scholarly journals IDENTIFICATION OF THE FUEL ROD CLADDING DESTRUCTION FROM THE CHANGE OF THE SNF STORAGE CASKS RADIATION

2020 ◽  
pp. 111-119
Author(s):  
V.G. Rudychev ◽  
N.A. Azarenkov ◽  
I.O. Girka ◽  
Y.V. Rudychev
Keyword(s):  
Numerical Modeling ◽  
Gamma Radiation ◽  
Dose Rate ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Side Surface ◽  
Neutron Radiation ◽  
External Radiation ◽  
Fuel Rods ◽  
Fuel Rod

Two options for changing the distribution of spent nuclear fuel due to the possible destruction of the cladding of fuel rods, which causes a change in radiation outside the cask, are considered for VSC-24 casks used for storage of spent nuclear fuel by the dry method. The effect of height reduction due to the destruction of the fuel rods of all 24 SFAs and 10 central SFAs on external radiation is studied analytically and by numerical modeling in the MCNP package. The destruction of 24 SFA is shown to lead to a significant decrease in the dose rate of neutrons and gamma-radiation from 60Co on the weather lid of the cask, and of gamma-radiation from SNF isotopes at the mid-height of the side surface of the cask. The destruction of the ten central SFAs can be determined only from a change in the neutron radiation in the air inlets of the cask.

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 DEVELOPMENT OF A MODELING APPROACH TO ESTIMATE RADIATION FROM A SPENT FUEL ROD QUIVER

2021 ◽  
Vol 247 ◽  
pp. 16006
Author(s):  
Zs. Elter ◽  
V. Mishra ◽  
S. Grape ◽  
E. Branger ◽  
P. Jansson ◽  
...  
Keyword(s):  
Gamma Radiation ◽  
Spent Nuclear Fuel ◽  
Radiation Transport ◽  
Neutron Radiation ◽  
Nuclear Safeguards ◽  
Monte Carlo Code ◽  
Spent Fuel ◽  
Fuel Rods ◽  
History Of ◽  
New Type

Before encapsulation of spent nuclear fuel in a geological repository, the fuels need to be verified for safeguards purposes. This requirement applies to all spent fuel assemblies, including those with properties or designs that are especially challenging to verify. One such example are quivers, a new type of containers used to hold damaged spent fuel rods. After placing damaged rods inside the quivers, they are sealed with a thick lid and the water is removed. The lid is thick enough to significantly reduce the amount of the gamma radiation penetrating through it, which can make safeguards verification from the top using gamma techniques difficult. Considering that the number of quivers at storage facilities is foreseen to increase in near future, studying the feasibility of verification is timely. In this paper we make a feasibility study related to safeguards verification of quivers, aimed at investigating the gamma and neutron radiation field around a quiver designed by Westinghouse AB and filled with PWR fuel rods irradiated at the Swedish Ringhals site. A simplified geometry of the quiver and the detailed operational history of each rod are provided by Westinghouse and the reactor operator, respectively. The nuclide inventory of the rods placed in the quiver and the emission source terms are calculated with ORIGEN-ARP. The radiation transport is modeled with the Serpent2 Monte Carlo code. The first objective is to assess the capability of the spent fuel attribute tester (SFAT) to verify the content for nuclear safeguards purposes. The results show that the thick quiver lid attenuates the gamma radiation, thereby making gamma radiation based verification from above the quiver difficult.

scholarly journals Development of Equivalent Beam Model of High Burnup Spent Nuclear Fuel Rods under Lateral Impact Loading

Metals ◽  
2020 ◽  
Vol 10 (4) ◽  
pp. 470
Author(s):  
Sanghoon Lee ◽  
Seyeon Kim
Keyword(s):  
Nuclear Fuel ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Model Parameters ◽  
Beam Model ◽  
Spent Fuel ◽  
Analysis Model ◽  
Fuel Rods ◽  
Integrity Assessment ◽  
Fuel Rod

Spent nuclear fuel (SNF) is nuclear fuel that has been irradiated and discharged from nuclear reactors. During the whole management stages of SNF before it is, in the end, disposed in a deep geological repository, the structural integrity of fuel rods and the assemblies should be maintained for safety and economic reasons. In licensing applications for the SNF storage and transportation, the integrity of SNF needs to be evaluated considering various loading conditions. However, this is a challenging task due to the complexity of the geometry and properties of SNF. In this paper, a simple and equivalent analysis model for SNF rods is developed using model calibration based on optimization and process integration. The spent fuel rod is simplified into a hollow beam with a homogenous isotropic material, and the model parameters thus found are not dependent on the length of the reference fuel rod segment that is considered. Two distinct models with different interfacial conditions between the fuel pellets and cladding are used in the calibration to account for the effect of PCMI (Pellet-Clad Mechanical Interaction). The feasibility of the models in dynamic impact simulations is examined, and it is expected that the developed models can be utilized in the analysis of assembly-level analyses for the SNF integrity assessment during transportation and storage.

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.

Occupational Exposure to Neutrons at the Beloyarsk NPP

ANRI ◽  
2021 ◽  
Vol 0 (3) ◽  
pp. 16-26
Author(s):  
Mariya Pyshkina ◽  
Aleksey Vasil'ev ◽  
Aleksey Ekidin ◽  
Evgeniy Nazarov ◽  
Anton Pudovkin ◽  
...  
Keyword(s):  
Occupational Exposure ◽  
Energy Distribution ◽  
Flux Density ◽  
Nuclear Fuel ◽  
Neutron Irradiation ◽  
Radiation Flux ◽  
Spent Nuclear Fuel ◽  
Neutron Radiation ◽  
Dose Equivalent ◽  
Equivalent Rate

Studies of the energy distribution of neutron radiation at the workplaces of the Beloyarsk NPP were carried out. At 1 and 2 power units, occupational exposure of neutron irradiation occurs during operations for loading spent nuclear fuel into special railway carriage. At power units 3 and 4, operations accompanied by neutron irradiation can be divided into 3 groups: (1) work in rooms adjacent to the reactor core; (2) manipulation of radioisotope neutron sources; (3) work with fresh and spent nuclear fuel. Based on the data obtained on the energy distribution of the neutron radiation flux density, the ‘true’ values of the ambient dose equivalent rate H*(10), the individual dose equivalent rate Hp(10) and the integral neutron radiation flux density at individual workplaces were determined. For each group of workplaces, Fluence-toambient dose equivalent conversion coefficients are determined, which lie in the range from 12 to 295 pSv⋅cm2. Correction factors for individual thermoluminescent dosimeters, taking into.

Numerical Analysis of Transient Temperature Distribution in a Partially Cooled Nuclear Fuel Rod

2015 ◽  
Author(s):  
Sandeep Patil ◽  
Siddarth Chintamani ◽  
Rajeev Kumar ◽  
Ratan Kumar ◽  
Brian H. Dennis
Keyword(s):  
Temperature Distribution ◽  
Numerical Analysis ◽  
Nuclear Fuel ◽  
Learning Curves ◽  
Coolant Flow ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Coolant Water ◽  
Nuclear Fuel Rods ◽  
Temperature Levels

Critical safety studies of a nuclear power plants are often associated with inadequate and improper cooling of the reactor core or the spent fuel rods. Coolant flow over the hot nuclear fuel rods often gets stalled during major accidents resulting in high temperature levels. These elevated temperature levels can potentially melt the fuel rod material and cause the release of radioactive gases. Research activities, both numerical and experimental in nature to explore these rare but potentially catastrophic possibilities have resulted in sophisticated numerical codes capable of simulating the various post-accident scenarios. These codes, although reasonably accurate and reliable have steep learning curves and are not often very user-friendly. A fast and accurate prediction of the critical temperature conditions using popular commercially available software packages is the subject of current study. Results from this parametric study of temperature distribution over a partially cooled fuel rod carried out using ANSYS as the numerical analysis tool is reported. Nuclear fuel rods being inadequately cooled inside a stagnant pool of coolant water in an accident scenario resulting in disrupted coolant flow has been simulated. This situation can arise within the reactor (design-basis accidents) or in the waste-fuel storage (as faced in Fukushima). In these situations, the fuel rod is often left partially immersed in the coolant water resulting in immersed portion of the rod cooled by water and the exposed portion cooled by air leading to non-uniform and improper cooling of the system. Realistic dimensions and materials as in commercial nuclear fuel rod have been used in the study. Taking advantage of the symmetry, an axisymmetric radial plane sliced longitudinally has been analyzed. Variations in the tangential direction have been neglected. The heat transfer problem uses homogeneous convective boundary conditions and assumes temperature dependent thermal conductivity. The parameters varied are the coolant level and the heat generation rate inside the fuel rod. A macro to automatically capture the transients in the temperatures was written in ANSYS (a finite element package). The governing energy equations were implicitly solved using finite volume scheme in MATLAB. ANSYS results are in close agreement with those obtained using MATLAB. The centerline temperature of the fuel rod shows a sharp rise below a certain coolant level.

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