  
Ladislav Klimša ◽  
Jaromír Kopeček ◽  
Kateřina Aubrechtová Dragounová ◽  
...  
Keyword(s):  
Microwave Plasma ◽  
Electrochemical Impedance ◽  
Nuclear Reactors ◽  
Deep Understanding ◽  
Diamond Coating ◽  
Carbon Phase ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Surface Electrochemistry ◽  
Soft Carbon

If we want to decrease the probability of accidents in nuclear reactors, we must control the surface corrosion of the fuel rods. In this work we used a diamond coating containing <60% diamond and >40% sp2 “soft” carbon phase to protect Zr alloy fuel rods (ZIRLO ®) against corrosion in steam at temperatures from 850 °C to 1000 °C. A diamond coating was grown in a pulse microwave plasma chemical vapor deposition apparatus and made a strong barrier against hydrogen uptake into ZIRLO® (ZIRLO) under all tested conditions. The coating also reduced ZIRLO corrosion in hot steam at 850 °C (for 60 min) and at 900 °C (for 30 min). However, the protective ability of the diamond coating decreased after 20 min in 1000 °C hot steam. The main goal of this work was to explain how diamond and sp2 “soft” carbon affect the ZIRLO fuel rod surface electrochemistry and semi conductivity and how these parameters influence the hot steam ZIRLO corrosion process. To achieve this goal, theoretical and experimental methods (scanning electron microscopy, Raman spectroscopy, electrochemical impedance spectroscopy, carrier gas hot extraction, oxidation kinetics, ab initio calculations) were applied. Deep understanding of ZIRLO surface processes and states enable us to reduce accidental temperature corrosion in nuclear reactors.

Rod Bow Ultrasonic Measurements Possibilities

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Marcin Kopeć ◽  
Martina Malá
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactors ◽  
Research Centre ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Assembly ◽  
Fuel Assemblies ◽  
Ultrasonic Measurements ◽  
History Of ◽  
Fuel Particles

The ultrasonic (UT) measurements have a long history of utilization in the industry, also in the nuclear field. As the UT transducers are developing with the technology in their accuracy and radiation resistance, they could serve as a reliable tool for measurements of small but sensitive changes for the nuclear fuel assembly (FA) internals as the fuel rods are. The fuel rod bow is a phenomenon that may bring advanced problems as neglected or overseen. The quantification of this issue state and its probable progress may help to prevent the safety-related problems of nuclear reactors to occur—the excessive rod bow could, in the worst scenario, result in cladding disruption and then the release of actinides or even fuel particles to the coolant medium. Research Centre Rez has developed a tool, which could serve as a complementary system for standard postirradiation inspection programs for nuclear fuel assemblies. The system works in a contactless mode and reveals a 0.1 mm precision of measurements in both parallel (toward the probe) and perpendicular (sideways against the probe) directions.

Modeling of Fuel Rod Behavior and Recent Advances of the JASMINE Code

2017 ◽  
Author(s):  
Xiaoyan Wei ◽  
Xin Jin ◽  
Yongjun Deng ◽  
Guoliang Zhang ◽  
Xiangang Fu
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactors ◽  
Physical Models ◽  
Mathematical Framework ◽  
Great Effort ◽  
Experiment Data ◽  
Running Speed ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Comprehensive Performance

JASMINE is a computer program for the comprehensive performance analysis of fuel rods in nuclear reactors and was developed at China General Nuclear Power Cooperation (CGNPC). The JASMINE code, which is built around a quasi-two-dimensional analysis of fuel rod, consists of a clearly defined mechanical/mathematical framework into which physical models can easily be incorporated. In convenience, the important inputs and principal outputs could be shown directly from the interface due to the special platform in the code. During its development great effort was spent on obtaining an extremely flexible platform which is easy to handle, exhibiting much faster running speed for the code. In recent years, much more experiment data is used for the validation to confirm its correctness.

Fretting Wear of the Zr Fuel Rod Cladding

2014 ◽  
Author(s):  
Ladislav Pecinka ◽  
Jaroslav Svoboda ◽  
Vladimír Zeman
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactors ◽  
Fretting Wear ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Assemblies ◽  
Flow Induced Vibrations ◽  
Water Reactors ◽  
Fretting Damage

Fretting wear is a particular type of wear that is expected to occur in fuel assemblies of pressurized water cooled nuclear reactors. Fretting damage of fuel rods may cause Nuclear Power Plant (NPP) operations problems and are very expensive to repair. As utilities and fuel vendors adopt higher utilization of uranium and improved thermal margins plants, burned fuel rods will be loaded at core the periphery as part of the margin mechanisms. Pressurized Water Reactors (PWRs) have experienced fuel rods fretting wear failures due to flow induced vibrations. This study describes basic results of theoretical analysis and describes experiments to predict thinning of the Zr cladding wall thickness performed.

Identification of Non-Linear Vibration Parameters of a Nuclear Fuel Rod

2018 ◽  
Author(s):  
Marco Amabili ◽  
Prabakaran Balasubramanian ◽  
Giovanni Ferrari ◽  
Stanislas Le Guisquet ◽  
Kostas Karazis ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Harmonic Balance Method ◽  
Resonance Phenomenon ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Safety Margins ◽  
Fuel Assemblies ◽  
Non Linear

In Pressurized Water Reactors (PWR), fuel assemblies are composed of fuel rods, long slender tubes filled with uranium pellets, bundled together using spacer grids. These structures are subjected to fluid-structure interactions, due to the flowing coolant surrounding the fuel assemblies inside the core, coupled with large-amplitude vibrations in case of external seismic excitation. Therefore, understanding the non-linear response of the structure and, particularly, its dissipation, is of paramount importance for the choice of safety margins. To model the nonlinear dynamic response of fuel rods, the identification of nonlinear stiffness and damping parameters is required. The case of a single fuel rod with clamped-clamped boundary conditions was investigated by applying harmonic excitation at various force levels. Different configurations were implemented testing the fuel rod in air and in still water; the effect of metal pellets simulating nuclear fuel pellets inside the rods was also recorded. Non-linear parameters were extracted from some of the experimental response curves by means of a numerical tool based on the harmonic balance method. The axisymmetric geometry of fuel rods resulted in the presence of a one-to-one internal resonance phenomenon, which has to be taken into account modifying accordingly the numerical identification tool. The internal motion of fuel pellets is a cause of friction and impacts, complicating further the linear and non-linear dynamic behavior of the system. An increase of the equivalent viscous-based modal damping with excitation amplitude is often shown during geometrically non-linear vibrations, thus confirming previous experimental findings in the literature.

The Application of the Electron Microprobe in the Analysis of Nuclear Fuel Meltdown Experiments*

1966 ◽  
Vol 10 ◽  
pp. 422-430
Author(s):  
S. J. Stachura ◽  
L. Cooper
Keyword(s):  
Electron Microprobe ◽  
Test Analysis ◽  
Fuel Rods ◽  
Failure Characteristics ◽  
Uranium Carbide ◽  
Fuel Rod ◽  
Post Test ◽  
Nuclear Heating ◽  
Carbide Fuel ◽  
Unique Capability

AbstractTransient (nuclear) heating experiments were conducted with uranium carbide fuel rods to study the failure characteristics of typical rod designs. Extremely unusual changes in microstructure were observed and the electron microprobe was employed to establish the disposition of materials resulting from the meltdown experiments. The probe results indicated substantial fuel-clad interactions and permitted the resolution of several uncertainties regarding the course of fuel rod failure and material redistribution. The electron microprobe represents a unique capability in the post-test analysis of such meltdown tests.

A Comparison of Fast Thorium Breeder Reactor Designs With Oxide and Metallic Fuels

2008 ◽  
Author(s):  
V. Jagannathan ◽  
Usha Pal ◽  
R. Karthikeyan ◽  
Devesh Raj
Keyword(s):  
Unit Volume ◽  
Core Region ◽  
Metallic Fuel ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies ◽  
Depletion Rate ◽  
Metallic Fuels ◽  
Core Characteristics

Loading of seedless thoria rods in internal blanket regions and using them later as part of seeded fuel assemblies is the central theme of the thorium breeder reactor (ATBR) concept [1]. The fast reactors presently consider seedless blanket region surrounding the seeded core region. This results in slower fissile production rate in comparison to fissile depletion rate per unit volume. The overall breeding is achieved mainly by employing blanket core with more than double the volume of seeded core. The blanket fuel is discharged with fissile content of ∼30g/kg, which is much less than the asymptotic maximum possible fissile content of 100g/kg. This is due to smaller coolant flow provided for in the blanket regions. In a newly proposed fast thorium breeder reactor (FTBR) [2], the blanket region is brought in and distributed through out the core. By this the fissile depletion and production rates per unit volume become comparable. The core considered simultaneous breeding from both fertile thoria and depleted uranium and hence the concept can be called as fast twin breeder reactor as well. Sodium is used as coolant. The blanket fuel rods achieve nearly 80% of the seed fuel rod burnup and also contain nearly the maximum possible fissile content at the time of discharge. In this paper a comparison of FTBR core characteristics with oxide and metallic fuel are compared.

Projection of the Neutronic and Thermal Fuel Rod Behavior in a BWR

2004 ◽  
Author(s):  
Hector Hernandez Lopez ◽  
Javier Ortiz Villafuerte
Keyword(s):  
Fuel Assembly ◽  
Mechanical Performance ◽  
Nuclear Research ◽  
Operating Conditions ◽  
Fast Neutrons ◽  
Two Phase ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Thermal Limits ◽  
Operation Conditions

Currently, at the Instituto Nacional de Investigaciones Nucleares (National Institute for Nuclear Research) in Mexico, it is being developed a computational code for evaluating the neutronic, thermal and mechanical performance of a fuel element at several different operation conditions. The code is referred as to MCTP (Multigrupos con Temperaturas y Potencia), and is benchmarked against data from the Laguna Verde Nuclear Power Plant (LVNPP). In the code, the neutron flux is approximated by six groups of energy: one group in the thermal region (E &lt; 0.625 eV), four in the resonances region (0.625 eV &lt; E &lt; 0.861 MeV), and one group in the fast region (E &gt; 0.861 MeV). Thus, the code is able to determine the damage to the cladding due to fast neutrons. The temperature distribution is approximated in both axial and radial directions taking into account the changes in the coolant density, for both the single and two-phase regions in a BWR channel. It also considerate the changes in the thermal conductivity of all materials involved for the temperature calculations, as well as the temperature and density effects in the neutron cross sections. In the code, fuel rod burnup is evaluated. Also, plutonium production and poison production from fission. In this work, the neutronic and thermal performance of fuel rods in a 10×10 fuel assembly is evaluated. The fuel elements have a content of 235U. The fuel assembly was introduced to the unit 1 of LVNPP reactor core in the cycle 9 of operation, and will stay in during three cycles. In the analysis of fuel rod performance, the operating conditions are those for the cycle 9 and 10, whereas for the current cycle (cycle 11) the reactor is projected to operate during 460 days. The analysis for cycle 11 uses the actual location of the fuel assembly that will have in the core. The results show that the fuel rods analyzed did not reach the thermal limits during the cycles 9 and 10, as expected, and for cycle 11 the same thermal limits are not predicted to be reached.

scholarly journals Experimental Approaches for Manufacturing of Simulated Cladding and Simulated Fuel Rod for Mechanical Decladder

2020 ◽  
Vol 2020 ◽  
pp. 1-12
Author(s):  
Young-Hwan Kim ◽  
Yung-Zun Cho ◽  
Jin-Mok Hur
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Energy Research ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Compressive Stresses ◽  
Simulated Fuel ◽  
Ceramic Fuel ◽  
Experimental Approaches

We are developing a practical-scale mechanical decladder that can slit nuclear spent fuel rod-cuts (hulls + pellets) on the order of several tens of kgf of heavy metal/batch to supply UO2 pellets to a voloxidation process. The mechanical decladder is used for separating and recovering nuclear fuel material from the cladding tube by horizontally slitting the cladding tube of a fuel rod. The Korea Atomic Energy Research Institute (KAERI) is improving the performance of the mechanical decladder to increase the recovery rate of pellets from spent fuel rods. However, because actual nuclear spent fuel is dangerously toxic, we need to develop simulated spent fuel rods for continuous experiments with mechanical decladders. We describe procedures to develop both simulated cladding tubes and simulated fuel rod (with physical properties similar to those of spent nuclear fuel). Performance tests were carried out to evaluate the decladding ability of the mechanical decladder using two types of simulated fuel (simulated tube + brass pellets and zircaloy-4 tube + simulated ceramic fuel rod). The simulated tube was developed for analyzing the slitting characteristics of the cross section of the spent fuel cladding tube. Simulated ceramic fuel rod (with mechanical properties similar to the pellets of actual PWR spent fuel) was produced to ensure that the mechanical decladder could slit real PWR spent fuel. We used castable powder pellets that simulate the compressive stress of the real spent UO2 pellet. The production criteria for simulated pellets with compressive stresses similar to those of actual spent fuel were determined, and the castables were inserted into zircaloy-4 tubes and sintered to produce the simulated fuel rod. To investigate the slitting characteristics of the simulated ceramic fuel rod, a verification experiment was performed using a mechanical decladder.

scholarly journals Microstructures Manufactured in Diamond by Use of Laser Micromachining

Materials ◽  
2020 ◽  
Vol 13 (5) ◽  
pp. 1199
Author(s):  
Mariusz Dudek ◽  
Adam Rosowski ◽  
Marcin Kozanecki ◽  
Malwina Jaszczak ◽  
Witold Szymański ◽  
...  
Keyword(s):  
Chemical Vapor Deposition ◽  
Raman Spectra ◽  
Vapor Deposition ◽  
Microwave Plasma ◽  
Chemical Vapor ◽  
Polycrystalline Diamond ◽  
Scanning System ◽  
Carbon Phase ◽  
Diamond Plate ◽  
Diode Pumped

Different microstructures were created on the surface of a polycrystalline diamond plate (obtained by microwave plasma-enhanced chemical vapor deposition—MW PECVD process) by use of a nanosecond pulsed DPSS (diode pumped solid state) laser with a 355 nm wavelength and a galvanometer scanning system. Different average powers (5 to 11 W), scanning speeds (50 to 400 mm/s) and scan line spacings (“hatch spacing”) (5 to 20 µm) were applied. The microstructures were then examined using scanning electron microscopy, confocal microscopy and Raman spectroscopy techniques. Microstructures exhibiting excellent geometry were obtained. The precise geometries of the microstructures, exhibiting good perpendicularity, deep channels and smooth surfaces show that the laser microprocessing can be applied in manufacturing diamond microfluidic devices. Raman spectra show small differences depending on the process parameters used. In some cases, the diamond band (at 1332 cm−1) after laser modification of material is only slightly wider and shifted, but with no additional peaks, indicating that the diamond is almost not changed after laser interaction. Some parameters did show that the modification of material had occurred and additional peaks in Raman spectra (typical for low-quality chemical vapor deposition CVD diamond) appeared, indicating the growing disorder of material or manufacturing of the new carbon phase.

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