• Annual Conference of KIPS
• /
• 2014.04a
• /
• pp.807-810
• /
• 2014

로봇 시스템의 제어 및 이를 이용한 환경 인식에는 많은 전자 광학 소자들이 사용되고 있다. 로봇 제어회로에 사용되고 있는 Si CMOS 공정의 CPU, ASIC, FPGA 소자는 고 선량의 감마선에 취약하다. 환경정보 수집용으로 로봇에 탑재되는 CMOS/CCD 카메라의 관측영상에는 고선량 감마선으로 인한 speckle (백색잡음, white noise) 들이 나타나며, 이들이 카메라의 관측성능을 저하시킨다. 후쿠시마 원자력발전소 사고와 같이 원자력시설에서 제어불능의 심각한 사고가 발생되면 고선량 감마선이 방출된다. 이러한 고선량 감마선방출은 사람에 의한 사고수습을 불가능하게 하며, 사고 수습을 위해서는 로봇의 활용이 불가피하다. 그러나, 방출되는 고선량 감마선의 세기(선량율)가 지나치게 높을 경우, 로봇 전자회로가 장애를 일으키기 때문에 로봇의 적절한 임무수행이 가능한 감마선 세기에 대한 고려가 필요하다. 본 논문에서는 고선량 감마선 환경하에서의 로봇 탑재 CCD/CMOS 카메라의 관측 성능을 고려하여 100 Gy/h 를 감마선 선량율 제한조건으로 설정한다. 그리고, 재 가동 승인심사를 받기 위해 일본의 원전 운영자들이 제시한 PWR (가압경수로) 원전의 중대사고 대책 적합성 평가문서에 나타난 노심용융개시 시점의 원자로 격납건물내 감마선 선량율 추이 계산결과를 활용하여 로봇의 대응시간을 계산하였다. 문서 (PDF) 에 표현된 감마선 선량율 추이 그래프를 영상 판독하여, 격납건물내 감마선 선량율이 100 Gy/h 제한조건에 도달하는 시간을 계산하였다. 이를 로봇의 대응시간으로 설정한다.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.35 no.3
• /
• pp.255-262
• /
• 2011

In a loss-of-coolantaccident, considerable debris may be generated and transported to the recirculation sump. The accumulation of debris will reduce the netpositivesuctionhead and threaten the safety of thenuclear power plant. Both NEI 04-07 and USNRC SER suggesteda CFD methodology. However, additional investigation is needed to consider the unique characteristics of nuclear power plants. The transport of the generated debris is strongly influenced by the break location and the plant characteristics, including the configuration.In this paper, a CFD methodology for blow-down transport evaluation is proposed and applied to an OPR1000 nuclear power plant. The results show that the percentage of small debris transported to the upper containment is 32%, which is 7% larger than the valuegiven in the NEI 04-07 baseline analysis. This result may be used as a point of reference in future analytical studies.

• Journal of Radiation Protection and Research
• /
• v.34 no.3
• /
• pp.129-136
• /
• 2009

During the maintenance period at Korean nuclear power plants, internal exposure of radiation workers occurred by the inhalation of $^{131}I$ released to the reactor building when primary system opened. The internal radioactivity of radiation workers contaminated by $^{131}I$ was measured using a whole body counter. Intake estimation and the calculation of committed effective dose were also conducted conforming to the guidance of internal dose assessments from publications of International Commission on Radiological Protection. Because the uptake and excretion of $^{131}I$ in a body occur quickly and $^{131}I$ is accumulated in the thyroid gland, the estimated intakes showed differences depending on the counting time after intake. In addition, since ICRP publications do not provide the intake retention fraction (IRF) for whole body of $^{131}I$, the IRF for thyroid was substitutionally used to calculate the intake and subsequently this caused more error in intake estimation. Thus, intake estimation and the calculation of committed effective dose were conducted by manual calculation. In this study, the IRF for whole body was also calculated newly and was verified. During this process, the estimated intake and committed effective dose were reviewed and compared using several computer codes for internal dosimetry.

• Journal of Radiation Protection and Research
• /
• v.14 no.1
• /
• pp.34-45
• /
• 1989

The habitability of a reactor control room in a French 1300 MWe P'4 type PWR has been evaluated through the exposure dose assessment for the reactor operator following a Loss of Coolant Accident. The main hypotheses adopted in this evaluation are based on the French Standard Safety Analysis Report. A simple computer program, named COREX(Control Room EXposure), was developed to calculate : the time-integrated radioactivities released from the reactor building, the volume factors for radionuclides concerned and the resulting time-integrated external whole body and internal thyroid doses to the reactor operators staying in the control room up to 30 days following the LOCA. The results obtained in this study, on the whole, well agreed with those proposed by the EDF(Electricite de France) except for the case of the whole body exposure, which was attributed to the differences in the volume factors for the radionuclides concerned.

• Magazine of the Korean Society of Agricultural Engineers
• /
• v.39 no.3
• /
• pp.72-82
• /
• 1997

In all inelastic deformations time rate effects are always present to some degree. Whether or not their exclusion has a significant influence on the prediction of the material behaviour depends upon several factors. In the study of structural components under static loading conditions at normal temperature it is accepted that time rate effects are generally not important. However metals, especially under high temperatures, exhibit simultaneously the phenomena of creep and viscoplasticity. In this study, elastoplastic and elasto-viscoplastic models include nonlinear geometrical effects were developed and several numerical examples are also included to verify the computer programming work developed here in this work. Comparisons of the calculated results, for the elasto-viscoplastic analysis of an internally pressurised thick cylinder under plane strain condition, have shown that the model yields excellent results. The results obtained from the numerical examples for an elasto-viscoplastic analysis of the Nuclear Reinforced Concrete Containment Structure(NRCCS) subjected to an incrementally applied internal pressure were summarized as follows : 1. The steady state hoop stress distribution along the shell layer of dome and dome wall junction part of NRCCS were linearly behave and the stress in interior surfaces was larger than that in exterior. 2.However in the upper part of the wall of NRCCS the steady state hoop stress in creased linearly from its inner to outer surfaces, being the exact reverse to the previous case of dome/dome-wall junction part. 3.At the lower part of wall of NRCCS, the linear change of steady state hoop stress along its wall layer began to disturb above a certain level of load increase.

• Annual Conference of KIPS
• /
• 2014.04a
• /
• pp.811-814
• /
• 2014

일본 후쿠시마 제일 원자력발전소의 대지진/쓰나미에 이은 원자로 건물 수소폭발 사고의 수습 과정에서 사용후 핵연료 저장조에 보관되어 있는 핵연료의 안전문제가 대두되었다. 사용후 핵연료의 잔열 성분을 냉각시키고, 그리고 사용후 핵연료가 방출하는 고선량 방사선을 차폐시키기 위해서 일정 깊이 이상의 수조에 사용후 핵연료를 저장한다. 사용후 핵연료 저장조에 냉각수 공급이 중단되면, 사용후 핵연료의 고유 잔열에 의해 수조의 물이 증발하여 수위가 감소하게 된다. 계속해서 냉각수 공급이 되지 않으면, 사용후 핵연료의 잔열은 증가하게 되고, 수조의 물은 비등하여 증발은 가속화 된다. 사용후 핵연료 저장조의 수위가 고갈되면 고선량의 감마선이 방출된다. 수조의 수위가 정상적일 경우 사용후 핵연료 저장조의 공기중 감마선 선량율은 0.15mSv/h 이다. 수조의 수위가 사용후 핵연료 상부 꼭대기를 기준으로 2m, 1m, 및 0m (핵연료 노출) 로 감소하게 되면, 사용후 핵연료 저장조의 공기중 감마선 선량율은 500mSv/h, 50Sv/h, 및 5kSv/h 로, 급격히 증가한다. 본 논문에서는 사용후 핵연료 저장조 감시카메라의 관측 성능을 평가하기 위해, 고성능 칼라 CCD 카메라에 대해서 1 kGy/h 의 고선량율로 감마선 조사실험을 수행하였다. 이에 대한 실험결과를 기술한다.

• Journal of the Korea institute for structural maintenance and inspection
• /
• v.23 no.3
• /
• pp.136-142
• /
• 2019

Nuclear power plant concrete structures are in contact with the coast, and durability due to chloride attack is very important because it is used as cooling water by taking seawater. For this purpose, a 3-year long-term saltwater immersion test was carried out to evaluate chloride ion diffusion coefficient and age apponent (m) The m values of the foundation with 4,000 class was 0.35 ~ 0.39, similar to KCI or ACI suggested values. essential service water constructions and tunnels of 5,000 class were 0.44 ~ 0.53 and 6,000 class, and 0.62 of reactor containment buildings were similar to the proposed values of FIB. As a result of the prediction of the service life with the measured age coefficient, all the safety related concrete structures of the nuclear power plants satisfied the service life of more than 60 years.

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.33 no.1
• /
• pp.63-72
• /
• 2020

The purpose of this study is to investigate the effects of the application of various numerical models and frequency contents of earthquakes on the performances of the reactor containment building (RCB) in a nuclear power plant (NPP) equipped with an advanced power reactor 1400. Two kinds of numerical models are developed to perform time-history analyses: a lumped-mass stick model (LMSM) and a full three-dimensional finite element model (3D FEM). The LMSM is constructed in SAP2000 using conventional beam elements with concentrated masses, whereas the 3D FEM is built in ANSYS using solid elements. Two groups of ground motions considering low- and high-frequency contents are applied in time-history analyses. The low-frequency motions are created by matching their response spectra with the Nuclear Regulatory Commission 1.60 design spectrum, whereas the high-frequency motions are artificially generated with a high-frequency range from 10Hz to 100Hz. Seismic responses are measured in terms of floor response spectra (FRS) at the various elevations of the RCB. The numerical results show that the FRS of the structure under low-frequency motions for two numerical models are highly matched. However, under high-frequency motions, the FRS obtained by the LMSM at a high natural frequency range are significantly different from those of the 3D FEM, and the largest difference is found at the lower elevation of the RCB. By assuming that the 3D FEM approximates responses of the structure accurately, it can be concluded that the LMSM produces a moderate discrepancy at the high-frequency range of the FRS of the RCB.

• Magazine of the Korean Society of Agricultural Engineers
• /
• v.39 no.4
• /
• pp.106-113
• /
• 1997

Dynamic loading of structures often causes excursions of stresses well into the inelastic range, and the influence of the geometric changes on the dynamic response is also significant in many cases. Therefore, both material and geometric nonlinearity effects should be considered in case that a dynamic load acts on the structure. A structure in a nuclear power plant is a structure of importance which puts emphasis on safety. A nuclear container is a pressure vessel subject to internal pressure and this structure is constructed by a reinforced concrete or a pre-stressed concrete. In this study, the material nonlinearity effect on the dynamic response is formulated by the elasto-viscoplastic model highly corresponding to the real behavior of the material. Also, the geometrically nonlinear behavior is taken into account using a total Lagrangian coordinate system, and the equilibrium equation of motion is numerically solved by a central difference scheme. The constitutive relation of concrete is modeled according to a Drucker-Prager yield criterion in compression. The reinforcing bars are modeled by a smeared layer at the location of reinforcements, and the steel layer model under Von Mises yield criteria is adopted to represent an elastic-plastic behavior. To investigate the dynamic response of a nuclear reinforced concrete containment structure, the steel-ratios of 0, 3, 5 and 10 percent, are considered. The results obtained from the analysis of an example were summarized as follows 1. As the steel-ratio increases, the amplitude and the period of the vertical displacements in apex of dome decreased. The Dynamic Magnification Factor(DMF) was some larger than that of the structure without steel. However, the regular trend was not found in the values of DMF. 2. The dynamic response of the vertical displacement and the radial displacement in the dome-wall junction were shown that the period of displacement in initial step decreased with the steel-ratio increases. Especially, the effect of the steel on the dynamic response of radial displacement disapeared almost. The values of DMF were 1.94, 2.5, 2.62 and 2.66, and the values increased with the steel-ratio. 3. The characteristics of the dynamic response of radial displacement in the mid-wall were similar to that of dome-wall junction. The values of DMF were 1.91, 2.11, 2.13 and 2.18, and the values increased with the steel-ratio. 4. The amplitude and the period of the hoop-stresses in the dome, the dome-wall junction, and the mid-wall were shown the decreased trend with the steel-ratio. The values of DMF were some larger than those of the structure without steel. However, the regular trend was not found in the values of DMF.