• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.257-264
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• 1996

The conventional Equipment Environment Qualification (EEQ) envelope is developed based on the containment responses during the design basis events. The Safety Depressurization System (SDS) design without In-containment Refueling Water Storage Tank (IRWST) adopted in the Ulchin 3&4 challenges the conventional EEQ envelope during long term Feed and Bleed (F&B) operation due to the direct discharge of high mass and energy into the containment. Therefore, it is necessary to confirm that the containment pressure and temperature history during the long term F&B operation does not violate the conventional EEQ envelope. However, this subject has never been quantitatively assessed before. To investigate the success path of long term F&B operation this paper analyzes the thermal hydraulic response of the containment and Reactor Coolant System (RCS) until the completion of depressurization and cooldown of RCS into Shutdown Cooling System (SCS) entry condition. It is found that the SCS entry condition can be reached within 6 hours without violating the EEQ curve by proper operation of SDS valves, High Pressure Safety Injection (HPSI) pumps and active Containment Heat Removal System (CHRS). The suggested strategy not only demonstrates the feasibility of long term F&B operation but also can be utilized in the preparation of Emergency Procedure Guidelines (EPGs)

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.249-260
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• 2012

Main focus of the combined nuclear research activities at Aachen University (RWTH) and the Research Center J$\ddot{u}$lich (J$\ddot{U}$LICH) is the experimental and analytical investigation of containment phenomena and processes. We are deeply convinced that reliable simulations for operation, design basis and beyond-design basis accidents of nuclear power plants need the application of so-called lumped-parameter (LP) based codes as well as computational fluid dynamics (CFD) codes in an indispensable manner. The LP code being used at our institutions is the GRS code COCOSYS and the CFD tool is ANSYS CFX mostly used in German nuclear research. Both codes are applied for safety analyses especially of beyond design accidents. Focal point of the work is containment thermal-hydraulics, but source term relevant investigations for aerosol and iodine behavior are performed as well. To increase the capability of COCOSYS and CFX detailed models for specific features, e.g. recombiner behavior including chimney effect, building condenser, and wall condensation are developed and validated against facilities at different scales. The close connection between analytical and experimental activities is notable and identifying feature of the RWTH/J$\ddot{U}$LICH activities.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1742-1756
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• 2023

The hydrogen behavior in a nuclear containment vessel is a significant issue when discussing the potential of hydrogen combustion during a severe accident. After the Fukushima-Daiichi accident in Japan, we have investigated in-depth the hydrogen transport mechanisms by utilizing experimental and numerical approaches. Computational fluid dynamics is a powerful tool for better understanding the transport behavior of gas mixtures, including hydrogen. This paper describes a Large-eddy simulation of gas mixing driven by a high-buoyancy flow. We focused on the interaction behavior of heat and mass transfers driven by the horizontal high-buoyant flow during density stratification. For validation, the experimental data of the Containment InteGral effects Measurement Apparatus (CIGMA) facility were used. With a high-power heater for the gas-injection line in the CIGMA facility, a high-temperature flow of approximately 390 ℃ was injected into the test vessel. By using the CIGMA facility, we can extend the experimental data to the high-temperature region. The phenomenological discussion in this paper helps understand the heat and mass transfer induced by the high-buoyancy flow in the containment vessel during a severe accident.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2628-2641
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• 2023

Nuclear safety is the lifeline for the development and application of nuclear energy. In severe accidents of pressurized water reactor (PWR), aerosols, as the main carrier of fission products, are suspended in the containment vessel, posing a potential threat of radioactive contamination caused by leakage into the environment. The gas-phase aerosols suspended in the containment will settle onto the wall or sump water through the natural deposition mechanism, thereby reducing atmospheric radioactivity. Aiming at the low accuracy of the aerosol model in the ISAA code, this paper improves the natural deposition model of aerosol in the containment. The aerosol dynamic shape factor was introduced to correct the natural deposition rate of non-spherical aerosols. Moreover, the gravity, Brownian diffusion, thermophoresis and diffusiophoresis deposition models were improved. In addition, ABCOVE, AHMED and LACE experiments were selected to validate and evaluate the improved ISAA code. According to the calculation results, the improved model can more accurately simulate the peak aerosol mass and respond to the influence of the containment pressure and temperature on the natural deposition rate of aerosols. At the same time, it can significantly improve the calculation accuracy of the residual mass of aerosols in the containment. The performance of improved ISAA can meet the requirements for analyzing the natural deposition behavior of aerosol in containment of advanced PWRs in severe accident. In the future, further optimization will be made to address the problems found in the current aerosol model.

• Proceedings of the Korea Concrete Institute Conference
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• 2000.04a
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• pp.805-810
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• 2000

The purpose of this study is to predict long-term structural safety on the Yonggwang Unit 3 prestressed concrete containment building. The aging-related degradations of its main structural materials are investigated and the effects of the property variation of time-dependent materials on the structural behavior of containment building are also assessed through the analysis on the ultimate pressure capacity. The nonlinear finite element analyses for both the design criteria condition a the present aging condition are conducted to assess the present structural capacity of the containment building As a result, it is verified that the structural capacity of the Yonggwang Unit 3 containment building under the present aging condition is judged to be still rugged. n addition, the sensitivity of the ultimate pressrue capacity of containment building according to th degradation levels of the structural materials are assessed. Finally, it is showed that the sensitivity levels are in the order of the tendon, rebar and concrete in case of individual material degradations, and the tendon-rebar, tendon-concrete and rebar-concrete in case of coupled material degradations.

• 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.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2006.03a
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• pp.597-604
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• 2006

The seismic safety of the prestressed concrete containment building was evaluated by the seismic fragility analysis based on the nonlinear dynamic time-history analyses. Four kinds of earthquake ground motions were used for the seismic fragility analysis of the containment building to consider the potential earthquake hazard. The conventional seismic fragility analysis of the safety related structures in nuclear pouter plant have been performed by using the linear elastic analysis results for the seismic design. In this study, the displacement based seismic fragility analysis method was proposed.

• Journal of the Korea institute for structural maintenance and inspection
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• v.8 no.3
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• pp.225-233
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• 2004

In this study, the seismic safety of containment building is assessed using response surface method. The structural analyses considering random variables such as load, resistance and analysis by ABAQUS are performed to obtain the structural response. The structural response is represented by polynomial of random variables, and the reliability analysis is performed by Level II method. Drucker-Prager failure criterion is applied as limit state function to take bi-axial stress states into account in the concrete. The lifetime probability of failure is evaluated by considering the lifetime of containment building, the annual occurrence rate of earthquake and the conditional probability of failure. Also the sensitivity analysis on the selection of sampling points is performed to obtain the steady results from response surface method.

• Computers and Concrete
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• v.3 no.1
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• pp.1-15
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• 2006

The seismic safety of reinforced concrete containment building can be evaluated by probabilistic analysis considering randomness of earthquake, which is more rational than deterministic analysis. In the safety assessment of earthquake-resistant structures by the deterministic theory, it is not easy to consider the effects of random variables but the reliability theory and random vibration theory are useful to assess the seismic safety with considering random effects. The reliability assessment of reinforced concrete containment building subjected to earthquake load includes the structural analysis considering random variables such as load, resistance and analysis method, the definition of limit states and the reliability analysis. The reliability analysis procedure requires much time and labor and also needs to get the high confidence in results. In this study, random vibration analysis of containment building is performed with random variables as earthquake load, concrete compressive strength, modal damping ratio. The seismic responses of critical elements of structure are approximated at the most probable failure point by the response surface method. The response surface method helps to figure out the quantitative characteristics of structural response variability. And the limit state is defined as the failure surface of concrete under multi-axial stress, finally the limit state probability of failure can be obtained simply by first-order second moment method. The reliability analysis for the multiaxial strength limit state and the uniaxial strength limit state is performed and the results are compared with each other. This study concludes that the multiaxial failure criterion is a likely limit state to predict concrete failure strength under combined state of stresses and the reliability analysis results are compatible with the fact that the maximum compressive strength of concrete under biaxial compression state increases.