Tarapur atomic power station: analysis of station blackout scenario
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A. Srivastava
Abstract
India is currently operating two BWR built by General Electric Company. The design features of these reactors are similar to the Fukushima's BWR except some better containment features in Indian BWR. This paper discusses the enveloping scenario of station blackout of infinite duration with no operator action and no component failure. The paper describes the details of modelling the TAPS-BWR plant model including SCDAP modelling of reactor core in system code RELAP5 and further thermal hydraulic safety assessment of station blackout scenario. The analysis brought out effectively the response of the plant to this high-pressure severe accident scenario. The time line of the severe accident progression will give details of various stages of accident progression along with hydrogen generation, which will be useful in evolving suitable severe accident management guidelines.
Kurzfassung
In diesem Beitrag werden Rechnungen mit RELAP5/SCDAP zum Notstromfall in den zwei KKW Tarapur vorgestellt. Beide KKW sind Siedewasserreaktoren und wurden von General Electric errichtet. Das Design ähnelt den SWR in Fukushima. Der zeitliche Ablauf des gewählten Szenarios wird im Beitrag vorgestellt.
References
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© 2015, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Evaluation of advanced displacement cross-sections for the major EUROFER constituents based on an atomistic modelling approach
- Tarapur atomic power station: analysis of station blackout scenario
- Estimating steady state and transient characteristics of molten salt natural circulation loop using CFD
- Thermal hydraulic analysis of ETRR-2 using RELAP5 code
- Transient performance during stopping the research reactor primary loop pump
- Neutronic performance calculations with alternative fluids in a hybrid reactor by using the Monte Carlo method
- A study on the criticality of modified neutron transport equation by using alternative scattering phase functions
- Production cross–section calculations of medical 32P, 117Sn, 153Sm and 186,188Re radionuclides used in bone pain palliation treatment
- Study on carbonated hydroxyapatite as a thermoluminescence dosimeter
- Social impact theory based modeling for security analysis in the nuclear fuel cycle
- Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Evaluation of advanced displacement cross-sections for the major EUROFER constituents based on an atomistic modelling approach
- Tarapur atomic power station: analysis of station blackout scenario
- Estimating steady state and transient characteristics of molten salt natural circulation loop using CFD
- Thermal hydraulic analysis of ETRR-2 using RELAP5 code
- Transient performance during stopping the research reactor primary loop pump
- Neutronic performance calculations with alternative fluids in a hybrid reactor by using the Monte Carlo method
- A study on the criticality of modified neutron transport equation by using alternative scattering phase functions
- Production cross–section calculations of medical 32P, 117Sn, 153Sm and 186,188Re radionuclides used in bone pain palliation treatment
- Study on carbonated hydroxyapatite as a thermoluminescence dosimeter
- Social impact theory based modeling for security analysis in the nuclear fuel cycle
- Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor
