• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.189-190
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• 2005

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.651-661
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• 2017

The Korea Institute of Nuclear Safety, as a representative organization of Korea, in February 2012 participated in an international Program to Assess the Reliability of Emerging Nondestructive Techniques initiated by the U.S. Nuclear Regulatory Commission. The goal of the Program to Assess the Reliability of Emerging Nondestructive Techniques is to investigate the performance of emerging and prospective novel nondestructive techniques to find flaws in nickel-alloy welds and base materials. In this article, Korean round-robin test results were evaluated with respect to the test blocks and various nondestructive examination techniques. The test blocks were prepared to simulate large-bore dissimilar metal welds, small-bore dissimilar metal welds, and bottom-mounted instrumentation penetration welds in nuclear power plants. Also, lessons learned from the Korean round-robin test were summarized and discussed.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.738-744
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• 2018

Fuel failure due to debris is a major cause of failure in pressurized water reactors. Fuel vendors have developed various filtering devices to reduce debris-induced failure and have evaluated filtering performance with their own test facilities and methods. Because of the different test facilities and methods, it is difficult to compare filtering performances objectively. This study presents an improved filtering test and an efficiency calculation method to fairly compare fuel-filtering efficiency regardless of the vendor's filtering features. To enhance the reliability of our evaluation, we established requirements for the test method and had a facility constructed according to the requirements. This article describes the debris specimens, the amount of debris, and the replicates for the proposed test method. A calculation method of comprehensive debris-filtering efficiency using a weighted mean is proposed. The test method was verified by repeated tests, and the tests were carried out using the PLUS7 and 17ACE7 test fuels to calculate the comprehensive debris-filtering efficiencies. The evaluation results revealed that the filtering performance of PLUS7 is better than that of 17ACE7. The proposed method can be used on any kind of debris-filtering devices and is appropriate for use as a standard.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1858-1868
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• 2021

Resisting an accidental impact of large commercial aircrafts is an important aspect of advanced nuclear power plant (NPP) design. Especially after the 9·11 event, some regulations were enacted, which required the design of NPPs should consider the accidental impact of large commercial aircrafts. Normal working of equipment is important for stopping reactor under an impact when an NPP is in operation. However, there is a lack of reliable analysis and research on the impact test of nuclear prototype equipment. Therefore, in order to study the response of the equipment under high acceleration impact, a centrifugal pump is selected as the research object to perform the impact test. A horizontal half-sinusoidal pulse wave was applied to the working pump. The test results show that the horizontal response of the motor and flange is greater compared to other parts, as well as the vertical response of the coupling. The stress response of the pump body support and motor support is high, hence these parts should be considered in the design of the pump. Finally, combined with the damage and stress evaluation results of the pump under different amplitudes, the ultimate impact acceleration that the pump can withstand is given.

• Nuclear Engineering and Technology
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• v.43 no.4
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• pp.355-360
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• 2011

This paper is a study on a nuclear water chiller. It presents a test-verified finite element model of a water chiller to be used at a Nuclear Power Plant. The test-verified model predicts natural frequencies within 5% for all major modes below 50 Hz. This model accurately represents the dynamic characteristics of the actual hardware and is qualified for its use in the final stress analysis for seismic verification.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.33 no.9
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• pp.643-657
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• 2009

This paper introduces the current nuclear experimental research activities in KAERI, the unique nuclear research institute in Korea, and the universities in Korea to solve and assess the issues which have been faced in the design of new reactors such as APR1400, SMART, GEN-IV reactors as well as fusion reactor. Also the experimental evaluations of safety for operating nuclear plants have been presented. The nuclear thermalhydraulic experiments performed in such organizations are classified the fundamental test, the separated effect test, and the integral effect test with ATLAS and SNUF. Introduction is deployed according to institutes. Finally, the future works and the direction of research voyage in the nuclear thermal-hydraulic field were suggested.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.10-16
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• 2000

Preliminary experiment was peformed to investigate the leaching characteristics of paraffin waste forms that had been recently generated in large quantities at domestic nuclear power plants. At first, waste simulants whose compositions were different in mixing ratio of paraffin to boric acid were prepared. Their compressive strengths were measured and ninety-day leaching test of specimen including cobalt was carried out according to ANSI/ANS-16.1 test procedure. Water immersion test was also conducted keeping pace with leaching test and the weight change and the compressive strength of specimen were observed after ninety days. The compressive strength of waste form exhibited 666 psi (4.53 MPa) in the case where mixing ratio of boric acid to paraffin was 78/22, which was adopted in concentrate waste drying system of domestic nuclear power plants. The leaching test resulted in about 50% of the cumulative fraction leached for boric acid and cobalt, respectively. The specific gravity of waste form was 0.87 ［g/g］whose value was less than that of water because the weight loss of about 39% occurred after the water immersion test of ninety days. It was also observed that the waste form which had undergone ninety-day water immersion test exhibited the compressive strength of 203 psi (1.38 MPa).

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.135.2-135
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• 1999

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.776-792
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• 2021

Conventional integral effect test facilities were constructed to enable the precise observation of thermal-hydraulic phenomena and reactor behaviors under postulated accident conditions to prove reactor safety. Although these facilities improved the understanding of thermal-hydraulic phenomena and reactor safety, applications of new technologies and their performance tests have been limited owing to the cost and large scale of the facilities. Various nuclear technologies converging 4th industrial revolution technologies such as artificial intelligence, drone, and 3D printing, are being developed to improve plant management strategies. Additionally, new conceptual passive safety systems are being developed to enhance reactor safety. A new integral effect test facility having a noticeable scaling ratio, i.e., the (UNIST reactor innovation loop (URI-LO), is designed and constructed to improve the technical quality of these technologies by performance and feasibility tests. In particular, the URI-LO, which is constructed using a transparent material, enables better visualization and provides physical insights on multidimensional phenomena inside the reactor system. The facility design based on three-level approach is qualitatively validated with preliminary analyses, and its functionality as a test facility is confirmed through a series of experiments. The design feature, design validation, functionality test, and future utilization of the URI-LO are introduced.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.142-142
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• 1999