• Nuclear Engineering and Technology
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• 제39권4호
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• pp.287-298
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• 2007

As part of the $AP1000^{TM}$ pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.

• 시스템엔지니어링학술지
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• 제15권2호
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• pp.39-46
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• 2019

The purpose of this paper is to present the results of the study, applying Systems Engineering(SE) method, on the feasibility of developing a Melt-down Proof Modular Micro Reactor(MDP-MMR) for its future deployment in Korea. The reactor is being developed by NCSU (North Carolina State University) due to its advantage of melt-down proof nature of the reactor core. For this paper, the characteristics of the MDP-MMR has been studied in terms of fuel characteristics, inherent safety features and passive safety system. The NCSU's development process has been reviewed applying the SE method, and further research is recommended for the feasibility study on deploying such a modular micro reactor in Korea.

• 한국안전학회지
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• 제32권6호
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• pp.81-88
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• 2017

This study deals with the optimal design of a hybrid control system composed of a combination of active control system and passive control system for effective seismic performance improvement of two adjacent structures. The proposed hybrid control system adopts a configuration of installing an active control device in one building and connecting two adjacent structures with a passive control device so that the one-side active control force can be bi-directionally applied to both buildings through the passive connecting devices. In order to derive the optimal performance of the proposed system, the design parameters of the passive and active control systems were searched using the genetic algorithm. Numerical simulations of 10-story and 8-story buildings have been performed to verify the effectiveness of the proposed technique. For the purpose of comparison, the conventional independent control system with two identical active control systems being installed separately for each structure was also optimally designed and its seismic response has been evaluated as well. From the comparative results of the two control systems, it is demonstrated that the proposed hybrid control system requires larger control force for its one-side active control device than the conventional independent control system does for each of both-side active devices, but quite less than the total control force required for both-side devices of the independent control system, while maintaining similar seismic performance. Therefore, the proposed system is more economical and reliable than the conventional independent control system with two identical active devices.

• Nuclear Engineering and Technology
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• 제49권2호
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• pp.360-372
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• 2017

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

• 한국안전학회지
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• 제19권1호
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• pp.102-107
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• 2004

Performance based design is a recent evolutionary step in the process of designing fire protection systems. In essence, it is a logical design process resulting in a solution that achieves a specified performance. Sometimes the prescriptive solutions presented in various codes and standards are too expensive or inflexible. Often the solutions do not and enables optimization of a solution for cost and function. In this study, performance based design was carried out to determine the extent of passive fire protection for oil terminal facilities. The results of performance based design were compared with those of prescriptive code based design. Performance based design is not always more economic than prescriptive code based design but provides more reliable and effective design that is fit for the purpose.

• 시스템엔지니어링학술지
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• 제12권2호
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• pp.91-99
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• 2016

The safety of nuclear power plants has received much attention; this safety largely depends on the continuous availability of electrical energy source during all modes of nuclear power plant operation. A station blackout (SBO) describes the loss of the off-site electric power, the failure of the emergency diesel generators, and the unavailability of the alternate AC (AAC) power. Consequently, all systems that are AC powered such as the safety injection, shutdown cooling, component cooling water, and essential service water systems are unavailable. The aim of this study is to investigate the deficiencies of the existing alternatives for coping with an extended SBO for APR1400 design. The method is analyzing the existing deficiencies and proposing an optimal solution for the NPP design during the extended SBO. This study, established a new passive system, called passive decay heat removal system (PDHRS), using systems engineering approach.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1059-1070
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• 2016

The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.84-88
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• 2014

A long term passive cooling system is considered as the most important safety feature for the nuclear design after the Fukushima Daiichi nuclear power plant accident in 2011. The conventional active pump driven safety systems are not available during a station Black Out (SBO) accident. The current design requirement on cooling time of the Passive Auxiliarly Feedwater System (PAFS) is about 8 hours only. To meet the 72 hours cooling time, the pool capacity of cooling water tank should be increased as much as 3~4 times larger than that of current water cooling tank. In order to extend the cooling time for 72 hours, a new passive air-water combined cooling system is proposed. This paper provides the feasibility of the combined passive air-water cooling system. The current pool capacity of water cooling system is preserved, and the cooling capability is extended by an additional air cooler.

• 한국안전학회지
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• 제26권4호
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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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• 제47권1호
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• pp.11-25
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• 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and $60m^3$ volume) are discussed in the light of the Fukushima accident.