• Nuclear Engineering and Technology
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• v.41 no.9
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• pp.1157-1170
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• 2009

A variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.676-682
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• 2019

After the Fukushima Daiichi accident, there has been an increasing preference for passive safety features in the nuclear power industry. Some passive safety systems require limited active components to trigger subsequent passive operation. Under very serious accident conditions, passive safety features could be rendered inoperable or damaged. This study evaluates (i) the performance and effectiveness of the passive safety features of iPOWER (innovative Power Reactor), and (ii) whether a severe accident condition could be reached if the passive safety systems are damaged, namely the case of heat exchanger tube rupture. Analysis results show that the reactor coolant system remains in the hot shutdown condition without operator actions or electricity for over 72 h when the passive auxiliary feedwater systems (PAFSs) are operable without damage. However, heat exchanger tube rupture in the PAFS leads to core damage after about 18 h. Such results demonstrate that, to enhance the safety of iPOWER, maintaining the integrity of the PAFS is critical, and therefore additional protections for PAFS are necessary. To improve the reliability of iPOWER, additional battery sets are necessary for the passive safety systems using limited active components for accident mitigation under such extreme circumstances.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.995-1004
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• 2009

The Advanced Power Reactor 1400 (APR1400) is an evolutionary advanced light water reactor (ALWR) based on the Optimized Power Reactor 1000 (OPR1000), which is in operation in Korea. The APR1400 incorporates a variety of engineering improvements and operational experience to enhance safety, economics, and reliability. The advanced design features and improvements of the APR1400 design include a pilot operated safety relief valve (POSRV), a four-train safety injection system with direct vessel injection (DVI), a fluidic device (FD) in the safety injection tank, an in-containment refueling water storage tank (IRWST), an external reactor vessel cooling system, and an integrated head assembly (IHA). Development of the APR1400 started in 1992 and continued for ten years. The APR1400 design received design certification from the Korean nuclear regulatory body in May of2002. Currently, two construction projects for the APR1400 are in progress in Korea.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.602-607
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• 1995

This article presents safety review experience of microprocessor-based Interposing Logic System(ILS) of Engineering Safety Feature Actuation System(ESFAS). The ILS is the first application of computerized logic design to safety system in Korean nuclear power plants without verification of the system reliability by proven technology concept. As a result of evaluation for the ILS, Korea Institute of Nuclear Safety(KINS) concluded that the microprocessor-based ILS is not acceptable in some features detailed enough to defend against software common mode failures(CMF). Therefore, we required licensee to install hardwired interlock signal configuration and a Hardwired Backup Panel to control safety-related equipment. We believe that the microprocessor-based ILS with the hardwired backup panel and inter-connection of interlock signal by hardwired configuration will improve the plant safety.

• Journal of the Korean Society of Safety
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• v.26 no.4
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• pp.7-13
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• 2011

A study on reliability estimation of sequential-ordered multiple failure modes, which are sequentially ordered between failure modes in a considering system, was performed. Especially, an approach to estimate the probabilities of failure modes has been proposed under an assumption that failure modes are mutually exclusive and sequentially ordered by only a critical variable. A feasibility of the proposed approach were studied by a practical example, which is a reliability estimation of passive safety systems for a probabilistic safety assessment(PSA) of a very high temperature reactor(VHTR) that is under development as a future nuclear system with enhanced safety features. It is difficult to define a robust failure state of this nuclear system because of its enhanced radiation release characteristics, so the new approach is a useful concept to estimate not only its safety but also a PSA. A feasibility study applied two failure modes(e.g., small and large release of radioactive materials) with considering the integrated behavior of this nuclear system. It is expected that the multiple release states for a practical estimation can be easily extended to the aforementioned example. It was found out that the proposed approach was a useful technique to cover the unfavorable features of this nuclear system as to performing a VHTR PSA.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2096-2106
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• 2023

Rotating machinery is widely applied in important equipment of nuclear power plants (NPPs), such as pumps and valves. The research on intelligent fault diagnosis of rotating machinery is crucial to ensure the safe operation of related equipment in NPPs. However, in practical applications, data-driven fault diagnosis faces the problem of small and imbalanced samples, resulting in low model training efficiency and poor generalization performance. Therefore, a deep convolutional conditional generative adversarial network (DCCGAN) is constructed to mitigate the impact of imbalanced samples on fault diagnosis. First, a conditional generative adversarial model is designed based on convolutional neural networks to effectively augment imbalanced samples. The original sample features can be effectively extracted by the model based on conditional generative adversarial strategy and appropriate number of filters. In addition, high-quality generated samples are ensured through the visualization of model training process and samples features. Then, a deep convolutional neural network (DCNN) is designed to extract features of mixed samples and implement intelligent fault diagnosis. Finally, based on multi-fault experimental data of motor and bearing, the performance of DCCGAN model for data augmentation and intelligent fault diagnosis is verified. The proposed method effectively alleviates the problem of imbalanced samples, and shows its application value in intelligent fault diagnosis of actual NPPs.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.368-372
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• 2016

Background: Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. Materials and Methods: It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. Results and Discussion: An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. Conclusion: The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

• Proceedings of the Safety Management and Science Conference
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• 2011.11a
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• pp.85-91
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• 2011

Safety in the nuclear facility has been a growing interest due to recent recurrences of the fatal accidents such as Fukushima accident and Chernobyl accident. It is not easy to determine the extent to what technical requirements of nuclear facility such as nuclear power plant are be likely applicable to the laboratory for use of nuclear material. All of workers in nuclear shall be recognized for the generic features of safety according to the related laws. This study surveys a status of safety control laws to enhance safety in laboratory for use of nuclear material.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2246-2255
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• 2023

Evacuation time estimation (ETE) is crucial for the effective implementation of resident protection measures as well as planning, owing to its applicability to nuclear emergencies. However, as confirmed in the Fukushima case, the ETE performed by nuclear operators does not reflect behavioral features, exposing thus, gaps that are likely to appear in real-world situations. Existing research methods including surveys and interviews have limitations in extracting highly feasible behavioral features. To overcome these limitations, we propose a VR-based immersive experiment system. The VR system realistically simulates nuclear emergencies by structuring existing disasters and human decision processes in response to the disasters. Evacuation behavioral features were quantitatively extracted through the proposed experiment system, and this system was systematically verified by statistical analysis and a comparative study of experimental results based on previous research. In addition, as part of future work, an application method that can simulate multi-level evacuation dynamics was proposed. The proposed experiment system is significant in presenting an innovative methodology for quantitatively extracting human behavioral features that have not been comprehensively studied in evacuation. It is expected that more realistic evacuation behavioral features can be collected through additional experiments and studies of various evacuation factors in the future.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.862-868
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• 1998

Korean Next Generation Reactor(KNGR) is a standardized evolutionary Advanced Light Water Reactor design under development Korea Power Company(KEPCO). It incorporates design enhncements such as active and passive advanced design features(ADFs) to increase the plant safety. A Preliminary level 1 Probabilistic Safety Assessment(PSA) has been performed for KNGR to examine the effect of these safety features. The preliminary PSA result shows that it meets the KNGR safety goal on core damage frequency(CDF). The result of this safety assessment shows that the four-train safety systems, and the ADFs such as Passive Secondary Cooling System (PSCS) contributes greatly to the reduction the CDF. Furthermore, several design changes are made or proposed for detailed review based on the PSA insights.