• Nuclear Engineering and Technology
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• v.35 no.2
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• pp.154-164
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• 2003

The explicit consideration of different pressurization rates in estimating the probabilities of containment failure modes has a profound effect on the confidence of containment performance evaluation that is so critical for risk assessment of nuclear power plants. Except for the sophisticated NUREG-1150 study, many of the recent containment performance analyses (through Level 2 PSAs or IPE back-end analyses) did not take into account an explicit distinction between slow and fast pressurization in their analyses. A careful investigation of both approaches shows that many of the approaches adopted in the recent containment performance analyses exactly correspond to the NUREG-1150 approach for the prediction of containment failure mode probabilities in the presence of fast pressurization. As a result, it was expected that the existing containment performance analysis results would be subjected to greater or less conservatism in light of the ultimate failure mode of the containment. The main purpose of this paper is to assess potential conservatism of a plant-specific containment performance analysis result in light of containment failure mode probabilities.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2005.04a
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• pp.183-189
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• 2005

Nuclear containment structure is the last barrier for being secure from any nuclear power plant accident. Even though the safety requirements of nuclear power plant have been focused on removing accidental situations, nuclear containment structure must reserve the sufficient resisting capacity to any accident because it works as the last barrier. The acceptable nuclear containment structure makes possible to limit the effect of internal accidents and to avoid radioactive release. In this study, to conduct the numerical analysis for the structural safety of a containment structure, loss of coolant accident (LOCA) is considered as the basic accidental load, and Wolsong containment structure is considered as a target structure. The CANDU containment structure, such as Wolsong containment structure, is a prestressed concrete shell structure which has dome and is reinforced with bonded tendons. The evaluation of ultimate pressure capacity was conducted by nonlinear analysis of a prestressed concrete containment structure.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.125-137
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• 2019

The purpose of this paper is to study the source term behavior after severe accidents by using a semi-kinetic model for simulation and calculation of in-containment activity. The reactor containment specification and the safety features of the containment under different accident conditions play a great role in evaluating the in-containment activity. Assuming in-vessel and instantaneous release of radioactivity into the containment, the behavior of in-containment isotopic activity is studied for noble gasses (Kr and Xe) and the more volatile elements of iodine, cesium, and aerosols such as Te, Rb and Sr as illustrative examples of source term release under LOCA conditions. The results of the activity removal mechanisms indicates that the impact of volumetric leakage rate for noble gasses is important during the accident, while the influence of deposition on the containment surfaces for cesium, mainly iodine isotopes and aerosol has the largest contribution in removal of activity during evolution of the accident.

• Nuclear Engineering and Technology
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• v.30 no.1
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• pp.26-39
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• 1998

The containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5 is modified to incorporate the passive containment cooling models. The correlations are selected from the existing experimental heat transfer correlations to model the natural and mixed convection in annular space between the containment shell and the shield building. The evaporative heat transfer of the water film on the outer shell of the containment is modeled using the correlations derived from the analogy between the heat and mass transfer. The modified code is applied to the Ap600 containment transient analysis for the model verification and the results are compared to the results of GOTHIC calculation done by Westinghouse. Also, d series of parametric sensitivity studies of heat transfer correlations, water film ratio and delay time of the wet cooling on the containment peak pressure and temperature following LOCA are performed for the containment of 1000MWe passive plant, KP1000.

• Nuclear Engineering and Technology
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• v.33 no.5
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• pp.468-482
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• 2001

As a process of Direct Containment Heating (DCH) issue resolution for Korean Standard Nuclear Power Plants (KSNPs), a containment load/strength assessment with two different approaches, the probabilistic and the deterministic, was performed with all plant-specific and phenomena-specific data. In case of the probabilistic approach, the framework developed to support the Zion DCH study, Two-Cell Equilibrium (TCE) coupled with Latin Hypercubic Sampling (LHS), provided a very efficient tool to resolve DCH issue. In case of the deterministic approach, the evaluation methodology using the sophisticated mechanistic computer code, CONTAIN 2.0 was developed, based on findings from DCH-related experiments or analyses. For three bounding scenarios designated as Scenarios V, Va, and VI, the calculation results of TCE/LHS and CONTAIN 2.0 with the conservatism or typical estimation for uncertain parameters, showed that the containment failure resulted from DCH loads was not likely to occur. To verify that these two approaches might be conservative , the containment loads resulting from typical high-pressure accident scenarios (SBO and SBLOCA) for KSNPs were also predicted. The CONTAIN 2.0 calculations with boundary and initial conditions from the MAAP4 predictions, including the sensitivity calculations for DCH phenomenological parameters, have confirmed that the predicted containment pressure and temperature were much below those from these two approaches, and, therefore, DCH issue for KSNPS might be not a problem.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.358-363
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• 1995

A study on passive cooling systems for concrete containment of advanced pressurized water reactors has been performed. The proposed passive containment cooling system (PCCS) consist of (1) condenser units located inside containment, (2) a steam condensing pool outside containment at higher elevation, and (3) downcommer/riser piping systems which provide coolant flow paths. During an accident causing high containment pressure and temperature, the steam/air mixture in containment is condensed on the outer surface of condenser tubes transferring the heat to coolant flowing inside tubes. The coolant transfers the heat to the steam condensing pool via natural circulation due to density difference. This PCCS has the following characteristic: (1) applicable to concrete containment system, (2) no limitation in plant capacity expansion, (3) efficient steam condensing mechanism (dropwise or film condensation at the surface of condenser tube), and (4) utilization of a fully passive mechanism. A preliminary conceptual design work has been done based on steady-state assumptions to determine important design parameter including the elevation of components and required heat transfer area of the condenser tube. Assuming a decay power level of 2%, the required heat transfer area for 1,000MWe plant is assessed to be about 2,000 ㎡ (equivalent to 1,600 of 10 m-long, 4-cm-OD tubes) with the relative elevation difference of 38 m between the condenser and steam condensing pool and the riser diameter of 0.62 m.

• Structural Engineering and Mechanics
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• v.53 no.1
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• pp.73-87
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• 2015

In case of aircraft impact on nuclear containment structures, the initial kinetic energy of the aircraft is transferred and absorbed by the outer containment, may causing either complete or partial failure of containment structure. In the present study safety analysis of BWR Mark III type containment has been performed. The total height of containment is 67 m. It has a circular wall with monolithic dome of 21m diameter. Crash analysis has been performed for fighter jet Phantom F4. A normal hit at the crown of containment dome has been considered. Numerical simulations have been carried out using finite element code ABAQUS/Explicit. Concrete Damage Plasticity model have been incorporated to simulate the behaviour of concrete at high strain rate, while Johnson-Cook elasto-visco model of ductile metals have been used for steel reinforcement. Maximum deformation in the containment building has reported as 33.35 mm against crash of Phantom F4. Deformations in concrete and reinforcements have been localised to the impact region. Moreover, no significant global damage has been observed in structure. It may be concluded from the present study that at higher velocity of aircraft perforation of the structure may happen.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2477-2487
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• 2021

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.

• Journal of the Korea Safety Management & Science
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• v.7 no.4
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• pp.85-99
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• 2005

The main objective of this paper is to search whether containment vessel's best pressure may increase until how long when loss of coolant accident (LOCA) happened in containment vessel of Ulchin nuclear power plant 1 and 2. Another goal of this research is to find the influential factors that increase containment vessel pressure. Model for this research is Ulchin nuclear power plant 1 with 10 cycles. Data were collected by simulator of Ulchin nuclear power plant 1 and design of experiment was used for data analysis. For the experiment, seven factors that are going to influence in containment vessel pressure were chosen. It was found that fatter which influences in early rise of containment vessel pressure after LOCA is only explosion size. Also, containment vessel's best pressure (3.74 bar.a) was much lower than limit (4.86 bar.a) of FSAR (Final Safety Analysis Report).

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3130-3139
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• 2022

Direct containment heating (DCH) is one of the potential factors leading to early containment failure. DCH is closely related to safety analysis and containment performance evaluation of nuclear power plants. In this study, a DCH prediction program was developed to analyze the DCH loads of containment vessel. The phenomenological model of debris dispersal, metal oxidation reaction, debris-atmospheric heat transfer and hydrogen jet burn was established. Code assessment was performed by comparing with several separate effect tests and integral effect tests. The comparison between the predicted results and experimental data shows that the program can predict the key parameters such as peak pressure, temperature, and hydrogen production in containment well, and for most comparisons the relative errors can be maintained within 20%. Among them, the prediction uncertainty of hydrogen production is slightly larger. The analysis shows that the main sources of the error are the difference of time scale and the oxidation of cavity debris.