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.

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 Simplified Model of a High Burnup Spent Nuclear Fuel Rod under Lateral Impact Considering a Stress-Based Failure Criterion

Metals ◽  
2021 ◽  
Vol 11 (10) ◽  
pp. 1631
Author(s):  
Seyeon Kim ◽  
Sanghoon Lee
Keyword(s):  
Nuclear Fuel ◽  
Failure Criterion ◽  
Power Plants ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Simplified Model ◽  
Beam Model ◽  
Detailed Model ◽  
Fuel Rod ◽  
High Burnup

The inventory of spent nuclear fuel (SNF) generated in nuclear power plants is continuously increasing, and it is very important to maintain the structural integrity of SNF for economical and efficient management. The cladding surrounding nuclear fuel must be protected from physical and mechanical deterioration, which causes fuel rod breakage. In this study, the material properties of the simplified beam model of a SNF rod were calibrated for a drop accident evaluation by considering the pellet–clad interaction (PCI) of the high burnup fuel rod. In a horizontal drop, which is the most damaging during a drop accident of SNF, the stress in the cladding caused by the inertia action of the pellets has a great effect on the integrity of the fuel rod. The failure criterion for SNF was selected as the membrane plus bending stress through stress linearization in the cross-sections through the thickness of the cladding. Because the stress concentration in the cladding around the vicinity of the pellet–pellet interface cannot be simulated in a simplified beam model, a stress correction factor is derived through a comparison of the simplified model and detailed model. The applicability of the developed simplified model is checked through dynamic impact simulations. The developed model can be used in cask level analyses and is expected to be usefully utilized to evaluate the structural integrity of SNF under transport and in storage conditions.

scholarly journals The Limitations of an Air-Oxidation Breakaway Model to Predict a Zirconium Fire in a Spent Nuclear Fuel Pool Accident

2019 ◽  
Vol 11 (22) ◽  
pp. 6364
Author(s):  
Sanggil Park ◽  
Min Bum Park
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Critical Phenomenon ◽  
Fission Products ◽  
Model Parameters ◽  
Spent Fuel ◽  
Air Oxidation ◽  
Loss Of Coolant Accident ◽  
Spent Nuclear Fuel Pool ◽  
Spent Fuel Pool

The OECD/NEA Spent Fuel Pool (SFP) project was conducted to investigate consequences of spent nuclear fuel pool accident scenarios. From the project, it was observed that cladding temperature could abruptly increase at a certain point and the cladding was completely oxidized. This phenomenon was called a “zirconium fire”. This zirconium fire is one of the crucial concerns for spent fuel pool safety under a postulated loss of coolant accident scenario, since it would lead to an uncontrolled mass release of fission products into the environment. To capture this critical phenomenon, an air-oxidation breakaway model has been implemented in the MELCOR code. This study examines this air-oxidation breakaway model by comparing the SFP project test data with a series of MELCOR code sensitivity calculation results. The air-oxidation model parameters are slightly altered to investigate their sensitivities on the occurrence of the zirconium fire. Through such sensitivity analysis, limitations of the air-oxidation breakaway model are identified, and needs for model improvement is recommended.

Japanese Codes and Standards for Dual Purpose Metal Casks for Spent Nuclear Fuel

2014 ◽  
Author(s):  
Toshiari Saegusa ◽  
Makoto Hirose ◽  
Norikazu Irie ◽  
Masashi Shimizu
Keyword(s):  
Nuclear Fuel ◽  
Power Plants ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Spent Fuel ◽  
Dual Purpose ◽  
Service Conditions ◽  
Interim Storage ◽  
And Storage

The first Japanese spent fuel interim storage facility away from a reactor site is about to be commissioned in Mutsu City, Aomori Prefecture. In designing, licensing and construction of the Dual Purpose Casks (DPCs, for transport and storage) for this facility, codes and standards established by the Atomic Energy Society of Japan (AESJ) and by the Japan Society of Mechanical Engineers (JSME) have been applied. The AESJ established the first standard for DPCs as “Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facilities” in 2002 (later revised in 2010). The standard provides the design requirements to maintain the basic safety functions of DPCs, namely containment, heat removal, shielding, criticality prevention and the structural integrity of the cask itself and of the spent fuel cladding during transport and storage. Inspection methods and criteria to ensure maintenance of the basic safety functions and structural integrity over every stage of operations involving DPCs including pre-shipment after storage are prescribed as well. The structural integrity criteria for major DPC components refer to the rules provided by the JSME. JSME completed the structural design and construction code (the Code) for DPCs as “Rules on Transport/Storage Packagings for Spent Nuclear Fuel” in 2001 (later revised in 2007). Currently, the scope of the rules cover the Containment Vessel, Basket, Trunnions and Intermediate Shell as major components of DPCs. Rules for these components are based on those for components of nuclear power plants (NPP) with similar safety functions, but special considerations based on their shapes, loading types and required functions are added. The Code has differences from that for NPP components with considerations to DPC characteristics; - The primary stress and the secondary stress generated in Containment Vessels shall be evaluated under Service Conditions A to D (from ASME Sec III, Div.1). - Stress generated in the seal region lid bolts of Containment Vessels shall not exceed yield strength under Service Conditions A to D in order to maintain the containment function. - Fatigue analysis on Baskets is not required, and Trunnions can be designed only for Service Conditions A and B with special stress limits consistent with conventional assessment methods for transport packages. - Stress limits for earthquakes during storage are specified. - Ductile cast iron with special fracture toughness requirements can be used as a material for Containment Vessels. DPC specific considerations in standards and rules will be focused on in this paper. Additionally, comparison with the ASME Code will be discussed.

scholarly journals Structural integrity investigations of spent nuclear fuel with finite element modeling

2021 ◽  
Vol 1 ◽  
pp. 13-14
Author(s):  
Efstathios Vlassopoulos ◽  
Susanne Pudollek ◽  
Olympios Alifieris ◽  
Dimitrios Papaioannou ◽  
Ramil Nasyrow ◽  
...  
Keyword(s):  
Finite Element ◽  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Structural Integrity ◽  
Mechanical Response ◽  
High Level Waste ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Final Disposal ◽  
National Cooperative

Abstract. Radioactive waste in Switzerland will be disposed of in a deep geological repository (DGR). Responsible for the planning and preparation of realization of this task is National Cooperative for the Disposal of Radioactive Waste (Nagra). Spent fuel assemblies (SFA) constitute the main high-level waste (HLW) stream that will be disposed in the DGR. Prior to final disposal they will be transferred or transported to an encapsulation plant, where they will be loaded into final disposal canisters. To ensure that the structural integrity of SFAs is not compromised during handling and transportation, it is desirable to characterize the expected mechanical parameters of SFAs after long-term interim storage. Experimental research activities performed at the JRC Karlsruhe include safety aspects of radioactive waste management, encompassing also spent fuel storage and spent fuel/HLW disposal activities. Nagra and JRC have established a collaboration to jointly study relevant properties and behaviours of spent fuel rods, with the support of the Gösgen nuclear power plant and of Framatome, and in collaboration with other partners in Europe and internationally. As part of this collaboration, 3-point bending and impact tests were performed at the hot-cell facilities of JRC Karlsruhe, to determine the mechanical response of spent fuel rodlets under quasi-static and dynamic loads. The structural integrity of fuel rods was also evaluated under different handling scenarios using finite element (FE) analysis. Starting with the construction of a static 3D FE model of a Pressurized Water Reactor (PWR) nuclear fuel rodlet in ANSYS Mechanical, Nagra has developed a series of FE models over the years. Mechanical properties of the original rodlet model were derived through an extensive validation process, using experimental data from the 3-point bending tests. To evaluate the mechanical response of an SFA in different loading scenarios, this model was expanded using 1D beam modeling approach. The development of the simplified 1D models is shown in this presentation. In particular, the effect of the contact formulation between the spacer grid and the rods is discussed. Finally, preliminary results of the bending response of a 15×15 PWR SFA sub-model are presented.

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 The Buckling of Fuel Rods Under Inertia Loading

2009 ◽  
Author(s):  
Gordon S. Bjorkman
Keyword(s):  
Structural Integrity ◽  
Buckling Analysis ◽  
Buckling Load ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Critical Buckling Load ◽  
Fuel Mass ◽  
Fuel Rod ◽  
The Subject ◽  
G Value

The buckling analysis of fuel rods during an end drop impact of a spent fuel transportation cask has traditionally been performed to demonstrate the structural integrity of the fuel rod cladding or the integrity of the fuel geometry in criticality evaluations for a cask drop event. The actual calculation of the fuel rod buckling load, however, has been the subject of some controversy, with estimates of the critical buckling load differing by as much as a factor of 5. Typically, in the buckling analysis of a fuel rod, assumptions are made regarding the percentage of fuel mass that is bonded to or that participates with the cladding during the buckling process, with estimates ranging from 0 to 100%. The greater the percentage of fuel mass that is assumed to be bonded to the cladding the higher the inertia loads on the cladding, and, therefore, the lower the “g” value at which buckling occurs. However, these solutions do not consider displacement compatibility between the fuel and the cladding during the buckling process. By invoking displacement compatibility between the fuel column and the cladding column, this paper presents an exact solution for the buckling of fuel rods under inertia loading. The results show that the critical inertia load magnitude for the buckling of a fuel rod depends on the weight of the cladding and the total weight of the fuel, regardless of the percentage of fuel mass that is assumed to be attached to or participate with the cladding in the buckling process. Therefore, 100% of the fuel always participates in the buckling of a fuel rod under inertia loading.

Commercial Spent Nuclear Fuel Integrity During Long-Term Dry Storage

2001 ◽  
Author(s):  
Leroy Stewart ◽  
Mikal A. McKinnon
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Dry Storage ◽  
Environmental Laboratory ◽  
Cooperative Project

Abstract The United States Department of Energy (DOE) Office of Civilian Radioactive Waste Management conducted spent nuclear fuel integrity and cask performance tests from 1984–1996 at the Idaho National Engineering and Environmental Laboratory (INEEL). Between 1994 and 1998, DOE also initiated a Spent Fuel Behavior Project that involved enhanced surveillance, monitoring, and gas-sampling activities for intact fuel in a GNS CASTOR V/21 cask and for consolidated fuel in a Sierra Nuclear VSC-17 cask. The results of these series of tests are reported in this paper. Presently, DOE is involved in a cooperative project to perform destructive evaluations of fuel rods that have been stored in the CASTOR V/21 cask. The results of those evaluations are presented elsewhere in these proceedings in a paper entitled “Examination of Spent PWR Fuel Rods after 15 years in Dry 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.

Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel

2006 ◽  
Vol 985 ◽  
Author(s):  
Jeffrey A. Fortner ◽  
A. Jeremy Kropf ◽  
James L. Jerden ◽  
James C. Cunnane
Keyword(s):  
Transition Zone ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Aqueous Environment ◽  
Spent Fuel ◽  
Aqueous Corrosion ◽  
Reduction Potentials ◽  
Chemical Effects ◽  
Before And After ◽  
X Ray Absorption

AbstractPerformance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO2 matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO2 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO22+) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of ∼ 50 micrometers. We find evidence of a thin (∼ 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO2+/Np4+ and UO22+/U4+ couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions.

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