• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2670-2677
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• 2023

In the nuclear safety studies, a new trend refers to the evaluation of uncertainties as a mandatory component of best-estimate safety analysis which is a modern and technically consistent approach being known as BEPU (Best Estimate Plus Uncertainty). The major objectives of this study consist in performing a study of uncertainties/sensitivities of the major analysis results for a generic CANDU 6 Nuclear Power Plant during Station Blackout (SBO) progression to understand and characterize the sources of uncertainties and their effects on the key figure-of-merits (FOMs) predictions in severe accidents (SA). The FOMs of interest are hydrogen mass generation and event timings such as the first fuel channel failure time, beginning of the core disassembly time, core collapse time and calandria vessel failure time. The outcomes of the study, will allow an improvement of capabilities and expertise to perform uncertainty and sensitivity analysis with severe accident codes for CANDU 6 Nuclear Power Plant.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2682-2695
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• 2021

This paper aims to evaluate the structural dynamic responses and damage/failure of the nuclear fuel reprocessing plant under the free drop impact of spent fuel cask (SFC) and fuel assembly (FA) during the on-site transportation. At the present Part I of this paper, the large-scale SFC model free drop test and the corresponding numerical simulations are performed. Firstly, a composite target which is composed of the protective structure, i.e., a thin RC plate (representing the inverted U-shaped slab in the loading shaft) and/or an autoclaved aerated concrete (AAC) blocks sacrificial layer, as well as a thick RC plate (representing the bottom slab in the loading shaft) is designed and fabricated. Then, based on the large dropping tower, the free drop test of large-scale SFC model with the mass of 3 t is carried out from the height of 7 m-11 m. It indicates that the bottom slab in the loading shaft could not resist the free drop impact of SFC. The composite protective structure can effectively reduce the damage and vibrations of the bottom slab, and the inverted U-shaped slab could relieve the damage of the AAC blocks layer dramatically. Furthermore, based on the finite element (FE) program LS-DYNA, the corresponding refined numerical simulations are performed. By comparing the experimental and numerical damage and vibration accelerations of the composite structures, the present adopted numerical algorithms, constitutive models and parameters are validated, which will be applied in the further assessment of drop impact effects of full-scale SFC and FA on prototype nuclear fuel reprocessing plant in the next Part II of this paper.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2009-2016
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• 2020

A nuclear power plant is a large complex system with tens of thousands of components. To ensure plant safety, the early and accurate diagnosis of abnormal situations is an important factor. To prevent misdiagnosis, operating procedures provide the anticipated symptoms of abnormal situations. While the more severe emergency situations total less than ten cases and can be diagnosed by dozens of key plant parameters, abnormal situations on the other hand include hundreds of cases and a multitude of parameters that should be considered for diagnosis. The tasks required of operators to select the appropriate operating procedure by monitoring large amounts of information within a limited amount of time can burden operators. This paper aims to develop a system that can, in a short time and with high accuracy, select the appropriate operating procedure and sub-procedure in an abnormal situation. Correspondingly, the proposed model has two levels of prediction to determine the procedure level and the detailed cause of an event. Simulations were conducted to evaluate the developed model, with results demonstrating high levels of performance. The model is expected to reduce the workload of operators in abnormal situations by providing the appropriate procedure to ultimately improve plant safety.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.25-30
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• 2013

The periodic inspection of piping and pressure vessels welds in nuclear power plant has to provide reliable result related to weld flaws, such as location, maximum amplitude response, ultrasonic length, height and finally the nature or flaw pattern. The founded flaw in ultrasonic inspection is accepted or rejected based on these data. Specially, the amplitude of flaw response is used as basic parameter for flaw sizing and it may cause some deviation in length sizing result. Currently the ultrasonic inspections in nuclear power plant components are performed by specific inspection procedure which describing inspection technique include inspection system, calibration methodology and flaw characterizing. To perform ultrasonic inspection during in-service inspection, reference gain should be established before starting ultrasonic inspection by the requirement of ASME code. This reference gain used as basic criteria to evaluate flaw sizing. Sometimes, a little difference in establishing reference gain between calibration and field condition can lead to deviation in flaw sizing. Due to this difference, the inspection result may cause flaw sizing error. Therefore, the objective of this study is to compare and evaluate the ultrasonic amplitude difference between air filled and water filled pipe in nuclear power plant. Additionally, the accuracy of flaw sizing is estimated by comparing both conditions.

• Journal of the Korean Society of Systems Engineering
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• v.11 no.1
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• pp.1-7
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• 2015

The purpose of this paper is to present the results of reengineering of the YGN 3&4 (Yonggwang Nuclear Power Plant, Units 3&4) SDS (Safety Depressurization System) retrofit design and to make recommendations for the improvement in design and design procedure implementing the Systems Engineering (SE) process. YGN 3&4 is a basic model for OPR1000 (the Korean standard 1000 MWe plant). The basic model, herein, represents the reference plant for the OPR1000 development. In the middle of the YGN 3&4 construction, the Korean Nuclear Regulatory Body requested a retrofit of this plant with a means to rapidly depressurize the plant in conformance with a severe accident mitigation requirement. For the reengineering of the SDS in YGN 3&4, V-model and functional and physical architectures have been developed. A SE decision making method has been used for the selection of SDS valves. Finally, recommendations have been made to improve OPR1000 design for the improved operation and enhanced safety.

• Fashion & Textile Research Journal
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• v.18 no.5
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• pp.668-676
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• 2016

The purposes of this study were to investigate the satisfaction of working uniform in Nuclear Power Plant and to suggest the improvement of dissatisfaction. 150 workers in control area of Nuclear Power Plant were participated in survey and 30 questionnaires were asked and subjective description was allowed. 65%/35% Poly/Rayon coverall type working uniform was investigated for this survey which had been wearing usually and basically in control area in Nuclear Power Plant. According to the results, respondents were most highly dissatisfied with the wearing convenience aspect of current coverall among any other aspects, like textile and management. In wearing convenience aspects, wrist and ankle opening band system, the design, number and placement of pocket, waist belt design, the width of sleeve and pants, and ADR opening system were dissatisfied and requested for improvement. In textiles aspects, weight, protection from radiation materials, prevention from static electricity, moisture absorption, ventilation and flexibility/elasticity were dissatisfied and requested for improvement. In management aspect, washing uniform and size variation were dissatisfied and requested for improvement. Therefore, for more comfortable human interfaced working uniform, wearing convenience system as well as textile and management system must be compensated and should be newly developed for improving worker comfort, mobility, and productivity.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.417-422
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• 1998

he Reactor Coolant Gas Vent System(RCGVS) design for Ulchin Nuclear Power Plant Units 3&4(UCN 3&4) has been improved from the Yonggwang Nuclear Power Plant Units 3&4(YGN 3&4) based on the evaluation results for depressurization capability tests performed at YGN 3&4. There has been a series of plant safety analyses for Natural Circulation Cooldown(NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown Phenomena in order to optimize the orifice size of UCN 3&4 RCGVS. Baesd on these analyses results, the RCGVS orifice size for UCN 3&4 has been reduced to 9/32 inch from the l1/32 inch for YGN 3&4. The depressurization capability tests, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3&4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation.

• Journal of the Korea institute for structural maintenance and inspection
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• v.6 no.1
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• pp.149-158
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• 2002

The carbonation process is affected by both the concrete material properties such as W/C ratio, types of cement and aggregates, admixture characteristics and the environmental factors such as $CO_2$ concentration, temperature, humidity. Based on results of preliminary study on carbonation, this study is to develop a carbonation prediction model by taking account of $CO_2$ concentration, temperature, humidity ad W/C ratio among major factor affecting the carbonation process. And to constitute a model formula which correspond to the mix design of the nuclear power plant, test coefficient that correspond to the design of the nuclear power plant is obtained based on the results of accelerated carbonation test. Also a field coefficient which is obtained based on results of the field examination is included to improve the conformity of the actual structures of nuclear power plant.

• Journal of Korean Society for Quality Management
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• v.47 no.3
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• pp.553-569
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• 2019

Purpose: The purpose of this study is to identify the impact of conflicts and cooperation on project performance in the nuclear power plant construction project, thus contributing to the successful performance of project participants and the achievement of project objectives. Methods: This study used a method of conducting a survey of participants in the nuclear power plant construction project and verifying the hypothesis statistically. Results: The results of this study are as follows; First, role conflicts and relationship conflicts have negative effects on collaboration. Second, role conflicts have a negative effects on project performance, but conflict in relation has no significant effects on project performance. Third, collaboration has a positive effects on project performance. Finally, the communication quality in the relationship between conflict and collaboration has been shown to have a Moderated effect. Conclusion: In order to achieve the goal in the nuclear power plant construction project, the level of collaboration should be improved based on conflict management among project participants.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.269-276
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• 2008

Operational license of WWER-440/213 units at Paks NPP, Hungary is limited to the design lifetime of 30 years. Prolongation by additional 20 years of the operational lifetime is feasible. Moreover, enhancement of the reactor thermal power by 8% will increase both the net power output and the competitiveness of the plant. Paks NPP is a pioneer considering the power up-rate and preparation of long-term operation of WWER-440/213 design. Systematic preparatory work for long-term operation of Paks NPP has been started in 2000. A regulatory framework and a comprehensive engineering practice have been developed. According to the authors view, creation of a gapless engineering system via consequent application of best practices, and feed-back of experiences together with proper consideration of WWER-440/V213 features are the decisive elements of ensuring the safety of long-term operation. That systematic engineering approach is in the focus of recent paper. Key elements of justification and measures for ensuring the safety of long-term operation of Paks NPP WWER-440/213 units are identified and discussed. These are the assessment of plant condition and review of adequacy of ageing management programmes, also the review, validation and reconstitution of time limited ageing analyses as core tasks of licence renewal.