Quantitative Risk Assessment of Induced Steam Generator Tube Rupture in Severe Accident

When the primary reactor system remain pressurized during core meltdown for a typical PWR plant, loop seals formed in the primary reactor system would lead to natural circulations in hot leg and steam generator. In this case, the hot gas released from the reactor core moves to a steam generator, and a steam generator tube would be failed with cumulative creep damage. From such phenomena, a high-pressure scenario during core meltdown may lead to large release of fission products to the environment. In the present study, natural circulation and creep damage in the primary reactor system accompanying the hot gas generation in the reactor core were discussed and the combining analysis with MELCOR and FLUENT codes were performed to examine the natural circulation behavior. For a typical 4 loop PWR plant, MELCOR code which can analyze for the severe accident progression was applied to the accident analyses from accident initiation to reactor vessel failure for the accident sequence of the main steam pipe break which is maintained at high pressure during core meltdown. In addition, using the CFD code FLUENT, fluid dynamics in the reactor vessel plenum, hot leg and steam generator of one loop were simulated with three-dimensional coordinates. And the hot gas natural circulation flow and the heat transfer to adjoining structures were analyzed using results provided by the MELCOR code as boundary conditions. The both ratios of the natural circulation flow calculated in the hot leg and the steam generator using MELCOR code and FLUENT code were obtained to be about 2 (two). And using analytical results of thermal hydraulic analysis with both codes, creep damage analysis at hottest temperature points of steam generator tube and hot leg were carried out. The results in both cases showed that a steam generator tube would be failed with creep rupture earlier than that of hot leg rupture.

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MELCOR Analysis of Steam Generator Tube Creep Rupture in Station Blackout Severe Accident

Nuclear Technology ◽

10.13182/nt05-5 ◽

2005 ◽

Vol 152 (3) ◽

pp. 302-313 ◽

Cited By ~ 13

Author(s):

Y. Liao ◽

K. Vierow

Keyword(s):

Steam Generator ◽

Creep Rupture ◽

Severe Accident ◽

Steam Generator Tube ◽

Station Blackout

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A Modeling Method of the Branch Probability of the Temperature-Induced SGTR Phenomenon

Volume 4: Nuclear Safety, Security, Non-Proliferation and Cyber Security; Risk Management ◽

10.1115/icone25-66378 ◽

2017 ◽

Author(s):

Zibin Liu ◽

Dingqing Guo ◽

Bing Zhang ◽

Jinkai Wang

Keyword(s):

Steam Generator ◽

Nuclear Power ◽

Power Plants ◽

Modeling Method ◽

Radioactive Material ◽

Severe Accident ◽

Steam Generator Tube ◽

Modeling Methods ◽

Radioactive Release ◽

Level 2

The phenomenon of temperature-induced steam generator tube rupture (TI-SGTR) is a typical phenomenon in the severe accident process of nuclear power plants. The occurrence of the phenomenon may result in the radioactive material bypass the containment, causing a large radioactive release. This paper investigates modeling methods of the phenomenon of temperature-induced SGTR in level 2 PSA and presents an optimizing modeling method to calculate the probability of branching probability of TI-SGTR, aiming at improving the rationality and veracity of level 2 PSA.

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CFD Predictions of Severe Accident Steam Generator Flows in a 1/7th Scale Pressurized Water Reactor

10th International Conference on Nuclear Engineering, Volume 3 ◽

10.1115/icone10-22466 ◽

2002 ◽

Cited By ~ 1

Author(s):

Christopher Boyd ◽

Kelly Hardesty

Keyword(s):

Steam Generator ◽

Natural Circulation ◽

Severe Accident ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Steam Generator Tube ◽

Flow Phenomena ◽

Experimental Values ◽

High Degree ◽

Side Flow

Computational Fluid Dynamics (CFD) is applied to steam generator inlet plenum mixing as part of a larger plan covering steam generator tube integrity. The technique is verified by comparing predicted results with severe accident natural circulation data [1] from a 1/7th scale Westinghouse facility. This exercise demonstrates that the technique can predict the natural circulation and mixing phenomena relevant to steam generator tube integrity issues. The model includes primary side flow paths for a single hot leg and steam generator. Qualitatively, the experimentally observed flow phenomena are predicted. The paths of the natural circulation flows and the relative flow proportions are correctly predicted. Quantitatively, comparisons are made with temperatures, mass flows, and other parameters. All predictions are generally within 10% of the experimental values. Overall, there is a high degree of confidence in the CFD technique for prediction of the relevant flow phenomena associated with this type of severe accident sequence.

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The PSI Artist Project: Aerosol Retention and Accident Management Issues Following a Steam Generator Tube Rupture

10th International Conference on Nuclear Engineering, Volume 2 ◽

10.1115/icone10-22048 ◽

2002 ◽

Author(s):

Salih Gu¨ntay ◽

Abdel Dehbi ◽

Detlef Suckow ◽

Jon Birchley

Keyword(s):

Steam Generator ◽

Tube Bundle ◽

Severe Accident ◽

Pressurized Water ◽

Steam Generator Tube ◽

Full Size ◽

Management Actions ◽

Accident Management ◽

Dry Conditions ◽

Secondary Side

Steam generator tube rupture (SGTR) incidents, such as those, which occurred in various operating pressurized, water reactors in the past, are serious operational concerns and remain among the most risk-dominant events. Although considerable efforts have been spent to understand tube degradation processes, develop improved modes of operation, and take preventative and corrective measures, SGTR incidents cannot be completely ruled out. Under certain conditions, high releases of radionuclides to the environment are possible during design basis accidents (DBA) and severe accidents. The severe accident codes’ models for aerosol retention in the secondary side of a steam generator (SG) have not been assessed against any experimental data, which means that the uncertainties in the source term following an unisolated SGTR concurrent with a severe accident are not currently quantified. The accident management (AM) procedures aim at avoiding or minimizing the release of fission products from the SG. The enhanced retention of activity within the SG defines the effectiveness of the accident management actions for the specific hardware characteristics and accident conditions of concern. A sound database on aerosol retention due to natural processes in the SG is not available, nor is an assessment of the effect of management actions on these processes. Hence, the effectiveness of the AM in SGTR events is not presently known. To help reduce uncertainties relating to SGTR issues, an experimental project, ARTIST (AeRosol Trapping In a Steam generaTor), has been initiated at the Paul Scherrer Institut to address aerosol and droplet retention in the various parts of the SG. The test section is comprised of a scaled-down tube bundle, a full-size separator and a full-size dryer unit. The project will study phenomena at the separate effect and integral levels and address AM issues in seven distinct phases: Aerosol retention in 1) the broken tube under dry secondary side conditions, 2) the near field close to break under dry conditions, 3) the bundle far-field under dry conditions, 4) the separator and dryer under dry conditions, 5) the bundle section under wet conditions, 6) droplet retention in the separator and dryer sections and 7) the overall SG (integral tests). Prototypical test parameters are selected to cover the range of conditions expected in severe accident as well as DBA scenarios. This paper summarizes the relevant issues and introduces the ARTIST facility and the provisional test program which will run between 2003 and 2007.

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