• 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.

• Journal of the Korea Concrete Institute
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• v.28 no.6
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• pp.673-683
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• 2016

In recent years, frequent terror or military attack by explosion, impact, fire accidents have occurred. Particularly, World Trade Center collapse and US Department of Defense Pentagon attack on Sept. 11 of 2001. Also, nuclear power plant incident on Mar. 11 of 2011. These attacks and incidents were raised public concerns and anxiety of potential terrorist attacks on major infrastructures and structures. Therefore, the extreme loading researches were performed of prestressed concrete (PSC) member, which widely used for nuclear containment vessel and gas tank. In this paper, to evaluate the blast resistance capacity and its protective performance of bi-directional unbonded prestressed concrete member, blast tests were carried out on $1,400{\times}1,000{\times}300mm$ for reinforced concrete (RC), prestressed concrete without rebar (PSC), prestressed concrete with rebar (PSRC) specimens. The applied blast load was generated by the detonation of 55 lbs ANFO explosive charge at 1.0 m standoff distance. The data acquisitions not only included blast waves of incident pressure, reflected pressure, and impulse, but also included displacement, acceleration, and strains at steel, concrete, PS tendon. The results can be used as basic research references for related research areas, which include protective design and blast simulation under blast loading.

• Journal of Energy Engineering
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• v.24 no.1
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• pp.42-50
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• 2015

To evaluate the heat removal capability of a condenser tube in the PCCS of an advanced nuclear power plant, a steam condensation experiment in the presence of noncondensable gas on a vertical tube is performed. The average heat transfer coefficient is measured on a vertical tube of 40 mm in O.D. and 1.0 m in length. The experiments covers the pressures of 2-4 bar, and the mass fraction of air ranges from 0.1 up to 0.7. From the experimental results, the effects of the total pressure and the concentration of air on the condensation heat transfer coefficient are investigated. The measured data are compared with the predictions by Uchida's and Tagami's correlations, and it is revealed that these models underestimate the condensation heat transfer coefficient of the steam-air mixture.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2005.04a
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• pp.151-158
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• 2005

For the safety analysis of large structures such as nuclear containment buildings, we conventionally prefer to use analytical approach using finite element method rather than empirical test. Therefor, this paper is mainly focused to develop low-order solid finite element model with the elasto-plastic material model for the safety analysis of nuclear containment building. Drucker-Prager failure criteria in uncracked concrete and maximum tensile stress criteria in cracked concrete are used to model the constitutive behavior of concrete. The concrete material model takes into account the aspects of tensile strain, compression strength reduction of concrete and shear transfer to improve the accuracy of the finite element analysis. Finally, numerical simulation to compare the performance of the developed model with experimental results is employed. The numerical results in this study agree very well with the experimental data.

• Proceedings of the Korea Concrete Institute Conference
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• 2004.05a
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• pp.758-761
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• 2004

Prestressing tendons of the nuclear containment building dome are arranged in a non-axisymmetric manner. However, simple axisymmetric modeling of the containment building is often employed to estimate the structural behavior for, e.g. the ultimate pressure, which requires the axisymmetric approximation of the actual tendon arrangements of the dome. A procedure is proposed that can devise the actual 3-dimensional tendon stiffness and prestressing effect into the axisymmetric model. A numerical example of the CANDU type is presented to verify the procedure and to estimate the amount of approximation.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2004.10a
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• pp.312-319
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• 2004

This paper is mai y focused to develop new concrete material model such as ultimate failure surface in compression-compression region, hardening rule and cracking criteria which are basically used in the nonlinear finite element analysis of nuclear prestressed concrete containment building. From the Kepri's experimental results, failure surface of the concrete based on the elasto-plastic material model is modified and new cracking criteria is proposed. Nonlinear FE analysis program using a new material model is implemented to analysis plane concrete. Finally, numerical simulation to compare the performance of the new material model with experimental results is employed. The numerical results by the proposed model in this study agree very well with the experimental data.

• Proceedings of the Korea Concrete Institute Conference
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• 1999.04a
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• pp.229-232
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• 1999

Nuclear power plant structures may be exposed to aggressive environmental effects that may cause their damage mechanisms are reasonably well understood and quantitative evaluation of their effects on time-dependent structural behavior is possible in some instances, such evaluations are generally very difficult and remain novel. The assessment of existing steel containment in nuclear power plants for continued service must provide quantitative evidence that they are able to withstand future extreme loads during a service period with an acceptable level of reliability. Rational methodologies to perform the reliability assessment can be developed from mechanistic models of structural deterioration, using time-dependent structural reliability analysis to take loading and strength uncertainties into account. The final goal of this study is to develop the analysis method for the analysis for the reliability of containment structures. The cause and mechanism of corrosion is first clarified and the reliability assessment method has been established. By introducing the equivalent normal distribution, the procedure of reliability analysis which can determine the failure probabilities has been established.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2006.04a
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• pp.639-646
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• 2006

A nonlinear finite element analysis is carried out to predict the ultimate internal pressure and failure mechanism of a 1/4 scale prestressed concrete containment vessel(PCCV) model using the commercial code ABAQUS. Therefore, this paper is mainly focused to compare the influence of concrete material model, tension stiffening parameter, uplift phenomenon and basemat. From the analysis results, nonlinear behavior of the PCCV showed a substantially different aspects in accordance with the nonlinear material model for the concrete as well as tension stiffening parameter. The boundary conditions beneath the basemat are considered to be a fixed condition and a nonlinear spring element to compare the influence of the uplift. The finite element analysis is considered with and without a basemat to find out the influence of the basemant itself. From the analysis results, the nonlinear behavior of the PCCV is entirely similar for the two cases.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2007.04a
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• pp.694-697
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• 2007

In this paper, a nonlinear finite element analysis program NUCAS, which has been developed for assessment of 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.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2014.11a
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• pp.205-206
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• 2014

It is very important to reduce the construction duration of the Reactor Containment Building (RCB) when considering the more than 50 months on average from concrete placement to completion. The purpose of this study attempts to evaluate the single-stage workability of the system given a change in the height of the setting of RCB exterior wall formwork to be used in nuclear power plant construction. As a result of this study, it is possible height of 3.5m~4m uses formwork when analyzing the construction period and material costs an increase in formwork by concrete lateral pressure, to ensure the workability of the RCB exterior wall formwork. Through this study, I want to provide as basic data for the improvement of workability and RCB shortening the construction period.