• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.2
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• pp.87-96
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• 2012

With a view to determine a safe speed the limit of a manipulator arm, several experiments was performed with a multi-jointed manipulator in maintenance and decommissioning tasks of nuclear facilities. Under the simulated emergency conditions, which were generated with random combinations of manipulator arm speed, failure probability and failure type, response characteristics of human operators to various malfunctions of a manipulator arm were measured in terms of reaction time, number of false alarm, and number of misses. This paper demonstrated that failure type, manipulator axes and manipulator arm speed has significant effects on human reaction time. As a whole the reaction time was slightly increased with manipulator arm speed, which is showed somewhat different pattern due to failure type. The reaction time to an axis acting on a workpiece directly, which could flex and extend, was fastest and much more its standard deviation was small. Various factors which may affect safe speed were also described.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.1015-1021
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• 2016

Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.2
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• pp.201-210
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• 2024

Because most spent nuclear fuel storage casks have been designed for low burnup fuel, a safety-significant high burnup dry storage cask must be developed for nuclear facilities in Korea to store the increasing high burnup and damaged fuels. More than 20% of fuels generated by PWRs comprise high burnup fuels. This study conducted a structural safety evaluation of the preliminary designs for a high burnup storage cask with 21 spent nuclear fuels and evaluated feasible loading conditions under normal, off-normal, and accident conditions. Two types of metal and concrete storage casks were used in the evaluation. Structural integrity was assessed by comparing load combinations and stress intensity limits under each condition. Evaluation results showed that the storage cask had secured structural integrity as it satisfied the stress intensity limit under normal, off-normal, and accident conditions. These results can be used as baseline data for the detailed design of high burnup storage casks.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.429-447
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• 2020

The considered simulation tasks are based on an electrometallurgical process development strategy and associated telemanipulator simulation systems are proposed with various scales of experimental facilities. Fundamentally, target facilities are assumed to be operated only by remote handling systems because the considered process is operated in hazardous environments. Futhermore, the feasibility at various scales should be experimentally verified with gradual increase in throughput. In this regard, bench, engineering, and pilot-scale simulation systems are important early-stage tools for assessing the practical operability of the target process with the material handling systems. Such simulation systems are highly customized for applications and are a precursor to larger pilot and demonstration-scale plants. This paper introduced and classified the developed simulator systems for this approach at various scales using remote handling systems which were assembled inside a virtual target facility, and the manmachine interface was included for a more realistic operation of the simulator. The results obtained for each simulator show the feasibility and requirement for improvement of the systems for the considered test issues with respect to the operation and maintenance of the process.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2008.11a
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• pp.27-28
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• 2008

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.446-447
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.05a
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• pp.163-164
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• 2012

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.661-669
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• 1995

A mobile robot, named KAEROT, has been developed for inspection and maintenance operations in nuclear facilities. The main feature of locomotion system is the planetary wheel assembly with small wheels. This mechanism has been designed to be able to go over the stairs and obstacles with stability. This paper presents the inverse kinematic solution that is to be operated by remote control. The automatic stair climbing algorithm is also proposed. The. proposed algorithms generates the moving pathes of small wheels and calculates the angular velocity of 3 actuation wheels. The results of simulations and experiments are given for KAEROT peformed on the irregular stairs in laboratory. It is shown that the proposed algorithm provides the lower inclination angle of the robot body and increases its stability during navigation.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3229-3235
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• 2021

The key process used in nuclear industries for the management of radiotoxicity associated with spent fuel in a closed fuel cycle is solvent extraction. An understanding of hydrodynamics and mass transfer is of primary importance for the design of mass transfer equipment used in solvent extraction processes. Understanding the interfacial phenomenon and the associated hydrodynamics of the liquid drops is essential for model-based design of mass transfer devices. In this work, the phenomenon of drop formation at the tip of a nozzle submerged in quiescent immiscible liquid phase is revisited. Previously reported force balance based models and empirical correlations are analyzed. Experiments are carried out to capture the process of drop formation using high-speed imaging technique. The images are digitally processed to measure the average drop diameter. A correlation based on the force balance model is proposed to estimate drop diameter and jet length. The average drop diameter obtained from the proposed model is in good agreement with experimental data with an average error of 6.3%. The developed model is applicable in both the necking as well as jetting regime and is validated for liquid-liquid systems having low, moderate and high interfacial tension.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1999.11a
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• pp.119-124
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• 1999

A selection methodology of liquid levelmeters, especially, level sensors in non-nuclear category, to be installed in nuclear energy facilities is developed using linguistic fuzzy approaches such as fully-linguistic and semi-linguistic methods. Depending on defuzzification techniques, the linguistic fuzzy methodology leads to either linguistic (exactly, fully-linguistic) or cardinal (i.e., semi-linguistic) evaluation. For the linguistic method, for each alternative, fuzzy preference index is converted to linguistic utility value by means of a similarity measure determining the degree of similarity between fuzzy index and linguistic ratings. For the cardinal method, the index is translated to cardinal overall utility value. According to these values, alternatives of interest are linguistically or numerically evaluated and a suitable alternative can be selected. Under given selection criteria, the suitable selections out of some liquid levelmeters for nuclear facilities are dealt with using the linguistic fuzzy methodology proposed. Then, linguistic fuzzy evaluation results are compared with qualitative result available in the literature. It is found that as to a suitable option the linguistic fuzzy selection is in agreement with the qualitative selection. Additionally, the comparative study shows that the fully-linguistic method using adequate scale system facilitates linguistic interpretation regarding evaluation results.