Local Creep Damage Analysis of Coolant Piping in Nuclear Power Plant under Severe Accident Condition : Deformation and Distribution of Creep Damage

2000 ◽  
Vol 2000 (0) ◽  
pp. 453-454
Author(s):  
Seiya HAGIHARA ◽  
Noriyuki MIYAZAKI ◽  
Kazuichiro Hashimoto
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Creep Damage ◽  
Severe Accident ◽  
Damage Analysis ◽  
Accident Condition

Classification method of severe accident condition for the development of severe accident instrumentation and monitoring system in nuclear power plant

2015 ◽  
Vol 53 (6) ◽  
pp. 870-877 ◽  
Author(s):  
Akira Murata ◽  
Koichiro Isoda ◽  
Takeshi Ikeuchi ◽  
Tetsuya Matsui ◽  
Fujio Shiraishi ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Monitoring System ◽  
Nuclear Power ◽  
Classification Method ◽  
Severe Accident ◽  
Accident Condition

scholarly journals Radionuclide Disperse To Environtment From Pwr Reactor On Severe Accident Condition Using Maccs Program

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Sri Kuntjoro
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Dispersion Coefficient ◽  
Computer Code ◽  
Dispersion Analysis ◽  
Severe Accident ◽  
Plant Site ◽  
Accident Condition

<p>The atmosphere is an important pathway in the transfer of radionuclides from nuclear power plants into the environment and population. Acceptance of radiation dose to the environment and population affected by the radionuclides release and site conditions surrounding of the nuclear power plant. The radionuclides release in the atmosphere is determined by the dispersion coefficient parameter. The aim of this paper is to obtain dispersion coefficient and radionuclide released in Sebagin (West Bangka district) caused by severe accident condition from the PWR Nuclear Power Plant. Dispersion analysis of radionuclides into the environment from nuclear power PWR on severe accident conditions have been done using MACCS program. Reference for the calculation of source term fraction is selected from calculation results of the MELCOR computer code and it is implemented to PWR reactors Westinghouse 3411 MWth subject. The calculation of radionuclides release performed using MACCS program for aspiring nuclear power plant site in West Bangka. Simulation calculations for the area radius from 0.80 kmup to 20 km from the nuclear power plant site are performed. Meteorological datas used in calculation are the meteorology data from Sebagin meteorological stations for the years of 2012 period. The result is the dispersion coefficient decreases as a function of time and distance. The concentration of radionuclides through soil pathway decreases as a function of the distance, and the dominant contributor of radionuclide radiation Xe-133 and   I-131. Radionuclide concentrations obtained through the air pathway decreases as a function of distance, and dominant contributors of radionuclide radiation is contributed also from I-131 and Xe-133. The presence of I-131 radionuclides are giving dangerous to humans, it is necessary to further treatment for prevent its impacts. </p>

scholarly journals Corrections to “Irradiation Tests of Optical Fibers and Cables Devoted to Corium Monitoring in Case of a Severe Accident in a Nuclear Power Plant”

2020 ◽  
Vol 67 (6) ◽  
pp. 1195-1195
Author(s):  
G. Cheymol ◽  
L. Maurin ◽  
L. Remy ◽  
V. Arounassalame ◽  
H. Maskrot ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Optical Fibers ◽  
Nuclear Power ◽  
Severe Accident

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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