• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1132-1141
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• 2019

Steel-concrete-steel (SCS) sandwich structures have important advantages over conventional concrete structures, however, bond-slip between the steel plate and concrete may lead to a loss of composite action, resulting in a reduction of stiffness and fatigue life of SCS sandwich structures. Due to the inaccessibility and invisibility of the interface, the interfacial performance monitoring and debonding detection using traditional measurement methods, such as relative displacement between the steel plate and core concrete, have proved challenging. In this work, two methods using piezoelectric transducers are proposed to detect the bond-slip between steel plate and core concrete during the test of the beam. The first one is acoustic emission (AE) method, which can detect the dynamic process of bond-slip. AE signals can be detected when initial micro cracks form and indicate the damage severity, types and locations. The second is electromechanical impedance (EMI) method, which can be used to evaluate the damage due to bond-slip through comparing with the reference data in static state, even if the bond-slip is invisible and suspends. In this work, the experiment is implemented to demonstrate the bond-slip monitoring using above methods. Experimental results and further analysis show the validity and unique advantage of the proposed methods.

• Nuclear Engineering and Technology
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• v.51 no.7
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.49-58
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• 2018

The regulatory periodic inspection program (PSI) conducted at every overhaul period is the most important process for confirming the safety of nuclear power plants. The PSI for operating nuclear power plants in Korea mainly consist of component level performance check that had been developed based on deterministic approach putting the same degree of importance to all the inspection items. This inspection methodology is likely to be effective for preoperational inspection. However, once the plant is put into service, the PSI must be focused on whether to minimize the risk of accident using defense-in-depth concept and risk insight. The incorporation of defense-in-depth concept and risk insight into the deterministic based safety inspection has not been well studied so far. In this study, two track approaches are proposed to make sure that core damage be avoided: one is to secure success path and the other to block the failure path in a specific event tree of PSA. The investigation shows how to select safety important components and how to set up inspection group to ensure that core damage would not occur for a given initiating event, which results in strengthening defense-in-depth level 3.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.402-409
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• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• Journal of the Korean Society of Safety
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• v.37 no.2
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• pp.43-53
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• 2022

This paper describes an attempt to improve and optimize the operational safety level of a domestic research reactor by conducting a probabilistic safety assessment (PSA) under full-power operating conditions. The PSA was undertaken to assess the level of safety at an operating research reactor in Korea, to evaluate whether it is probabilistically safe and reliable to operate, and to obtain insights regarding the requisite procedural and design improvements for achieving safer operation. The technical objectives were to use the PSA to identify the accident sequences leading to core damage, and to conduct sensitivity analyses based thereon to derive insights regarding potential design and procedural improvements. Based on the dominant accident sequences identified by the PSA, eight types of sensitivity analysis were performed, and relevant insights for achieving safer operation were derived. When these insights were applied to the reactor design and operating procedure, the risk was found to be reduced by approximately ten times, and the safety was significantly improved. The results demonstrate that the PSA methodology is very effective for improving reactor safety in the full-power operating phase. In particular, it is a highly suitable approach for identifying the deficiencies of a reactor operating at full power, and for improving the reactor safety by overcoming those deficiencies.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3324-3335
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• 2022

Several countries show interest in the Generation-IV power reactor innovative small module (PRISM), including: Canada, Japan, Korea, Saudi Arabia and the United Kingdom. Generation IV International Forum (GIF) has recommended the utilizing of probabilistic safety assessment (PSA) in evaluating the safety of Generation-IV reactors. This paper reviews the PSA performed for PRISM using SAPHIRE 7.27 code. This work shows that the core damage frequency (CDF) of PRISM for a single module is estimated by 8.5E-8/year which is lower than the Generation-IV target that is 1E-6 core damage per year. The social risk of PRISM (likelihood of latent cancer fatality) with evacuation is estimated by 9.0E-12/year which is much lower than the basic safety objective (BSO) that is 1E-7/year. The social risk without evacuation is estimated by 1.2E- 11/year which is also much lower than the BSO. For the individual risk (likelihood of prompt fatality), it is concluded that it can be considered negligible with evacuation (1.0E-13/year). Assuming no evacuation, the individual risk is 2.7E-10/year which is again much lower than the BSO. In comparison with other PSAs performed for similar sodium fast reactors (SFRs), it shows that PRISM concept has the lowest CDF.

• Journal of the Korean Society of Safety
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• v.38 no.3
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• pp.110-120
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• 2023

Following the Fukushima accident, portable equipment employed as accident mitigating systems have been installed and operated to reduce core damage and large early release frequencies. In addition, the establishment of an accident management strategy has gained importance. This study investigated the current status of portable equipment including the international portable equipment FLEX (diverse and flexible coping strategies), and domestic portable equipment multi-barrier accident coping strategy (MACST). Research on optimal utilization of MACST remains insufficient. As a preliminary study for establishing an optimal strategy, sensitivity studies were conducted to facilitate the priority of use on portable equipment, number of portable equipment, and dependency of operator actions based on a multi-unit probabilistic safety assessment model. The results revealed the conditions that reduced the multi-unit and site conditional core damage probabilities, indicating the optimal strategy of MACST. The results of this study can be used as a reference for establishing an optimal strategy that utilizes domestic safety equipment in the future.

• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.434-442
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• 2015

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

• Structural Engineering and Mechanics
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• v.32 no.2
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• pp.351-370
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• 2009

This paper presents a computational study for the structural response of blast loaded metallic sandwich panels, with the emphasis placed on their failure behaviours. The fully-clamped panels are square, and the honeycomb core and skins are made of the same aluminium alloy. A material model considering strain and strain rate hardening effects is used and the blast load is idealised as either a uniform or localised pressure over a short duration. The deformation/failure procedure and modes of the sandwich panels are identified and analysed. In the uniform loading condition, the effect of core density and face-sheets thicknesses is analysed. Likewise, the influence of pulse shape on the failure modes is investigated by deriving a pressure-impulse (P-I) diagram. For localised loading, a comparative study is carried out to assess the blast resistant behaviours of three types of structures: sandwich panel with honeycomb core, two face-sheets with air core and monolithic plate, in terms of their permanent deflections and damage degrees. The finding of this research provides a valuable insight into the engineering design of sandwich constructions against air blast loads.

• Structural Engineering and Mechanics
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• v.28 no.3
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• pp.295-316
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• 2008

In the present paper a mechanical model to predict the compressive response of high strength short concrete columns with square cross-section confined by transverse steel is presented. The model allows one to estimate the equivalent confinement pressures exercised by transverse steel during the loading process taking into account of the interaction of the stirrups with the inner core both in the plane of the stirrups and in the space between two successive stirrups. The lateral pressure distributions at hoop levels are obtained by using a simple model of elastic beam on elastic medium simulating the interaction between stirrups and concrete core, including yielding of steel stirrups and damage of concrete core by means of the variation in the elastic modulus and in the Poisson's coefficient. Complete stress-strain curves in compression of confined concrete core are obtained considering the variation of the axial forces in the leg of the stirrup during the loading process. The model was compared with some others presented in the literature and it was validated on the basis of the existing experimental data. Finally, it was shown that the model allows one to include the main parameters governing the confinement problems of high strength concrete members such as: - the strength of plain concrete and its brittleness; - the diameter, the pitch and the yielding stress of the stirrups; - the diameter and the yielding stress of longitudinal bars; - the side of the member, etc.