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

2005 ◽  
Vol 152 (3) ◽  
pp. 302-313 ◽  
Author(s):  
Y. Liao ◽  
K. Vierow
Keyword(s):  
Steam Generator ◽  
Creep Rupture ◽  
Severe Accident ◽  
Steam Generator Tube ◽  
Station Blackout

Evaluation of the Creep Rupture Behavior of Alloy 690 Steam Generator Tubes Considering the Pressure Ramp Rate

2021 ◽  
Author(s):  
Jongmin Kim ◽  
Min-Chul Kim ◽  
Joonyeop Kwon
Keyword(s):  
Steam Generator ◽  
Creep Rupture ◽  
Parameter Method ◽  
Alloy 600 ◽  
Steam Generator Tube ◽  
Failure Pressure ◽  
Alloy 690 ◽  
Rupture Model ◽  
Temperature And Pressure ◽  
Steam Generator Tubes

Abstract The materials used previously for steam generator tubes around the world have been replaced and will be replaced by Alloy 690 given its improved corrosion resistance relative to that of Alloy 600. However, studies of the high- temperature creep and creep-rupture characteristics of steam generator tubes made of Alloy 690 are insufficient compared to those focusing on Alloy 600. In this study, several creep tests were conducted using half tube shape specimens of the Alloy 690 material at temperatures ranging from 650 to 850C and stresses in the range of 30 to 350 MPa, with failure times to creep rupture ranging from 3 to 870 hours. Based on the creep test results, creep life predictions were then made using the well-known Larson Miller Parameter method. Steam generator tube rupture tests were also conducted under the conditions of a constant temperature and pressure ramp using steam generator tube specimens. The rupture test equipment was designed and manufactured to simulate the transient state (rapid temperature and pressure changes) in the event of a severe accident condition. After the rupture test, the damage to the steam generator tubes was predicted using a creep rupture model and a flow stress model. A modified creep rupture model for Alloy 690 steam generator tube material is proposed based on the experimental results. A correction factor of 1.7 in the modified creep rupture model was derived for the Alloy 690 material. The predicted failure pressure was in good agreement with the experimental failure pressure.

An analysis on the consequences of a severe accident initiated steam generator tube rupture

2019 ◽  
Vol 348 ◽  
pp. 14-23 ◽  
Author(s):  
JinHo Song ◽  
ByungHee Lee ◽  
SungIl Kim ◽  
GwangSoon Ha
Keyword(s):  
Steam Generator ◽  
Severe Accident ◽  
Steam Generator Tube

Natural Circulation and Creep Damage Analyses of Severe Accident for a Typical PWR Plant

2009 ◽  
Author(s):  
Osamu Kawabata ◽  
Masao Ogino
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Reactor Core ◽  
Creep Damage ◽  
Reactor Vessel ◽  
Reactor System ◽  
Severe Accident ◽  
Steam Generator Tube ◽  
Hot Gas ◽  
Core Meltdown

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.

Evaluation of Creep Properties of Alloy 690 Steam Generator Tubes at High Temperature Using Tube Specimen

2019 ◽  
Author(s):  
Jongmin Kim ◽  
Woogon Kim ◽  
Minchul Kim
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Creep Test ◽  
Rupture Life ◽  
Creep Rupture ◽  
Severe Accident ◽  
High Temperature Creep ◽  
Creep Life ◽  
Temperature Creep ◽  
Alloy 690

Abstract Thermally induced steam generator (SG) tube failures caused by hot gases from a damaged reactor core can result in a containment bypass event and may lead to release of fission products to the environment. A typical severe accident scenario is a station blackout (SBO) with loss of auxiliary feedwater. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690. Based on reported creep data and creep test results of Alloy 690 in this study, creep life extrapolation was carried out using Larson-Miller Parameter (LMP), Orr-Sherby-Dorn (OSD), Manson-Haferd Parameter (MHP), and Wilshire’s approach. And a hyperbolic sine (sinh) function to determine master curves in LMP, OSD and MHP methods was used for improving the creep life estimation of Alloy 690 material.

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

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.

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

2002 ◽  
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.

scholarly journals Evaluation of the Potential for Containment Bypass due to Steam Generator Tube Rupture in VVER-1000/V320 Reactor during Extended SBO sequence using SCDAP/RELAP5 code

2021 ◽  
Vol 9 (4) ◽  
pp. 9-15
Author(s):  
Van Thai Nguyen ◽  
Manh Long Doan ◽  
Chi Thanh Tran
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Severe Accident ◽  
Pressure Relief ◽  
Plant Behavior ◽  
Pressure Relief Valves ◽  
Station Blackout ◽  
Coolant System ◽  
Relap5 Code ◽  
Tube Failure

A severe accident-induced of a Steam Generator (SG) tube releases radioactivity from the Reactor Coolant System (RCS) into the SG secondary coolant system from where it may escape to the environment through the pressure relief valves and an environmental release in this manner is called “Containment Bypass”. This study aims to evaluate the potential for “Containment Bypass” in VVER/V320 reactor during extended Station Blackout (SBO) scenarios that challenge the tubes by primarily involving a natural circulation of superheated steam inside the piping loop and then induce creep rupture tube failure. Assessments are made of SCDAP/RELAP5 code capabilities for predicting the plant behavior during an SBO event and estimates are made of the uncertainties associated with the SCDAP/RELAP5 predictions for key fluid and components condition and for the SG tube failure margins. 

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