• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.632-637
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• 2017

Even though the potential for an accident in nuclear power plants is very low, multiple emergency plans are necessary because the impact of such an accident to the public is enormous. One of these emergency plans involves a robotic system for investigating accidents under conditions of high radiation and contaminated air. To develop a robot suitable for operation in a nuclear power plant, we focused on eliminating the three major obstacles that challenge robots in such conditions: the disconnection of radio communication, falling on uneven floors, and loss of localization. To solve the radio problem, a Wi-Fi extender was used in radio shadow areas. To reinforce the walking, we developed two- and four-leg convertible walking, a floor adaptive foot, a roly-poly defensive falling design, and automatic standing recovery after falling methods were developed. To allow the robot to determine its location in the containment building, a bar code landmark reading method was chosen. When a severe accident occurs, this robot will be useful for accident condition monitoring. We also anticipate the robot can serve as a workman aid in a high radiation area during normal operations.

• Fire Science and Engineering
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• v.32 no.1
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• pp.57-65
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• 2018

In this study, a post-fire safe shutdown analysis (PFSSA) including multiple spurious operation (MSO) treatments for cables was conducted with an imaginary nuclear power plant for training using a deterministic fire analysis code. The imaginary nuclear power plant for the training consisted of a reactor containment building and an auxiliary building, including a total of 22 fire areas. The equipment including valves, pumps, emergency diesel generators, switch gears, motor control centers, and logic controllers were located in each fire area of the imaginary plant. It was assumed that each equipment is connected with two cables and that each cable passes through the fire areas along the cable trays. A database containing the information on the equipment and cables for the imaginary plant was constructed for the fire area analysis. The fire area analysis was performed for several assumed MSO scenarios, equipment logics, and cable logics. A mitigation measure using a three hour rated wrap was applied to the failed cables and cable trays after the fire area analysis.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.20 no.3
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• pp.353-364
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• 2007

In this paper, a nonlinear finite element analysis program NUCAS, which has been developed for assessment of ultimate pressure capacity and failure mode for nuclear containment building is described. Degenerated shell element with assumed strain method and low-order solid element with enhanced assumed strain method is adapted to microscopic material and elasto-plastic material model, respectively. Finally, the performance of the developed program is tested and demonstrated with several examples. From the numerical tests, the present results show a good agreement with experimental data or other numerical results.

• Journal of Radiation Protection and Research
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• v.40 no.4
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• pp.244-251
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• 2015

The containment filtered venting system reduces the range of the contamination area around the nuclear power plant by strengthening the integrity of the containment building. In this study, the probabilistic assessment code MACCS2 was used to assess the effect of the CFVS to off-site. The accident source term was selected from a Probabilistic Safety Analysis report of SHINKORI 1&2 Nuclear Power Plant. The three source term categories from 19 STC were chosen to evaluate the effective dose and thyroid dose of residents around the power plant and the dose with CFVS and without CFVS were compared. The dose was calculated according to the distance from the nuclear power plant, so the damage scale based on the distance that exceeds the IAEA criteria for effective dose (100 mSv per 7 days) and thyroid dose (50 mSv per 7 days) were compared. The effective dose reduction rates of the STC-3, STC-4, STC-6 were about 95-99% in the whole range (0~35 km), 96-98% for the thyroid dose. There are similar results between effective dose and thyroid dose. After applying the CFVS, the damage scale that exceeds the effective dose criteria was about 1 km (mean). Especially, the STC-4 damage scale was decreased from 26 km (mean) to 1.2 km (mean) significantly. The damage scale that exceed the thyroid dose criteria was decreased to 2~3 km (mean). The STC-4 damage scale was also decreased significantly as compared to STC-3, STC-6 in terms of effective dose.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.23 no.10
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• pp.1274-1284
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• 1999

The Regulatory Guide 1.82 recommends an analysis of hydraulic performance for sump of ECCS (Emergency Core Cooing System) when LOCA(Loss of Coolant Accident) occurs in a nuclear power plant. The present study deals with 3-dimensional, unsteady, turbulent and two-phase flow simulation to examine the behavior of mixture of reactor coolant and debris near the floor of containment building in conjunction with appropriate assumptions. The dispersed solid model has been adjusted to the interfacial momentum transfer between reactor coolant and debris. According to the results, the counterclockwiserecirculation zone had been formed in the region between sump and connection aisle about 376 second after LOCA occurs. The debris thickness accumulated on a sump screen periodically increases or decreases up to 2000 second, afterwards its peak decreases.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.36 no.3
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• pp.155-163
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• 2023

Real-time monitoring technology is critical for ensuring the safety and reliability of nuclear power plant structures. However, the current seismic monitoring system has limited system identification capabilities such as modal parameter estimation. To obtain global behavior data and dynamic characteristics, multiple sensors must be optimally placed. Although several studies on optimal sensor placement have been conducted, they have primarily focused on civil and mechanical structures. Nuclear power plant structures require robust signals, even at low signal-to-noise ratios, and the robustness of each mode must be assessed separately. This is because the mode contributions of nuclear power plant containment buildings are concentrated in low-order modes. Therefore, this study proposes an optimal sensor placement methodology that can evaluate robustness against noise and the effects of each mode. Indicators, such as auto modal assurance criterion (MAC), cross MAC, and mode shape distribution by node were analyzed, and the suitability of the methodology was verified through numerical analysis.

• Journal of the Korean Society of Safety
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• v.23 no.2
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• pp.21-29
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• 2008

To evaluate the structural integrity of the NPP containment building more rigorously, the effective prestress, which is one of the most affecting elements, needs to be estimated exactly. This paper presents the results of an experimental study to determine the effective prestress force in prestressed concrete beams. It is possible to improve the effective prestress measuring method by test beam, which is being applied for the investigation of the nuclear power plant in operation. If experimentally evaluated Lift-Off method in this study can be coupled with test beam test currently being used in in-service nuclear power plant, it is possible to measure prestress loss of the tendon and the level of the effective prestress load.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.3068-3084
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• 2021

Investigations of large commercial aircraft impact effect on nuclear power plant (NPP) buildings have been drawing extensive attentions, particularly after the 9/11 event, and this paper aims to experimentally assess the damage and vibrations of NPP buildings subjected to aircraft crash. In present Part I, two shots of reduce-scaled model test of aircraft impacting on NPP building were carried out. Firstly, the 1:15 aircraft model (weighs 135 kg) and RC NPP model (weighs about 70 t) are designed and prepared. Then, based on the large rocket sled loading test platform, the aircraft models were accelerated to impact perpendicularly on the two sides of NPP model, i.e., containment and auxiliary buildings, with a velocity of about 170 m/s. The strain-time histories of rebars within the impact area and acceleration-time histories of each floor of NPP model are derived from the pre-arranged twenty-one strain gauges and twenty tri-axial accelerometers, and the whole impact processes were recorded by three high-speed cameras. The local penetration and perforation failure modes occurred respectively in the collision scenarios of containment and auxiliary buildings, and some suggestions for the NPP design are given. The maximum acceleration in the 1:15 scaled tests is 1785.73 g, and thus the corresponding maximum resultant acceleration in a prototype impact might be about 119 g, which poses a potential threat to the nuclear equipment. Furthermore, it was found that the nonlinear decrease of vibrations along the height was well reflected by the variations of both the maximum resultant vibrations and Cumulative Absolute Velocity (CAV). The present experimental work on the damage and dynamic responses of NPP structure under aircraft impact is firstly presented, which could provide a benchmark basis for further safety assessments of prototype NPP structure as well as inner systems and components against aircraft crash.

• Earthquakes and Structures
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• v.15 no.3
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• pp.295-306
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• 2018

Reinforced concrete containment (RCC) building has long been considered as the last barrier for keeping the radiation from leaking into the environment. It is important to quantify the performance of these structures and facilities considering extreme conditions. However, the preceding research on evaluating nuclear power plant (NPP) structures, particularly considering mainshock-aftershock seismic sequences, is deficient. Therefore, this manuscript serves to investigate the seismic fragility of a typical RCC building subjected to mainshock-aftershock seismic sequences. The implementation of the fragility assessment has been performed based on the incremental dynamic analysis (IDA) method. A lumped mass RCC model considering the tri-linear skeleton curve and the maximum point-oriented hysteretic rule is employed for IDA analyses. The results indicate that the seismic capacity of the RCC building would be overestimated without taking into account the mainshock-aftershock effects. It is also found that the seismic capacity of the RCC building decreases with the increase of the relative intensity of aftershock ground motions to mainshock ground motions. In addition, the effects of artificial mainshock-aftershock ground motions generated from the repeated and randomized approaches and the polarity of the aftershock with respect to the mainshock on the evaluation of the RCC are also researched, respectively.

• Journal of the Earthquake Engineering Society of Korea
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• v.23 no.3
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• pp.149-157
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• 2019

The purpose of this study is to investigate the effects of the significant duration of ground motions on responses of base-isolated nuclear power plants (NPPs). Two sets of ground motion records with short duration (SD) and long duration (LD) motions, scaled to match the target response spectrum, are used to perform time-history analyses. The reactor containment building in the Advanced Power Reactor 1400 (APR1400) NPP is numerically modeled using lumped-mass stick elements in SAP2000. Seismic responses of the base-isolated NPP are monitored in forms of lateral displacements, shear forces, floor response spectra of the containment building, and hysteretic energy of the lead rubber bearing (LRB). Fragility curves for different limit states, which are defined based on the shear deformation of the base isolator, are developed. The numerical results reveal that the average seismic responses of base-isolated NPP under SD and LD motion sets were shown to be mostly identical. For PGA larger than 0.4g, the mean deformation of LRB for LD motions was bigger than that for SD ones due to a higher hysteretic energy of LRB produced in LD shakings. Under LD motions, median parameters of fragility functions for three limit states were reduced by 12% to 15% compared to that due to SD motions. This clearly indicates that it is important to select ground motions with both SD and LD proportionally in the seismic evaluation of NPP structures.