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.

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.

Prediction of Severe Accident Counter Current Natural Circulation Flows in the Hot Leg of a Pressurized Water Reactor

2006 ◽  
Author(s):  
Christopher F. Boyd
Keyword(s):  
Steam Generator ◽  
Discharge Coefficient ◽  
Current Flow ◽  
Natural Circulation ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Flow Rates ◽  
Surge Line ◽  
Counter Current Flow ◽  
Counter Current

During certain phases of a severe accident in a pressurized water reactor (PWR), the core becomes uncovered and steam carries heat to the steam generators through natural circulation. For PWR’s with U-tube steam generators and loop seals filled with water, a counter current flow pattern is established in the hot leg. This flow pattern has been experimentally observed and has been predicted using computational fluid dynamics (CFD). Predictions of severe accident behavior are routinely carried out using severe accident system analysis codes such as SCDAP/RELAP5 or MELCOR. These codes, however, were not developed for predicting the three-dimensional natural circulation flow patterns during this phase of a severe accident. CFD, along with a set of experiments at 1/7th scale, have been historically used to establish the flow rates and mixing for the system analysis tools. One important aspect of these predictions is the counter current flow rate in the nearly 30 inch diameter hot leg between the reactor vessel and steam generator. This flow rate is strongly related to the amount of energy that can be transported away from the reactor core. This energy transfer plays a significant role in the prediction of core failures as well as potential failures in other reactor coolant system piping. CFD is used to determine the counter current flow rate during a severe accident. Specific sensitivities are completed for parameters such as surge line flow rates, hydrogen content, as well as vessel and steam generator temperatures. The predictions are carried out for the reactor vessel upper plenum, hot leg, a portion of the surge line, and a steam generator blocked off at the outlet plenum. All predictions utilize the FLEUNT V6 CFD code. The volumetric flow in the hot leg is assumed to be proportional to the square root of the product of normalized density difference, gravity, and hydraulic diameter to the 5th power. CFD is used to determine the proportionality constant in the range from 0.11 to 0.13 and termed a discharge coefficient. The value is relatively unchanged for typical surge line flow rates as well as the hydrogen content in the flow. Over a significant range of expected temperature differences for the steam generator and reactor vessel upper plenum, the discharge coefficient also remained consistent. The discharge coefficient is a suitable model for determining the hot leg counter current flow rates during this type of severe accident.

Assessment of RELAP/SCDAPSIM for Turbine Trip Transient in NuScale-SMR

2018 ◽  
Author(s):  
Katarzyna Skolik ◽  
Anuj Trivedi ◽  
Marina Perez-Ferragut ◽  
Chris Allison
Keyword(s):  
Natural Circulation ◽  
System Modeling ◽  
Thermal Power ◽  
Reactor Core ◽  
Operating Conditions ◽  
Severe Accident ◽  
Circulation System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Secondary Steam

The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe). Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses. The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.

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

Experimental and Numerical Investigation of Solidification of Gallium in an Initially Emptied Horizontal Pipe Flow

2020 ◽  
Author(s):  
Shawn Somers-Neal ◽  
Alex Pegarkov ◽  
Edgar Matida ◽  
Vinh Tang ◽  
Tarik Kaya
Keyword(s):  
Numerical Model ◽  
Molten Metal ◽  
Reactor Core ◽  
Experimental Tests ◽  
Severe Accident ◽  
Penetration Length ◽  
Horizontal Pipe ◽  
Cooling Pipe ◽  
Severe Accidents ◽  
Core Meltdown

Abstract In a reactor core meltdown under postulated severe accidents, the molten material (corium) could be ejected or relocated through existing vessel penetrations (cooling pipe connections), thus potentially contaminating other locations in the power plant. There exists, however, a potential for plugging of melt flow due to its complete solidification, providing the availability of an adequate heat sink. Therefore, a numerical model was created to simulate the flow of molten metal through an initially empty horizontal pipe. The numerical model was verified using a previously developed analytical model and validated against experimental tests with gallium (low melting temperature) as a substitute for corium. The numerical model was able to predict the penetration length (length of distance travelled by the molten metal) after a complete blockage occurred with an average percent error range of 9%. Since the numerical model has been verified and validated, the model can be updated to predict the penetration length in the cooling pipe in case of a severe accident.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

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.

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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