Experimental Investigation in LIFUS5/Mod2 Facility of Spiral-Tube Steam Generator Rupture Scenarios for ELFR

2017 ◽  
Author(s):  
Alessio Pesetti ◽  
Alessandro Del Nevo ◽  
Andrea Neri ◽  
Stefano Cati ◽  
Valerio Sermenghi ◽  
...  
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Quality Data ◽  
Water Tank ◽  
Spiral Tube ◽  
Pressure Transducers ◽  
Experimental Campaign ◽  
Evolution Path ◽  
Pressure Peak ◽  
Lower Temperature

In the framework of the European Commission LEADER project, an experimental campaign of seven tests was performed in the LIFUS5/Mod2 facility, at ENEA CR Brasimone, for investigating the postulated Steam Generator Tube Rupture (SGTR) event in a relevant configuration for the Spiral-Tube Steam Generator (STSG) of the European Lead Fast Reactor (ELFR). The LIFUS5/Mod2 facility is composed by a water tank of 15 L injecting subcooled water up to 200 bar into the reaction tank of 100 L (420 mm of diameter), which is connected by a 3 inch pipe to the dump tank of 2 m3. A dedicated test section was designed, assembled and implemented in the reaction tank. It is composed by 188 tubes, vertically disposed with triangular pitch inside a cylindrical support. This tube bundle is representative of a portion of the STSG of ELFR. The cylindrical support is closed at the lower and upper end by two tube plates and has a perforated lateral shell (300 mm of diameter and 400 mm high). The reaction tank is filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C up to the top tube plate, with an argon cover gas at about 2 bar. The water is injected at about 180 bar and 270°C through the central tube, at middle height of the bundle. The water-LBE interaction is characterised by high quality data acquisition system: 6 fast Pressure Transducers (PTs) working at 10 kHz for precisely characterize the first narrow injection peaks, 70 low constant time Thermocouples (TCs) to understand the vapour evolution path and 13 strain gages (SGGs) for measuring the strain of the bundle and main vessel. The overall LEADER experimental campaign is constituted by seven tests, divided in three series (B1, B2 and B3), characterized by different injection orifice diameters of 4, 8.9 and 12.6 mm, respectively. This paper presents the experimental results of the first two tests of series B2 (B2.1 and B2.2) having 8.9 mm of injection orifice. The first test analysed showed a first narrow pressure peak of about 32 bar, some milliseconds after the cap rupture instant. The following pressurization due to the evaporation of water entered into the reaction vessel was of an analogues magnitude for both the tests (about 50 bar) and lasted some tenths of second. The water/LBE interaction lower temperature was reached on the inner ranks of tubes, about 150°C. The outer rank was cooled down to about 300°C. The strain gage measurements showed a decreasing deformation on the tubes toward the outer positions. No ruptures were observed on tubes surrounding the injector. The amount of LBE transported into the dump tank was strongly dependent on the LBE level in the reaction tank at the start of the tests and about 200 kg.

Experimental Investigation of Spiral Tubes Steam Generator Rupture Scenarios in LIFUS5/Mod2 Facility for ELFR

2016 ◽  
Author(s):  
Alessio Pesetti ◽  
Alessandro Del Nevo ◽  
Nicola Forgione
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Injection Pressure ◽  
Strain Gages ◽  
Cylindrical Wall ◽  
Pressure Transducers ◽  
Evolution Path ◽  
High Injection Pressure ◽  
Pressure Peak ◽  
Lower Temperature

In the framework of the EC FP7 LEADER project, an experimental campaign was performed in the LIFUS5/Mod2 facility, at ENEA CR Brasimone, for investigating the postulated Steam Generator Tube Rupture (SGTR) event in a relevant configuration for the spiral tube Steam Generator (SG) of the European Lead Fast Reactor (ELFR). Two tests are analysed. The LIFUS5/Mod2 facility implemented a test section composed by 188 tubes, vertically disposed with triangular pitch, in a shell closed by top and bottom flanges and having a perforated cylindrical wall. The central tube injected water at about 180 bar and 270°C, at middle height of the tube bundle, in the reaction tank partially filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C with an argon cover gas at about 2 bar. It was connected to a 2 m3 dump tank, due to the high injection pressure. In the reaction tank fast instrumentation was set: 6 fast Pressure Transducers (PTs) acquiring data at 10 kHz for precisely characterize the first injection peaks; 70 low constant time Thermocouples (TCs) to understand the vapour evolution path; and 13 strain gages (SGGs) to measure the strain of the bundle and main vessel. The first test analysed showed a first pressure peak of about 25 bar, due to pressure wave propagation at the cap rupture instant. It did not appear in the second test as consequence of a leakage from the cap before the complete rupture. The following pressurization caused by the entering of water into the reaction vessel was of an analogues magnitude for both the tests (about 30 bar). The water/LBE interaction lower temperature was reached on the inner ranks of tubes, about 160°C. The outer rank was cooled down to 340°C. The strain gage measurements showed a decreasing deformation on the tubes toward the outer positions. No ruptures were observed on tubes surrounding the injector. The amount of LBE transported into the dump tank was strongly dependent on the LBE level in the reaction tank at the start of the tests.

Simulation of Fluid Flow Using Reduced-Order Modeling by POD Approach Applied to a Fixed Tube Bundle System

2010 ◽  
Author(s):  
Marie Pomarede ◽  
Erwan Liberge ◽  
Aziz Hamdouni ◽  
Elisabeth Longatte ◽  
Jean-Franc¸ois Sigrist
Keyword(s):  
Dynamic Analysis ◽  
Steam Generator ◽  
Tube Bundle ◽  
Fluid Structure Interaction ◽  
Reduced Model ◽  
Steam Generator Tube ◽  
Fluid Structure ◽  
Reduced Order ◽  
Structure Interaction ◽  
Tube Bundles

Tube bundles in steam boilers of nuclear power plants and nuclear on-board stokehold are known to be exposed to high levels of vibrations. This coupled fluid-structure problem is very complex to numerically set up, because of its three-dimensional characteristics and because of the large number of degrees of freedom involved. A complete numerical resolution of such a problem is currently not viable, all the more so as a precise understanding of this system behaviour needs a large amount of data, obtained by very expensive calculations. We propose here to apply the now classical reduced order method called Proper Orthogonal Decomposition to a case of 2D flow around a tube bundle. Such a case is simpler than a complete steam generator tube bundle; however, it allows observing the POD projection behaviour in order to project its application on a more realistic case. The choice of POD leads to reduced calculation times and could eventually allow parametrical investigations thanks to a low data quantity. But, it implies several challenges inherent to the fluid-structure characteristic of the problem. Previous works on the dynamic analysis of steam generator tube bundles already provided interesting results in the case of quiescent fluid [J.F. Sigrist, D. Broc; Dynamic Analysis of a Steam Generator Tube Bundle with Fluid-Structure Interaction; Pressure Vessel and Piping, July 27–31, 2008, Chicago]. Within the framework of the present study, the implementation of POD in academic cases (one-dimensional equations, 2D-single tube configuration) is presented. Then, firsts POD modes for a 2D tube bundle configuration is considered; the corresponding reduced model obtained thanks to a Galerkin projection on POD modes is finally presented. The fixed case is first studied; future work will concern the fluid-structure interaction problem. Present study recalls the efficiency of the reduced model to reproduce similar problems from a unique data set for various configurations as well as the efficiency of the reduction for simple cases. Results on the velocity flow-field obtained thanks to the reduced-order model computation are encouraging for future works of fluid-structure interaction and 3D cases.

Vibration Excitation Forces in a Rotated Triangular Tube Bundle Subjected to Two-Phase Cross Flow

2010 ◽  
Author(s):  
H. Senez ◽  
N. W. Mureithi ◽  
M. J. Pettigrew
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Cross Flow ◽  
Fretting Wear ◽  
Experimental Program ◽  
Phase Flow ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Fluid Forces ◽  
Vibration Excitation

Two-phase cross flow exists in many shell-and-tube heat exchangers. Flow-induced vibration excitation forces can cause tube motion that will result in long-term fretting wear or fatigue. Detailed flow and vibration excitation force measurements in tube bundles subjected to two-phase cross flow are required to understand the underlying vibration excitation mechanisms. Studies on this subject have already been done, providing results on flow regimes, fluidelastic instabilities, and turbulence-induced vibration. The spectrum of turbulence-induced forces has usually been expected to be similar to that in single-phase flow. However, a recent study, using tubes with a diameter larger than that in a real steam generator, showed the existence of significant quasi-periodic forces in two-phase flow. An experimental program was undertaken with a rotated-triangular array of cylinders subjected to air-water cross-flow, to simulate two-phase mixtures. The tube bundle here has the same geometry as that of a real steam generator. The quasi-periodic forces have now also been observed in this tube bundle. The present work aims to understand turbulence-induced forces acting on the tube bundle, providing results on drag and lift force spectra and their behaviour according to flow parameters, and describing their correlations. Detailed experimental test results are presented in this paper. Comparison is also made with previous measurements with larger diameter tubes. The present results suggest that quasi-periodic fluid forces are not uncommon in tube arrays subjected to two-phase cross-flow.

Proven Concepts for LLW-Treatment of Large Components for Free-Release and Recycling

2007 ◽  
Author(s):  
Lena Bergstro¨m ◽  
Maria Lindberg ◽  
Anders Lindstro¨m ◽  
Bo Wirendal ◽  
Joachim Lorenzen
Keyword(s):  
Steam Generator ◽  
Heat Exchangers ◽  
Tube Bundle ◽  
Pilot Project ◽  
Initial Weight ◽  
Steam Generators ◽  
Huge Number ◽  
Material Recycling ◽  
Waste Production ◽  
Sampling Program

This paper describes Studsvik’s technical concept of LLW-treatment of large, retired components from nuclear installations in operation or in decommissioning. Many turbines, heat exchangers and other LLW components have been treated in Studsvik during the last 20 years. This also includes development of techniques and tools, especially our latest experience gained under the pilot project for treatment of one full size PWR steam generator from Ringhals NPP, Sweden. The ambition of this pilot project was to minimize the waste volumes for disposal and to maximize the material recycling. Another objective, respecting ALARA, was the successful minimization of the dose exposure to the personnel. The treatment concept for large, retired components comprises the whole sequence of preparations from road and sea transports and the management of the metallic LLW by segmentation, decontamination and sorting using specially devised tools and shielded treatment cell, to the decision criteria for recycling of the metals, radiological analyses and conditioning of the residual waste into the final packages suitable for customer-related disposal. For e.g. turbine rotors with their huge number of blades the crucial moments are segmentation techniques, thus cold segmentation is a preferred method to keep focus on minimization of volumes for secondary waste. Also a variety of decontamination techniques using blasting cabinet or blasting tumbling machines keeps secondary waste production to a minimum. The technical challenge of the treatment of more complicated components like steam generators also begins with the segmentation. A first step is the separation of the steam dome in order to dock the rest of the steam generator to a specially built treatment cell. Thereafter, the decontamination of the tube bundle is performed using a remotely controlled manipulator. After decontamination is concluded the cutting of the tubes as well as of the shell is performed in the same cell with remotely controlled tools. Some of the sections of steam dome shell or turbine shafts can be cleared directly for unconditional reuse without melting after decontamination and sampling program. Experience shows that the amount of material possible for clearance for unconditional use is between 95 – 97% for conventional metallic scrap. For components like turbines, heat exchangers or steam generators the recycling ratio can vary to about 80–85% of the initial weight.

ALFRED Steam Generator Assessment: Design and Pre-Test Analysis of HERO Experiment

2018 ◽  
Author(s):  
Pierdomenico Lorusso ◽  
Alessio Pesetti ◽  
Mariano Tarantino
Keyword(s):  
High Pressure ◽  
Steam Generator ◽  
Working Conditions ◽  
Test Section ◽  
Tube Bundle ◽  
Secondary System ◽  
Test Analysis ◽  
Pressurized Water ◽  
Pressure Tests ◽  
Secondary Loop

In the framework of the ALFRED design (Advanced Lead Fast Reactor European Demonstrator) for DEMO-LFR, a new concept of steam generator (SG) has been proposed consisting in a double wall bayonet tube bundle which improves the plant safety reducing the possibility of water-lead interaction thanks to a double physical separation between them, and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas. In order to support the development of this innovative SG configuration, the ENEA has designed and realized the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section, a mock-up (1:1 in length) which represents the ALFRED SG. This test section, implemented in the CIRCE pool facility, aims to investigate on the thermal-hydraulic features of the system, providing a database for STH codes validation. The experimental campaign consists of high pressure tests at about 180 bar carried out in the framework of the HORIZON2020 SESAME project (Simulations and Experiments for the Safety Assessment of MEtal cooled reactors). The secondary loop has been realized for the HERO SG feeding, consisting in an open loop circuit fed by demineralized water. The system is equipped with a volumetric pump and a heater in order to reach the water nominal working conditions of 335°C at the SG inlet and about 180 bar at the outlet. A preliminary test analysis is carried out by RELAP5-3D© thermal-hydraulic system code. A numerical 1-D model of the HERO SG and the secondary loop has been realized in order to test the loop layout and to characterize the main components from a thermal-hydraulic point of view, defining the start-up procedures for the achievement of the working conditions of the water for the high pressure tests. Furthermore, several simulations are carried out to investigate on the secondary system behavior both for steady states and transients.

Flow-Induced Vibration Responses of U-Tube Bundle in Air-Water Flow

2007 ◽  
Author(s):  
In-Cheol Chu ◽  
Heung June Chung ◽  
Chang Hee Lee ◽  
Hyung Hyun Byun ◽  
Moo Yong Kim
Keyword(s):  
Steam Generator ◽  
Two Phase Flow ◽  
Tube Bundle ◽  
Damping Ratio ◽  
Geometrical Parameters ◽  
Phase Flow ◽  
Elastic Instability ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Vibration Responses

In the present study, a series of experiments have been performed to investigate a fluid-elastic instability of a nuclear steam generator U-tube bundle in an air-water two-phase flow condition. A total of 39 U-tubes are arranged in a rotated square array with a pitch-to-diameter ratio of 1.633. The diameter and other geometrical parameters of U-bend region are the same to those of an actual steam generator, but the vertical length of U-tubes are reduced to 2-span in contrast to 9-span of an actual steam generator. The following parameters were experimentally measured to evaluate a fluid-elastic instability of U-tube bundles in a two-phase flow: a general tube vibration response, a critical gap velocity, a damping ratio and a hydrodynamic mass. Based on the experimental measurements, the instability factor, K, of Connors’ relation was preliminary assessed with some assumptions on the velocity and density profiles of the two-phase flow.

Study on the Simplified Method of Special Supporting Structure for Steam Generator Tubes

2018 ◽  
Author(s):  
Wei Tan ◽  
Songyuan Jiang ◽  
Zhao Li ◽  
Zhanbin Jia ◽  
Liyan Liu
Keyword(s):  
Natural Frequency ◽  
Steam Generator ◽  
Tube Bundle ◽  
Safety Margin ◽  
Contact Length ◽  
Contact Condition ◽  
Special Support ◽  
Simplified Method ◽  
Simple Support ◽  
Supporting Structures

The flow induced vibration of tubes in the steam generator gradually attracts great attention. The natural frequency of the tube is the basic parameter for vibration analysis. The supporting structures of the tube bundle in steam generator are the cinquefoil orifice-baffle and the anti-vibration bar which is different from the common baffle plate. The issue that researchers focus on is how these supporting structures affect the natural frequency. A simplified method of the special support in the tube bundle was studied based on the numerical simulation. According to the characteristics of supporting structures, the effect of stiffness of the supporting structures on the natural frequency was studied by the spring element constraints. The results show that when the stiffness of the support structure is larger than the magnitude of 105 N/m, the stiffness has no influence on the natural frequency. The frictional force of the circumferential constraint inside the plane is too weak to constrain the tube so we need to pay more attention on the vibration inside the plane. Through changing the contact length of the support component and tubes, the effect of the contact condition on the natural frequency is studied. The results show that the contact condition has a certain effect on the natural frequency. When the support is simplified as simple support, the influence on natural frequency is small and the deviation is less than 1.5%. And there is a certain safety margin under the simplified method. In the calculation process, the special support can be simplified as simple support and the calculation results are relatively accurate and conservative.

scholarly journals Stern Twin-Propeller Effects on Harbor Infrastructures. Experimental Analysis

Water ◽  
2018 ◽  
Vol 10 (11) ◽  
pp. 1571 ◽  
Author(s):  
Anna Mujal-Colilles ◽  
Marcel·la Castells ◽  
Toni Llull ◽  
Xavi Gironella ◽  
Xavier Martínez de Osés
Keyword(s):  
Experimental Analysis ◽  
Water Tank ◽  
Subsequent Increase ◽  
Medium Scale ◽  
Erosion Rates ◽  
Experimental Campaign ◽  
Marine Traffic ◽  
Scouring Depth ◽  
Experimental Laboratory ◽  
Basin Area

The growth of marine traffic in harbors, and the subsequent increase in vessel and propulsion system sizes, produces three linked problems at the harbor basin area: (i) higher erosion rates damaging docking structures; (ii) sedimentation areas reducing the total depth; (iii) resuspension of contaminated materials deposited at the seabed. The published literature demonstrates that there are no formulations for twin stern propellers to compute the maximum scouring depth. Another important limitation is the fact that the formulations proposed only use one type of maneuvering during the experimental campaign, assuming that vessels are constantly being undocked. Trying to reproduce the real arrival and departure maneuvers, 24 different tests were conducted at an experimental laboratory in a medium-scale water tank using a twin propeller model to estimate the consequences and the maximum scouring depth produced by stern propellers during the backward/docking and forward/undocking scenarios. Results confirm that the combination of backward and forward scenario differs substantially from the experiments performed so far in the literature using only an accumulative forward scenario, yielding deeper scouring holes at the harbor basin area. The results presented in this paper can be used as guidelines to estimate the effects of regular vessels at their particular docking location.

Analysis of Heat Transfer and Flow Characteristics of AP1000 Passive Residual Heat Removal Heat Exchanger

2014 ◽  
Author(s):  
Xu Xie ◽  
Changhua Nie ◽  
Li Zhan ◽  
Hua Zheng ◽  
Pengzhou Li ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Temperature Difference ◽  
Tube Bundle ◽  
Natural Circulation ◽  
Heat Removal ◽  
Large Temperature ◽  
Water Tank ◽  
Secondary Side ◽  
Residual Heat

In this paper, the computational fluid dynamics (CFD) method is applied to the thermal-hydraulic analysis, while the porous media model is used to simplify AP1000 passive residual heat removal heat exchanger tube. The temperature as well as flow distribution in the secondary side of the heat exchanger are obtained, aiming at analysis of natural circulation ability. It can be noted that the fluid in the secondary side of heat exchanger moves driven by the effect of thermal buoyancy, forming the natural cycle, which takes away heat in tube bundle region. The heat transfer in water tank is mainly enhanced by vortex and turbulent flow, caused by the large resistance of tube bundle region as well as large temperature difference. This phenomenon is obvious especially for the recirculation of flow near the tube bundle. The enduring change of flow rate and direction enhance the heat transfer. Besides, the big temperature difference helps to increase the driving effect of natural circulation. Consequently, the heat transfer of the tank is enhanced by above mechanism. The results of this study contribute to the capacity analysis of passive residual heat removal of natural circulation system, providing valuable information for safe operation of AP1000.

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