• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.241.1-241
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• 2002

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2771-2782
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• 2022

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-of-the-art semi-empirical approach (ANS 5.4-2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusion-decay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as interrelated phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1914-1928
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• 2022

To improve the heat transfer efficiency of the reactor fuel assembly, it is necessary to accurately calculate the two-phase flow boiling characteristics and the critical heat flux (CHF) in the fuel assembly. In this paper, a Eulerian two-fluid model combined with the extended wall boiling model was used to numerically simulate the 5 × 5 fuel rod bundle with spacer grids (four sets of mixing vane grids and four sets of simple support grids without mixing vanes). We calculated and analyzed 11 experimental conditions under different pressure, inlet temperature, and mass flux. After comparing the CHF and the location of departure from the nucleate boiling obtained by the numerical simulation with the experimental results, we confirmed the reliability of computational fluid dynamic analysis for the prediction of the CHF of the rod bundle and the boiling characteristics of the two-phase flow. Subsequently, we analyzed the influence of the spacer grid and mixing vanes on the void fraction, liquid temperature, and secondary flow distribution. The research in this article provides theoretical support for the design of fuel assemblies.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2887-2900
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• 2020

Cladding failure in a fuel rod during operation in a BWR is analyzed using a FALCON code-based model. Comparative calculation with a neighbouring, intact rod is presented, as well. A considerable 'hot spot' effect in cladding temperature is predicted with the SLICE-DO model due to a thermal barrier caused by the localized crud deposition. Particularly significant overheating is expected to occur on the affected side of the cladding of the failed rod, in agreement with signs of significant localized crud deposition as revealed by Post Irradiation Examination. Different possibilities (criteria) are checked, and Pellet-Cladding Mechanical Interaction (PCMI) is shown to be one of the plausible potential threats. It is shown that PCMI could lead to discernible concentrated inelastic deformation in the overheated part of the cladding. None of the specific mechanisms considered can be experimentally or analytically identified as an only cause of the rod failure. However, according to the current calculation, a possibility of cladding failure by PCMI cannot be excluded if the crud thickness exceeded 300 ㎛.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.135-147
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• 2020

A methodology for evaluation of mechanical and thermal effects of localized non-axisymmetric oxidation in zircaloy claddings on LWR fuel reliability is proposed. To this end, the basic capabilities of the FALCON fuel behaviour code are used. Examples of methodology application to adjustment of selected operational limits for modern BWR fuel rods, to capture effects of the excess local oxidation, are presented. Specifically, the limiting rod internal pressure for the onset of cladding lift-off is reduced, depending on initial excess oxidation spot sizes. Also, the power limits for Anticipated Operational Occurrences are adjusted, to preclude fuel melting and cladding failure due to PCMI and PCI-SCC in the affected fuel rods.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.130-136
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• 1998

In the PANBOX 2 system an excore detector module simulating the excore signal responses during a short term transient is implemented in order to simulate the reaction of the flux detector and control system upon rapid power changes as it occurs e. g, in rod drop events. This module has been verified in the past by comparison calculations with the PANBOX 1 system. This report describes additional PANBOX 2 validation calculations which have bee compared with experiment data measured at german plant KKG, cycle 1, for a rod drop event. In general, the PANBOX 2 results are in very good agreement with the KKG experiments. Therefore it is concluded that the excore detector model of PANBOX 2 is successfully validated.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2714-2719
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• 2022

The present paper proposes some new computational methods and results in the framework of flow computation through congested domains seen as porous media, as it can be found in the core of a Pressurized Water Reactor (PWR). The flow is thus mostly governed by the distribution of pressure losses, both through the porous structures, such as fuel assemblies, and in the thin fluid layers between them. The purpose of the present paper is to consider the question of the interaction of a flow and a rod bundle from an analytical point of view gathering all the contributions through a set of equations as simple and representative as possible. It aims at demonstrating a sound understanding of the relevant phenomena governing the flow establishment in the geometry of interest instead of relying mainly on a posteriori observations obtained both experimentally and numerically. Comparison with two set of experimental results showed good agreement. The model proposed being analytical it appears easily implementable for studies needing an expression of fluid forces in a rod array as for fuel assembly bowing issue. It would be interesting to test the reliability of the model on other geometry with different P/R ratios.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.31-39
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• 2013

Nuclear fuel can be damaged under unexpected circumstances in a nuclear reactor. Fuel rod failure can be occurred due to debris fretting or excessive hydriding or PCI (Pellet-to-clad Interaction) etc. It is important to identify the causes of such failed fuel rods for the safe operation of nuclear power plants. If a fuel rod failure occurs during the operation of a nuclear power plant, the coolant water is contaminated by leaked fission products, and in some case the power level of the plant may be lowered or the operation stopped. In addition, all spent fuels must be transferred to a dry storage. But failed fuel can not be transferred to a dry storage. Therefore, the purpose of this study is to develop a system which is capable of inspecting whether the spent fuel in the storage pool is failed or not. The sipping technology is to analyze the leakage of fission products in state of gas and liquid. The failed fuel inspection system with gamma analyzer has successfully demonstrated that the system is enough to find the failed fuel at Wolsong plant.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.28 no.5
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• pp.186-194
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• 2016

As a turbulence-enhancing device, a mixing vane installed at a spacer grid of the fuel assembly plays a role in improving the convective heat transfer by generating either swirl flow in the subchannels or cross flow between fuel rod gaps. Therefore, both configuration and arrangement pattern of a mixing vane are important factors that determine the performance of a mixing vane. In this study, in order to examine the flow distribution features inside $5{\times}5$ fuel assembly with swirl-type mixing vanes used in benchmark calculation of OECD/NEA, simulations were conducted with commercial CFD software ANSYS CFX R.14. Predicted results were compared to data measured from MATiS-H (Measurement and Analysis of Turbulent Mixing in Subchannels-Horizontal) test facility. In addition, the effect of swirl-type mixing vanes on flow pattern inside the fuel assembly was described.

• Proceedings of the Acoustical Society of Korea Conference
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• 1998.06e
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• pp.247-250
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• 1998

The scattering of plane ultrasonic waves by the nuclear fuel pin of liquid metal reactor in sodium is studied. According to the internal composition in the cladding tube, the fuel pin has three cross sections, i.e. helium gas plenum, sodium-filled section, and fuel insertion section. The scattering spectra for each section of the fuel pin are different. The circumnavigating ultrasonic waves of each section are analyzed by the resonance scattering method. The whispering gallery wave modes are generated in the sodium-filled plenum section and the fuel rod insertion section with a sodium-gap. The circumferential wave modes are propagated in the cladding tube of the helium gas plenum section. The annular gap between the cladding tube and metal uranium pellet rod affects the scattering spectra. The different propagation characteristics can be utilized for the nondestructive method of detecting the unbonded area and measuring the level of the sodium-filled section of the fuel pin.