• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.3
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• pp.347-354
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• 2020

Site characterization for decommissioning Kori Unit 1 is ongoing in South Korea after 40 years of successful operation. Kori Unit 1's containment building is assumed to be mostly radioactively contaminated, and therefore radiation exposure management and detailed contamination investigation are required for decommissioning and dismantling it safely. In this study, site-specific Derived Concentration Guideline Levels (DCGLs) were derived using the residual radioactivity risk evaluation tool, RESRAD-BUILD code. A conceptual model of containment building for Kori Unit 1 was set up and limited occupational worker building inspection scenario was applied. Depending on the source location, the maximum contribution source and exposure pathway of each radionuclide were analyzed. The contribution of radionuclides to dose and exposure pathways, by source location, is expected to serve as basic data in the assessment criteria of survey areas and classification of impact areas during further decommissioning and decontamination of sites.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2002.04a
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• pp.377-384
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• 2002

In this palter, a finite element technique is applied to both reinforced concrete and prestressed concrete containment vessels to predict the ultimate pressure capacity of the vessels subjected to internal pressure due to accident. The so-called volume-control technique is utilized to control the change in volume enclosed by the cylindrical containment vessels and layered shell elements equipped with a pressure node is utilizing to model the PSC vessels. The finite element analysis is carried out to obtain both global and local failure behavior of prestressed concrete nuclear containment vessels. nalytical results are verified by comparison with experimental data.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.188.1-188
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• 1999

• Journal of the Korean Society of Safety
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• v.35 no.3
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• pp.96-104
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• 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
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• v.38 no.4
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• pp.189-193
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• 2013

Annual dose on the containment building wall of the interim storage facility at normal condition was calculated to estimate the dose rate transition of the facility of PWR spent nuclear fuel. In this study, source term was generated by ORIGEN-ARP with 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facility and the containment building and radiation shielding evaluations were conducted by MCNP code depending on the distance between the wall and the facility in the building. In the case of the centralized storage system, the distance required for the annual dose rate limit from 10CFR72 was estimated to be 50 m.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2696-2707
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• 2021

The purpose of this study is to investigate the efficiency of various structural modeling schemes for evaluating seismic performances and fragility of the reactor containment building (RCB) structure in the advanced power reactor 1400 (APR1400) nuclear power plant (NPP). Four structural modeling schemes, i.e. lumped-mass stick model (LMSM), solid-based finite element model (Solid FEM), multi-layer shell model (MLSM), and beam-truss model (BTM), are developed to simulate the seismic behaviors of the containment structure. A full three-dimensional finite element model (full 3D FEM) is additionally constructed to verify the previous numerical models. A set of input ground motions with response spectra matching to the US NRC 1.60 design spectrum is generated to perform linear and nonlinear time-history analyses. Floor response spectra (FRS) and floor displacements are obtained at the different elevations of the structure since they are critical outputs for evaluating the seismic vulnerability of RCB and secondary components. The results show that the difference in seismic responses between linear and nonlinear analyses gets larger as an earthquake intensity increases. It is observed that the linear analysis underestimates floor displacements while it overestimates floor accelerations. Moreover, a systematic assessment of the capability and efficiency of each structural model is presented thoroughly. MLSM can be an alternative approach to a full 3D FEM, which is complicated in modeling and extremely time-consuming in dynamic analyses. Specifically, BTM is recommended as the optimal model for evaluating the nonlinear seismic performance of NPP structures. Thereafter, linear and nonlinear BTM are employed in a series of time-history analyses to develop fragility curves of RCB for different damage states. It is shown that the linear analysis underestimates the probability of damage of RCB at a given earthquake intensity when compared to the nonlinear analysis. The nonlinear analysis approach is highly suggested for assessing the vulnerability of NPP structures.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.493-498
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• 1996

A PC window-based computer code, CONPAS(CONtainment Performance Analysis System), has been developed to integrate the numerical, graphical and results-operation aspects of Level 2 probabilistic safety assessments (PSA) for nuclear power plants automatically. As a main logic for accident progression analysis, it employs a concept of the small containment phenomenological event tree(CPET) helpful to trace out visually individual accident progressions and of the large supporting event tree(LSET) for its detailed quantification. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, and sensitivity analysis, reporting aspects including tabling and graphic, and user-friend interface.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.669-674
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• 1996

The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction(MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass, The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 1997.10a
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• pp.90-97
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• 1997

The purpose of this study is to evaluate the seismic response characteristics of Wolsong nuclear power plant (NPP) structures for the Kyeongju earthquake(ML=4.3) occurred on June 26, 1997. The seismograms are obtained from five accelerographs of nuclear power plant at Wolsong, Kyeongbuk. The distance from the epicenter is about 25km. The peak acceleration (PA) due to the earthquake is 0.0235g, which is far lower value than that of design basis earthquake(DBE). The PA at the containment wall is about twice as large as that at free field. Also, the higher the accelerograph is located in, the larger the PA is measured to be From the response spectrum analysis, the dominant frequency of the response is close to 4 Hz, which is similar to the free field is poor because of contamination by high frequency waves as a result of reflection and diffraction between ground and NPP structure. We are of opinion that the accelerograph at the free field should be moved approximately twice the building dimension away from the containment structure.

• Nuclear Engineering and Technology
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• v.28 no.6
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• pp.551-562
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• 1996

This paper introduces the five-step methodology for identifying information needs and assessing instrument availability during the course of severe accidents in nuclear power plants. The methodology is applied to the Yonggwang (YGN) 3&4 to shed light on accident management. It constructs three safety objective trees to prevent the reactor vessel failure, to prevent the containment failure, and to mitigate the fission product release from the containment. The study assesses information needs and instrument availability under severe conditions for preventing the reactor vessel failure of YGN 3&4, and recommends additional instrument that m8y prove to be of vital importance in managing the accident.