• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2847-2851
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• 2020

A simple water level measurement method based on wavelength division multiplexing (WDM) is proposed and demonstrated. The measurement principle is based on the change of Fresnel reflection occurring at the end facet of the optical fiber tip (OFT). To increase the spatial resolution of water level sensing, a broadband light source (BLS) and an arrayed waveguide grating (AWG) are employed. The OFTs are multiplexed with the dedicated wavelength channels of AWG. By measuring all of the reflection powers reflected at the OFTs with a proposed on-site reflectometer, the water level can be monitored continuously for a fast emergency response. Moreover, it can be implemented easily with the commercially available optical components and devices with the simple configuration.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3628-3638
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• 2023

This paper uses 135 Licensed Operator Event Reports (LOER) from Chinese nuclear plants to analyze how safety culture affects unsafe behaviors in nuclear power plants. On the basis of a modified human factors analysis and classification system (HFACS) framework, structural equation model (SEM) is used to explore the relationship between latent variables at various levels. Correlation tests such as chi-square test are used to analyze the path from safety culture to unsafe behaviors. The role of latent error is clarified. The results show that the ratio of latent errors to active errors is 3.4:1. The key path linking safety culture weaknesses to unsafe behaviors is Organizational Processes → Inadequate Supervision → Physical/Technical Environment → Skill-based Errors. The most influential factors on the latent variables at each level in the HFACS framework are Organizational Processes, Inadequate Supervision, Physical Environment, and Skill-based Errors.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1758-1764
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• 2019

As the safety level of nuclear power plants (NPPs) relates to the safety of individuals, society, and the environment, it is important to establish NPP safety goals. In Korea, two quantitative health objectives and one large release frequency (LRF) criterion were formally set as quantitative safety goals for NPPs by the Nuclear Safety and Security Commission in 2016. The risks of prompt and cancer fatalities from NPPs should be less than 0.1% of the overall risk, and the frequency of nuclear accidents releasing more than 100 TBq of Cs-137 should not exceed 1E-06 per reactor year. This paper reviews the hierarchical structure of safety goals in Korea, its relationship with those of other countries, and the relationships among safety goals and subsidiary criteria like core damage frequency and large early release frequency. By analyzing the effect of the release of 100 TBq of Cs-137 via consequence analysis codes in eight different accident scenarios, it was shown that meeting the LRF criterion results in negligible prompt fatalities in the surrounding area. Hence, the LRF criterion dominates the safety goals for Korean NPPs. Safety goals must be consistent with national policy, international standards, and the goals of other counties.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.997-1005
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• 2018

The regulatory periodic safety inspection system is one of the most important methods for confirming the safety of nuclear power plants and the defense in depth in nuclear safety is the most important basic means for accident prevention and mitigation. Recently, a new regulatory technology based on risk-informed and safety performance has been developed and used in advanced countries. However, since the domestic periodic inspection system is being used in the same way over 30 years, it is necessary to know how the inspection contributes to the safety confirmation of the nuclear power plants. In this study, the domestic periodic inspection system currently in use was analyzed from the perspective of defense in depth in nuclear safety. In addition, the analysis results were compared to the U.S. NRC's safety inspection system to obtain consistency and lessons in this study. As a result of analysis, the NRC's safety inspections were distributed almost evenly at the all levels of defense in depth, while in the case of domestic inspection, they were heavily focused on the level 1 of defense in depth. Therefore, it appeared urgent to improve the inspection system to strengthen the other levels of defense in depth in nuclear safety.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1698-1711
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• 2022

A combined safety approach, which uses a best-estimate computer code and adopts conservative assumptions for safety systems availability, is developed and applied to the safety margin evaluation for the Loss of Condenser Vacuum (LOCV) of the 1000 MWe Korean Nuclear Power Plant. The Multi-dimensional Analysis of Reactor Safety-KINS standard (MARS-KS) code is selected as a best-estimate code and the PAPIRUS program is used to obtain different initial operational conditions through random sampling of control variables. During an LOCV event, fuel integrity is not threatened by the increase in Departure from Nuclear Boiling Ratio (DNBR). However, the high pressure in the primary coolant system and the secondary system might affect the system integrity. Thus, the peak pressure becomes a major safety concern. Transient analyses are performed for 124 cases of different initial conditions and the most conservative case, which results in the highest system pressure is selected. It is found the suggested methodology gives similar peak pressures when compared to those predicted from existing methodologies. The proposed approach is expected to minimize the time and efforts required to identify the conservative plant conditions in the existing conservative safety methodologies.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2438-2446
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• 2023

To cool down a nuclear reactor core and prevent the fuel damage without a pump-driven active component during any anticipated accident, the passive emergency core cooling system (PECCS) was designed and adopted in an advanced light water reactor, i-POWER. In this study, for a validation of the cooling capability of PECCS, thermal-hydraulic integral effect tests were performed with the ATLAS facility by simulating intermediate and small break loss-of-coolant accidents (IBLOCA and SBLOCA). The test result showed that PECCS could effectively depressurize the reactor coolant system by supplying the safety injection water from the safety injection tanks (SITs). The result pointed out that the safety injection from IRWST should have been activated earlier to inhibit the excessive core heat-up. The sequence of the PECCS injection and the major thermal hydraulic transient during the SBLOCA transient was similar to the result of the IBLOCA test with the equivalent PECCS condition. The test data can be used to evaluate the capability of thermal hydraulic safety analysis codes in predicting IBLOCA and SBLOCA transients under an operation of passive safety system.

• 항공우주시스템공학회지
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• 제18권3호
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• pp.41-47
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• 2024

최근에 무인항공기의 실용화 및 사업화가 추진됨에 따라 무인항공기의 안전성 확보에 관한 관심이 증가하고 있다. 무인항공기의 사고는 재산 및 인명 피해를 발생시키기 때문에 사고를 예방할 수 있는 기술의 개발은 중요하다. 이러한 이유로 AutoEncoder 모델을 이용한 비정상 비행 상태 탐지 기법이 개발되었다. 그러나 기존 탐지 기법은 성능과 실시간 처리 측면에서 한계를 지닌다. 본 논문에서는 U-Net 기반 비정상 비행 탐지 기법을 제안한다. 제안하는 기법에서는 U-Net 모델에서 얻어지는 재구성 오차에 대한 마할라노비스 거리 증가량에 기반하여 비정상 비행이 탐지된다. 모의실험을 통해 제안 탐지 기법이 기존 탐지 기법에 비해 탐지 성능이 우수하며 온보드 환경에서 실시간으로 구동될 수 있음을 알 수 있다.

• Nuclear Engineering and Technology
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• 제41권1호
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• pp.91-98
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• 2009

The "3+3 Process" for safety critical software for nuclear power plants' I&C (Instrumentation and Control system) has been developed in this work. The main idea of the "3+3 Process" is both to simplify the software development and safety analysis in three steps to fulfill the requirements of a software safety plan [1]. The "3-Step" software development process consists of formal modeling and simulation, automated code generation and coverage analysis between the model and the generated source codes. The "3-Step" safety analysis consists of HAZOP (hazard and operability analysis), FTA (fault tree analysis), and DV (design validation). Put together, these steps are called the "3+3 Process". This scheme of development and safety analysis minimizes the V&V work while increasing the safety and reliability of the software product. For assessment of this process, validation has been done through prototyping of the SDS (safety shut-down system) #1 for PHWR (Pressurized Heavy Water Reactor).

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.695-700
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• 2013

Thirty-eight radiation safety reports for brachytherapy equipment were evaluated to determine the current status of brachytherapy units in Korea and to assess how radiation oncology departments in Korea complete radiation safety reports. The following data was collected: radiation safety report publication year, brachytherapy unit manufacturer, type and activity of the source that was used, affiliation of the drafter, exposure rate constant, the treatment time used to calculate workload and the HVL values used to calculate shielding design goal values. A significant number of the reports (47.4%) included the personal information of the drafter. The treatment time estimates varied widely from 12 to 2,400 min/week. There was acceptable variation in the exposure rate constant values (ranging between 0.469 and 0.592 ($R{\cdot}m^2/Ci{\cdot}hr$), as well as in the HVLs of concrete, steel and lead for Iridium-192 sources that were used to calculate shielding design goal values. There is a need for standard guidelines for completing radiation safety reports that realistically reflect the current clinical situation of radiation oncology departments in Korea. The present study may be useful for formulating these guidelines.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.