• KSCE Journal of Civil and Environmental Engineering Research
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• v.32 no.3A
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• pp.183-188
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• 2012

The bonded tendon was adopted to the reactor containment building of some operating nuclear power plants in Korea and the assessment of the prestress force on the bonded tendon is very important for the evaluation of the structural integrity. The prestress force of the bonded tendon at real reactor containment building, was evaluated using the SI technique and impact signal analysis technique which were developed to improve the existing indirect assessmment technique. For these techniques, the strain of the reactor containment building and the stress wave velocity of the bonded tendon were measured. Both SI technique and impact signal analysis technique give the highly reliable results comparison with the existing theoretical approach. Therefore, it is confirmed that the developed techniques are very useful for the evaluation of the prestress force on the bonded tendon.

• Journal of the Korea Concrete Institute
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• v.17 no.4 s.88
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• pp.527-533
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• 2005

Axisymmetric modeling of the nuclear containment building has been often employed in practice to estimate structural behavior for the axisymmetric loadings, where the axisymmetric approximation is required for the actual non-axisymmetric tendon arrangements in the dome. In the preceding companion paper, some procedures are proposed for the domestic CANDU and KSNP type containments that can implement the actual 3-dimensional tendon stiffness and prestressing effect into the axisymmetric model. In this paper, the proposed schemes are verified through some numerical examples comparing the results of the actual 3-dimensional model with those of some axisymmetric models. The results of the proposed axisymmetric analyses show relatively good agreements with the actual structural behavior especially for the CANDU type. Also, it is shown that proper level of the prestressing in a hoop direction plays an important role to predict the actual prestressing effect in the axisymmetric dome modeling. Finally, correction factors are discussed that can revise some approximations introduced in the derivations.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.34 no.6
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• pp.385-392
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• 2021

Interest in the seismic performance of nuclear facilities under strong earthquakes has increased because their nonlinear response is important. In this paper, we proposed appropriate parameters for the nonlinear finite element analysis of a concrete material model, for a reinforced concrete (RC) shear wall in nuclear facilities: maximum tensile strength, dilation angle, and damage parameter. The study of the effects of the important parameters, on the nonlinear behavior and shear failure mode of the RC shear wall having low aspect ratio, was conducted using ABAQUS finite element analysis program. Based on the study results the nonlinear response of a nuclear reactor containment building (RCB) subjected to a strong earthquake was evaluated using nonlinear time-history analysis.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2748-2754
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• 2022

Because existing reactant coolant system (RCS) leakage detection mechanisms are insensitive to small leaks, a real-time, direct detection system with a detection threshold below 0.5 gpm·hr^-1 was studied. A beta-ray detection system using a silicon detector with good energy resolution for beta rays and a low gamma-ray response was proposed. The detection performance in the leakage condition was evaluated through experiments and simulations. The concentration of ¹⁶N in the coolant corresponding to a coolant leakage of 0.5 gpm was calculated using the analytic method and ORIGEN-ARP. Based on the concentration of ¹⁶N and the measurement of the silicon detector with ⁹⁰Sr/⁹⁰Y, the beta-ray count rate was estimated using MCNPX. To evaluate the effect of gamma rays inside the containment building, the signal-to-noise ratio (SNR) was calculated. To evaluate the count rate ratio, the radiation field inside the containment building was simulated using MCNPX, and response evaluation experiments were performed using beta and gamma rays on the silicon detector. The expected beta-ray count rate at 0.5 gpm leakage was 7.26 × 10⁵ counts/sec, and the signal-to-background count rate ratio exceeded 88 for a transport time of 10 s, demonstrating its suitability for operation inside a reactor containment building.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.29 no.6A
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• pp.577-585
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• 2009

A 9-node degenerated shell finite element(FE), which has been developed for assessment of ultimate pressure capacity and nonlinear analysis for nuclear containment building is described in this paper. Reissner-Midnlin(RM) assumptions are adopted to develop the shell FE so that transverse shear deformation effects is considered. Material model for concrete prior to cracking is constructed based on the equivalent stress-equivalent strain relationship. Tension stiffening model, shear transfer mechanism and compressive strength reduction model are used to model the material behavior of concrete after cracking. Niwa and Aoyagi-Yamada failure criteria have been adapted to find initial cracking point in compression-tension and tension-tension region, 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.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2014.11a
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• pp.212-213
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• 2014

The purpose of this study is to suggest analysis model of RCB construction in nuclear power plant. For the objective, This study drew the risk factors of RCB construction from existing literature. The results of the study proposed analysis model made hierarchy in rebar, form, and concrete work. These will be baseline data for risk management in construction project of nuclear power plant.

• Journal of Energy Engineering
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• v.24 no.3
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• pp.96-108
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• 2015

A passive containment cooling system has been designed to remove the heat inside a containment during accidents without external power supply. In this work, the PCCS was introduced in the APR1400 plant to replace the containment spray system and, then, the thermal-hydraulic performance of the PCCS was analyzed using the system thermal-hydraulic computer code, MARS. A double-ended cold-leg break accident, which is known to induce the maximum pressure in the containment, is simulated, where the thermal hydraulics of the PCCS, the reactor coolant system, and the containment are simultaneously simulated. The results of the calculations showed that the PCCS can replace the existing spray system and that the containment building and its internal structure also play a very important role for the heat removal during the accident. Some sensitivity calculations were carried out to evaluate the model uncertainty and the effects of design parameters. The limitations of the PCCS are also discussed.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1327-1328
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.583-584
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.423.1-423
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• 2002