Evaluation of Fretting Damage of Zircaloy Cladding Tubes

1996 ◽  
Vol 118 (4) ◽  
pp. 705-710 ◽  
Author(s):  
Olof Vingsbo ◽  
Ali R. Massih ◽  
Stig Nilsson
Keyword(s):  
Fuel Assembly ◽  
Laboratory Testing ◽  
Nuclear Reactors ◽  
Full Scale ◽  
Endurance Test ◽  
Fuel Rods ◽  
Damage Characteristics ◽  
Actual Service ◽  
Service Conditions ◽  
Fretting Damage

The work is part of the search for a technique to study the fretting properties of the cladding of fuel rods in nuclear reactors. The strategy is to compare the fretting scars of specimens tested in a near-full-scale, nonradioactive simulator (so called fuel assembly endurance test) with scars on specimens tested under well controlled conditions in the laboratory. By systematic variations of the laboratory testing parameters it has proved possible to achieve scars with the same type of damage characteristics as those from the fuel assembly endurance test. This makes it possible to estimate, by deduction, the parameters that cause fretting damage under actual service conditions in a reactor.

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.

scholarly journals Full-Scale Laboratory Testing Of A Geosynthetically Reinforced Soil Railway Structure

2021 ◽  
pp. 100526
Author(s):  
A.F. Esen ◽  
P.K. Woodward ◽  
O. Laghrouche ◽  
T.M. Čebašek ◽  
A.J. Brennan ◽  
...  
Keyword(s):  
Laboratory Testing ◽  
Reinforced Soil ◽  
Full Scale

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.

Endurance test of full-scale mock-up helium circulator for HTR-PM

2018 ◽  
Vol 329 ◽  
pp. 20-24 ◽  
Author(s):  
Gang Zhao ◽  
Ping Ye ◽  
Hong Wang ◽  
Liang Zhu ◽  
Qinzhao Zhang ◽  
...  
Keyword(s):  
Full Scale ◽  
Endurance Test

2D Natural Convection and Radiation Heat Transfer Simulations of a PWR Fuel Assembly Within a Constant Temperature Support Structure

2006 ◽  
Author(s):  
Pablo E. Araya Go´mez ◽  
Miles Greiner
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Fuel Assembly ◽  
Heat Generation ◽  
Temperature Difference ◽  
Wall Temperature ◽  
Radiation Heat Transfer ◽  
Radiation Heat ◽  
Fuel Rods ◽  
Water Reactor

Two-dimensional simulations of steady natural convection and radiation heat transfer for a 14×14 pressurized water reactor (PWR) spent nuclear fuel assembly within a square basket tube of a typical transport package were conducted using a commercial computational fluid dynamics package. The assembly is composed of 176 heat generating fuel rods and 5 larger guide tubes. The maximum cladding temperature was determined for a range of assembly heat generation rates and uniform basket wall temperatures, with both helium and nitrogen backfill gases. The results are compared with those from earlier simulations of a 7×7 boiling water reactor (BWR). Natural convection/radiation simulations exhibited measurably lower cladding temperatures only when nitrogen is the backfill gas and the wall temperature is below 100°C. The reduction in temperature is larger for the PWR assembly than it was for the BWR. For nitrogen backfill, a ten percent increase in the cladding emissivity (whose value is not well characterized) causes a 4.7% reduction in the maximum cladding to wall temperature difference in the PWR, compared to 4.3% in the BWR at a basket wall temperature of 400°C. Helium backfill exhibits reductions of 2.8% and 3.1% for PWR and BWR respectively. Simulations were performed in which each guide tube was replaced with four heat generating fuel rods, to give a homogeneous array. They show that the maximum cladding to wall temperature difference versus total heat generation within the assembly is not sensitive to this geometric variation.

scholarly journals Field verification of a new asphalt compactor, AMIR

1991 ◽  
Vol 18 (3) ◽  
pp. 465-471 ◽  
Author(s):  
Otto J. Svec ◽  
A. O. Abd El Halim
Keyword(s):  
Key Words ◽  
Laboratory Testing ◽  
Research Council ◽  
Joint Venture ◽  
National Research Council ◽  
Field Trials ◽  
Full Scale ◽  
Field Verification ◽  
Asphalt Mix

A prototype of a new asphalt compactor termed "asphalt multi-integrated roller (AMIR)" was built as a joint venture between the National Research Council of Canada (NRC) and a Canadian manufacturer, Lovat Tunnel Equipment, Inc. The purpose of this project was to prove this new compaction concept in a full-scale environment. This paper describes one of the field trials carried out on the campus of the NRC and reports the results quantifying the quality of the AMIR compaction. Key words: compactor, asphalt mix, field trials, laboratory testing.

scholarly journals Numerical Investigation on the Flow Characteristics in a 17 × 17 Full-Scale Fuel Assembly

Energies ◽  
2020 ◽  
Vol 13 (2) ◽  
pp. 397 ◽  
Author(s):  
Zihao Tian ◽  
Lixin Yang ◽  
Shuang Han ◽  
Xiaofei Yuan ◽  
Hongyan Lu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Flow Characteristics ◽  
Full Scale ◽  
Pressurized Water Reactor ◽  
Lateral Velocity ◽  
Pressurized Water ◽  
Spacer Grids ◽  
Rod Bundle ◽  
Real Size ◽  
Cfd Method

In a previous study, several computational fluid dynamics (CFD) simulations of fuel assembly thermal-hydraulic problems were presented that contained fewer fuel rods, such as 3 × 3 and 5 × 5, due to limited computer capacity. However, a typical AFA-3G fuel assembly consists of 17 × 17 rods. The pressure drop levels and flow details in the whole fuel assembly, and even in the pressurized water reactor (PWR), are not available. Hence, an appropriate CFD method for a full-scale 17 × 17 fuel assembly was the focus of this study. The spacer grids with mixing vanes, springs, and dimples were considered. The polyhedral and extruded mesh was generated using Star-CCM+ software and the total mesh number was about 200 million. The axial and lateral velocity distribution in the sub-channels was investigated. The pressure distribution downstream of different spacer grids were also obtained. As a result, an appropriate method for full-scale rod bundle simulations was obtained. The CFD analysis of thermal-hydraulic problems in a reactor coolant system can be widely conducted by using real-size fuel assembly models.

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 < 0.625 eV), four in the resonances region (0.625 eV < E < 0.861 MeV), and one group in the fast region (E > 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.

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