scholarly journals FACILITY FOR HYDROGENATION AND THERMAL TESTING OF INTERNALLY PRESSURIZED DUMMY FUEL RODS

2020 ◽  
pp. 195-197
Author(s):  
V.A. Dolgiy ◽  
V.A. Chunosov ◽  
D.L. Kokosha ◽  
Т.P. Chernyayeva ◽  
V.M. Grytsyna ◽  
...  
Keyword(s):  
Hydrogen Concentration ◽  
Thermal Testing ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Tangential Stresses ◽  
Test Parameters ◽  
Measurement Devices ◽  
Hydride Reorientation ◽  
Test Requirements ◽  
And Storage

A facility allowing to conduct experiments on dummy fuel rods up to 250 mm long, at temperatures up to 700 °C is presented. The designed facility is unique in that the conditions for the tests to be conducted on it most closely resemble (except for irradiation) those of fuel rods operation, loading and storage in SFDSF. All test parameters are programmed and regulated by special sensors, which brings the experiment on fuel rod temperature effect and pressure under the cladding as close as possible to the conditions of fuel rod operation in reactor and further storage in SFDSF. Hydride Reorientation Test (HRT) was conducted on dummy fuel rod sunder internal pressure of 3...5 MPa (at room temperature) and with hydrogen concentration of 50...300 ppm in the modes that simulate SNF handling with limiting heating to 410 °С and accidents with seven 410↔300 and 410↔180 °С thermal cycles. It has been demonstrated that the effectiveness of the influence of the test conditions in the specified modes on hydride reorientation increases with increasing hydrogen concentration and tangential stresses in the dummy fuel rod claddings. It has been shown that the test samples design, control and measurement devices, as well as the parameters estimated during the test and further investigations fully meet the test requirements.

The Design of Actuating Cylinder Salt Spray Testing Device Based on Lab Windows/CVI and PLC Control

2011 ◽  
Vol 411 ◽  
pp. 269-273
Author(s):  
Geng He Luo ◽  
Yu Wang ◽  
Bin Gao Yu
Keyword(s):  
Salt Spray ◽  
Computer Software ◽  
Experimental Device ◽  
Control Precision ◽  
Aircraft Landing ◽  
Plc Control ◽  
Test Parameters ◽  
Aircraft Landing Gear ◽  
Test Requirements ◽  
And Storage

The performance tests and the requirements of experiments on aircraft alighting gear actuating cylinder in the salt spray and hot and humid environment are presented in this paper, and the hydraulic system of experimental device is designed. The host computer software is developed by Lab Windows/CVI for actuating cylinder on test parameters settings, monitoring, and storage and record, The Siemens PLC are used at the sub-processors to control the salt spray experimental device. The field-test shows that the device control precision is higher to pressure, flow rate, and the parameters such as temperature, and it can realize the simulation for different salt spray and hot and humid environments. The test on aircraft landing gear by using actuating cylinder has a good human-computer interaction and communication function. The device is safe, reliable and it can reach the test requirements.

scholarly journals Effect of Hydrogen, Hydride Orientation and Temperature on Low-Cycle Fatigue Resistance of Zr-1%Nb Fuel Rod Claddings

2021 ◽  
pp. 53-59
Author(s):  
G. Riedkina ◽  
V. Grytsyna ◽  
S. Klymenko ◽  
Т. Chernyayeva
Keyword(s):  
Fatigue Life ◽  
Fatigue Resistance ◽  
Hydrogen Concentration ◽  
Hydrogen Content ◽  
Fatigue Testing ◽  
Low Cycle Fatigue ◽  
Outer Diameter ◽  
Cycle Fatigue ◽  
Fuel Rod ◽  
Hydride Reorientation

Low-cycle fatigue testing was conducted on annular samples with an outer diameter of 9.13 mm, a wall thickness of 0.68 mm and a width of 2.7 mm, namely: non-hydrogenated samples (cut out of standard Zr‑1%Nb cladding tubes); hydrogenated samples with a hydrogen concentration of 50 ... 400 ppm; samples cut out from hydrogenated dummy claddings after hydride reorientation tests performed according to various test modes. The tests were conducted at the temperatures of 25, 180, 350, 400 and 450 °С. The results obtained demonstrate that with increasing the hydrogen content in Zr-1%Nb alloy claddings the fatigue life increases.

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.

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 < 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.

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.

Measurement of Two-Dimensional Void Fraction Distributions of Rising Bubbles in a Simulated Sub-Channel by Wire-Mesh Sensors at Conditions of Forced Convective and Stagnant Flows

2017 ◽  
Author(s):  
Yota Suzuki ◽  
Yusei Tanaka ◽  
Taku Sakka ◽  
Akinori Sato ◽  
Kazuyuki Takase ◽  
...  
Keyword(s):  
Flow Condition ◽  
Two Phase Flow ◽  
Flow Behavior ◽  
Wire Mesh ◽  
Phase Flow ◽  
Two Dimensional ◽  
Two Phase ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Stagnant Flow

Clarifying thermal-hydraulic characteristics in a nuclear reactor core is important in particular to enhance the thermo-fluid safety of nuclear reactors. Spacers installed in subchannels of fuel assemblies have the role of keeping the interval between adjacent fuel rods constantly. Similarly, in case of PWR the spacer has also the role as the turbulence promoter. When the transient event occurs, two-phase flow is generated by boiling of water due to heating of fuel rods. Therefore, it is important to confirm the two-phase flow behavior around the spacer. So, the effect of the spacer affecting the two-phase flow was investigated experimentally at forced convective flow condition. Furthermore, in order to improve the thermal safety of current light water reactors, it is necessary to clarify the two-phase flow behavior in the subchannels at the stagnant flow condition. So, the bubbly flow data around a simulated fuel rod were obtained experimentally at the stagnant flow condition. A wire-mesh sensor was used to obtain a detailed two-dimensional void fraction distribution around the simulated spacer and fuel rod. As a result of this research, the bubbly behavior around the simulated spacer and fuel rod was qualitatively revealed and also bubble dynamics in the sub-channels at the conditions of forced convective and stagnant flows were evaluated. The present experimental data are very useful for verifying the detailed three-dimensional two-phase flow analysis codes.

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