Effect of CRGT Cooling on Modes of Global Vessel Failure of a BWR Lower Head

2012 ◽  
Author(s):  
Walter Villanueva ◽  
Chi-Thanh Tran ◽  
Pavel Kudinov
Keyword(s):  
Heat Transfer ◽  
Vessel Wall ◽  
Reactor Vessel ◽  
Creep Model ◽  
Guide Tube ◽  
Mechanical Loads ◽  
Control Rod ◽  
Melt Pool ◽  
Creep Analysis ◽  
Lower Head

An in-vessel stage of a severe core melt accident in a Nordic type Boiling Water Reactor (BWR) is considered wherein a decay-heated pool of corium melt inflicts thermal and mechanical loads on the lower-head vessel wall. This process induces creep leading to a mechanical failure of the reactor vessel wall. The focus of this study is to investigate the effect of Control Rod Guide Tube (CRGT) and top cooling on the modes of global vessel failure of the lower head. A coupled thermo-mechanical creep analysis of the lower head is performed and cases with and without CRGT and top cooling are compared. The debris bed heat-up, re-melting, melt pool formation, and heat transfer are calculated using the Phase-change Effective Convectivity Model and transient heat transfer characteristics are provided for thermo-mechanical strength calculations. The creep analysis is performed with the modified time hardening creep model and both thermal and integral mechanical loads on the reactor vessel wall are taken into account. Known material properties of the reactor vessel as a function of temperature, including the creep curves, are used as an input data for the creep analysis. It is found that a global vessel failure is imminent regardless of activation of CRGT and top cooling. However, if CRGT and top cooling is activated, the mode and timing of failure is different compared to the case with no CRGT and top cooling. More specifically, with CRGT and top cooling, there are two modes of global vessel failure depending on the size of the melt pool: (a) ‘ballooning’ of the vessel bottom for smaller pools, and (b) ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool for larger pools. Without CRGT and top cooling, only a ballooning mode of global vessel failure is observed. Furthermore, a considerable delay (about 1.4 h) on the global vessel failure is observed for the roughly 30-ton debris case if CRGT and top cooling is implemented. For a much larger pool (roughly 200-ton debris), no significant delay on the global vessel failure is observed when CRGT and top cooling is implemented, however, the liquid melt fraction and melt superheat are considerably higher in non-cooling case.

Stratification and Heat Transfer of Molten Corium Pool for In-Vessel Retention

2020 ◽  
Author(s):  
Qingan Xiang ◽  
Jian Deng ◽  
Dahuan Zhu ◽  
Xiaoli Wu ◽  
Jinsheng Bi ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Thermal Loading ◽  
Metal Layer ◽  
Safety Margin ◽  
Light Metal ◽  
Reactor Vessel ◽  
Physicochemical Interaction ◽  
Thermal Failure ◽  
Mechanical Loads ◽  
Lower Head

Abstract In-vessel retention (IVR) consists in cooling the molten corium contained in the lower head of reactor vessel by natural convection and reactor cavity flooding. The general approach which is used to study IVR problems is a “bounding” approach which consists in assuming a specified corium pool stratification in the vessel and then demonstrating that the vessel can cope with the resulting thermal and mechanical loads. Thermal loading on the vessel is controlled by the convective heat transfer inside the molten corium pool. Traditionally molten corium pool in the lower head was expected to stratify into two-layer with the dense oxide pool at the bottom and the light metal layer on the top. Based on the MASCA experiments, the increased density of the metal layer attributed to a transfer of uranium metal leads to inverse stratification with a heavy metal layer relocating below the oxide pool. This behavior can be explained by physicochemical interaction between the oxidic and metallic phases of the corium pool. Therefore, a methodology which couples physicochemical effects and thermal hydraulics has been developed to address the IVR issue. The main purpose of this paper is to present this methodology and its application for calculate stratification probability of two-layer and three-layer configuration, analyze the safety margin of IVR in two-layer and three-layer configuration, and evaluate the lower head heat thermal failure probability.

Core Melt Solidification Characteristics in PRV Lower Head-Experimental Results From LIVE Tests

2010 ◽  
Vol 132 (10) ◽  
Author(s):  
Xiaoyang Gaus-Liu ◽  
Alexei Miassoedov ◽  
Thomas Cron ◽  
Jerzy Foit ◽  
Thomas Wenz ◽  
...  
Keyword(s):  
Vessel Wall ◽  
Cooling Water ◽  
Reactor Vessel ◽  
Water Cooling ◽  
Melt Pool ◽  
External Cooling ◽  
Time Period ◽  
Crust Layer ◽  
Lower Head ◽  
Melt Solidification

Core melt solidification phenomena in the lower plenum of pressurized reactor vessel during external reactor vessel cooling is investigated in late in-vessel phase experiment tests under different external cooling conditions and melt pouring positions. The melt solidification behavior, which has not yet been given sufficient attention, is an important issue since it influences not only the transient but also the steady state of melt pool thermal hydraulics. A noneutectic melt (80 mol %KNO3–20 mol %NaNO3) was used to simulate the core melt. It has been found out that when the vessel is cooled with water during the whole test period (water cooling), the cooling is more effective than the case that the vessel lower head is first cooled with air and flooded by water (air/water cooling). Water cooling at the beginning leads to faster buildup of crust layer on the vessel inner wall and lower crust thermal conductivity compared with air/water cooling. In comparison with the air/water cooling, the water cooling also achieves shorter time period of crust growth. During the solidification period in all tests, the constitutional supercooling condition is fulfilled. Pouring position near the vessel wall results in considerable asymmetry in the heat flux distribution through the vessel wall.

Heat Transfer Coefficient Measurement for Downward-Facing Heat Transfer on Curved Rectangular Water Channel

2016 ◽  
Author(s):  
Jun Yeong Jung ◽  
Yong Hoon Jeong
Keyword(s):  
Heat Transfer ◽  
Test Section ◽  
Flow Boiling ◽  
Water Channel ◽  
Outer Wall ◽  
Reactor Vessel ◽  
Working Fluid ◽  
Heater Surface ◽  
Ir Camera ◽  
Lower Head

In-Vessel Retention by External Reactor Vessel Cooling (IVR-ERVC) is method of removing the decay heat by cooling reactor vessel after corium relocation, and is also one of severe accident management strategies. Estimating heat transfer coefficients (HTCs) is important to evaluate heat transfer capability of the ERVC. In this study, the HTCs of outer wall of reactor vessel lower head were experimentally measured under the IVR-ERVC situation of Large Loss of Coolant Accident (LLOCA) condition. Experimental equipment was designed to simulate flow boiling condition of ERVC natural circulation, and based on APR+ design. This study focused on effects of real reactor vessel geometry (2.5 m of radius curvature) and material (SA508) for the HTCs. Curved rectangular water channel (test section) was design to simulate water channel which is between the reactor vessel lower head outer wall and thermal insulator. Radius curvature, length, width and gap size of the test section were respectively 2.5 m, 1 m, 0.07 m and 0.15 m. Two connection parts were connected at inlet and outlet of the test section to maintain fluid flow condition, and its cross section geometry was same with one of test section. To simulate vessel lower head outer wall, thin SA508 plate was used as main heater, and test section supported the main heater. Thickness, width, length and radius curvature of the main heater were 1.2 mm, 0.07 m, 1 m and 2.5 m respectively. The main heater was heated by DC rectifier, and applied heat flux was under CHF value. The test section was changed for each experiment. The HTCs of whole reactor vessel lower head (bottom: 0 ° and top: 90 °) were measured by inclining the test section, and experiments were conducted at four angular ranges; 0–22.5, 22.5–45, 45–67.5 and 67.5–90 °. DI water was used as working fluid in this experiment, and all experiments were conducted at 400 kg/m2s of constant mass flux with atmospheric pressure. The working fluid temperatures were measured at two point of water loop by K-type thermocouple. The main heater surface temperatures were measured by IR camera. The main heater was coated by carbon spray to make uniform surface emissivity, and the IR camera emissivity calibration was also conducted with the coated main heater. The HTCs were calculated by measured main heater surface temperature. In this research, the HTC results of 10, 30, 60 and 90 ° inclination angle were presented, and were plotted with wall super heat.

In-Vessel Retention Modeling Capabilities of SCDAP/RELAP5-3D©

2002 ◽  
Author(s):  
D. L. Knudson ◽  
J. L. Rempe
Keyword(s):  
Heat Transfer ◽  
Molten Pool ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Complex Processes ◽  
Lower Head ◽  
Environmental Laboratory ◽  
Molten Materials ◽  
Core Materials

Molten core materials may relocate to the lower head of a reactor vessel in the latter stages of a severe accident. Under such circumstances, in-vessel retention (IVR) of the molten materials is a vital step in mitigating potential severe accident consequences. Whether IVR occurs depends on the interactions of a number of complex processes including heat transfer inside the accumulated molten pool, heat transfer from the molten pool to the reactor vessel (and to overlying fluids), and heat transfer from exterior vessel surfaces. SCDAP/RELAP5-3D© has been developed at the Idaho National Engineering and Environmental Laboratory to facilitate simulation of the processes affecting the potential for IVR, as well as processes involved in a wide variety of other reactor transients. In this paper, current capabilities of SCDAP/RELAP5-3D© relative to IVR modeling are described and results from typical applications are provided. In addition, anticipated developments to enhance IVR simulation with SCDAP/RELAP5-3D© are outlined.

Flow Mixing Inside a Control-Rod Guide Tube: Part II—Experimental Tests and CFD-Simulations

2010 ◽  
Author(s):  
Kristian Angele ◽  
Mathias Cehlin ◽  
Carl-Maikel Ho¨gstro¨m ◽  
Ylva Odemark ◽  
Mats Henriksson ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Temperature Fluctuations ◽  
Thermal Fluctuations ◽  
Experimental Results ◽  
Flow Structures ◽  
Scaling Analysis ◽  
Guide Tube ◽  
Cfd Simulations ◽  
Control Rod ◽  
Control Rods

A large number of control rod cracks were detected during the refuelling outage of the twin reactors Oskarshamn 3 and Forsmark 3 in the fall of 2008. The extensive damage investigation finally lead to the restart of both reactors at the end of 2008 under the condition that further studies would be conducted in order to clarify all remaining matters. Also, all control rods were inserted 14% in order to locate the welding region of the control rod stem away from the thermal mixing region of the flow. Unfortunately, this measure led to new cracks a few months later due to a combination of surface finish of the new stems and the changed flow conditions after the partial insertion of the control rods. The experimental evidence reported here shows an increase in the extension of the mixing region and in the intensity of the thermal fluctuations. As a part of the complementary work associated with the restart of the reactors, and to verify the CFD simulations, experimental work of the flow in the annular region formed by the guide tube and control rod stem was carried out. Two full-scale setups were developed, one in a Plexiglass model at atmospheric conditions (in order to be able to visualize the mixing process) and one in a steel model to allow for a higher temperature difference and heating of the control rod guide tube. The experimental results corroborate the general information obtained through CFD simulations, namely that the mixing region between the cold crud-removal flow and warm by-pass flow is perturbed by flow structures coming from above. The process is characterized by low frequent, high amplitude temperature fluctuations. The process is basically hydrodynamic, caused by the downward transport of flow structures originated at the upper bypass inlets. The damping thermal effects through buoyancy is of secondary importance, as also the scaling analysis shows, however a slight damping of the temperature fluctuations can be seen due to natural convection due to a pre-heating of the cold crud-removal flow. The comparison between numerical and experimental results shows a rather good agreement, indicating that experiments with plant conditions are not necessary since, through the existing scaling laws and CFD-calculations, the obtained results may be extrapolated to plant conditions. The problem of conjugate heat transfer has not yet been addressed experimentally since complex and difficult measurements of the heat transfer have to be carried out. This type of measurements constitutes one of the main challenges to be dealt with in the future work.

Large eddy simulation on turbulent heat transfer in reactor vessel lower head corium pools

2018 ◽  
Vol 111 ◽  
pp. 293-302 ◽  
Author(s):  
Luteng Zhang ◽  
Simin Luo ◽  
Yapei Zhang ◽  
Wenxi Tian ◽  
G.H. Su ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Large Eddy Simulation ◽  
Reactor Vessel ◽  
Turbulent Heat Transfer ◽  
Turbulent Heat ◽  
Eddy Simulation ◽  
Large Eddy ◽  
Lower Head

Flow Mixing Inside a Control-Rod Guide Tube: Part I—CFD Simulations

2010 ◽  
Author(s):  
Hernan Tinoco ◽  
Hans Lindqvist ◽  
Ylva Odemark ◽  
Carl-Maikel Ho¨gstro¨m ◽  
Kristian Angele
Keyword(s):  
Heat Transfer ◽  
Thermal Fatigue ◽  
Guide Tube ◽  
Cfd Simulations ◽  
Control Rod ◽  
Flow Simulations ◽  
Cfd Models ◽  
Flow Mixing ◽  
And Control ◽  
Control Rods

Two broken control rods and a large number of rods with cracks were found at the inspection carried out during the refueling outage of the twin reactors Oskarshamn 3 and Forsmark 3 in the fall of 2008. As a part of an extensive damage investigation, time dependent CFD simulations of the flow and the heat transfer in the annular region formed by the guide tube and control rod stem were carried out, [1]. The simulations together with metallurgical and structural analyses indicated that the cracks were initiated by thermal fatigue. The knowledge assembled at this stage was sufficient to permit the restart of both reactors at the end of year 2008 conditioned to that further studies to be carried out for clarifying all remaining matters. Additionally, all control rods were inserted 14% to protect the welding region of the stem. Unfortunately, this measure led to new cracks a few months later. This matter will be explained in the second part of this work, [2]. As a part of the accomplished complementary work, new CFD models were developed in conformity with the guidelines of references [3] and [4]. The new results establish the simulation requirements needed to accomplish accurate conjugate heat transfer predictions. Those requirements are much more rigorous than the ones needed for flow simulations without heat transfer. In the present case, URANS simulations, which are less resource consuming than LES simulations, seem to rather accurately describe the mixing process occurring inside the control rod guide tube. Structure mechanics analyses based on the CFD simulations show that the cracks are initiated by thermal fatigue and that their propagation and growth are probably enhanced by mechanical vibrations.

CFD Analysis of PWR Reactor Vessel Upper Plenum Sections: Flow Simulation in Control Rods Guide Tubes

2010 ◽  
Author(s):  
Min-Tsung Kao ◽  
Chung-Yun Wu ◽  
Ching-Chang Chieng ◽  
Yiban Xu ◽  
Kun Yuan ◽  
...  
Keyword(s):  
Stokes Equations ◽  
Reactor Vessel ◽  
Unit Cell ◽  
Guide Tube ◽  
Control Rod ◽  
Geometry Model ◽  
Mesh Sensitivity ◽  
Representative Unit Cell ◽  
Sensitivity Studies ◽  
Domain Models

The AP1000™ PWR reactor vessel upper plenum contains numerous control rod guide tubes and support columns. Below the upper plenum are the upper core plate and the top core region of the fuel assemblies. Before detailed CFD simulations of the flow in the entire upper plenum and top core regions are performed, conducting local simulations for smaller sections of the domain can provide crucial and detailed physical aspects of the flow. These sub-domain models can also be used to perform mesh sensitivities and to assess what geometrical details may be eliminated from the larger model in order to reduce mesh size and computational requirements. The study discussed in this paper focuses on the sections of the domain related to the control rod guide tubes. The commercial CFD code STAR-CCM+ is employed to generate the computational mesh, to solve the Reynolds-averaged Navier-Stokes equations for incompressible flow with a Realizable k-epsilon turbulence model, and to post-process the results. Two sub-domains are modeled and analyzed: (1) a 1/4 section of one control rod guide tube by itself and (2) a representative unit cell containing two sections of adjacent control rod guide tubes and one 1/4 section of a neighboring support column. For the 1/4 guide tube model (sub-domain 1), trimmed meshes of up to 16 million cells are generated to compute the flow and pressure fields in both complete and simplified (without chamfers and narrow gaps) models. Comparisons of the results lead to the conclusion that the simplified geometry model might be used when developing larger domain models in the future. The representative unit cell (sub-domain 2) is assumed to be positioned in the center of the upper plenum where the global lateral flow effects are minimal. At this position, the lateral flows are generated mainly by the flow as it exits the guide tubes. After flow enters the unit cell from the bottom, there are three potential locations for flow to leave the unit cell: (1) lower locations near the support column and the upper core plate, (2) side windows in the lower portion of the guide tubes, and (3) upper locations near the guide plates positioned inside the guide tubes. Both trimmed and polyhedral meshes are generated as part of the mesh sensitivity studies. Predicted flow rates at each of the outflow locations in conjunction with results from the mesh sensitivity studies provide guidance on (1) what geometry to preserve or remove, (2) what geometry can be simplified to reduce the required mesh, and (3) an estimate of the total mesh required to model the control rod guide tube and support column geometry in the much larger simulation of the entire upper plenum and top fuel domain.

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