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.

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.

Predicting Thermal Mixing and Fatigue Inside Control Rod Guide Tubes

2013 ◽  
Author(s):  
Eric Lillberg
Keyword(s):  
Thermal Fatigue ◽  
Turbulent Mixing ◽  
Low Frequency ◽  
Experimental Measurements ◽  
Guide Tube ◽  
Control Rod ◽  
Thermal Mixing ◽  
Purge Flow ◽  
Control Rods ◽  
Good Agreement

The cracked control rods shafts found in two Swedish NPPs were subjected to thermal fatigue due to mixing of cold purge flow with hot bypass water in the upper part of the top tube on which the control rod guide tubes rests. The interaction between the jets formed at the bypass water inlets is the main source of oscillation resulting in low frequency downward motion of hot bypass water into the cold purge flow. This ultimately causes thermal fatigue in the control rod shaft in the region below the four lower bypass water inlets. The transient analyses shown in this report were done to further investigate this oscillating phenomenon and compare to experimental measurements of water temperatures inside the control rod guide tube. The simulated results show good agreement with experimental data regarding all important variables for the estimation of thermal fatigue such as peak-to-peak temperature range, frequency of oscillation and duration of the temperature peaks. The results presented in this report show that CFD using LES methodology and the open source toolbox OpenFOAM is a viable tool for predicting complex turbulent mixing flows and thermal loads.

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.

Flow mixing inside a control-rod guide tube – Experimental tests and CFD simulations

2011 ◽  
Vol 241 (12) ◽  
pp. 4803-4812 ◽  
Author(s):  
Kristian Angele ◽  
Ylva Odemark ◽  
Mathias Cehlin ◽  
Bengt Hemström ◽  
Carl-Maikel Högström ◽  
...  
Keyword(s):  
Experimental Tests ◽  
Guide Tube ◽  
Cfd Simulations ◽  
Control Rod ◽  
Flow Mixing

scholarly journals Experimental investigation of flow-induced vibration in the control rod guide cylinder of pressurized-water reactor

2019 ◽  
Vol 97 (1) ◽  
pp. 14-22
Author(s):  
Dao-Gang Lu ◽  
Hui-Min Zhang ◽  
Yuan-Peng Wang
Keyword(s):  
Experimental Investigation ◽  
Flow Rate ◽  
Experimental Results ◽  
Pressurized Water Reactor ◽  
Flow Induced Vibration ◽  
Pressurized Water ◽  
Control Rod ◽  
Water Reactor ◽  
Experiment Platform ◽  
Control Rods

An experiment was conducted to investigate flow-induced vibration (FIV) in the control rods of a pressurized-water reactor (PWR). Control rods and a guide cylinder (full scale compared with the real structure) were installed on an FIV experiment platform, with which flow distributions were simulated according to actual situations. The vibration displacements of the control rods were observed in different flow rate ranges of transverse and vertical flows. Several FIV characteristics of the control rods under different transverse and vertical flows were determined by analyzing the experimental results. A formula was also proposed to predict vibration.

Comparison of DSMC and CFD Models of Heat Transfer in a Rarefied Two-Dimensional Geometry

2018 ◽  
Author(s):  
Dilesh Maharjan ◽  
Mustafa Hadj-Nacer ◽  
Miles Greiner ◽  
Stefan K. Stefanov
Keyword(s):  
Heat Transfer ◽  
Rarefied Gas ◽  
Complex Geometry ◽  
Direct Simulation Monte Carlo ◽  
Accurate Method ◽  
Vacuum Drying ◽  
Helium Pressure ◽  
Two Dimensional ◽  
Dsmc Method ◽  
Cfd Simulations

During vacuum drying of used nuclear fuel (UNF) canisters, helium pressure is reduced to as low as 67 Pa to promote evaporation and removal of remaining water after draining process. At such low pressure, and considering the dimensions of the system, helium is mildly rarefied, which induces a thermal-resistance temperature-jump at gas–solid interfaces that contributes to the increase of cladding temperature. It is important to maintain the temperature of the cladding below roughly 400 °C to avoid radial hydride formation, which may cause cladding embrittlement during transportation and long-term storage. Direct Simulation Monte Carlo (DSMC) method is an accurate method to predict heat transfer and temperature under rarefied condition. However, it is not convenient for complex geometry like a UNF canister. Computational Fluid Dynamics (CFD) simulations are more convenient to apply but their accuracy for rarefied condition are not well established. This work seeks to validate the use of CFD simulations to model heat transfer through rarefied gas in simple two-dimensional geometry by comparing the results to the more accurate DSMC method. The geometry consists of a circular fuel rod centered inside a square cross-section enclosure filled with rarefied helium. The validated CFD model will be used later to accurately estimate the temperature of an UNF canister subjected to vacuum drying condition.

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