• Journal of the Korean Society of Safety
• /
• v.26 no.4
• /
• pp.7-13
• /
• 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
• /
• v.47 no.1
• /
• pp.11-25
• /
• 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.

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

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.623-628
• /
• 1998

A new advanced integral reactor of 330 MWt thermal capacity named SMART (System-Integrated Modular Advanced Reactor) is currently under development in Korea Atomic Energy Research Institute (KAERI) for multi-purpose applications. Modular once-through steam generator (SG) and self-pressurizing pressurizer equipped with wet thermal insulator and cooler are essential components of the SMART. The SMART Provides safety systems such as Passive Residual Heat Removal System (PRHRS). In this study, a computer code for performance analysis of the PRHRS is developed by modeling relevant components and systems of the SMART. Using this computer code, a performance analysis of the PRHRS is performed in order to check whether the passive cooling concept using the PRHRS is feasible. The results of the analysis show that PRHRDS of the SMART has excellent passive heat removal characteristics.

• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.1963-1973
• /
• 2020

Liquid metal-cooled reactors use various passive safety systems driven by natural circulation. Investigating these safety systems experimentally is more advantageous by using a simulant. Although numerous experimental approaches have been applied to natural circulation-driven passive safety systems using simulants, there has been no clear validation of the similarity law. To validate the similarity law experimentally, SINCRO-V experiment was conducted using Wood's metal and water for simulant of the Wood's metal. A pair of SINCRO-V facilities with length-scale ratio of 14.1:1 for identical Bo' was investigated, which was the main similarity parameter in temperature field simulation. In the experimental range of 0.2-1.0% of decay heat, the temperature distribution characteristics of the small water facility were very similar to that of the large Wood's metal facility. The temperature of the Wood's metal predicted by the water experiment showed good agreement with the actual Wood's metal temperature. Despite some error factors like discordance of Gr' and property change along the temperature, the water experiment predicted the Wood's metal temperature with an error of 27%. The validity of the similarity law was confirmed by the SINCRO-V experiments.

• Nuclear Engineering and Technology
• /
• v.53 no.8
• /
• pp.2499-2508
• /
• 2021

Passive containment cooling systems (PCCSs) have been actively studied to improve the inherent safety of nuclear power plants. Hered, we present two concepts, open-loop PCCS (OL-PCCS) and closed-loop PCCS (CL-PCCS), applicable to the PWR with a concrete-type containment. We analyzed the heat-removal performance and flow instability of these PCCS concepts using the GOTHIC code. In both cases, PCCS performance improved when a passive containment cooling heat exchanger (PCCX) was installed in the lower part of the containment building. The OL-PCCS was found to be superior in terms of heat-removal performance. However, in terms of flow instability, the OL-PCCS was more vulnerable than the CL-PCCS. In particular, the possibility of flow instability was higher when the PCCX was installed in the upper part of the containment. Therefore, the installation location of the OL-PCCS should be restricted to minimize flow instability. Conversely, a CL-PCCS can be installed without any positional restriction by adjusting the initial system pressure within the loop, which eliminates flow instability. These results could be used as base data for the thermo-hydraulic evaluation of PCCS in PWR with a large dry concrete-type containment.

• Nuclear Engineering and Technology
• /
• v.56 no.8
• /
• pp.3435-3449
• /
• 2024

Considering that system analysis codes are used for the evaluation of the performance of Passive Safety Systems (PSSs), it is important to investigate the capability of the system analysis code to reliably predict the heat transfer and natural circulation flow, which are the main phenomena governing the performance of a PSS. Since MARS-KS has been widely validated for heat transfer models, this study focuses on evaluating its capability to predict the single and two-phase pressure drops and natural circulation flow. The straight pipe simulation results indicate that the pressure drop predictions are reliable within ±5 % error margin for the single-phase flow and the errors of pressure drop up to - 30 % for the two-phase flow. Through single-phase natural circulation flow analysis, it is concluded that the use of the appropriate K-factor modeling based on the flow regimes is important since the natural circulation flow rate in MARS-KS is mainly affected by the form loss factor modeling. With two-phase natural circulation flow analysis, this study emphasizes the behavior of the system could change significantly depending on the two-phase wall friction and pressure loss modeling. With the analysis results, modeling considerations for the PSS performance evaluation with the system analysis codes are proposed.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
• /
• v.23 no.6
• /
• pp.31-39
• /
• 2009

This paper proposes an ultrasonic sensor based intelligent automatic door system which improves the performance of conventional door systems by adding more intelligent functions such that it offers more convenience to passersby and reduces power loss. The conventional automatic door systems employed passive and active infrared sensors for detecting objects and human bodies. But, they have problems such as power loss in door closing, not sensing fast approaching objects, and safety. The proposed automatic door system with ultrasonic sensors prevents unnecessary door closings to save the power and senses fast approaching objects to open the door at proper time, and improves safety. Thus, the proposed system improves the performance of the conventional systems in terms of operation, economy, and safety.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.791-792
• /
• 2005

• Proceedings of the KSR Conference
• /
• 2011.10a
• /
• pp.2517-2526
• /
• 2011

As technologies are developed and systems are complicated, hazards embedded in the system are also increasing. proving safety and managing the safety is more scientific and organizational domain so that safety management system is pursuing to be active formation detecting the factors of hazard and managing them beyond passive-way. In the future, in order to establish and manage national safety management system, it is important to have effective system and manage it and also more important that all the people related to target system has to change their recognition and to play roles in it.