• Proceedings of the Korea Concrete Institute Conference
• /
• 1996.10a
• /
• pp.95-100
• /
• 1996

Creep, drying shrinkage, modulus of elasiticity and Poisson's ratio of concrete are influenced by a number of factors such as mix type, member thickness, curing condition and loading cases. Particularly, creep and shrinkage in concrete have yet to be studied due to its complicated time-dependent properties. In this study, the concrete creep tests were carried out at varous ages of loading-7, 28, 90, 180 and 365 days in order to investigate and quantify its long-term properties. The test procedures and analysis of the test results were also described herein. The results of this study will enable A/E to calculate effective prestressing forces considering time-dependent prestressing loss and evaluate the structural integrity of the prestressing system using the representative values derived from this property test.

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

• Journal of the Korean Society of Safety
• /
• v.35 no.3
• /
• pp.96-104
• /
• 2020

The numerical simulations were conducted to investigate the thermal-fluid phenomena occurred inside the experimental apparatus during a PCCS, used to remove heat released in accidents from a containment of light water nuclear power plant, operation. Numerical simulations of the flow and heat transfer caused by wall condensation inside the containment simulation vessel (CSV), which equipped with 18 vertical heat exchanger tubes, were conducted using the commercial computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the wall condensation model were used for turbulence closure and wall condensation, respectively. The simulation using the actual size of the apparatus. However, rather than simulating the whole experimental apparatus in consideration of the experimental cases, calculation resources, and calculation time, the simulation model was prepared only in CSV. Selective simulation was conducted to verify the effects of non-condensable gas(NC gas) concentration, CSV internal pressure, and wall sub-cooling conditions. First, as a result of the internal flow of CSV, it was observed that downward flow due to condensation occurred surface of the vertical tube and upward flow occurred in the distant place. Natural convection occurred actively around the heat exchanger tube. Due to this rising and falling internal flow, natural circulation occurred actively around the heat exchanger tubes. Next, in order to check the performance of built-in condensation model using according to the non-condensable gas concentration, CSV internal flow and wall sub-cooling, the heat flux values were compared with the experimental results. On average, the results were underestimated with and error of about 25%. In addition, the influence of CSV internal pressure and wall sub-cooling was small, but when the condensate was highly generated due to the low non-condensable gas concentration, the error was large compared to the experimental values. This is considered to be due to the nature of the condensation model of the CFX code. However, in spite of the limitations of CFD, it is valid to use the built-in condensation model of CFD for PCCS performance prediction from a conservative perspective.

• Journal of Radiation Protection and Research
• /
• v.35 no.4
• /
• pp.172-177
• /
• 2010

The radioactive effluents from pressurized heavy water reactors (PHWRs) are relatively larger than those from pressurized water reactors (PWRs). Futhermore, radioactive effluents from PHWRs are released continuously. Thus, the discharge of radioactive effluents is strictly controlled. To do this, radiation detectors are installed at stacks of reactor buildings to monitor the concentration of radioactive effluents in real-time. Derived release limits (DRLs) of annual discharge are also set up for each radionuclide and effluents are rigidly controlled not to exceed those limits. In this paper, the discharge process of radioactive effluents, the standard for establishment of DRL and its methodology, and currents status for PHWRs were reviewed.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.38 no.4
• /
• pp.355-361
• /
• 2014

Steam explosions can be caused by fuel-coolant interactions resulting from failure of the external vessel cooling system in a new nuclear power plant. This can threaten the integrity of structures, including the nuclear reactor and the containment building. In the present study, an improved technique for analyzing the steam explosion phenomenon was proposed on the basis of previous research and was verified by simulations involving alumina experiments. Also, the improved analysis technique was applied to determine the pressure history of the reactor cavity in accordance with postulated failure locations. The results of the analysis revealed that the effects of vessel side failure are more serious than those of vessel bottom failure, with approximately 70% higher maximum pressure.

• Proceedings of the Korean Association of Geographic Inforamtion Studies Conference
• /
• 2007.10a
• /
• pp.169-175
• /
• 2007

고리, 월성, 울진, 영광 등4개 원전부지와 하나로 연구용 원자로 부지에 대해 방사성물질의 대기 중 누출사고 발생 시 대축척 전자지도와 연계한 사고정보 파악, 예상피해분석, 방재시설 및 소개정보 활용 등을 통해 중앙정부 및 지방자치단체가 방사능 물질 피해지역관리 및 신속하고 효율적인 주민대응조치 수립을 위한 의사 결정 지원할 수 있는 방사능방재 지리정보시스템 구축이 필요하다. 본 연구에서는 고리, 월성, 울진, 영광, 대전지역의 원자력 발전소 및 연구용 원자로 반경 40km이내 지역의 행정경계, 도로, 등고, 수계, 건물 등의 일반지형지물정보와, 비상계획구역 내 마을의 상세정보, 집결지, 대피소, 교통통제소, 환경방사능감시기, TLD등의 방재시설물 위치 및 관련 상세정보, 관공서, 경찰서, 소방서, 보건소, 학교, 병원 등의 방재관련 지형지물 위치 및 관련 상세정보, 원전부지 내 인구분포, 보유 차량 분포, 농작물 재배 현황, 축산물 재배현황 등의 방재관련 사회통계정보를 포함하는 공간 및 속성 데이터베이스는 구축하였다. 이를 기반으로 방사선 피폭영향 평가시스템(FADAS)의 예상평가결과를 전자지도 상에 표출하고, 이에 근거한 예상피해를 분석하며, 소개단계 대상 마을 검색 및 바람장 분석을 활용한 소개경로 제시 등을 통해 주민보호조치 의사결정을 지원하며, 사고대응 및 소개현황 정보를 관리하는 웹 기반의 원자력방재 지리정보시스템을 확대 개발하였다. 방재시설물 및 방재관련 지형지물, 방재관련 사회통계자료의 검색기능 및 실시간 원전 바람장 정보조회, 실시간 ERMS 수집정보 조회, 수치예보 정보 조회, 온라인DB관리 등의 확대 구현을 통해 사고대응조치 및 피해분석업무를 지원하였다. 본 연구를 통한 원자력방재 지리정보시스템 완성을 통해 방사능 비상시 중앙본부와 지역본부 및 유관기관 간에 지리정보와 연계한 정확한 사고정보 및 방재정보의 신속한 공유를 제공하고, 적절한 비상대응조치 의사결정 및 주민보조조치 수행을 지원하여 효율적인 사고지역 관리 및 인적 물적 자원의 피해를 최소화하는데 기여할 것으로 기대된다.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.2 no.1
• /
• pp.1-10
• /
• 1998

In the dynamic soil-structure interaction(SSI) analysis, numerous uncertain parameters are involved. They include the uncertainties in the definition of input motions, modeling of soil-structure interaction systems. analysis techniques, etc. This paper presents the results of parametric studies of the seismic responses of a reactor containment structure built on the viscoelastic layered soil. Among the numerous parameter, this study concentrates on the effects of definition point of the input motion, embedment of structure to the base soil, thickness of the top soil layer, and rigidity of the base soil. The substructure method using frequency independent impedances is adopted. The method is based on the mode superposition method in time domain using the composite modal damping values of th SSI system computed from the ratio of dissipated energy to the strain energy for each model. From the study results, the sensitivity of each parameter on the earthquake responses has been suggested for the practical application of the substructure method of SSI analysis.

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.19 no.1 s.71
• /
• pp.93-103
• /
• 2006

It is absolutely essential that safety assessment of the containment buildings during service life because containment buildings are last barrier to protect radioactive substance due to the accidents. Therefore, this study describes an enhanced degenerated shell finite element(FE) which has been developed for nonlinear FE analysis of reinforced concrete(RC) containment buildings with elasto-plastic material model. For the purpose of the material nonlinear analysis, Drucker-Prager failure criteria is adapted in compression region and material parameters which determine the shape of the failure envelop are derived from biaxial stress tests. Reissner-Mindlin(RM) assumptions are adopted to develop the degenerated shell FE so that transverse shear deformation effects is considered. However, it is found that there are serious defects such as locking phenomena in RM degenerated shell FE since the stiffness matrix has been overestimated in some situations. Therefore, shell formulation is provided in this paper with emphasis on the terms related to the stiffness matrix based on assumed strain method. Finally, the performance of the present shell element to analysis RC containment buildings is tested and demonstrated with several numerical examples. From the numerical tests, the present results show a good agreement with experimental data or other numerical results.

• Journal of the Korea Academia-Industrial cooperation Society
• /
• v.10 no.12
• /
• pp.3748-3754
• /
• 2009

An integral head loss test in a test apparatus was conducted to simulate chemical effects on a head loss across a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). The test was conducted during 30 days in the condition of a short spray, a long spray, and no materials with chemical effects. The result exhibited that the head loss was affected on amounts of the exposed materials according to spray conditions. XRD analysis of the collected precipitates showed that the precipitates were phosphate compounds. Comparison of the head loss with dissolved species concentration showed that high increase rate of the head loss resulted from the corrosion of aluminum and zinc but slow increase rate of the head loss resulted from the precipitates induced by Si, Mg, and Ca from leaching reaction at NUKON and concrete after passivation of metal specimens.

• KSCE Journal of Civil and Environmental Engineering Research
• /
• v.13 no.3
• /
• pp.59-67
• /
• 1993

The variations of the natural frequencies and the peak response acceleration at the top of prestressed concrete reactor building due to random variability and/or model uncertainty of structural parameters are studied. The results may be used as essential input parameters in seismic probabilistic risk assessment or seismic margin assessment of the reactor building. The sensitivity test of each structural parameter is first performed to determine the most influential parameter upon the natural frequency of structure model. Then Monte Carlo simulation technique is applied to evaluate the effect of parameter variation on the natural frequencies and the peak response acceleration. The acceleration time history is obtained by direct integration scheme. As the study results, it is found that the fundamental natural frequency and the peak response acceleration at the top of the building are most strongly affected by Young's modulus among the structural parameters, in which the value of mean plus one standard deviation obtained by probabilistic approach deviates up to about (+)12% from the result of deterministic method. Considering the uncertainty of flexural rigidity, the structural responses vary in range of (-)4%~(+)14%.