• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.163-170
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• 2023

Since spent nuclear fuel assemblies (SFA) are transported to interim storage or final disposal facility after cooling the decay heat, finite element analysis (FEA) with simplification is widely used to show their integrity against cladding failure to cause dispersal of radioactive material. However, there is a lack of research addressing the comprehensive impact of shape and element simplification on analysis results. In this study, for the optimization of a typical pressurized water reactor SFA, different types of finite element models were generated by changing number of fuel rods, fuel rod element type and assembly length. A series of FEA in use of these different models were conducted under a shock load data obtained from surrogate fuel assembly transportation test. Effects of number of fuel rods, element type and length of assembly were also analyzed, which shows that the element type of fuel rod mainly affected on cladding strain. Finally, an optimal finite element model was determined for other practical application in the future.

• Nuclear Engineering and Technology
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• v.10 no.2
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• pp.65-72
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• 1978

In this paper, it is attempted to investigate the phenomena of iodine stress corrosion cracking of Zircaloy-4 cladding failures in reactor through the results of similar out-of-pile test in iodine vapour. The main result of this experiment is a finding of the relation between the threshold stress which can lead to iodine stress corrosion cracking of Zircaloy-4 tube and the iodine concentration. The values of critical stress and the critical iodine concentration are also obtained. A model which relates failure time of Zircaley-4 tube to failure stress and iodine concentration is suggested as follows: log t$_{F}$ =5.5-(3/2)log$_{c}$-4log $\sigma$ where t$_{F}$ : failure time, minutes c: iodne concentration, mg/㎤ $\sigma$: stress, 10$^4$psi.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1333-1344
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• 2019

An integrated approach of model simplification for high burnup spent nuclear fuel is proposed based on material calibration using optimization. The spent fuel rods are simplified into a beam with a homogenous isotropic material. The proposed approach of model simplification is applied to fuel rods with two kinds of interfacial configurations between the fuel pellets and cladding. The differences among the generated models and the effects of interfacial bonding efficiency are discussed. The strategy of model simplification adopted in this work is to force the simplified beam model of spent fuel rods to possess the same compliance and failure characteristics under critical loads as those that result in the failure of detailed fuel rod models. It is envisioned that the simplified model would enable the assessment of fuel rod failure through an assembly-level analysis, without resorting to a refined model for an individual fuel rod. The effective material properties of the simplified beam model were successfully identified using the integrated optimization process. The feasibility of using the developed simplified beam models in dynamic impact simulations for a horizontal drop condition is examined, and discussions are provided.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.179-186
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• 2009

This paper presents a thermal creep test under internal pressure and post-test characterization performed on high burnup (68 MWd/kgU) ZIRLO. This research has been done by the CSN, ENRESA, and ENUSA in order to investigate the behavior of advanced cladding materials in contemporary PWRs at higher burnup under dry cask storage conditions. Also, to investigate the hydride reorientation, the cool-down of the samples after the test has been done in a coordinated manner with the internal pressure. The creep results obtained are consistent with the expected behavior from reference CWSR material, Zr-4. During the test, the material retained significant ductility: one specimen leaked during the test at an engineering strain of the tube section of 17%; remarkably, the crack closed due to de-pressurization. Although significant hydride reorientation occurred during the cool-down under pressure, no specimen failed during the cool-down.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2000.10a
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• pp.151-158
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• 2000

Cable dome that consists of three component such as cable, strut and fabric membrane has complex structural characteristics. Main structural system of cable dome is cable-strut tensegric system and fabric membrane element is conceived as cladding roof material. One of the important problem of cable dome is to investigate the structural response from external load effect such as snow and wind. When cable dome is subjected to load each structural component has various special structural characteristics. One is that geometrical nonlinearity should be considered because large deformation is occurred from their flexible characteristic. The other is that wrinkling occurs occasionally because cable and membrane elements can not transmit compressive forces. So this paper researches the physical structural response of cable dome structure and the structural behavior when failure occurred at a part of structure.

• Wind and Structures
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• v.14 no.4
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• pp.321-335
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• 2011

The vulnerability of roofing components of contemporary houses built in cyclonic regions of Australia is assessed for increasing wind speeds. The wind loads and the component strengths are treated as random variables with their probability distributions derived from available data, testing, structural analysis and experience. Design details including types of structural components of houses are obtained from surveying houses and analyzing engineering drawings. Wind load statistics on different areas of the roof are obtained by wind tunnel model studies and compared with Australian/New Zealand Standard, AS/NZS 1170.2. Reliability methods are used for calculating the vulnerability of roofing components independently over the roof. Cladding and batten fixings near the windward gable edge are found to experience larger negative pressures than prescribed in AS/NZS 1170.2, and are most vulnerable to failure.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2402-2409
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• 2020

Analytical criteria for the onset of fuel fragmentation and Burst Fission Gas Release in fuel rods with ballooned claddings are formulated. On that basis, the GRSW-A model integrated with a fuel behaviour code is updated. After modification, the updated code is successfully applied to simulation of the Halden LOCA test IFA-650.12. Specifically, the calculation with Burst Fission Gas Release during the test resulted in prediction of cladding failure, whereas it could not be predicted at the test planning, before new models were implemented. A good agreement of the current model with experimental data for transient Fission Gas Release in the tests IFA-650.12 and IFA-650.14 is shown, as well.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.728-744
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• 2024

The presence of hydrogen absorbed by zirconium-based cladding materials during reactor operation can trigger degradation mechanisms and endanger the rod integrity. Ensuring the durability of the rods in extended time-frames like dry storage requires anticipating hydrogen behavior using numerical modeling. In this context, the present paper describes a hydrogen post-processing tool for Falcon - HYPE, a PSI's in-house tool able to calculate hydrogen uptake, transport, thermochemistry, reorientation of hydrides and hydrogen-related failure criteria. The tool extracts all necessary data from a Falcon output file; therefore, it can be considered loosely coupled to Falcon. HYPE has been successfully validated against experimental data and applied to reactor operation and interim storage scenarios to present its capabilities.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.999-1008
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• 2023

Deformed fuel rods can cause a partial blockage of the flow area in a subchannel. Such flow blockage will influence the core coolant flow and further the core heat transfer during the reflooding phase and subsequent severe accidents. Nevertheless, most of the system analysis codes simulate the accident process based on the assumed flow blockage ratio, resulting in inconsistencies between simulated results and actual conditions. This paper aims to study the influence of flow blockage in severe accident scenario of the CAP1400 reactor. First, the flow blockage model of ISAA code is improved based on the FRTMB module. Then, the ISAA-FRTMB coupling system is adopted to model and calculate the QUENCH-LOCA-0 experiment. The correctness and validity of the flow blockage model are verified by comparing the peak cladding temperature. Finally, the DVI Line-SBLOCA accident is induced to analyze the influence of flow blockage on subsequent CAP1400 reactor core heat transfer and core degradation. From the results of the DVI Line-SBLOCA accident analysis, it can be concluded that the blockage ratio is in the range of 40%-60%, and the position of severe blockage is the same as that of cladding rupture. The blockage reduces the circulation area of the core coolant, which in turn impacts the heat exchange between the core and the coolant, leading to the early failure and collapse of some core assemblies and accelerating the core degradation process.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.