scholarly journals Simplified beam model of high burnup spent fuel rod under lateral load considering pellet-clad interfacial bonding influence

2019 ◽  
Vol 51 (5) ◽  
pp. 1333-1344 ◽  
Author(s):  
Sanghoon Lee ◽  
Seyeon Kim
Keyword(s):  
Lateral Load ◽  
Interfacial Bonding ◽  
Beam Model ◽  
Spent Fuel ◽  
Fuel Rod ◽  
High Burnup

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 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.

Chemical Durability of High Burnup Lwr-Spent Fuel in Concentrated Salt Solutions

1993 ◽  
Vol 333 ◽  
Author(s):  
A. Loida ◽  
B. Grambow ◽  
P. Dressier ◽  
K. Friese ◽  
H. Geckeis ◽  
...  
Keyword(s):  
Surface Area ◽  
Volume Ratio ◽  
Sample Surface ◽  
Solution Concentration ◽  
Reaction Rates ◽  
Spent Fuel ◽  
Intrinsic Dissolution ◽  
Fine Grained ◽  
High Burnup ◽  
Saturation Effects

ABSTRACTHigh-burnup (<50 MWd/kgU) spent fuel samples of various sizes were exposed to NaCl solutions under static, anaerobic and reducing conditions. The accumulated corrosion time was about 200 days. Gas phase and leach solutions were analyzed. By dissolving mm sized fragments in large volumes of solution, saturation effects were avoided and upper limits for intrinsic dissolution rates of about 5-20 mg/(m2d) were measured. Surface area normalized reaction rates were significantly lower when using fine grained fuel powder (estimated sample surface area to solution volume ratio S/V ca. 3000 m-1), indicating saturation effects. The maximum concentrations of Pu and Am in the tests are close to reported solubility limited concentrations in pure 5m NaCl solutions in the absence of radiolysis effects. The presence of iron effectively reduces the solution concentration of all measured radionuclides (except Cs).

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.

RIA Failure of High Burnup Fuel Rod Irradiated in the Leibstadt Reactor: Out-of-Pile Mechanical Simulation and Comparison with Pulse Reactor Tests

2010 ◽  
Vol 7 (9) ◽  
pp. 102988 ◽  
Author(s):  
V. Grigoriev ◽  
R. Jakobsson ◽  
D. Schrire ◽  
G. Ledergerber ◽  
T. Sugiyama ◽  
...  
Keyword(s):  
Pulse Reactor ◽  
Mechanical Simulation ◽  
Fuel Rod ◽  
High Burnup

Structural Integrity Demonstration of High Burnup Fuel Rod

2006 ◽  
Author(s):  
Tong Liu ◽  
Danrong Song ◽  
Hua Zhao ◽  
Kaiming Wang
Keyword(s):  
Structural Integrity ◽  
Design Criterion ◽  
Current Status ◽  
Design Codes ◽  
Fuel Rod ◽  
Fuel Models ◽  
High Burnup ◽  
Near Future

This paper introduces the current status and the work has been done for structural integrity demonstration of high burnup fuel in China. The FRAMATOME ANP COCCINEL, FLAMBE and MISTIGRI design codes are used in the structural integrity demonstration of fuel rod. This paper presents the results of structural integrity demonstration of AFA3G fuel rod in GNPS. All the design criterion are met. Based on CAFA program, the new fuel models are planning to be developed in the near future.

Liquid-Metal Fast Breeder Reactor Fuel Rod Performance and Modeling at High Burnup

1978 ◽  
Vol 39 (2) ◽  
pp. 167-185 ◽  
Author(s):  
P. Verbeek ◽  
H. Többe ◽  
N. Hoppe ◽  
B. Steinmetz
Keyword(s):  
Liquid Metal ◽  
Reactor Fuel ◽  
Fast Breeder Reactor ◽  
Fuel Rod ◽  
High Burnup

Packaging of High Burnup UO2 and MOX Spent Fuel Considering Temperature Restrictions in the Near Field of a Repository

1997 ◽  
Vol 506 ◽  
Author(s):  
D.F. McGinnes ◽  
J. W. Schneider
Keyword(s):  
Near Field ◽  
Management Program ◽  
High Level Waste ◽  
Heat Output ◽  
Spent Fuel ◽  
Ambient Temperatures ◽  
Heat Loading ◽  
High Burnup ◽  
Maximum Temperatures ◽  
High Level

ABSTRACTThe direct disposal of spent fuel is one of the options considered in the Swiss high level waste management program. One of the important questions, within this program, is the heat generation from high-burnup UO2and MOX spent fuels. Depending on the repository boundary conditions (e.g. ambient temperatures at depth, thermal properties of the host rock etc.), on the maximum temperatures allowed in the near field and on the heat output of the fuel, it may not always be possible to completely fill the conceptual waste canister. The aim of this paper is to address the potential loading of spent fuel into canisters for different possible repository heat loading restrictions

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