• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4798-4808
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• 2022

The porous corrosion deposits (known as CRUD) adhered to the cladding have an important effect on the heat transfer from fuel rods to coolant in PWRs. The vapor film is the main constituent in the two-phase film boiling model. This paper presents a vapor film thickness correlation, associated with CRUD porosity, CRUD chimney density, CRUD particle size, CRUD thickness and heat flux. The dependences of the vapor film thickness on the various influential factors can be intuitively reflected from this vapor film thickness correlation. The temperature, pressure, and boric acid concentration distributions in CRUD can be well predicted using the two-phase film boiling model coupled with the vapor film thickness correlation. It suggests that the vapor thickness correlation can estimate the vapor film thickness more conveniently than the previously reported vapor thickness calculation methods.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4465-4471
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• 2023

Hands-on education and training is a key part of fixing and developing technology knowledge and is an inherent part of many engineering and scientific curricula. However, access to large complex training facilities, such as nuclear reactor, could be limited by various factors, such as unavailability of those facilities in the region, high traveling costs or harmonization of the schedules of hands-on E&T with theoretical lectures and with the operational schedule of the facility. To handle the issue, several success stories have been reached with the introduction of the Internet Reactor Labs (IRL). The Internet Reactor Labs can strongly contribute to accessibility of training at research reactors and can contribute to improvements in their utilization. The paper describes the development of the Internet Reactor Lab at the VR-1 reactor of the Czech Technical University in Prague. Contrary to single-purpose IRLs, it presents various modalities of online teaching and training in experimental reactor physics and reactor operation in general as well as outreach activities that have been developed in recent years.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.219-225
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• 2015

In this paper, a three-dimensional model for the dynamic analysis of a flywheel based on the finite element method is presented. The static structure analysis for the model provides stress and strain distribution cloud charts. The modal analysis provides the basis of dynamic analysis due to its ability to obtain the natural frequencies and the vibration-made vectors of the first 10 orders. The results show the main faults are attrition and cracks, while also indicating the locations and patterns of faults. The harmonic response simulation was performed to gain the vibration response of the flywheel under operation. In this paper, we present a Hilbert-Huang transform (HHT) algorithm for flywheel vibration analysis. The simulation indicated that the proposed flywheel vibration signal analysis method performs well, which means that the method can lay the foundation for the detection and diagnosis in a reactor main coolant pump.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.64-64
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• 1998

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.637-646
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• 2021

The free-piston Stirling engine (FPSE) has been widely used in aerospace owing to its advantages of high efficiency, high reliability, and self-starting ability. In this paper, a 20-kW FPSE is proposed by analyzing the requirements of space nuclear power reactor. A code was developed based on an improved simple analysis method to evaluate the performance of the proposed FPSE. The code is benchmarked with experimental data, and the maximum relative error of the output power is 17.1%. Numerical results show that the output power is 21 kW, which satisfies the design requirements. The results show that: a) reducing the pressure shell's thickness can improve the output power significantly; b) the system efficiency increases with the wire porosity, while the growth of system efficiency decreases when the porosity is higher than 80%, and system efficiency exhibits a linear relationship with the temperatures of the cold and hot sides; c) the system efficiency increases with the compression ratio; the compression ratio increases by 16.7% while the system efficiency increases by 42%. This study can provide valuable theoretical support for the design and analysis of FPSEs for space nuclear power reactors.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.293-300
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• 2022

During the decommissioning of the nuclear facilities, the radioactive gases in pressure vessels may leak due to the demolition operations. The decommissioning site has large space, slow air circulation, and many large nuclear facilities, which increase the difficulty of workers' inhalation exposure assessment. In order to dynamically evaluate the activity distribution of radionuclides and the committed effective dose from inhalation in nuclear decommissioning environment, an inhalation exposure assessment method based on the modified eddy-diffusion model and the inhaled dose conversion factor is proposed in this paper. The method takes into account the influence of building, facilities, exhaust ducts, etc. on the distribution of radioactive gases, and can evaluate the influence of radioactive gases diffusion on workers during the decommissioning of nuclear facilities.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2803-2815
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• 2021

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as ²⁴¹Am, ¹³⁷Ba, ²⁴⁴Cm, ²³⁸Pu, and ⁹⁰Y.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.11-11
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• 1999

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.410-418
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• 2024

Study of flow behavior within rod bundles has been an active topic. Surface modification technologies are important parts of the design of the fourth generation reactor, which can increase the strength of the secondary flow within the rod bundle lattices. Quasi-periodic large-scale vortex structure (QLVS) is introduced by arranging micro ribs on the surface of rod bundles, which enhanced the scale of the secondary flow between the rod bundle lattices. Using computational fluid dynamics (CFD) and water experiments, the flow field distribution and drag coefficient of the rod-bundle lattices are studied. The secondary flow between the micro-ribbed rod-bundle lattice is significantly enhanced compared to the standard rod-bundle lattice. The numerical simulation results agree well with the experimental results.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.204-205
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• 2018