• Nuclear Engineering and Technology
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• v.37 no.3
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• pp.207-220
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• 2005

Nuclear power plants (NPP) are complex socio-technological systems that rely on the success of both hardware and human components. Empirical studies of plant operating experience show that human errors are important contributors to accidents and incidents, and that organizational factors play an important role in creating contexts for human errors. Current probabilistic safety assessments (PSA) do not explicitly model the systematic contribution of organizational factors to safety. As some countries, like the United States, are moving towards increased use of risk information in the regulation and operation of nuclear facilities, PSA quality has been identified as an area for improvement. The modeling of human errors, and underlying organizational weaknesses at the root of these errors, are important sources of uncertainty in existing PSAs and areas of on-going research. This paper presents a review of research into the following questions: Is there evidence that organizational factors are important to NPP safety? How do organizations contribute to safety in NPP operations? And how can these organizational contributions be captured more explicitly in PSA? We present a few past incidents that illustrate the potential safety implications of organizational deficiencies, some mechanisms by which organizational factors contribute to NPP risk, and some of the methods proposed in the literature for performing root-cause analyses and including organizational factors in PSA.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.896-904
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• 2017

Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.275-283
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• 2019

Rare isotope beam facilities require shielding data in early stage of their design. There is much less shielding data on neutrons from the reactions between heavy ion beams and matter than the data on neutrons produced by protons. The purpose of the present work is to produce and thus increase the amount of shielding data on neutrons generated by high-energy heavy ion beams based on the RAON in-flight fragment facility. Calculations were performed with the computational Monte Carlo codes PHITS and MCNPX. The secondary neutron source terms were evaluated at 550 MeV/u for Ca, Kr, and Sn and at 400 MeV/u for U ions on a graphite target. Source terms and attenuation lengths were obtained by fitting the ambient dose equivalent inside an ordinary concrete shield.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.493-500
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• 2022

The objective of this study was to derive derived concentration guideline levels (DCGLs) reflecting the site-specific characteristics of KRR-1&2. A total of 7 nuclides (H-3, C-14, Co-60, Sr-90, Cs-137, Eu-152, and Eu-154) were selected for DCGLs derivation. Radiation dose at the sites was evaluated with RESRAD-ONSITE program. The dose contribution due to direct external exposure was the highest during the entire evaluation period. Ingestion had the second effect. The DCGLs of Co-60 was derived to be 0.051 Bq/g, and DCGLs of Cs-137 was 0.193 Bq/g. The DCGLs of H-3 showed the highest value of 129 Bq/g. The ratio of DCGLs derived by applying site-specific values and default values ranged from 0.27 to 19.6. For six nuclides excluding H-3, KRR-1&2 sites and the overseas NPP sites showed similar DCGLs. H-3 showed large differences in DCGLs from this study and overseas NPPs. The large difference resulted from input parameter values applied to the sites. In conclusion, it is critical to apply site-specific parameter values reflecting the site characteristics to derive DCGLs for decommissioned site clearance. The result of this study can be used as a reference for nuclide selection and DCGLs derivation reflecting the site characteristics when decommissioning nuclear facilities, including nuclear power plants in Korea.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.315-327
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• 2023

In South Korea, the exporters of items related to nuclear power generation are diversified. Consequently, there is a risk of illegitimate export by companies failing to recognize the export control system because the awareness about this system for the strategic items among the subcontractors of nuclear power facilities is limited. To prevent illegitimate export of the strategic items, it is necessary to conduct outreach activities regarding the export control system for the related companies. Additionally, the exporters and export license examiners should consider whether an export target is on the Denial List, who may divert the strategic items to weapons of mass destruction. Therefore, the Korea Institute of Nuclear Nonproliferation and Control developed two systems for controlling illegitimate export of the Trigger List items. The first system, Nuclear Industry Information Collection and Analysis System, can gather information about the key nuclear industries in Korea and analyze the dealing of strategic items. The second system, Denied Persons Information Gathering System, can regularly gather information about the denied persons and provide the updated data to the exporters and regulatory examiners. These two systems can be used for outreach activities and export license examination to prevent illegitimate export of the strategic items.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.92-99
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• 2024

To minimize the cumulative radiation dose, various path-finding approaches for single agent have been proposed. However, for emergence situations such as nuclear power plant accident, these methods cannot be effectively utilized for evacuating a large number of workers because no multi-agent method is valid to conduct the mission. In this study, a novel algorithm for solving the multi-agent path-finding problem is proposed using the conflict-based search approach and the objective function redefined in terms of the cumulative radiation dose. The proposed method can find multi paths that all agents arrive at the destinations with reducing the overall radiation dose. To verify the proposed method, three problems were defined. In the single-agent problem, the objective function proposed in this study reduces the cumulative dose by 82% compared with that of the shortest distance algorithm in experiment environment of this study. It was also verified in the two multi-agent problems that multi paths with minimized the overall radiation dose, in which all agents can reach the destination without collision, can be found. The method proposed in this study will contribute to establishing evacuation plans for improving the safety of workers in radiation-related facilities.

• 제어로봇시스템학회:학술대회논문집
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• 1993.10a
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• pp.821-824
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• 1993

In hazardous conditions, where entry of human operators is restricted, such as high radiation regions in nuclear facilities, a lot of remote inspections and remote handling tasks must be performed. In this study, the stereo imaging system has been developed and the remote handling technique, has been studied to enhance the efficiency of tele-operation. The nozzle dam handling robot is one of the most important robots applied to nuclear facility. The robot will be equipped with the developed stereo imaging system. The stereo imaging system will outstandingly improve the tele installing/removal tasks for nozzle dam.

• Proceedings of the KWS Conference
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• 2004.05a
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• pp.296-298
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• 2004

Recently, various fields are building developed and applied system environment using the technology of the web-based virtual reality. This paper describes about the building of Web-based learning environment that supports efficient educational effectiveness throughout the effective 2-D and 3-D visual interface and the operation of main facilities and the navigation of it. The objective of this environmental building is to provide proper knowledge and technology of a nuclear power plant to those students (examine. and trainee), and help them gain effective experiences throughout such activities.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.939-940
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• 2005

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.92-100
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• 2015

For the purpose of improving the efficiency of the radiation impact assessment of dry interim storage facilities for the spent nuclear fuel of pressurized water reactors (PWRs), radiation impact assessment was performed after the application of sensitivity assessment according to the radiation source term designation method, development of a 2-step calculation technique, and cooling time credit. The present study successively designated radiation source terms in accordance with the cask arrangement order in the shielding building, assessed sensitivity, which affects direct dose, and confirmed that the radiation dosage of the external walls of the shielding building was dominantly affected by the two columns closest to the internal walls. In addition, in the case in which shielding buildings were introduced into storage facilities, the present study established and assessed the 2-step calculation technique, which can reduce the immense computational analysis time. Consequently, results similar to those from existing calculations were derived in approximately half the analysis time. Finally, when radiation source terms were established by adding the storage period of the storage casks successively stored in the storage facilities and the cooling period of the spent nuclear fuel, the radiation dose of the external walls of the buildings was confirmed to be approximately 40% lower than the calculated values; the cooling period was established as being identical. The present study was conducted to improve the efficiency of the Monte Carlo shielding analysis method for radiation impact assessment of interim storage facilities. If reliability is improved through the assessment of more diverse cases, the results of the present study can be used for the design of storage facilities and the establishment of site boundary standards.