Search results for: LOCA

Authors: R. Mahmoodi, A. R. Zolfaghari

Abstract:

In nuclear power plants, loss of coolant from the primary system is the type of reduced removed capacity that is given most attention; such an accident is referred as Loss of Coolant Accident (LOCA). In the current study, investigation of shell and tube THTL heat exchanger behavior during LOCA is implemented by ANSYS CFX simulation software in both steady state and transient mode of turbulent fluid flow according to experimental conditions. Numerical results obtained from ANSYS CFX simulation show good agreement with experimental data of THTL heat exchanger. The results illustrate that in large break LOCA as short term accident, heat exchanger could not fast response to temperature variables but in the long term, the temperature of shell side of heat exchanger will be increase.

Keywords: shell-and-tube heat exchanger, shell-side, CFD, flow and heat transfer, LOCA

Procedia PDF Downloads 416

Authors: Khaleel Sami Hamdan, Dong-Eok Kim, Sang-Ki Moon

Abstract:

An appropriate model to predict the size of the droplets resulting from the break-up with the structures will help in a better understanding and modeling of the two-phase flow calculations in the simulation of a reactor core loss-of-coolant accident (LOCA). A droplet behavior impacting on a hot surface above the Leidenfrost temperature was investigated. Droplets of known size and velocity were impacted to an inclined plate of hot temperature, and the behavior of the droplets was observed by a high-speed camera. It was found that for droplets of Weber number higher than a certain value, the higher the Weber number of the droplet the smaller the secondary droplets. The COBRA-TF model over-predicted the measured secondary droplet sizes obtained by the present experiment. A simple model for the secondary droplet size was proposed using the mass conservation equation. The maximum spreading diameter of the droplets was also compared to previous correlations and a fairly good agreement was found. A better prediction of the heat transfer in the case of LOCA can be obtained with the presented model.

Keywords: break-up, droplet, impact, inclined hot plate, Leidenfrost temperature, LOCA

Procedia PDF Downloads 374

Authors: J. R. Wang, H. T. Lin, Y. Chiang, H. C. Chen, C. Shih

Abstract:

Kuosheng Nuclear Power Plant (NPP) is located on the northern coast of Taiwan. Its nuclear steam supply system is a type of BWR/6 designed and built by General Electric on a twin unit concept. First, the methodology of Kuosheng NPP SPU (Stretch Power Uprate) safety analysis TRACE/SNAP model was developed in this research. Then, in order to estimate the safety of Kuosheng NPP under the more severe condition, the SBO (Station Blackout) + LOCA (Loss-of-Coolant Accident) transient analysis of Kuosheng NPP SPU TRACE/SNAP model was performed. Besides, the animation model of Kuosheng NPP was presented using the animation function of SNAP with TRACE/SNAP analysis results.

Keywords: TRACE, safety analysis, BWR/6, severe accident

Procedia PDF Downloads 683

Authors: J. R. Wang, W. Y. Li, H. T. Lin, J. H. Yang, C. Shih, S. W. Chen

Abstract:

Fuel rod analysis program transient (FRAPTRAN) code was used to study the fuel rod performance during a postulated large break loss of coolant accident (LBLOCA) in Maanshan nuclear power plant (NPP). Previous transient results from thermal hydraulic code, TRACE, with the same LBLOCA scenario, were used as input boundary conditions for FRAPTRAN. The simulation results showed that the peak cladding temperatures and the fuel center line temperatures were all below the 10CFR50.46 LOCA criteria. In addition, the maximum hoop stress was 18 MPa and the oxide thickness was 0.003 mm for the present simulation cases, which are all within the safety operation ranges. The present study confirms that this analysis method, the FRAPTRAN code combined with TRACE, is an appropriate approach to predict the fuel integrity under LBLOCA with operational ECCS.

Keywords: FRAPTRAN, TRACE, LOCA, PWR

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Authors: J. Iqbal, A. Shah, M. Zeeshan

Abstract:

UBC-97, USNRC, chines origin code GB50011-2011 and site response spectrum was used to make comparison between them for Chashma site and most conservative one was selected and the USNRC was the most conservative one. The dynamic analysis of CHASNUPP-2 containment building was performed using SAP-2000 for dead load, live load (crane), pre stressed loads, wind load, temperature load, accidental pressure during LOCA, earthquake loads and the conservative response spectrum. After applying selected response spectrum on model, detail comparison was made against area of steal calculated from the analysis and the actually provided. Then prepared curve of area of steal vs. g value which shows that if the particular site was design on that spectrum that much steel needed for structural integrity.

Keywords: response spectrum, USNRC, LOCA, area of steel, structure integrity

Procedia PDF Downloads 659

Authors: Raphaelle Guillou, Matthieu Le Saux, Jean-Christophe Brachet, Thomas Guilbert, Elodie Rouesne, Denis Menut, Caroline Toffolon-Masclet, Dominique Thiaudiere

Abstract:

In some hypothetical accidental situations, such as during a Loss Of Coolant Accident (LOCA) in pressurized water reactors, fuel cladding tubes made of zirconium alloys can be exposed for a few minutes to steam at High Temperature (HT up to 1200°C) before being cooled and then quenched in water. Under LOCA-like conditions, the cladding undergoes a number of metallurgical changes (phase transformations, oxygen diffusion and growth of an oxide layer...) and is consequently submitted to internal stresses whose state evolves during the transient. These stresses can have an effect on the oxide structure and the oxidation kinetics of the material. They evolve during cooling, owing to differences between the thermal expansion coefficients of the various phases and phase transformations of the metal and the oxide. These stresses may result in the failure of the cladding during quenching, once the material is embrittled by oxidation. In order to progress in the evaluation of these internal stresses, X-ray diffraction experiments were performed in-situ under synchrotron radiation during HT oxidation and subsequent cooling on Zircaloy-4 sheet samples. First, structural evolutions, such as phase transformations, have been studied as a function of temperature for both the oxide layer and the metallic substrate. Then, internal stresses generated within the material oxidized at temperatures between 700 and 900°C have been evaluated thanks to the 2θ diffraction peak position shift measured during the in-situ experiments. Electron backscatter diffraction (EBSD) analysis was performed on the samples after cooling in order to characterize their crystallographic texture. Furthermore, macroscopic strains induced by oxidation in the conditions investigated during the in-situ X-ray diffraction experiments were measured in-situ in a dilatometer.

Keywords: APRP, stains measurements, synchrotron diffraction, zirconium allows

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Authors: Shengjun Zhang, Xu Cheng, Feng Shen

Abstract:

The passive containment cooling system (PCCS) is widely used in the advanced nuclear reactor in case of the loss of coolant accident (LOCA) and the main steam line break accident (MSLB). The internal heat exchanger is one of the most important equipment in the PCCS and its heat transfer characteristic determines the performance of the system. In this investigation, a theoretical model is presented for predicting the heat and mass transfer which accompanies condensation. The conduction through the liquid condensate is considered and the interface temperature is defined by iteration. The parameter in the correlation to describe the suction effect should be further determined through experimental data.

Keywords: non-condensable gas, condensation, heat transfer coefficient, heat and mass transfer analogy

Procedia PDF Downloads 321

Authors: J. R. Wang, H. C. Chen, C. Shih, S. W. Chen, J. H. Yang, Y. Chiang

Abstract:

In this study, we focus on the establishment of the analysis model for Maanshan PWR nuclear power plant (NPP) by using RADTRAD and SNAP codes with the FSAR, manuals, and other data. In order to evaluate the cumulative dose at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) outer boundary, Maanshan NPP RADTRAD/SNAP model was used to perform the analysis of the DBA LOCA case. The analysis results of RADTRAD were similar to FSAR data. These analysis results were lower than the failure criteria of 10 CFR 100.11 (a total radiation dose to the whole body, 250 mSv; a total radiation dose to the thyroid from iodine exposure, 3000 mSv).

Keywords: RADionuclide, transport, removal, and dose estimation (RADTRAD), symbolic nuclear analysis package (SNAP), dose, PWR

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Authors: Kai Chun Yang, Shao-Wen Chen, Jong-Rong Wang, Chunkuan Shih, Jung-Hua Yang, Hsiung-Chih Chen, Wen-Sheng Hsu

Abstract:

In this study, we build RADionuclide Transport, Removal And Dose Estimation/Symbolic Nuclear Analysis Package (SNAP/RADTRAD) model of Kuosheng Nuclear Power Plant which is based on the Final Safety Evaluation Report (FSAR) and other data of Kuosheng Nuclear Power Plant. It is used to estimate the radiation dose of the Exclusion Area Boundary (EAB), the Low Population Zone (LPZ), and the control room following ‘release from the containment’ case in Loss Of Coolant Accident (LOCA). The RADTRAD analysis result shows that the evaluation dose at EAB, LPZ, and the control room are close to the FSAR data, and all of the doses are lower than the regulatory limits. At last, we do a sensitivity analysis and observe that the evaluation doses increase as the intake rate of the control room increases.

Keywords: RADTRAD, radionuclide transport, removal and dose estimation, snap, symbolic nuclear analysis package, boiling water reactor, NPP, kuosheng

Procedia PDF Downloads 317

Authors: Jefri Draup, Antoine Ambard, Chi-Toan Nguyen

Abstract:

Zirconium alloys exhibit solid-state phase transformations under thermal loading. These can lead to a significant evolution of the microstructure and associated mechanical properties of materials used in nuclear fuel cladding structures. Therefore, the ability to capture effects of phase transformation on the material constitutive behavior is of interest during conditions of severe transient thermal loading. Whilst typical Avrami, or Johnson-Mehl-Avrami-Kolmogorov (JMAK), type models for phase transformations have been shown to have a good correlation with the behavior of Zircaloy-4 under constant heating rates, the effects of variable and fast heating rates are not fully explored. The present study utilises the results of in-situ high energy synchrotron X-ray diffraction (SXRD) measurements in order to validate the phase transformation models for Zircaloy-4 under fast variable heating rates. These models are used to assess the performance of fuel cladding structures under loss of coolant accident (LOCA) scenarios. The results indicate that simple Avrami type models can provide a reasonable indication of the phase distribution in experimental test specimens under variable fast thermal loading. However, the accuracy of these models deteriorates under the faster heating regimes, i.e., 100Cs⁻¹. The studies highlight areas for improvement of simple Avrami type models, such as the inclusion of temperature rate dependence of the JMAK n-exponent.

Keywords: accident, fuel, modelling, zirconium

Procedia PDF Downloads 119