• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1781-1788
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• 2022

The containment building Kori Unit 1 may require sequential steps for full decommissioning. This study assumes that the containment building is to be used as an auxiliary building that handles nuclear power systems and materials during decommissioning before conversion into a greenfield. Through the derivation of guidelines and dose evaluation, it was confirmed whether the radiation workers were satisfied with the ALARA decision. The specific modeling of the external radiation exposure was performed based on the facility investigation procedures. The external radiation specific derived concentration guideline levels (DCGLs) for radiation workers in containment building were obtained using the RESRAD-BUILD code and were applied to the VISIPLAN 3D ALARA Planning Tool code to calculate the working dose and check worker safety. The derivation of site-specific and realistic DCGLs and dose evaluation via 3D modeling can contribute to the scenario development for the decommission and remediation of containment building.

• Journal of the Korean Institute of Gas
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• v.12 no.1
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• pp.36-41
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• 2008

Using the finite element analysis, this paper presents the strength safety of a side wall of an outer tank and a roof structures in a full containment LNG storage tank system. The outer tank structure in which is constructed with a prestressed concrete is forced by internal hydrostatic and hydrodynamic pressures of a leaked LNG and an external wind pressure including a typhoon one. The FEM computed results show that the ring beam between a side wall of an outer tank and a roof structure supports most of the internal and the external loads. This means that the design point of the outer tank system is a ring beam structure and the other one is a center part of the roof structure. In this FE analysis model of a full containment LNG tank system, the outer tank and the roof structures are safe for the given combined loads such as an internal leaked LNG pressure and an external typhoon pressure.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2859-2865
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• 2021

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.

• 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.

• KSTLE International Journal
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• v.7 no.2
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• pp.45-50
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• 2006

In this paper, the leakage safety of prestressed concrete structure including the insulation panels has been analyzed using a finite element analysis just after a collapse of 9% nickel inner tank. This FEM study shows that the outer tank may contain the leaked cryogenic liquid for the time being until the primary pump in the inner tank transports stored cryogenic liquids to the nearest LNG storage tank before the outer tank is demolished. This means that the total tank thickness from the insulation panel to the outer tank system safely may retain the leaked cryogenic fluids. The FE computed results indicate that the current structure in a full containment tank is obviously enough to securing the leak-proof safety of the tank system with two primary pumps.

• Journal of the Korean Institute of Gas
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• v.11 no.4
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• pp.85-90
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• 2007

This paper presents the finite element analysis on the strength safety of a full containment LNG storage tank system with damping safety structures. For the FEM analysis of the inner tank, the combined loads in which are related to a hydrostatic pressure, a cryogenic temperature load, BOG pressure, LNG weight, and a sinking force at the comer of the inner tank have been applied to the inner tank structure. The FEM computed results show that the conventional inner tank is safe for the given combined loads, but the damping safety structure such as compressive springs may be more useful structures to increase the safety of the tank system. The increased stiffness and the appropriate position of the springs are very important design parameters for increasing the damping strength safety of the tank system.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.767-790
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• 2013

MAAP4 is a computer code that can simulate the response of a light water reactor power plant during severe accident sequences, including actions taken as part of accident management. The code quantitatively predicts the evolution of a severe accident starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure. Furthermore, models are included in the code to represent the actions that could mitigate the accident by in-vessel cooling, external cooling of the reactor pressure vessel, or cooling the debris in containment. A key element tied to using a code like MAAP4 is an uncertainty analysis. The purpose of this paper is to present a MAAP4 based analysis to examine the sensitivity of a key parameter, in this case hydrogen production, to a set of model parameters that are related to a Level 2 PRA analysis. The Level 2 analysis examines those sequences that result in core melting and subsequent reactor pressure vessel failure and its impact on the containment. This paper identifies individual contributors and MAAP4 model parameters that statistically influence hydrogen production. Hydrogen generation was chosen because of its direct relationship to oxidation. With greater oxidation, more heat is added to the core region and relocation (core slump) should occur faster. This, in theory, would lead to shorter failure times and subsequent "hotter" debris pool on the containment floor.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.290-298
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• 1996

Direct Containment Heating, DCH, is supposed to occur during a core melt-down accident if the primary system pressure is still high at the time of vessel breach in a Nuclear Power Plant (NPP). In this case, DCH is considered to be one of very important severe phenomena during postulated severe accident scenario because of the fast heat transfer rate to atmosphere and the sharp pressure increase in a containment. To reduce the effect of this DCH phenomena, the capture volume wes designed at Ulchin NPP units 3 and 4. But, the effect of this has not been studied extensively. This work consists of experimental and numerical analyses of the effects of capture volume in the cavity on DCH phenomena. The experimental model is a 1/30 scaled-down model of Ulchin NPP units 3 and 4. We used three types of capture volumes to investigate the effect of size. Numerical analysis using CONTAIN 1.2 is performed with the correlation for the dispersed fraction of molten corium from the cavity into the containment derived from the experimental data to examine the effect of capture volume on DCH phenomena in full scale of Ulchin NPP units 3 and 4.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3677-3691
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• 2023

Since containment building is the major shielding structure to ensure safety of nuclear power plant, the structural behavior and ultimate pressure capacity of containments must be studied in depth. This paper addresses ambiguous issue of determining free-field position for liner failure by suggesting an expanded free-field region and comparing internal pressure capacities obtained by test data, conservative assumption and suggested free-field region. For this purpose, a practical approach to determine the free-field position for the evaluation of liner tearing is carried out. The maximum principal strain histories versus internal pressure capacities among different free-field positions at various azimuths and elevations are compared with those at the equipment hatch as a conservative assumption. The comparison shows that there are considerable differences in the internal pressure capacity at liner failure within the expanded free-field region compared to the vicinity of the equipment hatch. Additionally, this study proposes an approximate correlation with conservative factors by considering the expanded free-field ranges and material characteristics to determine realistic failure criteria for liner. The applicability of the proposed correlation is demonstrated by comparing the internal pressure capacities of full-scale containment buildings following liner failure criteria according to RG 1.216 and an approximate correlation.

• Journal of Energy Engineering
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• v.21 no.4
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• pp.339-345
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• 2012

RMembrane LNG storage tanks have been recently investigated to replace full-containment LNG storage tanks because of safety and cost aspects. Quantitative Risk Analysis (QRA) and Finite Element Method (FEM) were used to evaluate safety of membrane LNG storage tanks. In this study, structural safety evaluation results via FEM analysis showed that both membrane type and full-containment type cryogenic LNG storage tanks with 140,000 $m^3$ capacity were equivalently safe in terms of strength safety and leakage safety of a storage tank system. Also, Fault Tree Analysis (FTA) was used to improve the safety of membrane LNG storage tanks and membrane LNG tanks were modified by adding three safety equipments: impact absorber structure for the low part of the membrane, the secondary barrier to diminish the thermal stress of the corner part of the outer tank, and a pump catcher in case of falling of a pump. Consequently, the safety of the modified membrane LNG storage tanks were proved to be equivalent to that of full-containment LNG storage tanks.