ATWS severe accident analysis in the loss of flow scenario using the MELCOR code in Bushehr nuclear Power Plant

Kerntechnik ◽  
2019 ◽  
Vol 84 (2) ◽  
pp. 123-130 ◽  
Author(s):  
F. Janbazi ◽  
M. R. Rezaie
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Accident Analysis

Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code

2014 ◽  
Author(s):  
Atsuo Takahashi ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Masanori Naitoh
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Accident Analysis ◽  
Unknown Boundary ◽  
Long Time ◽  
Fukushima Daiichi ◽  
Nuclear Power Plant Unit

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.

Deposition Fractions of Fission Product and Heavy Elements on Soil in Fukushima Dai-Ichi Nuclear Power Plant Accident

2013 ◽  
Author(s):  
Toru Yamamoto
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Soil Samples ◽  
Heavy Elements ◽  
Severe Accident ◽  
Accident Analysis ◽  
Radioactivity Measurement ◽  
Power Station ◽  
The Core

Based on radioactivity measurement of soil samples in the site of Fukushima Dai-Ichi Nuclear Power Station, radioactivity of Sr, Nb, Mo, Tc, Ru, Ag, Te, I, Cs, Ba, La, Pu, Am, and Cm isotopes were compiled as radioactivity ratios to 137Cs. By exponentially fitting or averaging, the radioactivity ratios at the core shutdown were estimated. They were divided by those of the fuel of the core at the shutdown to obtain a deposited radioactivity fractions of the nuclides as relative values to 137Cs, which also correspond to deposition fractions of the elements as relative values to Cs. They were estimated to be orders of 10−4 to 10−3 for Sr, 10−4 for Nb, 10−2 to 10−1 for Mo, 10−1 for Ag, 10−1 to 100 for Te, 100 for I, 10−3 for Ba, 10−6 to 10−5 for Pu, 10−6 to 10−5 for Am, and 10−6 for Cm. The observed radioactivity ratios to 137Cs were compared with those obtained by severe accident analysis to assess the validation of the analysis.

Modeling of Dryout in Core Debris Beds of Conical and Cylindrical Geometries

2012 ◽  
Author(s):  
Eveliina Takasuo ◽  
Ville Hovi ◽  
Mikko Ilvonen ◽  
Stefan Holmström
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Field ◽  
Nuclear Power ◽  
Experimental Results ◽  
Severe Accident ◽  
Accident Analysis ◽  
Simulation Results ◽  
Particle Bed ◽  
Good Agreement

A porous particle bed consisting of core debris may be formed as a result of a core melt accident in a nuclear power plant. The coolability of conical (heap-like) and cylindrical (evenly-distributed) ex-vessel debris beds have been investigated in the COOLOCE experiments at VTT. The experiments have been modeled by using the MEWA severe accident analysis code. The main objectives of the modeling were (1) to validate the simulation results against the experiments by comparing the dryout power density predicted by the code to the experimental results and (2) to evaluate the effect of geometry on the coolability by examining the flow field and the development of dryout in the two geometries. In addition to the MEWA simulations, 3D demonstration calculations of the particle bed dryout process have been performed using the in-house code PORFLO. It was found that the simulation and experimental results are in a relatively good agreement. The results suggest that the coolability of the conical debris bed is poorer than that of the cylindrical bed due to the greater height of the conical configuration.

Analysis of Inertization Strategies for the Filtered Containment Venting System in Cofrentes Nuclear Power Plant

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Kevin Fernández-Cosials ◽  
Gonzalo Jiménez ◽  
César Serrano ◽  
Luisa Ibáñez ◽  
Ángel Peinado
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Computational Model ◽  
Nuclear Power ◽  
Inert Atmosphere ◽  
Severe Accident ◽  
Nitrogen Injection

During a severe accident (SA) in a nuclear power plant (NPP), there are several challenges that need to be faced. To coup with a containment overpressure, the venting action will lower the pressure but it will release radioactivity to the environment. In order to reduce the radioactivity released, a filtered containment venting system (FCVS) can be used to retain iodine and aerosols radioactive releases coming from the containment atmosphere. However, during a SA, large quantities of hydrogen can also be generated. Hydrogen reacts violently with oxygen and its combustion could impair systems, components, or structures. For this reason, to protect the integrity of the FCVS against hydrogen explosions, an inertization system is found necessary. This system should create an inert atmosphere previous to any containment venting that impedes the contact of hydrogen and oxygen. In this paper, the inertization system for Cofrentes NPP is presented. It consists of a nitrogen injection located in three different points. A computational model of the FCVS as well as the inertization system has been created. The results show that if the nitrogen sweeps and the containment venting are properly synchronized, the hydrogen risk could be reduced to a minimum and therefore, the integrity of the FCVS would be preserved.

Estimation of Radiological Consequences at a Proposed Nuclear Power Plant Site Using Atmospheric Dispersion Code

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Kwame Gyamfi ◽  
Sylvester Attakorah Birikorang ◽  
Emmanuel Ampomah-Amoako ◽  
John Justice Fletcher
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Atmospheric Dispersion ◽  
Atmospheric Stability ◽  
Assessment System ◽  
Severe Accident ◽  
Dispersion Modeling ◽  
Stability Class ◽  
Total Effective Dose Equivalent

Abstract Atmospheric dispersion modeling and radiation dose calculation have been performed for a generic 1000 MW water-water energy reactor (VVER-1000) assuming a hypothetical loss of coolant accident (LOCA). Atmospheric dispersion code, International Radiological Assessment System (InterRAS), was employed to estimate the radiological consequences of a severe accident at a proposed nuclear power plant (NPP) site. The total effective dose equivalent (TEDE) and the ground deposition were calculated for various atmospheric stability classes, A to F, with the site-specific averaged meteorological conditions. From the analysis, 3.7×10−1 Sv was estimated as the maximum TEDE corresponding to a downwind distance of 0.1 km within the dominating atmospheric stability class (class A) of the proposed site. The intervention distance for evacuation (50 mSv) and sheltering (10 mSv) were estimated for different stability classes at different distances. The intervention area for evacuation ended at 0.5 km and that for sheltering at 1.5 km. The results from the study show that designated area for public occupancy will not be affected since the estimated doses were below the annual regulatory limits of 1 mSv.

Onset of Gas Blowthrough During Melt Expulsion From a Hole in a RPV: Numerical Analysis Using the SIMMER Code

2006 ◽  
Author(s):  
Frank Kretzschmar
Keyword(s):  
Numerical Analysis ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Core Material ◽  
Test Facility ◽  
Computer Codes ◽  
Direct Containment Heating ◽  
Heating Up

In the case of a severe accident in a nuclear power plant there is a residual risk, that the Reactor Pressure Vessel (RPV) does not withstand the thermal attack of the molten core material, of which the temperature can be about 3000 K. For the analysis of the processes governing melt dispersal and heating up of the containment atmosphere of a nuclear power plant in the case of such an event, it is important to know the time of the onset of gas blowthrough during the melt expulsion through the hole in the bottom of the RPV. In the test facility DISCO-C (Dispersion of Simulant Corium-Cold) at the FZK /6/, experiments were performed to furnish data for modeling Direct Containment Heating (DCH) processes in computer codes that will be used to extrapolate these results to the reactor case. DISCO-C models the RPV, the Reactor Coolant System (RCS), cavity and the annular subcompartments of a large European reactor in a scale 1:18. The liquid type, the initial liquid mass, the type of the driving gas and the size of the hole were varied in these experiments. We present results for the onset of the gas blowthrough that were reached by numerical analysis with the Multiphase-Code SIMMER. We compare the results with the experimental results from the DISCO-C experiments and with analytical correlations, given by other authors.

Level 2 PSA Overview of HPR1000 Nuclear Power Plant

2020 ◽  
Author(s):  
Wang Ziguan ◽  
Lu Fang ◽  
Yang Benlin ◽  
Chen Shi ◽  
Hu Lingsheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Injection System ◽  
Risk Level ◽  
Level 2

Abstract Risk-informed design approaches are comprehensively implemented in the design and verification process of HPR1000 nuclear power plant. Particularly, Level 2 PSA is applied in the optimization of severe accident prevention and mitigation measures to avoid the extravagant redundancy of system configurations. HPR1000 preliminary level 2 PSA practices consider internal events of the reactor in the context of at-power condition. Severe accidents mitigation and prevention system and its impact on the overall large release frequency (LRF) level are evaluated. The results showed that severe accident prevention and mitigation systems, such as fast depressurization system, the cavity injection system and the passive containment heat removal system perform well in reducing LRF and overall risk level of HPR1000 NPP. Bypass events, reactor rapture events, and the containment bottom melt-through induced by MCCI are among the dominant factors of the LRF. The level 2 PSA analysis results indicate that HPR1000 design is reliable with no major weaknesses.

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