• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.169.1-169.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.201-201
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.84.2-84
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• 2002

• Nuclear Engineering and Technology
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• v.31 no.5
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• pp.476-485
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• 1999

In order to evaluate the gravity-injection capability to maintain core cooling after a loss-of-shutdown-cooling event during shutdown operation, the plant conditions of the Yong Gwang Units 3&4 were reviewed. The six cases of possible gravity-injection paths from the refueling water tank (RWT) were identified and the thermal-hydraulic analyses were performed using the RELAP5/MOD3.2 code. The core cooling capability was significantly dependent on the gravity-injection path, the RCS opening, and the injection rate. In the cases with the pressurizer manway opening higher than the RWT water level, the coolant was held up in the pressurizer and the system pressure continued increasing after gravity-injection. The gravity injection eventually stopped due to the high system pressure and the core was uncovered. In the cases with the injection path and opening on the same leg side, the core cooling was dependent on whether the water injected from the RWT passed the core region or not. However, in the cases with the injection path and opening on the different leg side, the system was well depressurized after gravity-injection and the core boiling was successfully prevented for a long-term transient. In addition, from the sensitivity study on the gravity-injection flow rate, it was found that about 54 kg/s of injection rate was required to maintain the core cooling and the core cooling could be provided for about 10.6 hours after event with that injection rate from the RWT. Those analysis results would provide useful information to operators coping with the event.

• The Journal of the Institute of Internet, Broadcasting and Communication
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• v.22 no.3
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• pp.171-176
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• 2022

The communication signal for quick cut-off specification is defined as "design to support fast cut-off requirements of all applicable photovoltaic(PV) systems" in NEC 2014, NEC 2017 or the corresponding UL standard regardless of the system configuration. On the other hand, if you look at the domestic regulations related to new and renewable energy, the standards, regulations, and guidelines set by each institution are general, or only the parts necessary for the institution are being established and operated. There are many insufficient points to apply these things to photovoltaic facilities, and there are cases where excessive facilities are installed according to the design, inspection standards of supervisors and inspection agencies, and the skill level of inspectors. The internationally accepted IEC standards deal with various facility standards in detail. In each European country, there are separate facility regulations based on IEC. In particular, the performance and safety of devices are dealt with in detail, and in the case of 'NEC Article 690' applied in North America such as the United States, each item is described in detail. Therefore, in this paper, we will look at the details of the PV shutdown technology that is currently used and applied internationally.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.211-221
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• 2020

One fundamental element of probabilistic safety assessment (PSA) is the initiating event (IE) analysis. Since IE frequencies can change over time, time-trend analysis is required to obtain optimized IE frequencies. Accordingly, such time-trend analyses have been employed to estimate industry-average IE frequencies for use in the PSAs of U.S. nuclear power plants (NPPs); existing PSAs of Korean NPPs, however, neglect such analysis in the estimation of IE frequencies. This article therefore provides the method for and results of estimating Korean industry-average IE frequencies using time-trend analysis. It also examines the effects of the IE frequencies obtained from this study on risk insights by applying them to recently updated internal events Level 1 PSA models (at-power and shutdown) for an OPR-1000 plant. As a result, at-power core damage frequency decreased while shutdown core damage frequency increased, with the related contributions from each IE category changing accordingly. These results imply that the incorporation of time-trend analysis leads to different IE frequencies and resulting risk insights. The IE frequency distributions presented in this study can be used in future PSA updates for Korean NPPs, and should be further updated themselves by adding more recent data.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.342-352
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• 1998

A nuclear design feasibility of soluble boron free(SBF core for the medium-sized(600MWe) PWR was investigated. The result conformed that soluble boron free operation could be performed by using current PWR proven technologies. Westinghouse advanced reactor, AP-600 was chosen as a design prototype. Design modification was applied for the assembly design with burnable poison and control rod absorber material. In order to control excess reactivity, large amount of gadolinia integral burnable poison rods were used and B4C was used as a control rod absorber material. For control of bottom shift axial power shape due to high temperature feedback in SBF core, axial zoning of burnable poison was applied to the fuel assemblies design. The combination of enrichment and rod number zoning for burnable poison could make an excess reactivity swing flat within around 1% and these also led effective control on axial power offset and peak pin power, The safety assessment of the designed core was peformed by the calculation of MTC, FTC and shutdown margin. MTC in designed SBF core was greater around 6 times than one of Ulchin unit 3&4. Utilization of enriched BIO(up to 50w1o) in B4C shutdown control rods provided enough shutdown margin as well as subcriticality at cold refueling condition.

• Journal of the Korean Society of Safety
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• v.32 no.1
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• pp.140-144
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• 2017

Fire protection for nuclear power plants should be designed according to the concept of "Defense in Depth" to achieve the reactor safety shutdown. This concept focuses on fire prevention, fire suppression and safe shutdown. Fire prevention is the first line of "Defense in Depth" and the licensee should establish administrative measures to minimize the potential for fire to occur. Administrative measures should include procedures to control handling and use of combustibles. Electrical cables is the major contributor of fire loads in nuclear power plants, therefore electrical cables should be fire retardant. Electrical cables installed in nuclear power plants should pass the flame test in IEEE-383 standard in accordance with NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants". To assure the fire retardant of electrical cables during design life, both aged and unaged cable specimens should be tested in accordance with IEEE-383. It can be generally thought that the flammability of electrical cables has been increased by wearing as time passed, however the results from fire retardant tests performed in U.S.A and Korea indicate the inconsistent tendency of aging and consequential decrease in flammability. In this study, it is expected that the effective methodology for validation of fire retardant performance would be identified through the review of the results from fire retardant tests.

• Journal of KIISE:Software and Applications
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• v.27 no.2
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• pp.120-133
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• 2000

Major obstruction of using formal methods for hybrid real-time systems in industry is the difficulty that engineers have in understanding and applying the quantitative methods in an abstract requirements phase. While formal methods technology in safety-critical systems can help increase confidence of software, difficulty and complexity in using them can cause another hazard. In order to overcome this obstruction, we propose a framework for qualitative requirements engineering of the hybrid real-time systems. It consists of a qualitative method for requirements specification, called QFM (Qualitative Formal Method), and a safety analysis method for the requirements based on a causality information, called CRSA (Causal Requirements Safety Analysis). QFM emphasizes the idea of a causal and qualitative reasoning in formal methods to reduce the cognitive burden of designers when specifying and validating the software requirements of hybrid safety systems. CRSA can evaluate the logical contribution of the software elements to the physical hazard of systems by utilizing the causality information that is kept during specification by QFM. Using the Shutdown System 2 of Wolsong nuclear power plants as a realistic example, we demonstrate the effectiveness of our approach.

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.1-11
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• 1989

Numerous losses of decay heat removal capability have occurred at U during stutodwn while its significance to safety is needless to say. A study is carried out as an attempt to assess what could be done to lower the frequency of these events and to mitigate their consequences in the unlikely event that one occurs. The shutdown risk model is developed and analyzed using Event/Fault Tree for the typical pressurized water reactor. The human cognitive reliability (HCR) model, two-stage bayesian approach and staircase function model are used to estimate human reliability, initiating event frequency and offsite power non-recovery probability given loss of offsite power, respectively. The results of this study indicate that the risk of a Pm at shutdown is not much lower than the risk when the plant is operating. By examining the dominant accident sequences obtained, several design deficiencies are identified and it is found that some proposed changes lead to significant reduction in core damage frequency due to loss of cooling events.