• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.121-122
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.166-172
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• 1996

The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for Westinghouse type PWR plants as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increase approximately by 1.2%. The resulting DNBR penalties can be covered within the current DNBR margin. Accident analyses are also investigated and the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation.

• Nuclear Engineering and Technology
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• v.56 no.10
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• pp.4062-4067
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• 2024

In this article, we propose and explore a novel step in the digitization of the mapping of the spent fuel pool of nuclear power plants, in which the audio signal from the operator's microphone is used to obtain the identification codes of those components that are in each of the cells of the pool. In this way, we have not only an acquisition system but also a verification system that can be used in combination with the outcome of the analysis of the video signal. We developed an algorithm that uses at its core one of the latest models of multilingual Automatic Speech Recognition to transcribe audio signal, and with a post-processing of the timed transcriptions we build the identification code of fuel heads and other components. Results show a very high accuracy in audios from real recording of Spanish nuclear facilities, and the methodology proposed is easily extensible to other nuclear facilities in the world.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1679-1686
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• 2024

Graphite material plays an important role in nuclear reactors especially the high-temperature gas-cooled reactors (HTGRs) by its outstanding comprehensive nuclear properties. The structural integrity of graphite pebble fuel elements is the first barrier to core safety under any circumstances. The correct knowledge of the stiffness coefficient of the graphite pebble fuel element inside the reactor's core is significant to ensure the valid design and inherent safety. In this research, a vertical extrusion device was set up to measure the stiffness coefficient of the graphite pebble fuel element by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. The stiffness coefficient equations of graphite pebble fuel elements at different temperatures are given (in a helium atmosphere). The result first provides the data on the high-temperature stiffness coefficient of pebbles in helium gas. The result will be helpful for the engineering safety analysis of pebble-bed nuclear reactors.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.29-35
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• 2013

The United States Department of Energy (US DOE) is conducting research and development (R&D) activities under the Used Fuel Disposition Campaign (UFDC) to support storage, transportation, and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. R&D activities are ongoing at nine national laboratories, and are divided into storage, transportation and disposal. Storage R&D focuses on closing technical gaps related to extended storage of UNF. Transportation R&D focuses on ensuring transportability of UNF following extended storage, and addressing data gaps regarding nuclear fuel integrity, retrievability, and demonstration of subcriticality. Disposal R&D focuses on identifying geologic disposal options and addressing technical challenges for generic disposal concepts in mined repositories in salt, clay/shale, and granitic rocks, and deep borehole disposal. UFDC R&D goals include increasing confidence in the robustness of generic disposal concepts, reducing generic sources of uncertainty that may impact the viability of disposal concepts, and developing science and engineering tools to support the selection, characterization, and licensing of a repository. The US DOE has also initiated activities in the Nuclear Fuel Storage and Transportation (NFST) Planning Project to facilitate the development of an interim storage facility and to support transportation infrastructure in the near term.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2792-2800
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• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.470-475
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• 1998

The marginal nuclear criticality analysis for the high density spent fuel storage at a PWR plant was carried out by using the HELIOS and CASMO-3 codes. More than 20 % of the calculated reactivity saving effect is observed in this analysis. This mainly comes from the adoption of some important fission products and B-10 in the criticality analysis. By taking burnup and boron credits, the high capacity of the spent fuel storage rack can be more fully utilized, reducing the space of storage. Larger storage for a given inventory of spent fuel should result in remarkable cost savings and mort importantly reduce the risks to the public and occupational workers.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.1982-1990
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• 2022

The purpose of this study was to verify the possibility of fuel rod pattern analysis in a fresh fuel assembly using the Yonsei single-photon emission computed tomography (YSECT) system. The YSECT system consisted of three main parts: four trapezoidal-shaped bismuth germanate scintillator-based 64-channel detectors, a semiconductor-based multi-channel data acquisition system, and a rotary stage. In order to assess the performance of the prototype YSECT, tomographic images were obtained for three representative fuel rod patterns in the 6 × 6 array using two representative image-reconstruction algorithms. The fuel-rod patterns were then assessed using an in-house fuel rod pattern analysis algorithm. In the experimental results, the single-directional projection images for those three fuel-rod patterns well discriminated each fuel-rod location, showing a Gaussian-peak-shaped projection for a single 10 mm-diameter fuel rod with 12.1 mm full-width at half maximum. Finally, we successfully verified the possibility of the fuel rod pattern analysis for all three patterns of fresh fuel rods with the tomographic images obtained by the rotational YSECT system.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3860-3865
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• 2023

A metallic microcell UO₂ pellet has a microstructure where a metal wall is connected to overcome the low thermal conductivity of the UO₂ fuel pellet. It has been verified that metallic microcell fuel pellets provide an impressive reduction of the fuel centerline temperature through a Halden irradiation test. However, it is difficult to predict the effective thermal conductivity of these pellets and researchers have had to rely on measurement and use of the finite element method. In this study, we designed a unit microcell model using a thermal resistance circuit to calculate the effective thermal conductivity on the basis of the microstructure characteristics by using the aspect ratio and compared the results with those of reported metallic UO₂ microcell pellets. In particular, using the thermal conductivity calculated by our model, the fuel centerline temperature of Cr microcell pellets on the 5th day of the Halden irradiation test was predicted within 6% error from the measured value.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3260-3267
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• 2023

Zr-Nb alloy have been widely used as fuel rods in nuclear power plants. However, from the Fukushima nuclear accident, the weakness of the rod was revealed under harsh conditions, and research on the safety of these types of rods was conducted after the disaster. The method of depositing chromium onto the existing Zr-Nb alloy fuel rods is being considered as a means by which to compensate for the weakness of Zr-Nb alloy rods because chromium is strong against oxidation at high temperatures and has high strength. In order to secure these advantages, it is important to maintain the Cr thickness of the rods and properly inspect the rods before and during their use in power generation. Eddy current testing is a typical means of evaluating the thickness of thin metals and detecting surface defects. Depending on the size and shape of the inspected object, various eddy current sensors can be applied. In particular, because pancake sensors can be manufactured in very small sizes, they can be used for inspections even in narrow spaces, such as a nuclear fuel assembly. In this study, an eddy current technique was developed to confirm the feasibility of Cr coating thickness evaluations. After determining the design parameters of the pancake sensor by means of a FEM simulation, a FPCB pancake sensor was manufactured and the optimal frequency was selected by measuring minute changes in the Cr-coating thickness using the developed sensor.